ML17054C229: Difference between revisions
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| number = ML17054C229 | | number = ML17054C229 | ||
| issue date = 01/17/2017 | | issue date = 01/17/2017 | ||
| title = | | title = License Amendment Request for the Transition to Westinghouse Core Design and Safety Analyses - Responses to NRC RAIs on August 2013 Methodology Transition LAR Submittal | ||
| author name = | | author name = | ||
| author affiliation = Wolf Creek Nuclear Operating Corp | | author affiliation = Wolf Creek Nuclear Operating Corp | ||
| Line 15: | Line 15: | ||
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50, Response to Request for Additional Information (RAI), Topical Report | | document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50, Response to Request for Additional Information (RAI), Topical Report | ||
| page count = 114 | | page count = 114 | ||
| | | project = | ||
| stage = Response to RAI | |||
}} | }} | ||
=Text= | |||
{{#Wiki_filter:Westinghouse Non-Proprietary Class 3 Page 1 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Responses to NRC RAIs on August 2013 Methodology Transition LAR Submittal (Non-Proprietary) | |||
November 2016 | |||
© 2016 Westinghouse Electric Company LLC All Rights Reserved | |||
Westinghouse Non-Proprietary Class 3 Page 2 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment WCNOC previously submitted a license amendment request (LAR) on August 13, 2013 (Reference 1) requesting approval of changes to the Wolf Creek Generating Station (WCGS) | |||
Technical Specifications (TSs). That application proposed the transition to Westinghouse core design and safety analysis methodologies, full scope implementation of Alternative Source Term (AST), and implementation of instrumentation setpoint and control uncertainty calculations based on the current Westinghouse Setpoint Methodology (including adoption of Option A of Technical Specification Task Force (TSTF) TSTF-493-A, Revision 4). WCNOC subsequently withdrew the LAR on June 18, 2014 (Reference 3) based on deficiencies discovered by WCNOC. | |||
References 2, 4, 24 and 25 are Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAI) related to Reference 1. WCNOC letters ET 14-0003 (Reference 20) and WO 14-0032 (Reference 22) provided responses to the RAIs received in Reference 24. WCNOC letter WO 14-0031 (Reference 21) provided responses to the RAIs received in Reference 25. | |||
WCNOC withdrew the LAR prior to providing a response to References 2 and 4. WCNOC has subsequently developed responses to the RAIs received in References 2 and 4. The responses are provided within this enclosure and the applicable information has been incorporated into the current LAR. | |||
Following the withdrawal of the LAR, two changes were made that affected the responses contained within References 20, 21, and 22. First, WCGS implemented the AST sump pH analysis (refer to Section 4.4 of Enclosure IV) as its licensing basis analysis under the 50.59 process. Second, it was determined that a modification would be implemented to support full implementation of AST. Specifically, the operator action to isolate the failed CREVS train (Discussed in Section 15A.3.1 of the USAR) has historically required local operator action. | |||
However, in order to reduce the difficulty of completing this time critical action and to eliminate dose exposure to a local operator, a design modification will be implemented to supply the CREVS control room isolation dampers with Class IE 120 VAC battery power. This modification will be implemented under the 50.59 process. Based upon these two changes, responses previously transmitted in References 20, 21, and 22 have been updated and are included within this enclosure (in addition to the unchanged responses previously provided). Specifically, the changes to the affected responses (EMCB-RAI-1, ESGB-RAI-1, and EEEB-RAI-2) are shown in bold. | |||
The responses to the RAIs are applicable to the current LAR and are therefore included in this Enclosure. Enclosure VIII provides supplemental documents that were requested by ARCB-RAI-33 and ARCB-RAI-36. The specific NRC request is provided in italics. | |||
Note that Reference 27 also provided NRC RAIs related to Reference 1. These RAIs are associated with the original proposed addition of TSTF-493 and, as such, no longer require a response. | |||
Westinghouse Non-Proprietary Class 3 Page 3 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 24 RAIs as clarified by Reference 26 EMCB-RAI-1 Implementation of an Alternate Source Term (AST) in accordance with Section 50.67, Accident source term, of Title 10 of the Code of Federal Regulations (10 CFR) could affect structures, systems, and components (SSCs) which were not previously evaluated for the consequences of a design-basis accident (DBA) and, as such, may not be seismically qualified. Appendix A to 10 CFR Part 100, Reactor Site Criteria, requires that SSCs necessary to assure the capability of the plant to mitigate the consequences of accidents, which could result in exposures comparable to the guideline exposures provided in 10 CFR Part 100, be designed to remain functional during and after a safe shutdown earthquake. In accordance with the 10 CFR Part 100 requirements, please identify SSCs which may be affected by the implementation of the proposed AST and address the following: | |||
a) Indicate whether any non-safety-related SSCs are being credited in the proposed AST license amendment. | |||
b) For any nonsafety-related SSCs credited in the AST, confirm that the SSCs have been seismically qualified in accordance with the plant licensing basis. | |||
c) Indicate whether the SSCs are new or existing. | |||
d) Describe the location of the SSCs and the seismic qualification method employed to demonstrate the seismic ruggedness of these SSCs, such as the plant licensing basis or an NRC-endorsed industry standard. | |||
e) Summarize the results of the seismic qualification of the equipment, indicating whether any modifications or re-design will be necessary in support of the AST. | |||
===Response=== | |||
The proposed license amendment request included the adoption of Alternative Source Term (AST) radiological analysis methodology consistent with 10 CFR 50.67, Accident source term. The design basis radiological consequence analyses performed for AST did not credit any non-safety related structures, systems, and components (SSCs). This is consistent with the current licensing basis analysis of record that does not credit any non-safety related SSCs. The design basis radiological consequences analyses performed for AST do not require the installation of any new SSCs (the damper modification being implemented will only change the damper power supply; it will not add new components). New cable trays or conduit may need to be installed to support the one modification being implemented, depending on cable tray and conduit availability. If the modification requires installation of a new cable tray or conduit for the power cables, the associated seismic evaluations will be performed in accordance with Wolf Creek methods used for existing cable trays and conduit. | |||
Westinghouse Non-Proprietary Class 3 Page 4 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ESGB-RAI-1 Describe the analysis methodology used to determine the pH in the sump water during the period of 30 days post-LOCA. Include detailed calculations of time dependent pH values in the sump during a 30 day period post-LOCA to demonstrate that the pH remains basic throughout this time period. | |||
===Response=== | |||
As discussed in Section 4.4.2 of Enclosure VI to Reference 1, the containment sump pH calculation did not include consideration of acid generation (nitric acid produced by the irradiation of water and air or hydrochloric acid produced by the radiolysis of chlorine bearing materials) as they were considered secondary effects. In Section 4.4.2 of Enclosure VI to Reference 1, WCNOC indicated that it was expected that the effect of acid generation on sump pH would decrease the pH value less than 0.1 pH based on a comparison of Byron Station information and the conclusion reached by the NRC in the safety evaluation for Amendment No. | |||
147 for the Byron Station and Amendment No. 140 for the Braidwood Station (Reference 23). | |||
WCNOC letter WO 14-0032 (Reference 22) provided the below details and results of the calculation. | |||
Methodology In order to calculate the minimum sump pH, the maximum amount of boric acid from the various sources of borated water that enter the containment sump and the strong acids that are generated from the effects of radiolysis post-LOCA combined with a minimum amount of caustic from the spray additive tank (SAT) will yield a minimum pH value. The concentrations of these substances are used to compute the value of the sump pH as a function of time using verified titration curve data for aqueous solutions of boric acid and sodium hydroxide. | |||
The current licensing basis (CLB) for the WCGS sump pH calculation (EN-03-W, Rev. 2, 18 Months Fuel Cycle, Cycle 4 Specific Boron - pH Calculations for TSA 20038-001, Rev. 3) forms the basis of existing Updated Safety Analysis Report (USAR) Figure 6.5-5. Two scenarios are analyzed: both containment spray trains operating with one NaOH eductor in service, and both containment spray trains operating with both NaOH eductors in service. | |||
Westinghouse Non-Proprietary Class 3 Page 5 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment No computer codes were used in the CLB calculation. To determine the pH in this calculation, NaOH and H3BO3 molarities were first calculated and then used with Oak Ridge National Laboratory (ORNL) titration curve data to determine the pH of the spray and sump solutions during the injection and recirculation phases of Emergency Core Cooling System (ECCS) operation. The CLB calculation did not consider the effects of the generation of strong acids inside containment post-LOCA. | |||
Westinghouse has subsequently performed an independent calculation (CN-SEE-I-13-10, Rev. 1, Wolf Creek Sump pH,) to update the conclusions of the CLB calculation for the transition to the Westinghouse Core Design and Safety Analysis methodologies. The Westinghouse calculation used pH data for boric acid/sodium hydroxide solutions from verified titration curve data for aqueous solutions of boric acid and sodium hydroxide. A revision of this calculation was performed in response to this RAI question (ESGB-RAI-1) and the effect of strong acid generation on the post-LOCA sump pH was considered. In accordance with the guidance of NUREG/CR-5950, Iodine Evolution and pH Control, the mass of nitric acid that is generated by the radiolysis of air and water inside containment and the mass of hydrochloric acid that is generated by the radiolysis of Hypalon electrical cable insulation inside containment were calculated. The amount of hydriodic acid that is generated from iodine released inside containment was judged to be negligible and was not considered in the sump pH calculation. | |||
The masses (moles) of the strong acids that are generated inside containment post-LOCA were assumed to instantaneously be neutralized by a molar equivalent of NaOH, which reduced the net mass of NaOH that was available to neutralize the boric acid injected from the Reactor Coolant System (RCS), accumulators and refueling water storage tank (RWST). The net effect of the neutralization of all acid species in the sump by the NaOH from the SAT determined the equilibrium sump pH. | |||
The CLB has been updated since the time the RAIs were sent and thus the current USAR figure is aligned with the AST analysis. | |||
Assumptions and Inputs Among the various assumptions that were made in the sump pH calculation is that of effective mixing of the water inventory in the sump. This is a valid assumption because of the long term operation of the Containment Spray System and the uniform dispersion of spray over the containment cross section. With respect to the generation of strong acids inside containment, the gaseous hydrochloric acid produced by electric cable insulation radiolysis was conservatively assumed to be instantly dissolved in the sump water, and the hydrochloric acid generated from the radiolysis of cable insulation and nitric acid generated by the irradiation of containment sump water and air was assumed to fully dissociate in the sump water. At the completion of the injection and mixing of all chemical species (acids and sodium hydroxide) inside containment post-LOCA, it was assumed that there would be no further change in the | |||
Westinghouse Non-Proprietary Class 3 Page 6 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment sump liquid inventory and, consequently, that there would be no further change in the calculated sump pH endpoint over the 30 day post-LOCA period. | |||
The minimum long-term sump pH in the revised Westinghouse calculation uses the following conservative bases: | |||
: 7. Maximum RWST and safety injection (SI) accumulator Technical Specification 3.5.1, Accumulators, and 3.5.4, Refueling Water Storage Tank (RWST), boron concentration of 2500 ppm. | |||
: 8. A conservatively high RCS boron concentration of 1900 ppm in the CLB, and 1980 ppm in the Westinghouse calculation. Given that the RCS is only approximately 15% of the sump fluid, this difference in boron concentration is a small effect. | |||
: 9. The SAT is assumed to contain the Technical Specification 3.6.7, Spray Additive System, minimum 28% by weight NaOH solution. | |||
: 10. The mass of Hypalon electrical cable insulation inside containment is 50,000 lbm. | |||
: 11. The 30-day integrated containment doses inside containment (beta and gamma sources) to which the Hypalon insulation is exposed post-LOCA were derived from the plant-specific report for the Environmental Qualification of Safety-Related Electrical Equipment. | |||
: 12. The maximum sump liquid mass is approximately 4.1x106 lbm, which corresponds to a liquid volume of 1.87x106 liters. | |||
While the Technical Specification 3.6.7 SAT minimum contained volume is 4340 gallons, the CLB analysis assumed a conservative minimum delivered volume of only 2960 gallons. The Westinghouse calculation used an adjusted delivered volume of 2752 gallons, which accounted for the neutralization of the strong acid generated inside containment post-LOCA. | |||
Results and Conclusions The Westinghouse calculation determined a long-term sump pH of 8.7, while the CLB calculation determined a minimum value of 8.6. The slightly higher pH in the Westinghouse calculation is attributed to the use of a different boron/NaOH/pH correlation. In either case, the results show that the sump pH remains above the CLB value of 8.5 and well above the NRC-required value of 7.0 after the minimum amount of NaOH is injected into the sump. | |||
In the CLB calculation, the limiting system alignment of both containment spray trains operating with one NaOH eductor in service results in a sump pH of 7 in approximately 15 minutes, with the final long-term pH of 8.6 in approximately 80 minutes. The less limiting case of both Containment Spray trains operating with both NaOH supplies in service results in a sump pH of 7 in approximately 11 minutes, with the final long-term pH of 8.6 in approximately 45 minutes. | |||
Westinghouse Non-Proprietary Class 3 Page 7 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment For the Westinghouse calculation, the limiting system alignment of both containment spray trains operating with one NaOH eductor in service results in a sump pH of 7 in approximately 11 minutes, with the final long-term pH of 8.7 in approximately 70 minutes. The less limiting case of both containment spray trains operating with both NaOH supplies in service results in a sump pH of 7 in approximately 9 minutes, with the final long-term pH of 8.7 in approximately 35 minutes. | |||
The figure below from the Wolf Creek USAR shows that the acceptance criteria of sump pH > | |||
8.5 (minimum) is met for one or two containment spray eductors in service, considering also the generation of strong acids inside containment post-LOCA, consistent with the guidance of NUREG/CR-5950. | |||
Westinghouse Non-Proprietary Class 3 Page 8 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 25 RAIs EEEB-RAI-1 Please confirm whether any loads are being added to the WCGS emergency diesel generators (EDGs). If so, describe their impact on the capability and capacity of the EDGs. Also, describe changes, if any, being made to the EDG loading sequence to support this license amendment request (LAR). | |||
===Response=== | |||
The proposed license amendment request included the adoption of Alternative Source Term (AST) radiological analysis methodology consistent with 10 CFR 50.67, Accident source term. | |||
The design basis radiological consequence analyses performed for AST did not result in the addition of loads to the onsite standby power source (train A and train B diesel generators (DGs)) and did not result in changes to the DG loading sequence. | |||
EEEB-RAI-2 Please confirm whether any components are being added to the Environmental Qualification (EQ) equipment list to comply with Title 10 of the Code of Federal Regulations, Section 50.49 (10 CFR 50.49) due to this LAR. If components are being added, describe the equipment qualification for the environmental conditions to which the components are expected to be exposed. | |||
===Response=== | |||
The design basis radiological consequences analyses performed for AST do not require the installation of any new SSCs. The one modification being implemented to support full implementation of AST will not add any components to the EQ list (the damper modification being implemented will only change the damper power supply; it will not add new components). As such, no components are added to the Environmental Qualification equipment list to comply with 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants. | |||
Westinghouse Non-Proprietary Class 3 Page 9 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 2 RAIs ARCB-RAI-1 Please justify all changes from the current licensing basis (see Issue 1 of NRC Regulatory Issue Summary 2006-04, Experience with Implementation of Alternative Source Terms, dated March 7, 2006 (ADAMS Accession No. ML053460347), for more detail). No justification is needed for changes that are consistent with Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ADAMS Accession No. ML003716792), or are provided in the submittal dated August 13, 2013 (ADAMS Accession No. ML13247A076), unless requested by these RAIs. | |||
===Response=== | |||
NRC Regulatory Issue Summary 2006-004, Experience with Implementation of Alternative Source Terms, indicates that the AST amendment request should (1) provide justification for each individual proposed change to the technical specifications (TS), (2) identify and justify each change to the licensing basis accident analyses, and (3) contain enough details (e.g., assumptions, computer analyses input and output) to allow the NRC staff to confirm the dose analyses results in independent calculations. Section 1 of Enclosure VI of Reference 1 identified the changes to the WCGS licensing basis as: | |||
The Control Room Habitability Envelope (CRHE) unfiltered inleakage is revised from 20 scfm to 50 scfm. | |||
The Control Building unfiltered inleakage is revised from 300 scfm to 400 scfm. | |||
Revise the USAR Chapter 15 dose analyses in accordance with the guidance in Regulatory Guide 1.183. | |||
Revise the Technical Specification (TS) to address the update of the accident source term and associated design basis accidents utilizing the guidance provided in Regulatory Guide 1.183 and the associated control room dose limits of General Design Criterion (GDC) 19, and offsite dose limits of 10 CFR 50.67. | |||
In addition, a detailed comparison of the analyses parameters for the AST versus the current licensing basis is included in Enclosure IV of this LAR, Section 4.3, Tables 4.3-5 through 4.3-16. | |||
These tables have been included below to facilitate the review. For the resubmittal of the LAR, the justification for the proposed changes to the TSs and the control room and control building unfiltered inleakages is justified in Section 2 of Enclosure IV. | |||
Westinghouse Non-Proprietary Class 3 Page 10 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Doses Analyses Parameters Summary Tables (excerpt from Enclosure IV) | |||
Table 4.3-5 Control Room and Control Building Parameters AST CLB Reason for Change Control room volume (ft3) 100,000 100,000 No change Control building volume (ft3) 239,000 239,000 No change Normal ventilation flow rates (cfm) | |||
Unfiltered makeup flow rate from 13,050 13,050 No change environment to control building Unfiltered makeup flow rate from 1950 1950 No change environment to control room Unfiltered inleakage to control room 50 10 Increased to allow for additional surveillance testing margin Emergency mode of operation flow rates prior to operator action (cfm) | |||
Filtered makeup flow rate from 1350 1350 No change environment to control building Filtered makeup flow rate from control 550 550 No change building to control room Unfiltered makeup flow rate from 400 300 Increased to allow for environment to control building additional surveillance testing margin Unfiltered makeup flow rate from control 550 550 No change building to control room Unfiltered inleakage to control room 50 20 Increased to allow for additional surveillance testing margin Filtered control room recirculation flow 1250 1250 No change Emergency mode of operation flow rates following operator action (cfm) | |||
Filtered makeup flow rate from 675 675 No change environment to control building Filtered makeup flow rate from control 550 550 No change building to control room Unfiltered makeup flow rate from 400 300 Increased to allow for environment to control building additional surveillance testing margin | |||
Westinghouse Non-Proprietary Class 3 Page 11 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-5 Control Room and Control Building Parameters (cont.) | |||
AST CLB Reason for Change Unfiltered makeup flow rate from control 0 0 No change building to control room Unfiltered inleakage to control room 50 20 Increased to allow for additional surveillance testing margin Filtered control room recirculation flow 1250 1250 No change Operator action time to terminate failed train 90 90 No change of filtered makeup flow from start of event (minutes) | |||
Filter efficiencies (%) | |||
Elemental iodine 95 95 No change Organic iodine 95 95 No change Particulates 95 95 No change Isolation setpoint for control room air supply 2.12E-03 N/A Change due to modeling of CR for all radiation monitors (GKRE0004 and accidents GKRE0005) (µCi/cc Xe-133) (not AST specific) | |||
Delay to switch to emergency mode of 60 N/A Change due to operation following receipt of isolation signal modeling of CR for all (seconds) accidents (not AST specific) | |||
Control room breathing rate for duration of the 3.5E-04 3.47E-04 Regulatory Guide event (m3/sec) update (not AST specific) | |||
Control room occupancy factors 0 - 24 hours 1.0 1.0 No change 1 - 4 days 0.6 0.6 No change 4 - 30 days 0.4 0.4 No change | |||
Westinghouse Non-Proprietary Class 3 Page 12 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis AST CLB Reason for Change RCS activity See Table 4.3-1a See Table 4.3-1b Updated calculations Initial secondary system activity See Table 4.3-1a See Table 4.3-1b Updated calculations Pre-accident iodine spike factor 60 60 No change Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 75 Modeling update (not AST specific) | |||
Letdown flow decontamination 100 N/A Modeling update | |||
(%) (not AST specific) | |||
RCS leakage (gpm) 11 1 Modeling update (not AST specific) | |||
Spike factor 500 500 No change Duration of accident-initiated 8 8 No change iodine spike (hr) | |||
RCS mass, maximum (lbm) 8.42E+05 4.94E+05 Modeling update (not AST specific) | |||
Equilibrium appearance rates (Ci/min) | |||
I-130 9.87E-03 N/A Modeling update (not AST specific) | |||
I-131 4.39E-01 N/A Modeling update (not AST specific) | |||
I-132 1.98E+00 N/A Modeling update (not AST specific) | |||
I-133 8.93E-01 N/A Modeling update (not AST specific) | |||
I-134 9.65E-01 N/A Modeling update (not AST specific) | |||
I-135 8.17E-01 N/A Modeling update (not AST specific) | |||
Westinghouse Non-Proprietary Class 3 Page 13 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis (cont.) | |||
AST CLB Reason for Change Iodine chemical form of releases (%) | |||
Elemental 97 N/A Regulatory Guide Update (not AST specific) | |||
Organic 3 N/A Regulatory Guide Update (not AST specific) | |||
Particulate 0 N/A Regulatory Guide Update (not AST specific) | |||
Approximate timing of events Safety injection (SI) signal 30 N/A Modeling update (sec) (not AST specific) | |||
Control room isolation 90 N/A Modeling update (including delay) (sec) (not AST specific) | |||
Faulted SG releases all initial 2 N/A Modeling update activity (min) (not AST specific) | |||
RHR cooling takes over 12 N/A Modeling update (releases from intact SGs (not AST specific) terminated) (hr) | |||
RCS cooled below 212°F 34 N/A Modeling update (releases from faulted SG (not AST specific) terminated) (hr) | |||
Mass transfer data Initial faulted SG release 165,000 164,500 Updated calculations (in first 2 minutes) (lbm) | |||
Total primary-to-secondary leakage Leakage through faulted 1 1 No change SG to atmosphere (gpm) | |||
Leakage into intact SGs 450 N/A Modeling update (gpd, total) (not AST specific) | |||
Steam Released from Intact SGs to Atmosphere 0 to 2 hours (lbm) 419,340 404,452 Updated calculations 2 to 12 hours (lbm) 1,310,269 945,973 Updated calculations | |||
Westinghouse Non-Proprietary Class 3 Page 14 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis (cont.) | |||
AST CLB Reason for Change RCS Mass, Minimum (lbm) 3.99E+05 4.94E+05 Input change (minimum CLB does not vs nominal) use maximum and minimum Faulted SG Mass, Maximum (lbm) 1.65E+05 164,500 Updated calculations Intact SGs Mass, Minimum 2.47E+05 286,500 Updated calculations (lbm, Total) | |||
SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Control room atmospheric dispersion factors (sec/m3) | |||
Intact SGs 0 - 0.025 hours 2.55E-02 N/A Not modeled in current AORs 0.025 - 2 hours 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours 1.39E-04 N/A Not modeled in current AORs Faulted SG 0 - 0.025 hours 2.11E-03 N/A Not modeled in current AORs 0.025 - 2 hours 6.12E-04 N/A Not modeled in current AORs 2 - 8 hours 4.38E-04 N/A Not modeled in current AORs 8 - 24 hours 1.79E-04 N/A Not modeled in current AORs 24 - 96 hours 1.14E-04 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 15 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis (cont.) | |||
AST CLB Reason for Change 96 - 720 hours 8.94E-05 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) | |||
Intact SGs 0 - 2 hours 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours 4.89E-05 N/A Not modeled in current AORs Faulted SGs 0 - 2 hours 2.80E-04 N/A Not modeled in current AORs 2 - 8 hours 1.80E-04 N/A Not modeled in current AORs 8 - 24 hours 6.44E-05 N/A Not modeled in current AORs 24 - 96 hours 4.42E-05 N/A Not modeled in current AORs 96 - 720 hours 3.22E-05 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 16 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-7 Assumptions Used for Loss of Non-Emergency AC Power Analysis Reason for AST CLB Change RCS activity See Table 4.3-1a See Table 4.3-1b Updated calculations Initial secondary system activity See Table 4.3-1a See Table 4.3-1b Updated calculations Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 75 Modeling update (not AST specific) | |||
Letdown flow decontamination (%) 100 N/A Modeling update (not AST specific) | |||
RCS leakage (gpm) 11 1 Modeling update (not AST specific) | |||
Spike factor 500 N/A Modeling update (not AST specific) | |||
Duration of accident-initiated iodine 8 N/A Modeling update spike (hr) (not AST specific) | |||
RCS mass, maximum (lbm) 8.42E+05 4.94E+05 Input change CLB does not use (maximum vs max and min nominal) | |||
Equilibrium appearance rates (Ci/min) | |||
I-130 9.87E-03 N/A Modeling update (not AST specific) | |||
I-131 4.39E-01 N/A Modeling update (not AST specific) | |||
I-132 1.98E+00 N/A Modeling update (not AST specific) | |||
I-133 8.93E-01 N/A Modeling update (not AST specific) | |||
I-134 9.65E-01 N/A Modeling update (not AST specific) | |||
I-135 8.17E-01 N/A Modeling update (not AST specific) | |||
Westinghouse Non-Proprietary Class 3 Page 17 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-7 Assumptions Used for Loss of Non-Emergency AC Power Analysis (cont.) | |||
Reason for AST CLB Change Iodine chemical form of releases (%) | |||
Elemental 97 N/A Regulatory Guide Update (not AST specific) | |||
Organic 3 N/A Regulatory Guide Update (not AST specific) | |||
Particulate 0 N/A Regulatory Guide Update (not AST specific) | |||
Time RHR cooling matched decay heat 12 8 Updated (SG releases terminated) (hr) calculations Mass transfer data Total primary-to-secondary leakage 1 1 No change (gpm) | |||
Steam released from SGs to atmosphere 0 to 2 hours (lbm) 419,846 549,000 Updated calculations 2 to 12 hours (lbm) 1,352,918 1,030,000 Updated calculations RCS mass, minimum (lbm) 3.99E+05 4.94E+05 Input change CLB does not (minimum vs use max and min nominal) | |||
Plant total SG mass, minimum (lbm) | |||
Until 2 hours (lbm) 3.30E+05 382,000 Updated calculations After 2 hours (lbm) 4.85E+05 382,000 Modeling update (not AST specific) | |||
SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Control room isolation None N/A No change | |||
Westinghouse Non-Proprietary Class 3 Page 18 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-7 Assumptions Used for Loss of Non-Emergency AC Power Analysis (cont.) | |||
Reason for AST CLB Change Control room atmospheric dispersion factors (sec/m3) 0 - 2 hours 2.55E-2 N/A Not modeled in current AORs 2 - 8 hours 1.84E-2 N/A Not modeled in current AORs 8 - 24 hours 7.46E-3 N/A Not modeled in current AORs 24 - 96 hours 4.72E-3 N/A Not modeled in current AORs 96 - 720 hours 3.43E-3 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours 4.89E-05 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 19 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-8 Assumptions Used for Locked Rotor Analysis AST CLB Reason for Change Core activity See Table 4.3-1a See Table 4.3-1b Updated calculations Failed fuel (% of Core) 5 5 No change Melted fuel (% of Core) 0 0 No change Peaking factor 1.65 1.65 No change Gap fractions I-131 0.08 0.12 Modeling change for AST Kr-85 0.10 0.30 Modeling change for AST Other iodines and noble gases 0.05 0.10 Modeling change for AST Iodine chemical form of releases (%) | |||
Elemental 97 N/A Regulatory Guide Update (not AST specific) | |||
Organic 3 N/A Regulatory Guide Update (not AST specific) | |||
Particulate 0 N/A Regulatory Guide Update (not AST specific) | |||
Time RHR cooling matched decay 12 8 Updated calculations heat (SG releases terminated) (hr) | |||
Mass transfer data Total primary-to-secondary 1 1 No change leakage (gpm) | |||
Steam released from SGs to atmosphere 0 to 2 hours (lbm) 419,846 5.49E+05 Updated calculations 2 to 12 hours (lbm) 1,352,918 1.03E+06 Updated calculations RCS mass, minimum (lbm) 3.99E+05 4.94E+05 Input change (minimum vs nominal) | |||
Plant total SG mass, minimum (lbm) | |||
Until 2 hours (lbm) 3.30E+05 3.82E+05 Updated calculations After 2 hours (lbm) 4.85E+05 3.82E+05 Updated calculations | |||
Westinghouse Non-Proprietary Class 3 Page 20 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-8 Assumptions Used for Locked Rotor Analysis (cont.) | |||
AST CLB Reason for Change SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Time of Control room isolation 120 N/A No change (including delays) (sec) | |||
Control room atmospheric dispersion factors (sec/m3) | |||
HVAC Flows Except for Control Room Unfiltered Inleakage 0 - 0.0333 hours 2.55E-02 N/A Not modeled in current AORs 0.0333 - 2 hours 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours 1.39E-04 N/A Not modeled in current AORs Control Room Unfiltered Inleakage 0 - 2 hours 2.55E-02 N/A Not modeled in current AORs 2 - 8 hours 1.84E-02 N/A Not modeled in current AORs 8 - 24 hours 7.46E-03 N/A Not modeled in current AORs 24 - 96 hours 4.72E-03 N/A Not modeled in current AORs 96 - 720 hours 3.43E-03 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 21 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-8 Assumptions Used for Locked Rotor Analysis (cont.) | |||
AST CLB Reason for Change TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours 4.89E-05 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 22 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis Reason for AST CLB Change Core activity See Table 4.3-1a See Table 4.3-1b Updated calculations Failed fuel (% of core) 10 10 No change Melted fuel (% of core) 0.25 0.25 No change Peaking factor 1.65 1.65 No change Gap fractions Iodines and noble gases 0.10 0.10 No change Alkali metals 0.12 N/A Modeling change for AST Containment Leakage Activity released to containment from failed fuel (%) | |||
Iodines and noble gases 10 10 No change Alkali metals 12 N/A Modeling change for AST Activity released to containment from melted fuel (%) | |||
Iodines and alkali metals 50 50 No change Noble gas 100 100 No change Iodine chemical form of releases (%) | |||
Elemental 4.85 91 Modeling change for AST Particulate 95 5 Modeling change for AST Organic 0.15 4 Modeling change for AST Containment leak rates (weight %/day) 0 - 24 hours 0.2 0.2 No change 1 - 30 days 0.1 0.1 No change 3 | |||
Containment volume (ft ) 2.5E+06 2.5E+06 No change Removal of airborne activity in None None No change containment (other than leakage or decay) | |||
Westinghouse Non-Proprietary Class 3 Page 23 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.) | |||
Reason for AST CLB Change SI signal (sec) 150 N/A Modeling update (not AST specific) | |||
Time of control room isolation (including 210 N/A Modeling update delays) (sec) (not AST specific) | |||
Control room atmospheric dispersion factors (sec/m3) 0 - 0.0583 hours 2.11E-03 N/A Not modeled in current AORs 0.0583 - 2 hours 6.12E-04 N/A Not modeled in current AORs 2 - 8 hours 4.38E-04 N/A Not modeled in current AORs 8 - 24 hours 1.81E-04 N/A Not modeled in current AORs 24 - 96 hours 1.29E-04 N/A Not modeled in current AORs 96 - 720 hours 9.65E-05 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 3.91E-04 N/A Not modeled in current AORs 2 - 8 hours 2.66E-04 N/A Not modeled in current AORs 8 - 24 hours 9.62E-05 N/A Not modeled in current AORs 24 - 96 hours 7.05E-05 N/A Not modeled in current AORs 96 - 720 hours 5.52E-05 N/A Not modeled in current AORs Primary-to-Secondary Leakage Activity released to RCS from failed fuel | |||
(%) | |||
Iodines and noble gases 10 10 No change Alkali metals 12 N/A Modeling change for AST | |||
Westinghouse Non-Proprietary Class 3 Page 24 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.) | |||
Reason for AST CLB Change Activity released to RCS from melted fuel (%) | |||
Iodines and alkali metals 50 50 No change Noble gas 100 100 No change Iodine chemical form of releases (%) | |||
Elemental 97 N/A Regulatory Guide Update (not AST specific) | |||
Organic 3 N/A Regulatory Guide Update (not AST specific) | |||
Particulate 0 N/A Regulatory Guide Update (not AST specific) | |||
Time RHR cooling matched decay heat 12 8 Updated (SG releases terminated) (hr) calculations Mass transfer data Total primary-to-secondary leakage 1 1 No change (gpm) | |||
Steam released from SGs to atmosphere 0 to 2 hours (lbm) 419,846 48,600 (140 sec) Updated calculations 2 to 12 hours (lbm) 1,352,918 N/A Updated calculations RCS mass, minimum (lbm) 3.99E+05 4.94E+05 Input change (minimum vs nominal) | |||
Plant total SG mass, minimum (lbm) | |||
Until 2 hours (lbm) 3.30E+05 4.16E+05 Updated calculations After 2 hours (lbm) 4.85E+05 4.16E+05 Updated calculations | |||
Westinghouse Non-Proprietary Class 3 Page 25 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.) | |||
Reason for AST CLB Change SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Time of Control room isolation (including 120 N/A No change delays) (sec) | |||
Control room atmospheric dispersion factors (sec/m3) | |||
HVAC Flows Except for Control Room Unfiltered Inleakage 0 - 0.0333 hours 2.55E-02 N/A Not modeled in current AORs 0.0333 - 2 hours 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours 1.39E-04 N/A Not modeled in current AORs Control Room Unfiltered Inleakage 0 - 2 hours 2.55E-02 N/A Not modeled in current AORs 2 - 8 hours 1.84E-02 N/A Not modeled in current AORs 8 - 24 hours 7.46E-03 N/A Not modeled in current AORs 24 - 96 hours 4.72E-03 N/A Not modeled in current AORs 96 - 720 hours 3.43E-03 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 4.83E-04 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 26 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.) | |||
Reason for AST CLB Change 2 - 8 hours 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours 4.89E-05 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 27 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-10 Assumptions Used for Letdown Line Break Analysis AST CLB Reason for Change RCS activity See Table 4.3-1a See Table 4.3-1b Updated calculations Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 195 Modeling update (not AST specific) | |||
Letdown flow decontamination 100 N/A Modeling update | |||
(%) (not AST specific) | |||
RCS leakage (gpm) 11 N/A Modeling update (not AST specific) | |||
Spike factor 500 N/A Modeling update (not AST specific) | |||
Duration of accident-initiated 8 N/A Modeling update iodine spike (hr) (not AST specific) | |||
Reactor coolant mass, maximum 8.42E+05 4.94E+05 Input change (lbm) (maximum vs nominal) | |||
Equilibrium appearance rates (Ci/min) | |||
I-130 9.87E-03 N/A Modeling update (not AST specific) | |||
I-131 4.39E-01 N/A Modeling update (not AST specific) | |||
I-132 1.98E+00 N/A Modeling update (not AST specific) | |||
I-133 8.93E-01 N/A Modeling update (not AST specific) | |||
I-134 9.65E-01 N/A Modeling update (not AST specific) | |||
I-135 8.17E-01 N/A Modeling update (not AST specific) | |||
Westinghouse Non-Proprietary Class 3 Page 28 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-10 Assumptions Used for Letdown Line Break Analysis (cont.) | |||
AST CLB Reason for Change Iodine chemical form of releases (%) | |||
Elemental 97 N/A Regulatory Guide Update (not AST specific) | |||
Organic 3 N/A Regulatory Guide Update (not AST specific) | |||
Particulate 0 N/A Regulatory Guide Update (not AST specific) | |||
Reactor coolant mass, minimum (lbm) 3.99E+05 N/A Modeling update (not AST specific) | |||
Flow rate out of broken line (gpm) 141 141 No change Iodine and alkali metal airborne 0.18 N/A Modeling update fraction (not AST specific) | |||
Maximum RCS letdown pressure 600 2200 Modeling update (psig) (not AST specific) | |||
Maximum RCS letdown 380 286 Modeling update temperature (°F) (not AST specific) | |||
Time to isolate break flow 30.167 30.167 No change (terminating releases) (min) | |||
Control room isolation None N/A No change Control room atmospheric dispersion N/A factor (sec/m3) 0 - 2 hours 2.11E-03 N/A Not modeled in current AORs TSC atmospheric dispersion factor (sec/m3) 0 - 2 hours 2.80E-04 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 29 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis Reason for AST CLB Change RCS activity See Table 4.3-1a See Table 4.3-1c Updated calculations Initial secondary system activity See Table 4.3-1a See Table 4.3-1c Updated calculations Pre-accident iodine spike factor 60 60 No change Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 120 Modeling update (not AST specific) | |||
Letdown flow decontamination 100 N/A Modeling update | |||
(%) (not AST specific) | |||
RCS leakage (gpm) 11 1 Modeling update (not AST specific) | |||
Spike factor 335 335 No change Reactor coolant mass, maximum 8.42E+05 5.05E+5 Input change (lbm) (maximum vs nominal) | |||
Duration of accident-initiated 8 8 No change iodine spike (hr) | |||
Equilibrium appearance rates (Ci/min) | |||
I-130 9.87E-03 N/A Modeling update (not AST specific) | |||
I-131 4.39E-01 3.54E-1 Modeling update (not AST specific) | |||
I-132 1.98E+00 1.36E+00 Modeling update (not AST specific) | |||
I-133 8.93E-01 7.72E+00 Modeling update (not AST specific) | |||
I-134 9.65E-01 7.02E-01 Modeling update (not AST specific) | |||
I-135 8.17E-01 6.63E-01 Modeling update (not AST specific) | |||
Westinghouse Non-Proprietary Class 3 Page 30 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.) | |||
Reason for AST CLB Change Iodine chemical form of releases (%) | |||
Elemental 97 N/A Regulatory Guide Update (not AST specific) | |||
Organic 3 N/A Regulatory Guide Update (not AST specific) | |||
Particulate 0 N/A Regulatory Guide Update (not AST specific) | |||
Approximate timing of events (sec) See Table 2.7.3-2 N/A Modeling update in Enclosure I of (not AST specific) this LAR (For dose input, exclude the 100 seconds of steady-state operation.) | |||
Time of control room isolation 120 N/A Modeling update (including delay) (sec) (not AST specific) | |||
Time to complete control room 600 N/A Modeling update isolation with SI signal from event (not AST specific) initiation (sec) | |||
Westinghouse Non-Proprietary Class 3 Page 31 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.) | |||
Reason for AST CLB Change Transient mass transfer data Non-flashed break flow (lbm) 0 - 52 seconds 2227.5 N/A Updated calculations 52 - 1102 seconds 43,129.9 N/A Updated calculations 1102 - 2902 seconds 88,387.2 N/A Updated calculations 2902 - 3502 seconds 32,991.2 N/A Updated calculations 3502 - 3846 seconds 18,224.8 N/A Updated calculations 3846 - 5155 seconds 61,523.0 N/A Updated calculations 5155 - 7527 seconds 41,166.4 N/A Updated calculations Flashed break flow (lbm) 0 - 52 seconds 438.9 N/A Updated calculations 52 - 1102 seconds 2,901.8 N/A Updated calculations 1102 - 2902 seconds 13,432.1 N/A Update calculations 2902 - 3502 seconds 2,635.6 N/A Update calculations 3502 - 3846 seconds 606.1 N/A Update calculations | |||
Westinghouse Non-Proprietary Class 3 Page 32 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.) | |||
Reason for AST CLB Change Steam released from ruptured SG (lbm) 0 - 52 seconds 188,100 N/A Update calculations 52 - 1102 seconds 27,469.2 N/A Update calculations 1102 - 2902 seconds 149,850.8 N/A Update calculations 2902 - 7527 seconds 0 N/A Update calculations 7527 - 43,200 seconds 2530 N/A Update calculations Steam released from intact SGs (lbm) 0 - 52 seconds 562,650 N/A Update calculations 52 - 1102 seconds 69,877.5 N/A Update calculations 1102 - 3502 seconds 0 N/A Update calculations 3502 - 3846 seconds 94,307.4 N/A Update calculations 3846 - 5155 seconds 130,799.9 N/A Update calculations 5155 - 7527 seconds 98,156.3 N/A Update calculations 7527 - 43,200 seconds 1,645,930 N/A Update calculations Reactor coolant mass, minimum (lbm) 3.99E+05 5.05E+05 CLB Input change does not use max (minimum vs and min nominal) | |||
Ruptured SG mass, minimum (lbm) 7.00E+04 N/A Updated calculations Intact SGs mass, minimum (lbm, 1.95E+05 N/A Updated total) calculations Condenser iodine and alkali metal 100 N/A Modeling update removal factor (not AST specific) | |||
SG iodine water/steam partition 100 N/A Regulatory Guide coefficient Update (not AST specific) | |||
Moisture carryover (%) 0.25 0.25 No change | |||
Westinghouse Non-Proprietary Class 3 Page 33 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.) | |||
Reason for AST CLB Change Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours 2.55E-02 N/A Not modeled in current AORs 0.0333 - 2 hours(1) 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours 1.39E-04 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours 4.89E-05 N/A Not modeled in current AORs Note: (1) The control room unfiltered inleakage continues to be associated with the normal mode /Q of 2.55E-02 sec/m3 until completion of control room isolation from a safety injection signal at 10 minutes. | |||
Westinghouse Non-Proprietary Class 3 Page 34 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis Reason for AST CLB Change Core activity See Table 4.3-1a See Table 4.3-1b Updated (containment leakage, calculations ECCS leakage, and RWST back-leakage) | |||
RCS activity See Table 4.3-1a See Table 4.3-1b Updated (containment purge) calculations Fuel release fractions 100% of the core activity is Modeling change and timing released immediately for AST following event initiation Nuclide Group Gap Early Gap Early Release In-Vessel Release In-Vessel Phase Phase Phase Phase Fraction Fraction Fraction Fraction Noble gases 0.05 0.95 N/A N/A Modeling change for AST Iodines 0.05 0.35 N/A N/A Modeling change for AST Alkali metals 0.05 0.25 N/A N/A Modeling change for AST Tellurium metals 0.00 0.05 N/A N/A Modeling change for AST Barium and 0.00 0.02 N/A N/A Modeling change strontium for AST Noble metals 0.00 0.0025 N/A N/A Modeling change for AST Cerium 0.00 0.0005 N/A N/A Modeling change for AST Lanthanides 0.00 0.0002 N/A N/A Modeling change for AST | |||
Westinghouse Non-Proprietary Class 3 Page 35 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.) | |||
Reason for AST CLB Change Duration of phases Phase Onset Duration Onset Duration Gap release 30 sec 0.49167 hr N/A N/A Modeling change for AST Early in-vessel 0.5 hr 1.3 hr N/A N/A Modeling change for AST SI signal (sec) 0 N/A Modeling update (not AST specific) | |||
Time of control room 120 0 Modeling update isolation (including (not AST delays) (sec) specific) | |||
Containment Leakage Iodine chemical form of releases (%) | |||
Elemental 4.85 91 Modeling change for AST Organic 0.15 4 Modeling change for AST Particulate 95 5 Modeling change for AST Containment volume, 2.7E+06 2.5E+06 Since a maximum (ft3) maximum volume is conservative, the larger value listed in USAR Section 6.2.1.5.3 was modeled | |||
% Sprayed 85 85 No change | |||
% Unsprayed 15 15 No change | |||
Westinghouse Non-Proprietary Class 3 Page 36 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.) | |||
AST CLB Reason for Change Mixing between sprayed and unsprayed 6.94E+04 8.50E+04 The CLB value corresponds containment volumes (cfm) to two hydrogen mixing fans operating in slow speed. | |||
While a mixing rate of 85,000 cfm will still be present following the event, the value has been conservatively reduced to the flow rate from one containment cooler fan Start of fan cooler mixing (min) 2 N/A Modeling update (not AST specific) | |||
Containment leak rates (weight %/day) 0 - 24 hours 0.2 0.2 No change 1 - 30 days 0.1 0.1 No change Spray timing Initiation (min) 2 0 The time to credit containment spray was conservatively increased to account for the delay to reaching full flow conditions Termination (hr) 5 9.55 Conservative time used Spray removal coefficients Organic iodine spray removal 0.0 0.0 No change coefficient (hr -1) | |||
Elemental iodine spray removal coefficient calculations Spray removal coefficient (hr-1), 10 10 No change DF < 200 Gas phase mass transfer 3 N/A Modeling update (not AST coefficient (m/min) specific) | |||
Time of fall of the spray drops 0.146 N/A Updated calculations (min) | |||
Volume flow rate of sprays 658.66 711.13 Conservative low flow (m3/hr) modeled. | |||
Westinghouse Non-Proprietary Class 3 Page 37 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.) | |||
AST CLB Reason for Change Containment sprayed volume 6.5E+04 6.0E+04 The maximum containment (m3) volume increased from 2.5E+06 ft3 to 2.7E+06 ft3. | |||
Since 85% of the containment is sprayed, the sprayed volume increased proportionally with the increase in total volume. | |||
Mass-mean diameter of the 0.00116 N/A Updated calculations spray drops (m) | |||
Particulate spray removal coefficient calculations Spray removal coefficient (hr-1), 5 0.45 Updated calculations DF < 50 (Used DFs up to 100) | |||
Spray removal coefficient (hr-1), 0.5 N/A Updated calculations DF > 50 Drop fall height (m) 35.966 36.017 Conservative rounding Volume flow rate of sprays 658.66 711.13 Conservative low flow (m3/hr) modeled. | |||
Containment sprayed volume 6.5E+04 6.0E+04 WCGS input change The (m3) maximum containment volume increased from 2.5E+06 ft3 to 2.7E+06 ft3. | |||
Since 85% of the containment is sprayed, the sprayed volume increased proportionally with the increase in total volume. | |||
Ratio of dimensionless collection efficiency to average spray drop diameter Prior to DF of 50 (m-1) 10 N/A Modeling update (not AST specific) | |||
After DF of 50 (m-1) 1 N/A Modeling update (not AST specific) | |||
Westinghouse Non-Proprietary Class 3 Page 38 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.) | |||
AST CLB Reason for Change Particulate sedimentation removal 0.1 N/A Modeling update (not AST coefficient (hr-1), DF < 1000 specific) pH of sump 7.0 8.5 The pH value was changed to 7.0 to be consistent with the iodine retention assumptions contained within the analysis. | |||
Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours 2.11E-03 5.30E-04 Updated calculations 0.0333 - 2 hours 6.12E-04 5.30E-04 Updated calculations 2 - 8 hours 4.38E-04 5.30E-04 Updated calculations 8 - 24 hours 1.81E-04 3.6E-04 Updated calculations 24 - 96 hours 1.29E-04 6.60E-05 Updated calculations 96 - 720 hours 9.65E-05 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 3.91E-04 2.2E-04 Updated calculations 2 - 8 hours 2.66E-04 2.2E-04 Updated calculations 8 - 24 hours 9.62E-05 1.17E-04 Updated calculations 24 - 96 hours 7.05E-05 2.04E-05 Updated calculations 96 - 720 hours 5.52E-05 0 Updated calculations ECCS Leakage Iodine chemical form of releases (%) | |||
Elemental 97 N/A Regulatory Guide Update Organic 3 N/A Regulatory Guide Update Particulate 0 N/A Regulatory Guide Update Sump volume (gal) 4.60E+05 4.60E+05 No change Time to initiate ECCS recirculation (min) 0 28.2 Modeling update (not AST specific) | |||
ECCS leakage to auxiliary building (gpm) 2 2 No change Iodine airborne fraction 0.10 0.10 No change | |||
Westinghouse Non-Proprietary Class 3 Page 39 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.) | |||
AST CLB Reason for Change Auxiliary building exhaust filter efficiency 90 90 No change (all forms of iodine) (%) | |||
Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours 2.11E-03 1.10E-04 Updated calculations 0.0333 - 2 hours 6.12E-04 1.10E-04 Updated calculations 2 - 8 hours 4.38E-04 1.10E-04 Updated calculations 8 - 24 hours 1.79E-04 6.80E-05 Updated calculations 24 - 96 hours 1.14E-04 1.70E-05 Updated calculations 96 - 720 hours 8.94E-05 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 2.80E-04 2.2E-04 Updated calculations 2 - 8 hours 1.80E-04 2.2E-04 Updated calculations 8 - 24 hours 6.44E-05 1.17E-04 Updated calculations 24 - 96 hours 4.42E-05 2.04E-05 Updated calculations 96 - 720 hours 3.22E-05 0 Updated calculations RWST Back-Leakage RWST initial gas volume, minimum (gal) 3.54E+05 N/A Modeling update Time to initiate ECCS recirculation (min) 0 28.2 Modeling update ECCS leakage to RWST (gpm) 3.8 3.8 No change Iodine airborne fraction 0.10 0.10 No change Release from RWST gas space (gpm) 3.8 3.8 No change Iodine chemical form of releases (%) | |||
Elemental 97 91 Regulatory Guide Update Organic 3 4 Regulatory Guide Update Particulate 0 5 Regulatory Guide Update | |||
Westinghouse Non-Proprietary Class 3 Page 40 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.) | |||
AST CLB Reason for Change Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours 1.03E-03 1.10E-04 Updated calculations 0.0333 - 2 hours 6.80E-04 1.10E-04 Updated calculations 2 - 8 hours 6.19E-04 1.10E-04 Updated calculations 8 - 24 hours 2.27E-04 6.80E-05 Updated calculations 24 - 96 hours 1.96E-04 1.70E-05 Updated calculations 96 - 720 hours 1.53E-04 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 1.87E-04 2.2E-04 Updated calculations 2 - 8 hours 1.25E-04 2.2E-04 Updated calculations 8 - 24 hours 4.51E-05 1.17E-04 Updated calculations 24 - 96 hours 3.33E-05 2.04E-05 Updated calculations 96 - 720 hours 2.61E-05 0 Updated calculations Containment Purge RCS activity released (%) 100 100 No change Iodine chemical form of releases (%) | |||
Elemental 97 91 Regulatory Guide Update Organic 3 4 Regulatory Guide Update Particulate 0 5 Regulatory Guide Update RCS mass, maximum (lbm) 8.42E+05 4.94E+05 Input change (maximum vs CLB does nominal) not use maximum and minimum Containment volume, minimum (ft3) 2.5E+06 2.5E+06 No change Maximum purge flow rate, unfiltered (cfm) 4,680 4,680 No change Duration of purge release (sec) 10 8 Updated input | |||
Westinghouse Non-Proprietary Class 3 Page 41 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.) | |||
AST CLB Reason for Change Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours 2.11E-03 1.10E-04 Updated calculations 0.0333 - 2 hours 6.12E-04 1.10E-04 Updated calculations 2 - 8 hours 4.38E-04 1.10E-04 Updated calculations 8 - 24 hours 1.79E-04 6.80E-05 Updated calculations 24 - 96 hours 1.14E-04 6.60E-05 Updated calculations 96 - 720 hours 8.94E-05 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours 2.80E-04 2.2E-04 Updated calculations 2 - 8 hours 1.80E-04 2.2E-04 Updated calculations 8 - 24 hours 6.44E-05 1.17E-04 Updated calculations 24 - 96 hours 4.42E-05 2.04E-05 Updated calculations 96 - 720 hours 3.22E-05 0 Updated calculations | |||
Westinghouse Non-Proprietary Class 3 Page 42 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-13 Assumptions Used for Waste Gas Decay Tank Failure Analysis AST CLB Reason for Change Activity in ruptured tank See See Updated calculations Table 4.3-2a Table 4.3-2b Iodine chemical form of releases (%) | |||
Elemental 100 N/A Modeling update (not AST specific) | |||
Duration of release (hr) 2 2 No change Control room isolation None N/A No change Control room atmospheric dispersion factor (sec/m3) 0 - 2 hours 7.17E-04 5.30E-04 Updated calculations TSC atmospheric dispersion factor (sec/m3) 0 - 2 hours 1.97E-04 N/A Not modeled in current AORs Table 4.3-14 Assumptions Used for Liquid Waste Tank Failure Analysis AST CLB Reason for Change Activity in ruptured tank Recycle holdup tank See See Updated calculations Table 4.3-2a Table 4.3-2b Hypothetical tank maximizing See Table N/A Updated calculations iodine 4.3-2a Iodine chemical form of releases (%) | |||
Elemental 100 N/A Modeling update (not AST specific) | |||
Duration of release (hr) 2 2 No change Control room isolation None N/A No change Control room atmospheric dispersion factor (sec/m3) 0 - 2 hours 7.17E-04 N/A Not modeled in current AORs TSC atmospheric dispersion factor (sec/m3) 0 - 2 hours 1.97E-04 N/A Not modeled in current AORs | |||
Westinghouse Non-Proprietary Class 3 Page 43 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-15 Assumptions Used for Fuel Handling Accident Analysis AST CLB Reason for Change Core activity for one assembly at minimum time prior to fuel movement (Ci) | |||
Kr-85m 1.10E+00 See Table 4.3-4c Updated calculations Kr-85 5.69E+03 See Table 4.3-4c Updated calculations Xe-131m 5.38E+03 See Table 4.3-4c Updated calculations Xe-133m 1.76E+04 See Table 4.3-4c Updated calculations Xe-133 8.19E+05 See Table 4.3-4c Updated calculations Xe-135m 5.58E+01 See Table 4.3-4c Updated calculations Xe-135 8.14E+03 See Table 4.3-4c Updated calculations I-130 1.45E+02 See Table 4.3-4c Updated calculations I-131 4.12E+05 See Table 4.3-4c Updated calculations I-132 3.95E+05 See Table 4.3-4c Updated calculations I-133 8.90E+04 See Table 4.3-4c Updated calculations I-135 3.42E+02 See Table 4.3-4c Updated calculations Number of fuel assemblies damaged 1.2 1.2 No change Peaking factor 1.65 1.65 No change Time of Control room isolation 120 N/A No change (including delays) (sec) | |||
Decay time prior to fuel movement, 76 76 No change minimum (hr) | |||
Gap fractions I-131 0.12 0.12 No change Kr-85 0.30 0.30 No change Other iodines and noble gases 0.10 0.10 No change Iodine chemical form in gap (%) | |||
Elemental 99.85 N/A Modeling change for AST Organic 0.15 N/A Modeling change for AST Fuel pool water depth, minimum (ft) 23 23 No change Fuel rod internal pressure, maximum 1500 1200 Modeling update (not (psig) AST specific) | |||
Westinghouse Non-Proprietary Class 3 Page 44 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-15 Assumptions Used for Fuel Handling Accident Analysis (cont.) | |||
AST CLB Reason for Change Overall pool iodine DF 200 100 Regulatory Guide Update Iodine airborne fractions (%) | |||
Elemental 70 N/A Modeling change for AST Organic 30 N/A Modeling change for AST Duration of release (hr) 2 2 No change Removal of airborne activity in No change containment/fuel building (other than None None decay) | |||
Control room atmospheric dispersion factor (sec/m3) | |||
Auxiliary Building Release HVAC Flows Except for Control Room Unfiltered Inleakage 0 - 0.0333 hours 2.11E-03 5.30E-04 Updated calculations 0.0333 - 2 hours 6.12E-04 5.30E-04 Updated calculations Control Room Unfiltered Inleakage 0 - 2 hours 2.11E-03 5.30E-04 Updated calculations Containment Release(1) 0 - 2 hours 1.38E-03 N/A Not modeled in current AORs TSC atmospheric dispersion factor (sec/m3)2 0 - 2 hours 3.91E-04 N/A Not modeled in current AORs Note: (1) The containment release /Q is lower that the auxiliary building release /Q; therefore the auxiliary building release /Q is used in the calculation of the doses. The containment release /Q is only used to determine the time the control room isolation setpoint is reached since it is lower than the containment leakage /Q and results in a conservative isolation time. | |||
Westinghouse Non-Proprietary Class 3 Page 45 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-16 Technical Support Center Parameters AST CLB Reason for Change TSC volume (ft3) 44,000 52,800 The CLB value is based on a simple bounding one volume calculation (length x width x height). The more accurate AST value was calculated by determining the volume of all individual TSC rooms and then summing them. | |||
Normal ventilation flow rates (cfm) | |||
Unfiltered makeup flow rate 550 550 No change Unfiltered inleakage 20 20 No change Emergency mode flow rates (cfm) | |||
Filtered makeup flow rate 550 550 No change Unfiltered makeup flow rate 0 N/A No change Unfiltered inleakage 20 20 No change Filtered recirculation flow 450 450 No change Filter efficiencies (%) | |||
Elemental iodine 95 90 Test procedure revised to increase acceptance criterion to 95% | |||
Organic iodine 95 90 Test procedure revised to increase acceptance criterion to 95% | |||
Particulates 95 90 Test procedure revised to increase acceptance criterion to 95% | |||
Delay to switch to emergency mode 60 0 A delay of 60 minutes was of operation after event initiation added to ensure the analysis (minutes) bounds the plant specific time to switch to the emergency mode of operation after event initiation TSC breathing rate for duration of 3.5E-04 3.5E-04 No change the event (m3/sec) | |||
TSC occupancy factors 0 - 24 hours 1.0 1.0 No change 1 - 4 days 0.6 0.6 No change 4 - 30 days 0.4 0.4 No change | |||
Westinghouse Non-Proprietary Class 3 Page 46 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-2: | |||
Enclosure VI, page 4-38 states, in part, that An additional fuel management multiplier is applied to the calculated core activity to account for anticipated variations in fuel cycle design. | |||
Please provide the value of the multiplier, how it was derived, and a justification for its use. | |||
===Response=== | |||
The analyses presented in Section 4.2 of Enclosure VI include a fuel management multiplier on the core inventory to accommodate anticipated changes in the values of fuel enrichment and fuel burnup, since the fuel cycle designs vary from cycle to cycle. | |||
The fuel management multiplier is equal to [ ]a,c. The fuel management multiplier increases the core activity calculated for the cycle 19 fuel cycle design, which approximates an equilibrium fuel cycle, to account for any differences in cycle 20 and future fuel cycles; therefore, it adds margin to the calculated doses. The addition of this margin is not required by Regulatory Guide 1.183 and provides an additional conservatism in the dose calculations. | |||
The multiplier is derived from a number of historical sensitivity analyses on core activity and post-accident dose consequence analyses. Variations in fuel enrichment, core loading (i.e. fuel mass), and cycle burnup were considered in the sensitivities analyses. Based on these sensitivity analyses, a value of [ ] a,c was used for the WCGS. | |||
ARCB-RAI-3 Enclosure VI, page 1-1 proposes an exception to the full scope implementation of the AST methodology. The exception to retain the current licensing basis (using Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, dated March 23, 1962 (not publicly available, proprietary information), for NUREG-0737, Clarification of TMI Action Plan Requirements, July 2000 (ADAMS Accession No. ML051400209), evaluations other than Control Room Habitability Envelope (CRHE) and Technical Support Center (TSC) doses) is based upon NRC Regulatory Guide (RG) 1.183, Section 1.3.5, Equipment Environmental Qualification and Section 6, Assumptions for Evaluating the Radiation Doses for Equipment Qualification. These RG 1.183 sections are only for equipment qualification and are not for NUREG-0737 evaluations. | |||
RG 1.183, Regulatory Position 4.3, under Other Dose Consequences, states, in part, that | |||
Westinghouse Non-Proprietary Class 3 Page 47 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737 .... Design envelope source terms provided in NUREG-0737 should be updated for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE | |||
[total effective dose equivalent]. | |||
b) Please confirm if the current licensing basis calculations are used to determine post-accident vital area access unaffected or bounding (using the TID-14844 source term) compared to those using the AST. | |||
b) Please explain if any new operator actions have been credited to support the AST methodology. | |||
c) Please explain if the current licensing basis calculations are used to determine the doses for operation of the post-accident sampling system or the containment high range radiation monitors used to monitor post-accident primary containment radiation levels (using the TID-14844 source term) unaffected or bounding compared to those using the AST. | |||
d) Please provide a detailed justification for the proposed exemption or follow Regulatory Position 4.3 of RG 1.183. | |||
===Response=== | |||
a) Regulatory Position 1.3.2 of Regulatory Guide 1.183 specifies that an evaluation of AST impact was performed on three representative operating reactors and concluded that the Technical Information Document (TID)-14844 based methodology and assumptions generally bound the AST methodology. Regulatory Position 1.3.2, states, in part: | |||
This evaluation determined that radiological analysis results based on TID-14844 source term assumptionsand the whole body and thyroid methodology generally bound the results from the analyses based on AST and TEDE methodology. | |||
Licensees may use the applicable conclusions of this evaluation in addressing the impact of the AST on design basis radiological analyses. | |||
Full implementation of AST at WCGS does not impact the assumptions or inputs to the current TID based analyses. As such, AST implementation would not impact the conclusions from Regulatory Position 1.3.2, the TID based analyses and results, or the ability to implement AST in other radiological analyses per the guidance in Regulatory Position 1.3.4. | |||
b) No new operator actions have been credited to support the AST methodology. | |||
Regarding the time critical action to isolate the failed CREVS train, the overall action | |||
Westinghouse Non-Proprietary Class 3 Page 48 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment described in USAR Section 15A.3.1 remains unchanged, i.e., operator action is required to ensure no bypass pathways exist for unfiltered air to enter the control room following the failure of a filtration fan. However, the damper modification being implemented will eliminate the possibility of having to locally isolate the failed CREVS train. Specifically, the modification will result in the system automatically isolating the CREVS by normal system operation when the control room AC fan unit is turned off, which can be performed remotely from the control room following the identification of the failed filtration fan. | |||
c) The requirements for the post-accident sampling system were eliminated in Amendment No. 137 (Reference 5) and thus is unaffected by the change to AST methodology. With regards to the containment high range radiation monitors, the controlling isotope for the TID-14844 source term, Xe133, remains the controlling isotope for the AST source term. Thus, the containment high range radiation monitors are also unaffected. | |||
d) WCNOC is only proposing exceptions to the full scope implementation of the AST for equipment qualification and NUREG-0737 evaluations other than Control Room Habitability Envelope (CRHE) and Technical Support Center (TSC) doses. With regards to NUREG-0737 evaluations, full implementation of AST at WCGS does not impact the assumptions or inputs to the current TID based analyses, as documented in the part a) response to this request. Thus, AST implementation would not impact their conclusions from Regulatory Position 1.3.2, the TID based analyses, their results, or the ability to implement AST in other radiological analyses per the guidance in Regulatory Position 1.3.4. Once AST is part of the sites licensing basis this regulatory position specifies that all characteristics of the AST and the TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an individual as-needed basis. WCNOC plans to incorporate AST into these remaining analyses on an as needed basis as described in Regulatory Position 1.3.4. | |||
ARCB-RAI-4: | |||
Enclosure VI, page 5-2 states, in part, that Core design parameters (enrichment, burnup, and MTU loading) are based on cycle 19 core design. | |||
c) Please describe how many batches or core regions were assumed to determine the source term. | |||
d) Please explain what period of irradiation (burnup) and specific power (i.e., megawatts days per metric ton of uranium (MWD/MTU)) was assumed for each region. | |||
c) Please explain if the activity from each batch was taken at the end of life for each cycle. | |||
Westinghouse Non-Proprietary Class 3 Page 49 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment d) Please provide the number of dose significant isotopes form the source term input for the RADTRAD dose evaluations. | |||
e) For the purpose of the design basis, please explain the maximum enrichment assumed and if the assumed period of irradiation allows for the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. | |||
===Response=== | |||
a) The core is modeled in seven fuel regions (batches). | |||
b) The length of cycle 19 was modelled as 511 effective full-power days (EFPD), which bounds the actual cycle length of 505 EFPD. Prior cycles were assumed to operate for 455 EFPD consistent with the WCGS fuel cycle operating experience. The region-specific data is tabulated below for cycle 19. | |||
Wt% MWt/FA MWt/FA Region # of FAs U-235 (pre- cycle 19) (cycle 19) a,c | |||
Westinghouse Non-Proprietary Class 3 Page 50 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment c) The activity reported in Enclosure IV, Table 4.2-1 is taken at the end of cycle 19 for each fuel region. (511 EFPD) d) The LOCA dose consequence analysis modeled the 63 nuclides presented in Enclosure IV, Table 4.3-1a. The main steamline break, loss of non-emergency AC power, locked rotor, rod ejection, letdown line break, and steam generator tube rupture dose consequence analyses consider the 21 noble gas, iodine, and alkali metal nuclides listed in Enclosure IV, Table 4.3-1a. The fuel handling accident dose consequence analysis (both inside containment and in the fuel building) considers the 12 noble gas and iodine nuclides listed in Enclosure IV, Table 4.3-15. The tank rupture dose consequence events (waste gas decay tank failure and liquid waste tank failure) consider the 16 noble gas and iodine nuclides listed in Enclosure IV, Table 4.3-2a. | |||
e) The maximum enrichment considered for a single fuel assembly is [ ]a,c wt% U-235, consistent with the cycle 19 fuel cycle design. | |||
In general, radionuclides with short half-lives are present in quantities that are proportional to core power. Those are conservatively treated by including the emergency core cooling system (ECCS) evaluation uncertainty in core thermal power. Quantities of longer-lived radionuclides will build up during the fuel cycle. Those are conservatively treated by modelling the longest anticipated cycle 19 length (511 EFPD). The end of life values are reported in Enclosure IV, Table 4.2-1. | |||
This is consistent with Regulatory Guide 1.183 Section C.3.1, which states: | |||
The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values.9 9 | |||
Note that for some radionuclides, such as Cs-137, equilibrium will not be reached prior to fuel offload. Thus, the maximum inventory at the end of life should be used. | |||
ARCB-RAI-5 Enclosure VI, page 15B-3 states that the analysis conforms to Regulatory Position 3.2 of RG 1.183. Please confirm that reactor fuel will have a peak burnup of less than 62,000 MWD/MTU of uranium and a maximum linear heat generation rate of 6.3 kilowatts per foot (kW/ft.) or less for peak rod average power for burnups exceeding 54,000 MWD/MTU. If not, please justify how the analysis conforms to Regulatory Position 3.2. | |||
Westinghouse Non-Proprietary Class 3 Page 51 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
===Response=== | |||
The fuel handling accident (FHA) analysis conservatively assumed that 100% of the fuel will not meet the Footnote 11 from Regulatory Position 3.2 of Regulatory Guide 1.183. Footnote 11 contains burnup and linear heat generation rate limits for the gap fractions contained in Table 3 of Regulatory Guide 1.183 as well as an alternative method for determining gap fractions provided they bound the limiting projected plant-specific power history for the specific load. The analysis conforms with the position such that alternative gap fractions were used which were appropriate for the assumption of not meeting the Footnote 11 limits. These gap fractions are obtained from Regulatory Guide 1.25, as modified by NUREG/CR-5009, which provides higher, i.e., more conservative gap fractions than Regulatory Guide 1.183, which are not constrained by the Footnote 11 burnup limits as they can be applied to higher burnup to bound power history. | |||
ARCB-RAI-6: | |||
Enclosure VI, page 4-63 states, in part, that Sedimentation is credited in the portion of containment that is not impacted by spray removal and in the sprayed portion when sprays are not on at a rate of 0.1 hr -1 until a DF [decontamination factor] of 1000 is reached at 23.5 hours. After this time sedimentation removal is terminated. | |||
Removal of aerosol by sprays and natural deposition are competing processes. Please justify crediting both spray removal and the proposed 0.1 hr -1 sedimentation rate. Describe how the natural deposition model accounts for removal due to the spray model used. If any further credit for a reduction in aerosols is taken for any pathway, please provide a justification for that credit while considering the impact of any other removal mechanism credited. Please confirm if the assumed sedimentation rate impacts the final dose. | |||
===Response=== | |||
The sprayed and unsprayed regions of containment were modeled as separate compartments. Activity removal via natural deposition was only modeled in the unsprayed region of containment and in the sprayed region of containment when sprays were not credited for removal of aerosols (e.g., after containment spray termination). Activity removal via the sprays and activity removal via natural deposition were not modeled concurrently in the same compartment of containment. Therefore, there was no double-accounting for the removal of aerosols in either the sprayed or unsprayed regions of containment. The user-defined coefficients in the RADTRAD model for sprays and natural deposition were used. | |||
There is no other credit for reduction of aerosols in any pathway except for removal by filters in the control building/control room ventilation systems. This removal does not | |||
Westinghouse Non-Proprietary Class 3 Page 52 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment compete with any other removal process and is justified due to the availability of the filters in these systems. | |||
The assumed sedimentation rate does impact the final dose results. | |||
ARCB-RAI-7: | |||
Enclosure VI, page 4-63 states, in part, that The resulting removal coefficient for elemental iodine is 22.9* hr - 1 SRP [Standard Review Plan] 6.5.2 allows for elemental iodine removal credit of up to 20 hr-1 during injection spray; however, to avoid sensitivities with spray switchover times from injection to recirculation and to conservatively address iodine loading in the spray fluid during recirculation, the removal is limited to 10 hr - 1 for either spray mode. | |||
Please explain in more detail why the removal coefficient is determined to be limited to 10 hr-1 and what is meant by "sensitivities with spray switchover times from injection to recirculation and to conservatively address iodine loading in the spray fluid during recirculation." | |||
===Response=== | |||
In accordance with Standard Review Plan (SRP) Section 6.5.2, the spray removal coefficient, if calculated to be greater than 20 hr-1, is to be limited to 20 hr-1. For the analysis performed for the WCGS, the removal coefficient is calculated to be greater than 20 hr-1; however, the removal coefficient is reduced further to 10 hr-1 to conservatively address iodine loading in the spray fluid during recirculation. During recirculation, the spray solution will gradually become loaded with elemental iodine, which will limit the capacity of the spray to remove airborne iodine. The spray removal coefficient would be inversely proportional to the decontamination factor achieved for elemental iodine. Thus, when recirculation spray is first credited, there is a low enough level of elemental iodine in the sump solution that it is appropriate to use the removal coefficient of 20 hr-1. However, when the decontamination factor approaches its defined limit of 200, the removal coefficient would be only a small fraction of its original value. The impact of this varying nature of the removal coefficient is approximated by setting the removal coefficient to one half of the calculated value. However, for conservatism and simplification, a value of 10 hr-1 (half of the 20 hr-1 limits specified in SRP 6.5.2) was used. | |||
The spray removal coefficient is a function of the spray flow rate. The analysis used a spray removal coefficient based on a conservatively low spray flow rate (658.66 m3/hr) to bound both modes (injection and recirculation, with spray flow rates of 665.86 m3/hr and 780.00 m3/hr, respectively) of spray operation. Therefore, sensitivities on timing on the spray switchover times are not necessary. | |||
Westinghouse Non-Proprietary Class 3 Page 53 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The lower spray flow rate and the factor of 2 reduction in the removal coefficient applied for the entire duration of spray operation bounds the impacts on the analysis results due to potential changes in switchover timing. | |||
The following wording is added to Enclosure IV, Section 4.3.9.2.2.1: | |||
During recirculation, the spray solution will gradually become loaded with elemental iodine, which will limit the capacity of the spray to remove airborne iodine. The spray removal coefficient would be inversely proportional to the decontamination factor (DF) achieved for elemental iodine. Thus, when recirculation spray is first credited, there is a low enough level of elemental iodine in the sump solution that it is appropriate to use the removal coefficient of 20 hr-1. However, when the DF approaches its defined limit of 200, the removal coefficient would be only a small fraction of its original value. The impact of this varying nature of the removal coefficient is approximated by setting the removal coefficient to one half of the calculated value. | |||
Also, the spray removal coefficient is a function of the spray flow rate. The analysis used a spray removal coefficient based on a conservatively low spray flow rate to bound both modes (injection and recirculation) of spray operation. | |||
ARCB-RAI-8: | |||
Enclosure VI, page 4-64 states, in part, that An adjustment is made to account for a reduction in the RWST [refueling water storage tank] gas volume available for dilution as the leakage into the RWST increase the water level. | |||
Please provide details regarding this adjustment so that the NRC staff can independently confirm the doses from the RWST back-leakage. | |||
===Response=== | |||
As leakage enters the RWST at a rate of 3.8 gpm (assumed to be all liquid), the gas volume that can be credited for dilution (initially 3.54E+05 gallons) is reduced, resulting in an increased concentration of activity in the RWST gas volume. The leakage is modeled for the duration of the analysis, resulting in a gas volume of 189,840 gallons at 720 hours. Because RADTRAD maintains a constant compartment volume, the effects of the increased concentration are accounted for via an increased flow rate from the RWST. The flow rate is increased in proportion to the expected increase in concentration resulting from the reduced RWST gas volume. | |||
The following wording is added to Enclosure IV, Section 4.3.9.2.2.3: | |||
The RWST gas volume is decreased by the rate of leakage into the RWST. | |||
Westinghouse Non-Proprietary Class 3 Page 54 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-9: | |||
Please provide the doses from each pathway analyzed for the loss-of-coolant accident (for the exclusion area boundary (EAB) (worst 2 hours), low population zone (LPZ), and control room (at 30 days)). Please explain whether the worst 2 hours EAB dose is determined using the sum of the worst 2 hours dose for each pathway to the EAB or by first summing all the time dependent dose pathways and then determining the worst 2 hour dose. | |||
===Response=== | |||
The following summarizes the doses for each pathway analyzed for the LOCA dose consequence analysis: | |||
Exclusion Area Boundary (EAB) (0.5 hr to 2.5 hr): | |||
Containment Leakage 4.73E+0 rem TEDE ECCS Leakage 4.63E-1 rem TEDE RWST Back-Leakage 4.36E-3 rem TEDE Containment Purge 0.00E+0 rem TEDE Low Population Zone (LPZ): | |||
Containment Leakage 2.00E+0 rem TEDE ECCS Leakage 1.26E+0 rem TEDE RWST Back-Leakage 1.57E+0 rem TEDE Containment Purge 8.24E-4 rem TEDE Control Room: | |||
Containment Leakage 1.01E+0 rem TEDE ECCS Leakage 8.48E-1 rem TEDE RWST Back-Leakage 2.69E+0 rem TEDE Containment Purge 6.52E-2 rem TEDE External Sources 1.53E-1 rem TEDE The worst 2-hour EAB dose was determined by summing all the time-dependent dose pathways and then determining the worst 2-hour dose. Note that releases via the mini-purge terminate prior to 0.5 hours and thus do not contribute to the limiting 2-hour EAB dose. | |||
This information is included in Enclosure IV, Section 4.3.9.4. | |||
ARCB-RAI-10: | |||
Enclosure VI, page 4-38 states, in part, that The FIPCO-V computer code calculates the buildup of fission product activities in plant systems and components, including the reactor coolant system, chemical and volume control | |||
Westinghouse Non-Proprietary Class 3 Page 55 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment system demineralizer resins, VCT [volume control tank] liquid and vapor phases, and waste decay tank (WGDT). | |||
Please provide details regarding the input and methodologies used in the FIPCO-V code for the staff to replicate the calculations performed by the FIPCO-V code. | |||
===Response=== | |||
The equations which are used to determine the concentrations of fission products at any time are developed from the general balance equation: | |||
Net Rate of = Rate of - Rate of Accumulation Formation Loss Reactor Core Fission Products The buildup of fission products in the fuel assemblies in the core is described by the following equations: | |||
[ | |||
]a,c | |||
Westinghouse Non-Proprietary Class 3 Page 56 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
[ | |||
]a,c | |||
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[ | |||
]a,c | |||
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[ | |||
]a,c | |||
Westinghouse Non-Proprietary Class 3 Page 59 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
[ | |||
]a,c | |||
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[ | |||
]a,c | |||
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[ | |||
]a,c | |||
Westinghouse Non-Proprietary Class 3 Page 62 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
[ | |||
]a,c | |||
Westinghouse Non-Proprietary Class 3 Page 63 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The inputs used in the FIPCO calculations are provided in the following two tables. The data below are taken from operating data, fuel cycle design data, and plant information. | |||
Key Input Parameters for WCGS RCS Activity Analyses Parameter Value Notes a,c Core thermal power, MWt Fuel Management multiplier RCS boron concentrations Fuel defect level, % | |||
RCS volume, ft3 RCS expansion factor Steam generator (S/G) tube plugging level, % | |||
Normal pressurizer liquid level @ | |||
full power, % | |||
Core Tavg, oF Letdown flow rate, gpm Volume control tank (VCT) volume, gallons/ft3 VCT liquid volume, % | |||
VCT conditions: | |||
Temperature, oF Pressure, psig VCT continuous purge rate, scfm . | |||
Gas decay tank volume, ft3 Number of gas decay tanks . | |||
Effective deborating demineralizer cut-in concentration, ppm Mixed bed resin volume, ft3 | |||
Westinghouse Non-Proprietary Class 3 Page 64 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Critical Boron Concentrations Burnup B Conc. Burnup B Conc. Burnup B Conc. | |||
(MWd/MTU) (ppm) (MWd/MTU) (ppm) (MWd/MTU) (ppm) a,c ARCB-RAI-11: | |||
For the containment purge pathway analyzed for the loss-of-coolant accident (LOCA), | |||
please describe what is the assumed form of radioiodine released from the reactor coolant system prior to isolation of the containment purge. Please justify your answer. | |||
===Response=== | |||
The forms of radioiodine assumed for the containment purge pathway is 97% elemental and 3% | |||
organic. This is an assumed split consistent with other releases specified in Regulatory Guide 1.183. There is no guidance on forms of radioiodine for purge releases in Regulatory Guide 1.183 and the split has no impact on the analysis results since no spray removal is credited for the RCS activity in containment. The Control Room Emergency Ventilation System (which includes the Control Building) filters are modeled with the same efficiency for all forms of iodine. | |||
The following wording is added to Enclosure IV, Section 4.3.9.2.2.4: | |||
Note that the assumed iodine chemical fractions do not impact the analysis results since spray removal is not credited for the RCS activity in containment. | |||
The following wording is added to Enclosure IV, Section 4.3.9.2.3: | |||
Note that the assumed iodine chemical fractions for the containment purge pathway do not impact the analysis results since the emergency mode ventilation system filters are modeled with the same efficiency for all forms of iodine. | |||
Westinghouse Non-Proprietary Class 3 Page 65 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-12 NRC Information Notice (IN) 2012-01: Seismic Considerations- Principally Issues Involving Tanks, dated January 26, 2012 (ADAMS Accession No. ML11292A175), provides examples and references to events in which licensees failed to recognize various seismic considerations and system alignment issues that could impact safety. The NRC staff has identified recent concerns about licensees who failed to recognize that aligning non-seismic piping to the RWST would require TS limiting condition for operation (LCO) action statement entry, system modifications, or license amendments. | |||
RG 1.183, Regulatory Position 5.1.2, Credit for Engineered Safeguard Features, states: | |||
Credit may be taken for accident mitigation features that are classified as safety related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences. | |||
During operations at WCGS, please describe if there is any non-seismic piping aligned to the RWST or systems which recirculate post-LOCA sump fluid. Also, provide details of how the AST analysis is modeled. Please provide enough detail so the NRC staff can independently model this configuration to assess its impact on design basis accident (DBA) doses. | |||
===Response=== | |||
WCNOC review of NRC Information Notice 2012-01 and associated operating experience is documented in the corrective action program. The Refueling Water Storage Tank (RWST) is described in USAR Section 6.3 as part of the Emergency Core Cooling System (ECCS) and its ECCS function is to contain an inventory of 394,000 gallons of water to provide for ECCS mitigation purposes. The RWST is required to perform its function during the short term (less than 24 hours) following a LOCA, MSLB, SGTR or any other accident requiring the ECCS following receipt of an safety injection (SI) signal. The ECCS consists of three separate subsystems: centrifugal charging (high head), SI (intermediate head), and residual heat removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains. | |||
A review of the interface piping connections with the RWST was performed. There are eight interface piping connections to the seismic category 1 piping of the RWST excluding the piping connections from RWST supply header to the suctions of the ECCS pumps and Containment Spray (CS) System pumps. The interface piping connections from RWST supply header to the suctions of the ECCS pumps and CS System pumps are seismic category 1 piping connections. | |||
Westinghouse Non-Proprietary Class 3 Page 66 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The remaining eight interface piping connections on the RWST were reviewed for seismic to non-seismic piping alignment concerns as follows: | |||
: 1. RWST to Spent Fuel Pool Cooling and Cleanup System Flow Path The Spent Fuel Cooling and Cleanup System consists of three subsystems: fuel pool cooling subsystem, fuel pool cleanup subsystem and fuel pool surface skimmer subsystem. The fuel pool cleanup subsystem is non-safety related and has no safety design basis. The fuel pool cleanup subsystem limits the fission and corrosion product concentrations in the refueling pool water, the transfer canal water, and the fuel storage pool water to permit operator access to the fuel storage area and for fuel handling operations. | |||
WCGS has two safety related air operated isolation valves (BNHV8800A/B) in series on the safety related piping from the RWST before it is connected to the non-safety related piping of the fuel pool cleanup subsystem. By design, valves BNHV8800A/B receive a safety injection signal to initiate automatic closure. The valves receive independent and redundant auto closure signals (i.e., each valve is actuated from a different SI actuation train) ensuring these safety related isolation valves are closed when the RWST may be required to support the ECCS. The fail-safe position of these valves is closed (the air solenoid fails open on a loss of power). The design features of the interconnecting piping between the RWST and the fuel pool cleanup subsystem meet GDC 35, Emergency core cooling. This criterion is satisfied by these two valves in that the system safety function can be accomplished with the failure of one valve as the other valve will fulfill the isolation function. Following a design basis accident that requires the ECCS mitigation function, a safety injection signal is generated and results in the closure of BNHV8800A and BNHV8800B, if the RWST was in the purification alignment mode. | |||
: 2. Safety Injection (SI) Pump Recirculation Flow Path to RWST The SI pumps are provided with a recirculation flow path to the top of the RWST to support pump operation. The SI pumps recirculation flow path to the RWST is not in use during normal plant operations as the SI pumps are in standby for ECCS mitigation. The SI pumps recirculation flow path is not used to bring the plant to a cold shutdown condition. The SI pumps recirculation flow path is typically used to support pump operation for pump testing during normal plant operations. The SI pump recirculation flow path to the RWST is seismically qualified piping (seismic category 1 and II/I Special Scope). II/I Special Scope (SS), is non-safety related piping that runs above safety related piping that has been supported in this area seismically so as not to degrade the function of the safety related piping below it. Piping with this qualification is designed and constructed to ensure that its failure will not reduce the functioning of a safety related component to an unacceptable safety level. | |||
: 3. Containment Spray (CS) Pump Test Flow Path to RWST | |||
Westinghouse Non-Proprietary Class 3 Page 67 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The CS pumps are provided with a flow path to the top of the RWST to provide a recirculation flow path to test each pump or to support system testing. The CS pumps recirculation flow path to the RWST is seismically qualified piping (seismic category 1 and II/I SS). | |||
: 4. Charging and Volume Control System (CVCS) Boric Acid Blending Tee Flow Path to RWST The CVCS boric acid blending tee flow path to the top of the RWST provides for the addition of blended boric acid solution to the RWST or a path for testing the boric acid transfer pumps. The CVCS boric acid blending tee flow path to the RWST is seismically qualified piping (seismic category 1 and II/I SS). | |||
: 5. SI System Test Flow Path to RWST The SI System test flow path to the top of the RWST allows for the testing of ECCS check valves, supports depressurization of ECCS piping, and provides dampening of the RHR and SI pump start pressure transients. The non-safety related piping in the SI System test flow path is 3/4 inch piping. Due to this small line size, a break or crack in this size of line is not postulated and isolation is provided by the containment isolation valves. | |||
: 6. RHR Pump Flow Path to RWST The RHR Pump flow path to the RWST is only used during refueling outages to transfer/drain the refueling pool water to the RWST. An isolation valve (BN8717) is normally locked closed and is opened under administrative controls when the plant is in a refueling outage. The RHR pump flow path to the RWST is seismic category 1 qualified piping. | |||
: 7. RWST Drain Flow Path to Waste Holdup Tank The RWST can be drained via a flow path to the Waste Holdup Tank in the Radwaste Building. | |||
This drain path is seismic category 1 qualified piping out to and including its isolation valve BNV0017 that is locked closed. Downstream of the isolation valve, the piping is classified as non-safety related. When the RWST is required to be drained for maintenance or internal inspection during refueling outages, this drain flow path is utilized. | |||
: 8. RWST Overflow Flow Path to Waste Holdup Tank The RWST is provided with an overflow path at the top of the RWST that is also connected to the Waste Holdup Tank. The overflow piping is seismic category 1 qualified near the RWST and connects with non-safety related piping going to the Waste Holdup Tanks. There are no isolation valve(s) in the overflow flow path as it is designed to provide overflow protection for the RWST. | |||
The piping aligned to systems which recirculate post-LOCA sump fluid is seismically qualified piping. | |||
Westinghouse Non-Proprietary Class 3 Page 68 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Relevant Details of AST Analysis Model: | |||
The AST analysis models RWST back-leakage at a rate of 3.8 gpm from the sump. The activity is modeled to be delivered directly to the gas filled portion of the RWST; however, only 10 percent of the activity becomes airborne and is available for release to the environment. The release rate from the RWST to the environment is based on the volume displacement from the incoming leakage. An adjustment to the release rate is made to account for a reduction in the RWST gas volume available for dilution as the leakage into the RWST increases the water level. | |||
The adjusted flow rate is based on increasing the modeled flow in order to exhaust the appropriate activity concentration. This is necessary as RADTRAD maintains a constant compartment volume and transfers are concentration-based. | |||
The RWST modeling description is included in Enclosure IV, Section 4.3.9.2.2.3 and also included as a response to ARCB-RAI-8 in Enclosures VI and VII. | |||
ARCB-RAI-13 According to Enclosure VI, Table 4.3-5, Control Room and Control Room Building Parameters, the delay due to switch to emergency mode operation following receipt of isolation signal in the current licensing basis is stated as N/A or not applicable. The submittal proposes to change this delay to 60 seconds. An isolation set point for the R-23 detector is also provided. | |||
Tables are provided for each accident with analysis parameters and assumptions (i.e., Tables 4.3-6 through Table 4.3-16.). These tables include either the time delay to switch to emergency mode of operation after event initiation, or the time delay to switch to emergency mode operation following receipt of isolation signal. | |||
a) For each change to the current licensing bases, please provide details about how the total time to isolate was calculated. | |||
b) Please state and justify the isolation signal assumed to initiate the switch to emergency mode. | |||
c) Please state whether the time to isolate includes the worst case single failure, control room isolation signal time, emergency diesel generator startup time, time to load the electrical bus, and damper closure time. If not, please justify why these assumptions and delays are not considered. | |||
d) Please describe how the R-23 detector is used to isolate the control room and for which accidents it is credited. Provide enough detail so that the NRC staff can independently calculate the isolation time credited. | |||
e) Please state if the R-23 detector is a general area monitor or is it located in the control room HVAC ductwork. | |||
Westinghouse Non-Proprietary Class 3 Page 69 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment f) Please state if the R-23 detector and initiation signal comply with Regulatory Position 5.1.2, Credit for Engineered Safeguard Features of RG 1.183 and explain how compliance with Regulatory Position 5.1.2 is met. | |||
===Response=== | |||
The R-23 detector discussed above is hereafter referred to as the control room air supply radiation monitors (GKRE0004 and GKRE0005). | |||
a) The total time to isolate was calculated by determining a time by which the isolation signal(s) would have been generated and then adding a delay to account for the Control Room Ventilation Isolation Signal (CRVIS) to actuate the Control Room Emergency Ventilation System (CREVS). Isolation signals include safety injection or high radiation in the control room air supply radiation monitors. To ensure that the time delay modeled in the analyses is bounding, new TS Surveillance Requirement (SR) 3.3.7.6 has been added for the performance of the required response time verification. | |||
For the loss of coolant accident (LOCA) radiological consequences analysis, the safety injection isolation signal is assumed to occur immediately after event initiation. For conservatism, the switchover to emergency mode ventilation was not assumed to occur until 2 minutes after the onset of the event (including the 1 minute delay for switchover). This is presented in Enclosure IV of this LAR, Section 4.3.9.2.3. | |||
For the main steam line break (MSLB) radiological consequences analysis, the safety injection isolation signal is assumed to occur immediately after event initiation. For conservatism, the switchover to emergency mode ventilation was not assumed to occur until 90 seconds after the onset of the event (including the 1 minute delay for switchover). This is presented in Enclosure IV of this LAR, Section 4.3.3.2.3. | |||
For the rod ejection containment leakage case, the safety injection isolation signal is assumed to occur within 150 seconds of event initiation. The switchover to emergency mode ventilation was not assumed to occur until 210 seconds after the onset of the event (including the 1 minute delay for switchover). This is presented in Enclosure IV of this LAR, Section 4.3.6.2.3. | |||
For the steam generator tube rupture (SGTR), rod ejection primary-to-secondary case, locked rotor, and fuel handling accident (FHA) radiological consequences analysis, the control room air supply radiation monitors setpoint is shown to be reached for each accident immediately after the onset of the event. This is based on the amount of Xe-133 activity initially in the primary system, the environmental release rate modeled for each analysis, and the atmospheric dispersion factor modeled for the control room. | |||
Since the control room air supply radiation monitors are located in the control room HVAC ductwork, the concentration of activity entering the control room is the concentration seen by the monitors. The delay from the control room air supply | |||
Westinghouse Non-Proprietary Class 3 Page 70 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment radiation monitors setpoint being reached to the switchover to emergency mode ventilation is 60 seconds. Each analysis conservatively assumed that switchover did not occur for at least 120 seconds. This is presented in Enclosure IV of this LAR, Sections 4.3.8.2.3, 4.3.6.2.3, 4.3.5.2.3, and 4.3.12.2.3. | |||
b) For the LOCA, rod ejection containment leakage case, and MSLB analyses, a safety injection signal is credited. For LOCA and rod ejection containment leakage, the safety injection signal is generated by the reduction in RCS pressure as a result of the respective breaks. For MSLB, the safety injection signal is generated by the low steamline pressure as a result of the break. This is presented in Enclosure IV of this LAR, Sections 4.3.9.2.3, 4.3.6.2.3, and 4.3.3.2.3. | |||
c) The CRVIS time, emergency diesel generator startup time, time to load the electrical bus and damper closure times have been considered and are bounded by the 60 second delay time assumed for each accident scenario as described in response to item a) above. | |||
The worst case single failure of a control room filtration fan is considered and results in a pathway for contaminated air from the control building to bypass the control room filtration system filter adsorber unit and thereby enter the control room unfiltered. The mitigative action for this single failure is an operator action to isolate this flow path by closing the appropriate train-specific damper. This single failure is the same as that assumed in the current licensing basis analysis of record (AOR), where the operator action is assumed to take place 1.5 hours after event initiation. Since the two CREVS trains are independent, this single failure only affects one train of the CREVS. | |||
d) The control room air supply radiation monitors are credited to isolate the control room in the SGTR, rod ejection primary-to-secondary leakage, locked rotor, and FHA analyses. The activity concentration seen by the control room air supply radiation monitors setpoint is determined by the following equation: | |||
Activity Concentration at Control Room Air Supply Radiation Monitors = Activity Conc | |||
* Environment Release Rate | |||
* CR /Q The following calculations determine the time to reach the setpoint for each accident. | |||
Note that the analyses modeled a control room radiation monitor setpoint of 2.12E-3 Ci/m3 Xe-133. | |||
Westinghouse Non-Proprietary Class 3 Page 71 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment SGTR Since the analysis assumes no delay in the release of activity and no delays in the transport of activity through the primary and secondary systems or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 4.19E4 Ci and the RCS mass is 3.99E5 lbm. The initial break flow rate is 3085 lbm/min and the 0-2 hour CR /Q for steaming releases from the SG for normal operation is 2.55E-2 sec/m3. | |||
The activity concentration at the detector is: | |||
(4.19E4 Ci / 3.99E5 lbm)(3085 lbm/min | |||
* 1 min/60 sec)(2.55E-2 sec/m3) = 1.4E-1 Ci/m3 The control room radiation monitor is shown above to exceed its setpoint immediately. | |||
This comparison is presented in Enclosure IV of this LAR, Section 4.3.8.2.3. | |||
Note that Enclosure IV, Section 4.3.2.1 describes that isolation based solely on the control room radiation monitor does not fully switch the HVAC systems such that the control room unfiltered inleakage continues to be associated with the normal mode X/Qs. The SGTR analysis credits the safety injection signal due to low pressurizer pressure to complete isolation of the control room. From Enclosure I, Section 2.7.3, the safety injection signal occurs less than 6 minutes after the onset of the event. For conservatism, the isolation of the control room is completed at 10 minutes after the onset of the event. | |||
Rod Ejection Primary-to-Secondary Leakage Since the analysis assumes no delay in the release of activity from the damaged fuel and no delays in the transport of activity through the primary and secondary systems or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 4.15E6 Ci and the RCS mass is 3.99E5 lbm. The primary to secondary leak rate is 8.342 lbm/min (applying a conservative analysis density of 62.4 lbm/ft3) and the 0-2 hour CR /Q for steaming releases from the SG for normal operation is 2.55E-2 sec/m3. | |||
The activity concentration at the detector is: | |||
(4.15E6 Ci / 3.99E5 lbm)(8.342 lbm/min | |||
* 1 min/60 sec)(2.55E-2 sec/m3) = 3.7E-2 Ci/m3 The control room air supply radiation monitors are shown above to exceed the setpoint immediately. | |||
This comparison is presented in Enclosure IV of this LAR, Section 4.3.6.2.3. | |||
Westinghouse Non-Proprietary Class 3 Page 72 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Locked Rotor Since the analysis assumes no delay in the release of activity from the damaged fuel and no delays in the transport of activity through the primary and secondary systems or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 8.30E5 Ci and the RCS mass is 3.99E5 lbm. The primary to secondary leak rate is 8.342 lbm/min (applying a conservative analysis density of 62.4 lbm/ft3) and the 0-2 hour CR /Q for steaming releases from the SG for normal operation is 2.55E-2 sec/m3. | |||
The activity concentration at the detector is: | |||
(8.30E5 Ci / 3.99E5 lbm)(8.342 lbm/min | |||
* 1 min/60 sec)(2.55E-2 sec/m3) = 7.4E-3 Ci/m3 The control room air supply radiation monitors are shown above to exceed the setpoint immediately. | |||
This comparison is presented in Enclosure IV of this LAR, Section 4.3.5.2.3. | |||
FHA Since the analysis assumes no delay in the release of activity from the damaged fuel and no delays in the transport of activity through the spent fuel pool or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 1.62E5 Ci. Given the linear release over a 2-hour period, the release rate is 8.1E4 Ci/hr or 22.5 Ci/sec. The 0-2 hour CR /Q for releases is 1.38E-3 sec/m3. (Note that this is the /Q associated with releases from a FHA inside containment. It is selected for this check as the lower | |||
/Q is conservative in determining when the setpoint is reached.) Therefore the concentration of Xe-133 at the control room air supply radiation monitors is: | |||
(22.5 Ci/sec)(1.38E-3 sec/m3) = 3.11E-2 Ci/m3 The control room air supply radiation monitors are shown above to exceed the setpoint immediately. | |||
This comparison is presented in Enclosure IV of this LAR, Section 4.3.12.2.3. | |||
Westinghouse Non-Proprietary Class 3 Page 73 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment e) The control room air supply radiation monitors are located in the control room HVAC ductwork. | |||
f) Yes, the control room air supply radiation monitors comply with Regulatory Position 5.1.2 of Regulatory Guide 1.183. Specifically, compliance is demonstrated by the following: | |||
: 1) The two redundant radiation monitors and associated actuation instrumentation are safety-related. | |||
: 2) The radiation monitors are required to be operable by TS limiting condition for operation (LCO) 3.3.7, CREVS Actuation Instrumentation. | |||
: 3) The radiation monitors are powered by Class 1E electrical equipment. | |||
: 4) The control room air supply radiation monitors provide input to initiate a CRVIS. | |||
Proper actuation of the CREVS is verified in the emergency operating procedures. | |||
ARCB-RAI-14 Please provide a diagram describing the model used for modeling the control building and control room for each design basis accident. Please provide the unfiltered in-leakage into the control building and justify the value. Please explain if this value has been confirmed by testing and if it will be confirmed periodically as part of a TS surveillance program. (Note this issue was previously identified in NRC Regulatory Issue Summary 2006-04, Issue Number 3.) | |||
===Response=== | |||
The figure displayed below (obtained from Section 15A.3 of the Wolf Creek USAR), provides a diagram of the flow rates modeled for the control building and control room. The Normal Mode HVAC flow rates are modeled prior to CREVS actuation. Following the CREVS actuation, the Emergency Mode HVAC Prior to Operator Action provides the flow rates considering a single failure of a control room filtration fan, which is consistent with the analysis of record single failure. The Emergency Mode HVAC after Operator Action flow rates are modeled after the failed train has been isolated by closing the train-specific damper. | |||
Westinghouse Non-Proprietary Class 3 Page 74 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Enclosure IV, Table 4.3-5 contains the flow rates input values used in the analysis. A value of 400 cfm was modeled as the inleakage into the control building. The inleakage value was increased from the analysis of record value of 300 cfm to allow additional testing margin. An acceptance criterion of 300 cfm was confirmed in October 2010 during performance of TS SR 3.7.10.4. Although the analysis acceptable inleakage increased, the previous TS surveillance test performance in accordance with the Control Room Envelope Habitability Program (TS 5.5.18) demonstrates that the control building inleakage is less than 400 cfm. The Frequency of SR 3.7.10.4 is also in accordance with the Control Room Envelope Habitability Program. This Frequency is procedurally controlled to 6 years. | |||
ARCB-RAI-15 In the proposed TS Bases B 3.3.8, Emergency Exhaust System (EES) Actuation Instrumentation, which ensures that radioactive materials in the fuel building atmosphere are filtered and absorbed prior to exhausting to the environment, the EES is not credited after 76 hours of decay from the fuel. Without the EES credited, the leakage (source) could occur anywhere there is a penetration or hole in the fuel building (rather than through the exhaust of the EES). Please explain how the worst-case source-receptor pairings are determined for the calculation of atmospheric dispersion factors when the EES is not credited. | |||
===Response=== | |||
In Reference 1, WCNOC proposed to adopt TSTF-51-A, Revision 2, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations. The adoption of TSTF A revised TS 3.3.8 by adding recently to the Applicability. The revised Applicability stated: | |||
During movement of recently irradiated fuel assemblies in the fuel building. In the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Revision 2, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations. are being | |||
Westinghouse Non-Proprietary Class 3 Page 75 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment withdrawn. Subsequently, WCNOC withdrew the license amendment request (Reference 1) in WCNOC letter dated June 18, 2014 (Reference 3). The resubmittal of the license amendment request excluded the adoption of TSTF-51-A, Revision 2. With the removal of changes proposed by TSTF-51-A, the Emergency Exhaust System (EES) actuation instrumentation is required to be OPERABLE during movement of irradiated fuel assemblies in the fuel building. | |||
As such, if a FHA were to occur, upon receipt of a fuel building ventilation isolation signal generated by the gaseous radioactivity monitors in the fuel building exhaust line, normal air discharges from the building are terminated, the fuel building is isolated, and the stream of ventilation air discharges through the EES filter trains. | |||
However, as described in Enclosure IV of this LAR, Section 4.3.12, while the EES is credited for the event and thus the release location will be the unit vent for a fuel handling accident within the fuel building, filtration from the EES was conservatively not credited for the FHA analysis as the dose limits were met without crediting EES filtration. | |||
Updates to the TS Bases are presented in Enclosure IV of this LAR, page B 3.3.8-1 of Section 11. | |||
ARCB-RAI-16 WCNOC assumes the control room does not isolate after a fuel handling accident. The normal unfiltered outside air makeup flow to the control building and the control room is 13,050 cubic feet per minute (cfm) and 1950 cfm, respectively. The unfiltered in-leakage to the control room is assumed to be 50 cfm. | |||
a) Please justify the use of 50 cfm for unfiltered in-leakage in the configuration when the control room ventilation system is not isolated. Please state if this value has been confirmed by testing and will it be confirmed periodically as part of a TS surveillance program. (Note this issue was previously identified in NRC Regulatory Issue Summary 2006-04, Issue Number 3.) | |||
b) In this case only, the normal makeup for the control building and control room is used to mitigate the consequences of the fuel handling accident. Please state if the normal control room heating and ventilation systems is credited meet the qualities, attributes, and performance criteria as described in RG 1.183, Regulatory Positions 4.2.4 and 5.1.2. (Note this issue was previously identified in NRC Regulatory Issue Summary 2006-04, Issue Number 3.) If so, please justify how the credited ventilation system complies with Regulatory Positions 4.2.4 and 5.1.2. If not, justify the proposed alternative assumptions used. | |||
c) Please state if there is a surveillance for these normal makeup flow rates in the TSs. If not, please explain if a sensitivity analysis has been performed to determine the limiting dose based upon the range of possible makeup flow rates. Please provide the results (dose versus makeup flow rates) of any sensitivity analyses performed. | |||
Westinghouse Non-Proprietary Class 3 Page 76 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment d) For other accidents that do not assume control room isolation is actuated (i.e., locked reactor coolant pump rotor, loss of alternating current power, letdown line break and tank ruptures) please explain if the assumptions used to model the control building or control room are the same as those used for the fuel handling accident. In your submittal for these accidents, where the control room isolation is not credited, the control room ventilation system is assumed to remain in the normal mode of operation. | |||
Please state whether this assumption yields more conservative doses than if the emergency mode is assumed to be actuated and justify your answer. For these accidents (other than the fuel handling accident) that do not credit control room isolation, please explain if a sensitivity analysis has been performed to determine the limiting dose based upon a range of possible makeup flows. If so, please provide the results of the analysis (dose vs. makeup flow). If not, justify why the assumed makeup flows are limiting. | |||
===Response=== | |||
In the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Rev. 2, are being withdrawn. As such, the normal HVAC lineup is not credited to mitigate the effects of a fuel handling accident. Specific responses are provided below: | |||
a) The 50 cfm of unfiltered inleakage in the configuration when the control room ventilation system is in the normal mode of operation is consistent with the unfiltered inleakage in the configuration when the CREVS is in service. The 50 cfm is conservatively added to the 1950 cfm air makeup flow to the control room to yield a total flow rate of 2000 cfm of unfiltered atmosphere into the control room. Since the normal makeup flow rates are not credited to mitigate the consequence of a fuel handling accident there is not a TS Surveillance Requirement to measure the normal makeup flow rates and inleakage. | |||
b) As stated in the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Rev. 2, are being withdrawn. As such, the normal makeup for the control building and control room is not credited to mitigate the consequences of a fuel handling accident. | |||
c) The normal makeup flow rates are no longer credited to mitigate the consequence of a fuel handling accident. Thus, there is not a TS Surveillance Requirement for the normal makeup flow rates. A sensitivity analysis was not performed for the normal flow rates. | |||
However, the actual normal makeup flow rates were measured and compared to the values modeled in the dose analyses. The normal makeup flow rates modeled in the dose analyses are greater than the measured plant flow rates for the control building and control room by more than 10%. | |||
Westinghouse Non-Proprietary Class 3 Page 77 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment d) The following analyses only credit normal makeup flow rates to mitigate the consequences of the accident: loss of alternating current power, letdown line break, and tank ruptures. The assumptions used in these analyses are consistent with the assumptions for the normal mode of operation used in each accident analyzed for the AST implementation, as described in Section 4.3.2.1 and Table 4.3.5 of Enclosure IV of this LAR. Remaining in normal mode of operation is conservative for the accidents since the normal mode flow rates are significantly higher than the emergency mode flows and no filtration is credited for activity entering the control building or control room during normal mode. In addition, filtered recirculation is credited for activity removal in the control room during emergency mode of operation. Thus, the activity in the control room is greater for normal mode of operation. | |||
The following wording is added to Enclosure IV of this LAR, Sections 4.3.4.2.3, 4.3.7.2.3, 4.3.10.2.3 and 4.3.11.2.3: This modeling is conservative since activity reduction due to filtration of inflow and filtered recirculation is not credited. | |||
A sensitivity analysis was not performed for the normal flow rates. However, the actual normal makeup flow rates were measured and compared to the values modeled in the dose analyses. The normal makeup flow rates modeled in the dose analyses are greater than the measured plant flow rates for the control building and control room by more than 10%. | |||
ARCB-RAI-17: | |||
The current licensing basis for the radioactive waste gas decay tank failure, from Updated Final Safety Analysis Report (UFSAR) Section 15.7.1.2, states that the tank is assumed to fail after 40 years, releasing the peak inventory expected in the tank. The proposed change requests a change to this assumption. Please justify this change and explain why it is conservative. | |||
===Response=== | |||
A detailed discussion of the waste gas decay tank failure calculation and justification for the assumption is provided in the response to ARCB-RAI-31 below. | |||
Westinghouse Non-Proprietary Class 3 Page 78 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-18 Page 15.7-13 of the UFSAR "markup" states that the gap fractions are obtained for high burnup fuel from Regulatory Guide (RG) 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Safety Guide 25)," dated March 1972 (ADAMS Accession No. ML083300022), as modified by NUREG/CR-5009, "Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors," dated February 1988 (not publicly available), to support the conservative assumption that 100 percent of the rods do not meet the burnup and kilowatts per foot (kW/ft) limits set forth in Footnote 11 of RG 1.183. Enclosure VI, page 4-69 of the submittal elaborates on the use of NUREG/CR-5009. | |||
a) Please justify the use and applicability of NUREG/CR-5009 instead of industry standards such as ANSI/ANS-5.4-2011, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel." | |||
b) Please justify the statement made on page 3A-12 of the UFSAR markup that "Use of this regulatory guide [RG 1.25] has been replaced by Regulatory Guide 1.183 for alternative source term application," in light of the statement that the gap fractions are obtained from RG 1.25. | |||
===Response=== | |||
a) The following table compares the nuclide gap fractions used in the WCGS Fuel Handling Accident analysis (which are obtained from RG 1.25, as modified by NUREG/CR-5009) versus those specified in Table 2.9 of PNNL-18212, Revision 1 (Update of Gap Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard, ADAMS Accession Number ML112070118), which calculated the gap release fractions utilizing ANSI/ANS-5.4-2011. | |||
WCGS Analysis Calculated Gap Fractions Nuclide Gap Fractions Utilizing ANSI/ANS-5.4-2011 Kr-85 0.30 0.38 I-131 0.12 0.08 I-132 0.10 0.09 Other Noble Gases 0.10 0.08 Other Halogens 0.10 0.05 | |||
Westinghouse Non-Proprietary Class 3 Page 79 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment As shown in the table above, the gap fractions used in the WCGS analysis are conservatively higher than those specified in ANSI/ANS-5.4-2011, with the exception of Kr-85. Kr-85 is not a significant dose contributor when compared to the other nuclides being modeled. Kr-85 has an effective dose equivalent (EDE) dose conversion factor (DCF) of 1.19E-16 Sv-m3/Bq-sec and a released activity of 3.38E+03 Ci, whereas Xe-133 (a major dose contributor) has an EDE DCF of 1.56E-15 Sv-m3/Bq-sec and a released activity of 1.62E+05 Ci. | |||
In the fuel handling accident radiological consequences analysis both release scenarios are considered, namely inside containment and inside the fuel building. For both release scenarios, Kr-85 contributes less than 0.03% of the total dose at each location (exclusion area boundary, low population zone, control room, and technical support center). The dose increase resulting from an increase of ~27% of Kr-85 activity (gap fraction of 0.30 vs 0.38), would be more than offset by decreasing the activity releases for other nuclides when modeling the ANSI/ANS-5.4-2011 gap fractions. | |||
Therefore, the gap fractions used for WCGS, which are based on RG 1.25, as modified by NUREG/CR-5009, are conservative in comparison to the gap fractions using ANSI/ANS-5.4-2011. | |||
b) The USAR text is revised to state that the use of RG 1.25 is being specified in Table 15B-1 [RG 1.183 Conformance Table]. | |||
ARCB-RAI-19 The current single failure taken for the fuel handling accident is the failure of the humidly control system for the engineered safety feature (ESF) emergency filtration system as stated on page 15.7-13 of the UFSAR. Please explain the worst case single failure that is assumed for the fuel handling accident in the proposed analysis. Please justify your answer. | |||
===Response=== | |||
The current single failure of the humidity control system resulted in a reduced iodine removal efficiency. The fuel handing accident analysis, documented in Section 4.3.12 of Enclosure IV of this LAR, modeled a direct release of activity to the environment with no credit for the filtration of the EES. Therefore, by conservatively not crediting the EES filtration, the analysis results documented in Section 4.3.12 of Enclosure IV of this LAR are more limiting than if the previous single failure of the humidity control system had been retained and a reduced iodine removal efficiency was credited. | |||
Consistent with the current control room radiological consequences calculation models, as discussed on page 15A-8 of the USAR, a failure of one of the filtration fans is assumed at the start of emergency mode of operation and a larger unfiltered inflow to the control room is assumed since only half of the makeup flow to the control room can pass through a filter. After a | |||
Westinghouse Non-Proprietary Class 3 Page 80 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment defined time of 90 minutes, operator action isolates the failed train and reduces the unfiltered inflow to the control room. | |||
ARCB-RAI-20 Enclosure VI, page 4-69 of the submittal states, in part, that Although not explicitly discussed, the specified overall DF [decontamination factor] also applies to rod internal pressures up to 1500 psig [pounds per square inch gauge]. | |||
The DF of 200 provided in RG 1.183 is based upon Reference B-1 ("Evaluation of Fission Product Release and Transport," dated October 5, 1971 (ADAMS Legacy Accession No. | |||
8402080322)) of RG 1.183. The data upon which the pool DF of 200 is based was developed in 1971 and was based on the Westinghouse fuel marketed at the time (the assumed internal fuel pressure of 1200 psig was used). Since higher pressures correlate to lower DFs, the NRC staff is concerned that a DF of 200 might not be sufficiently conservative for pressures higher than 1200 psig. | |||
Please provide the data for current fuel types used at WCNOC that justify a DF of 200 for fuel pressures up to 1500 psig. Also, please provide a detailed justification for using a DF of 200 for pressures up to 1500 psig. | |||
===Response=== | |||
The current fuel type for WCNOC (17x17 RFA-2) is generically addressed for a DF of 200 at higher rod internal pressures by WCAP-16072-P-A (Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, ADAMS Accession Number ML042510053). WCAP-16072-P-A was submitted to the NRC for review and approval, and included an evaluation of iodine decontamination factors for fuel rod pressures up to 1500 psig. In the Final Safety Evaluation for WCAP-16072-P-A (ADAMS Accession Number ML041270102), the NRC stated that, the staff has determined that there is reasonable assurance that fuel rod design pressures of up to 1500 psig will not invalidate analysis assumptions related to iodine decontamination. The staff has also determined that this conclusion remains valid for the decontamination factor of 200 provided in RG 1.183 and RG 1.195, which supersede SG 25 for alternative source terms and TID14844 source terms, respectively. | |||
WCAP-16072-P-A was prepared for CE NSSS plant fuel designs. However, the justification for the continued applicability of a DF of 200 was based on evaluations performed in WCAP-7518-L Radiological Consequences of a Fuel Handling Accident which did not distinguish between fuel types. WCAP-7518-L provides a method for calculating an iodine DF given a bubble rise time which is dependent on rod internal pressures. This method is not fuel type specific and was used and discussed in WCAP-16072-P-A to justify the continued applicability of a DF value of | |||
Westinghouse Non-Proprietary Class 3 Page 81 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment 200. Therefore, the justification provided in WCAP-16072-P-A is applicable to all Westinghouse fuel types. | |||
It should be noted that Regulatory Guide 1.183 refers to NRC Staff Technical Paper Evaluation of Fission Product Release and Transport for a Fuel Handling Accident (ADAMS Accession Number 8402080322). WCAP-7518-L is discussed in the NRC Staff Technical Paper as being submitted to the NRC to present new experimental data pertaining to the consequences of a refueling accident. | |||
Also, the NRC has previously approved the use of a DF of 200 for fuel pressures up to 1500 psig (reference Indian Point Nuclear Unit No. 3 - Issuance of Amendment Re: Full Scope Adoption of Alternative Source Term, ADAMS Accession Number ML050750431). In the NRC approval of this amendment, the NRC noted that the DF for elemental iodine would remain above 400 for fuel pressures at 1500 psig. This is greater than the DF of 285 for elemental iodine used in the analysis to obtain an overall iodine DF of 200. Therefore, the use of a DF of 200 remains conservative and is appropriate for use. | |||
ARCB-RAI-21 , page 4-69 of the submittal states: The decay time used in determining the inventory of the damaged rods is 76 hours. Thus, the analysis supports the TS limit [emphasis added] of 76 hours decay time prior to fuel movement. The staff agrees that the decay time appears to meet Criterion 2 of 10 CFR 50.36, Technical Specifications, but could not locate the TS limit of 76 hours. Please state where in the TSs this limit exists. | |||
===Response=== | |||
The sentence on page 4-69 of Enclosure VI of Reference 1 is in error. The 76 hour decay time prior to fuel movement limit is not specified in the Technical Specifications. In the conversion to the improved Technical Specifications (Reference 6), discussion of change (DOC) 3-01-R proposed the relocation of specification 3/4.9.3, Decay Time, to a licensee controlled document. The NRC approved the improved Technical Specifications in License Amendment No. 123 (Reference 7). The safety evaluation for License Amendment No. 123 states, in part: | |||
: 2. CN 3-01-R [current Technical Specifications (CTS)] 3/4.9.3, Reactor Decay Time (CTS 3/4.9) | |||
The requirements in CTS 3/4.9.3 on the decay time that the reactor core must be subcritical before there is movement of irradiated fuel in the reactor core are being relocated to the TRM. This LCO requires the reactor to be subcritical for 100 hours to allow the radioactive decay of the short-lived fission products. The screening criteria for including the requirements in the ITS have been satisfied for Criterion 2 since decay time is consistent with the assumptions used in an accident analysis; however, the activities necessary to be performed at WCGS before commencing movement of irradiated fuel | |||
Westinghouse Non-Proprietary Class 3 Page 82 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ensure that 100 hours of subcriticality will elapse before there is movement of irradiated fuel in the core. Therefore, because the CTS is not required to assure that 100 hours have elapsed prior to fuel movement, the decay time LCO and SRs in the CTS may be relocated to the TRM, a licensee-controlled document outside TS. The TRM is included by reference in the USAR and is an acceptable licensee-controlled document; therefore, the relocation is acceptable. | |||
The two relocated specifications from the CTS discussed above are not required to be in the ITS because they do not fall within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii). They are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety. In addition, the NRC staff finds that sufficient regulatory controls exist under 10 CFR 50.59 to maintain the effect of the provisions in these specifications. The NRC staff has concluded that appropriate controls have been established for all of the current specifications that are being moved to the USAR. | |||
The decay time requirement was relocated to USAR Section 9.1.4.2.3. Additionally, the TS 3.9.7 Bases, Refueling Pool Water Level, discussed the minimum decay time prior to fuel handling. | |||
ARCB-RAI-22: | |||
Please confirm that with the exception of different release points, the assumptions and inputs are identical for the fuel handling accident within the containment and the fuel handling accident outside the containment. | |||
===Response=== | |||
The assumptions and inputs for the fuel handling accident within containment and outside of containment are identical with the exception of control room and technical support center atmospheric dispersion factors associated with the different release points. | |||
ARCB-RAI-23: | |||
a) Please confirm that the most limiting combination of release point and receptor for the control room and technical support center (TSC) were used to determine atmospheric dispersion factors for each accident. | |||
b) Please explain why the control room atmospheric dispersion factors provided in the tables for individual accidents (i.e., Tables 4.3-11 and 4.3-15) do not correlate to values provide in Table 4.1.2-3. | |||
c) Please state and justify the release points that correlate to the atmospheric dispersion factors for each design basis accident. | |||
Westinghouse Non-Proprietary Class 3 Page 83 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
===Response=== | |||
a) Based on the NRC question, WCNOC re-reviewed the release point and receptor combinations and determined more limiting pairs should be used. Two additional release points were determined to need atmospheric dispersion factor calculations: (1) release from a reactor building wall diffuse area source and (2) release from the radwaste building. | |||
In addition, it was also discovered during the review (of release point and receptor combinations) that atmospheric dispersion factors needed to be determined for one additional receptor: the control room normal HVAC intake. | |||
Due to the new atmospheric dispersion factor calculations, four of the presented accident and transport path combinations were determined to not be using the most limiting combination of release point and receptor. The affected accident and transport path combinations are: (1) rod ejection accident containment leakage transport path to the control room emergency intake vent, (2) loss of coolant accident containment leakage transport path to the control room emergency intake vent, (3) waste gas decay tank failure transport path to the control room emergency intake vent and TSC intake vent, and (4) liquid waste tank failure transport path to the control room emergency intake vent and TSC intake vent. These accident and transport path combinations have been updated and are included in the response to item c) below, and also contained in Enclosure IV of this LAR, Section 4.1.2.4 and Table 4.1.2-5. | |||
b) The updated (refer to response a) above) Table 4.1.2-3 and updated individual accident tables (Table 4.3-6 through 4.3-15) have been included in Enclosure IV of this LAR, and have been confirmed to contain the same information. The updated version of the Table 4.1.2-3, which has been renumbered to 4.1.2-3(a), is included below for reference. | |||
Westinghouse Non-Proprietary Class 3 Page 84 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.1.2-3(a) Calculated /Q (sec/m3) for the Emergency Control Room Intake Vent Equipment Hatch to Emergency Control Room Air Intake 0 to 2 hours 5.44E-04 2 to 8 hours 4.35E-04 8 to 24 hours 1.62E-04 1 to 4 days 1.22E-04 4 to 30 days 8.70E-05 Unit Vent Stack to Emergency Control Room Air Intake 0 to 2 hours 6.12E-04 2 to 8 hours 4.38E-04 8 to 24 hours 1.79E-04 1 to 4 days 1.14E-04 4 to 30 days 8.94E-05 MSSVs/ARVs to Emergency Control Room Air Intake 0 to 2 hours 1.04E-03 2 to 8 hours 7.46E-04 8 to 24 hours 3.03E-04 1 to 4 days 1.90E-04 4 to 30 days 1.39E-04 RWST Vent to Emergency Control Room Air Intake 0 to 2 hours 6.80E-04 2 to 8 hours 6.19E-04 8 to 24 hours 2.27E-04 1 to 4 days 1.96E-04 4 to 30 days 1.53E-04 TDAFW Exhaust to Emergency Control Room Air Intake 0 to 2 hours 5.17E-04 2 to 8 hours 3.99E-04 8 to 24 hours 1.60E-04 1 to 4 days 1.00E-04 4 to 30 days 7.21E-05 Radwaste Building to Emergency Control Room Air Intake 0 to 2 hours 6.92E-04 2 to 8 hours 6.06E-04 8 to 24 hours 2.39E-04 1 to 4 days 2.00E-04 4 to 30 days 1.47E-04 Reactor Building Wall to Emergency Control Room Air Intake 0 to 2 hours 6.02E-04 2 to 8 hours 4.35E-04 8 to 24 hours 1.81E-04 1 to 4 days 1.29E-04 4 to 30 days 9.65E-05 | |||
Westinghouse Non-Proprietary Class 3 Page 85 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment c) Enclosure IV of this LAR, Table 4.1.2-5, included for reference below, provides the release points considered in the calculation of the /Q values for each transport path of each accident. The table is structured according to the accident type and associated transport paths. The accident types include (1) main steam line break (two transport paths), (2) loss of non-emergency AC power (one transport path), (3) locked rotor (one transport path), (4) rod ejection (two transport paths), (5) letdown line break (one transport path), (6) steam generator tube rupture (one transport path), (7) loss of coolant accident (four transport paths), (8) waste gas decay tank failure (one transport path), (9) liquid waste tank failure (one transport path), and (10) fuel handling accident (two transport paths). Some accidents have multiple transport paths. The /Q values assigned to a given accident depend on the transport paths associated with that accident. Accidents with the same transport path are assigned the same /Q values. | |||
As explained in Enclosure IV of this LAR, Section 4.1.2.4, for each accident, transport paths and release points are identified. Transport paths refer to paths such as primary-to-secondary leakage, primary-to-secondary break, containment leakage, ECCS leakage, RWST back-leakage, containment purge, direct release to auxiliary building, and direct release to atmosphere. The release points refer to the release points whose /Q values for CR and TSC intakes have been determined. These release points include the reactor building equipment hatch, reactor building wall diffuse source, unit vent stack, TDAFW exhaust vent, MSSV/ARV vents, RWST vent, and radwaste building. If a transport path has a single release point, the /Q associated with that release point is the limiting value. Some transport paths may have multiple release points. For such cases, the limiting /Q values are assumed to be the maximum of all release points associated with that transport path and release period. A site diagram with the considered release points and receptor points labeled is shown in Figure 4.1.2-1 below, and contained in Enclosure IV of this LAR, Section 4.1.2. | |||
The containment leakage path considers three release points (equipment hatch, unit vent stack, and diffuse reactor building wall). For the containment leakage path, the limiting /Q values applied to the emergency control room intake are from the unit vent stack for a period from 0 to 8 hours, and from the reactor building wall for a period from 8 hours to 30 days. The limiting /Q values for the normal control room intake are from the unit vent stack for all periods. The limiting | |||
/Q values for the technical support intake are from the equipment hatch for all periods. | |||
The primary-to-secondary leakage to the intact SGs is another example of a transport path with multiple release points. The path through the intact SGs leads to the release points at the MSSV/ARV vents and the TDAFW exhaust vent. For this case, the limiting /Q values are assigned to the larger values of the two release points. | |||
Westinghouse Non-Proprietary Class 3 Page 86 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment For the main steam line break accident, the primary-to-secondary leakage to the intact SGs and the faulted SG are the two transport paths that lead to different release points to the environment. The path through the intact SGs leads to the release points at the MSSV/ARV vents and the TDAFW exhaust vent. The limiting /Q values for the intact SG release path are assumed the larger values of the two release points. However, the path through the faulted SG releases radionuclides into the auxiliary building if a steam line break occurs outside containment. The main steam line break is assumed to occur outside of containment because a break outside of the containment will bound any break inside containment. The radionuclides released from the break are drawn by the building ventilation system to the unit vent stack and eventually exhausted to the environment through the unit vent stack. Hence, the /Q values for the broken steam line release path are the values for the unit vent. | |||
For the loss of non-emergency AC power accident, the primary-to-secondary leakage to intact SGs transport path leads to two release points at the MSSV/ARV vents and the TDAFW exhaust vent. The /Q values for this release path are assumed as the larger value of the two release points. | |||
For the locked rotor accident, the primary-to-secondary leakage to SGs transport path leads to two release points at the MSSV/ARV vents and the TDAFW exhaust vent. The /Q values for this release path are assumed as the larger value of the two release points. | |||
For the rod ejection accident there are two transport paths, containment leakage to the environment, and primary to secondary leakage. The /Q values for the containment leakage release path are assumed as the maximum value of the three release points: (1) equipment hatch, (2) reactor building wall, and (3) unit vent stack. The /Q values for the primary to secondary leakage release path are assumed as the larger value of the MSSV/ARV vents and the TDAFW exhaust vent release points. | |||
For the letdown line break accident, the line break is assumed to occur outside of containment and radionuclides are directly released into the auxiliary building. The /Q values for this case are those of the unit vent stack. | |||
For the steam generator tube rupture accident, the primary-to-secondary break to the ruptured SG is the major transport path to the environment. The /Q values for the ruptured SG release path are assumed the larger value of the MSSV/ARV vents and the TDAFW exhaust vent. | |||
For the loss of coolant accident, there are four transport paths. The first path is the containment leakage transport path to the environment. The /Q values for this transport path are assumed as the maximum value of the three release points: (1) equipment hatch, (2) reactor building wall, and (3) unit vent stack. The second transport path is the ECCS leakage path. The /Q values for this path are those of the unit vent stack. The third transport path is the RWST back-leakage path. The /Q values for this path are of the RWST vent. The fourth transport path is the containment purge path. The containment purge system vents the containment atmosphere | |||
Westinghouse Non-Proprietary Class 3 Page 87 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment through the containment exhaust penetrations up to the unit vent stack. Therefore, the /Q values for this path are those of the unit vent stack. | |||
For the waste gas decay tank failure, direct release from the radwaste building is the only credited transport path to the environment. The /Q values for this case are those of the radwaste building. | |||
For the liquid waste tank failure, direct release from the radwaste building is the only credited transport path to the environment. The /Q values for this case are those of the radwaste building. | |||
For the fuel handling accident, two possible accident types are considered: (1) in containment and (2) in fuel building. For a fuel handling accident occurring in containment, the transport path is through the open equipment hatch. For a fuel handling accident occurring in the fuel building, the radionuclides released are drawn by the building ventilation system to the unit vent stack and eventually exhausted to the environment through the unit vent stack. | |||
Westinghouse Non-Proprietary Class 3 Page 88 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
Westinghouse Non-Proprietary Class 3 Page 89 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Figure 4.1.2-1. Diagram of Source and Receptor Locations for Wolf Creek Table 4.1.2-5 Accident Release Sources Accident Transport Path Release Points Limiting /Q Main Steamline Primary to Secondary Leakage to Intact MSSVs/ARVs Maximum of all Break SG Associated TDAFW Exhaust Release Points Primary to Secondary Leakage to Unit Vent Unit Vent Faulted SG for a Break Outside of Containment Primary to Secondary Leakage to Equipment Hatch Maximum of all Faulted SG for a Break Inside of Associated Unit Vent Containment Release Points (This transport path is not used Reactor Building because the break outside of Wall (Diffuse containment is bounding) Source) | |||
Loss of Primary to Secondary Leakage MSSVs/ARVs Maximum of all Non-Emergency Associated TDAFW Exhaust AC Power Release Points Locked Rotor Primary to Secondary Leakage MSSVs/ARVs Maximum of all Associated TDAFW Exhaust Release Points Rod Ejection Containment Leakage Equipment Hatch Maximum of all Associated Unit Vent Release Points Reactor Building Wall (Diffuse Source) | |||
Primary to Secondary Leakage MSSVs/ARVs Maximum of all Associated TDAFW Exhaust Release Points Letdown Line Release Outside of Containment Unit Vent Unit Vent Break Steam Generator Primary to Secondary Break MSSVs/ARVs Maximum of all Tube Rupture Associated TDAFW Exhaust Release Points | |||
Westinghouse Non-Proprietary Class 3 Page 90 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.1.2-5 Accident Release Sources (cont.) | |||
Accident Transport Path Release Points Limiting /Q Loss of Coolant Containment Leakage Equipment Hatch Maximum of all Accident Associated Unit Vent Release Points Reactor Building Wall (Diffuse Source) | |||
ECCS Leakage Unit Vent Unit Vent RWST Back-Leakage RWST Vent RWST Vent Containment Purge Unit Vent Unit Vent Waste Gas Release to Atmosphere Radwaste Building Radwaste Decay Tank Building Failure Liquid Waste Release to Atmosphere Radwaste Building Radwaste Tank Failure Building Fuel Handling Release In Containment Equipment Hatch Equipment Hatch Accident Release In Fuel Building Unit Vent Unit Vent ARCB-RAI-24 In a letter dated November 7, 2013 (ADAMS Accession No. ML13246A358), the NRC informed the Technical Specifications Task Force of concerns that the NRC staff had recently identified during a review of plant-specific license amendments requesting adoption of three travelers including traveler TSTF-51, Revision 2, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations. | |||
Enclosure VI, page 2-4 of the submittal discusses the proposed TSTF-51 changes. TSTF-51 states, in part, that The addition of the term recently associated with handling irradiated fuel in all of the containment function Technical Specification requirements is only applicable to those licensees who have demonstrated by analysis that after sufficient radioactive decay has occurred, off-site doses resulting from a fuel handling accident remain below the Standard Review Plan limits (well within 10CFR100). [or 10 CFR 50.67] | |||
Westinghouse Non-Proprietary Class 3 Page 91 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment NUREG-0800, Standard Review Plan (SRP) 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, July 2000 (ADAMS Accession No. ML003734190), states, in part, that The models, assumptions, and parameter inputs used by the licensee should be reviewed to ensure that the conservative design basis assumptions outlined in RG-1.183 have been incorporated. | |||
Appendix B of RG 1.183, Regulatory Position 1.1 states: | |||
The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or weight of a dropped fuel assembly ... | |||
With regard to the WCGS submittal to adopt TSTF-51, please provide plant-specific information to verify that the limiting cases have been considered, e.g., a fuel handling accident analysis that evaluates the dropping of loads allowed over irradiated fuel assemblies (i.e., new fuel assembly, sources, or reactivity control components) onto irradiated fuel assemblies prior to and after the proposed 76-hour decay time. Such an analysis should only credit those safety systems required to be operable as required by TS. There must be reasonable assurance that the fuel handling accident analysis doses remain within regulatory limits when references to Core Alterations are removed from TSs and Engineered Safety Features are no longer required during movement of loads such as new fuel assemblies, sources or reactivity control components. | |||
===Response=== | |||
Reference 1 included the adoption of TSTF-51-A, Revision 2. WCNOC became aware of the NRC concerns with TSTF-51 in early October 2013 after Reference 1 had been submitted to the NRC. With the subsequent issuance of a letter dated November 7, 2013 (Reference 8), | |||
WCNOC communicated with the NRC Project Manager on November 14, 2013 by electronic mail that consideration would be given to withdrawing the changes proposed by TSTF-51-A depending on the significance of the impact to Reference 1. WCNOC withdrew the license amendment request (Reference 1) by WCNOC letter dated June 18, 2014 (Reference 3). The resubmittal of the license amendment request excludes the adoption to TSTF-51, Revision 2. | |||
Westinghouse Non-Proprietary Class 3 Page 92 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-25 Several calculations assume an iodine partition in the steam generators of 100 which is applied to the releases resulting from steaming of secondary side fluid (i.e., Enclosure VI, page 4-59). Please confirm that the partition factor is applied only to the elemental iodine or justify how organic iodine is partitioned. | |||
===Response=== | |||
Regulatory Position 5.5.4 in Appendix E of Regulatory Guide 1.183, which specifies the iodine partition of 100 in the steam generators, does not distinguish between elemental and organic iodine. As a result, the partition factor of 100 was applied to both forms of iodine being released from the steam generators for all applicable analyses. The NRC approved this modeling of partitioning on iodine releases from the steam generators in previous AST submittals (reference Point Beach Nuclear Plant (PBNP), Units 1 and 2 -Issuance of License Amendments Regarding Use of Alternate Source Term (TAC Nos. ME0219 and ME0220), ADAMS Accession Number ML110240054). | |||
ARCB-RAI-26 The column labeled "Comments" of Enclosure VI, Table A for Regulatory Position 4.1.1, states that progeny was not included in the dose calculations consistent with the two previously approved submittals. The NRC staffs review of the subject safety evaluations of these two submittals did not find explicit approval or a review of excluding the effects of progeny. Since this appears to conflict with Regulatory Position 4.1.1 and could potentially yield a non-conservative estimate of doses, please either provide justification for not including the progeny, or include progeny in these calculations. | |||
===Response=== | |||
The modeling of progeny has been included in the updated results presented in Enclosure IV of this LAR, Section 4.3. The comments in Enclosure IV of this LAR, Table A for Regulatory Position 4.1.1 have been updated to the following response: | |||
The dose calculations determine the TEDE and consider all radionuclides, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences. | |||
Westinghouse Non-Proprietary Class 3 Page 93 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-27 For those accidents that model accident and pre-existing spikes, RG 1.183 specifies the release from the fuel to be in the form of 95 percent cesium iodide. Please confirm that the spike modeled in the reactor coolant system models the increase in cesium in addition to the iodine modeled. | |||
===Response=== | |||
A spike in cesium associated with the iodine spikes (both pre-accident and accident-initiated) in the WCGS radiological consequences analyses (main steamline break, steam generator tube rupture, loss of AC power, and letdown line break) was not explicitly modeled, consistent with methods used in previously approved AST submittals (e.g., Point Beach Units 1 and 2 AST Approval, April 14, 2011 under ADAMS Accession Number ML110240054). However the initial cesium activity modeled in the RCS bounds the cesium which would be released to the RCS in association with iodine spiking. | |||
The cesium activity initially modeled in the RCS for the above accident analyses was that associated with the design basis fuel defect level of 1%. The initial RCS iodine activity was scaled down to the Technical Specification 3.4.16, RCS Specific Activity, limit of 1.0 µCi/gm dose equivalent (DE) I-131 as identified in Enclosure IV of this LAR, Section 4.3 and explained further in the response to ARCB-RAI 37. As shown in the response to ARCB-RAI 37, the total isotopic dose equivalence for the iodines considered is 4.22, which means that the 1% fuel defect iodine activity is reduced by a factor of 4.22 to be at the Technical Specification 3.4.16 limit of 1.0 µCi/gm DE I-131. Therefore, the fuel defect level corresponding to the Technical Specification limit is approximately 0.24% fuel defects. | |||
Tables 1 and 2 convert the fuel defect level activities to corresponding nuclide masses. | |||
The RCS mass used in the calculation is the mass used in the dose calculations (1.81E+08 gm). Note that the calculations were performed in a spreadsheet and additional decimal places were not included in the tables for ease of presentation. | |||
A sample calculation for Cs-134 is as follows: | |||
RCS Activity (Ci) = 1% Fuel Defect Activity | |||
* RCS Mass | |||
* 1 Ci / 1E6 µCi | |||
= 4.82E+00 µCi/gm | |||
* 1.81E+08 gm | |||
* 1 Ci / 1E6 µCi = 8.72E+02 Ci RCS Activity (Bq) = RCS Activity (Ci) | |||
* 3.7E+10 Bq/Ci | |||
= 8.72E+02 Ci | |||
* 3.7E+10 Bq/Ci = 3.23E+13 Bq Number of atoms = RCS Activity (Bq) | |||
* Half-life (sec) / ln(2) | |||
= 3.23E+13 Bq | |||
* 6.50E+07 sec / ln(2) = 3.03E+21 atoms Mass (gm) = Number of atoms | |||
* Atomic Mass Number (gm/mole) / 6.022E+23 (atoms/mole) | |||
= 3.03E+21 atoms | |||
* 134 gm/mole / 6.022E+23 atoms/mole = 6.74E-01 gm | |||
Westinghouse Non-Proprietary Class 3 Page 94 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 1 Cesium Mass at 1% Fuel Defect Level 1% Fuel Defect Activity RCS RCS Activity Half-life Number of Nuclide (µCi/gm) Activity (Ci) (Bq) (sec) atoms Mass (gm) | |||
Cs-134 4.82E+00 8.72E+02 3.23E+13 6.50E+07 3.03E+21 6.74E-01 Cs-136 4.35E+00 7.87E+02 2.91E+13 1.13E+06 4.75E+19 1.07E-02 Cs-137 2.68E+00 4.85E+02 1.79E+13 9.46E+08 2.45E+22 5.57E+00 Cs-138 1.16E+00 2.10E+02 7.77E+12 1.93E+03 2.16E+16 4.96E-06 Total: 6.26E+00 Table 2 Cesium Mass at Technical Specification Limit Fuel Defect Level 0.24% Fuel Defect Activity RCS RCS Half-life Number of Nuclide (µCi/gm) Activity (Ci) Activity (Bq) (sec) atoms Mass (gm) | |||
Cs-134 1.16E+00 2.09E+02 7.75E+12 6.50E+07 7.26E+20 1.62E-01 Cs-136 1.04E+00 1.89E+02 6.99E+12 1.13E+06 1.14E+19 2.57E-03 Cs-137 6.43E-01 1.16E+02 4.31E+12 9.46E+08 5.88E+21 1.34E+00 Cs-138 2.78E-01 5.04E+01 1.86E+12 1.93E+03 5.19E+15 1.19E-06 Total: 1.50E+00 As shown, the amount of excess cesium mass in modeling 1% fuel defects as opposed to the fuel defects of 0.24% is 4.76 gm. The amount of excess was determined for the radiocesium nuclides considered in the dose analyses (Cs-134, Cs-136, Cs-137, and Cs-138). | |||
For the pre-accident spike scenario, the total iodine mass modeled in the RCS at the spike limit of 60 µCi/gm DE I-131 is calculated. The mass contributions of both I-127 and I-129, as well as the mass contributions of iodine modeled in the dose analyses (I-130 through I-135), are calculated. Since I-127 is stable, its contribution to 1% fuel defects is given in units of grams of I-127 per gram of RCS coolant. | |||
Table 3 converts the fuel defect level activities to corresponding nuclide masses. The pre-accident spike activity level used is 60 µCi/gm DE I-131 and the RCS mass used in the calculation is the mass used in the dose calculations (1.81E8 gm). The calculations were performed in a spreadsheet and additional decimal places were not included in the tables for ease of presentation. | |||
Westinghouse Non-Proprietary Class 3 Page 95 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 3 Iodine Mass in RCS for Pre-Accident Iodine Spike Isotopic Activity at Activity Scaled Isotopic Isotopic Isotopic 1.0% Fuel to 1.0 µCi/gm Activity at Activity at Activity at Defect Level DE I-131 1.0 µCi/gm DE 60 µCi/gm 60 µCi/gm DE I- Half-life Number of Nuclide (µCi/gm) (µCi/gm) I-131 (Ci) DE I-131 (Ci) 131 (Bq) (sec) atoms Mass (gm) | |||
I-127 1.24E-10 2.94E-11 5.32E-03 (gm) 3.19E-01 (gm) N/A N/A N/A 3.19E-01 (gm/gm) (gm/gm) | |||
I-129 7.17E-08 1.70E-08 3.07E-06 1.84E-04 6.82E+06 4.95E+14 4.87E+21 1.04E+00 I-130 4.65E-02 1.10E-02 1.99E+00 1.20E+02 4.43E+12 4.45E+04 2.84E+17 6.13E-05 I-131 3.28E+00 7.77E-01 1.41E+02 8.44E+03 3.12E+14 6.95E+05 3.13E+20 6.81E-02 I-132 3.39E+00 8.03E-01 1.45E+02 8.72E+03 3.23E+14 8.28E+03 3.85E+18 8.45E-04 I-133 5.04E+00 1.19E+00 2.16E+02 1.30E+04 4.80E+14 7.49E+04 5.18E+19 1.14E-02 I-134 7.30E-01 1.73E-01 3.13E+01 1.88E+03 6.95E+13 3.16E+03 3.17E+17 7.05E-05 I-135 2.85E+00 6.75E-01 1.22E+02 7.33E+03 2.71E+14 2.38E+04 9.31E+18 2.09E-03 Total 1.45E+00 | |||
Westinghouse Non-Proprietary Class 3 Page 96 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Since the atomic masses of cesium and iodine are similar and 95% of the iodine is in particulate form (i.e., CsI), then the total cesium which would be modeled by a pre-accident iodine spike is 1.37 gm (1.45 gm of iodine | |||
* 0.95 fraction of CsI to total iodine), which is less than the excess radiocesium initially modeled in the RCS. | |||
For the accident-initiated spike scenario, the total iodine mass in the gap of the defective fuel rods is calculated. The mass contributions of both I-127 and I-129, as well as the mass contributions of iodine modeled in the dose analyses (I-130 through I-135), are calculated. Since I-127 is stable, its contribution in the core is given in units of grams of I-127. | |||
Table 4 converts the iodine activity available in the core to corresponding nuclide masses that are available for release from the fuel gap in the defective fuel rods. The fuel gap fractions of 12% for I-131 and 10% for the remaining iodines are used and are conservative gap fractions based on those used in the fuel handling accident analysis. Also, the fuel defect level of 0.24% | |||
(described above) is used. Note that the calculations were performed in a spreadsheet and additional decimal places were not included in the tables for ease of presentation. | |||
Westinghouse Non-Proprietary Class 3 Page 97 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4 Iodine Mass Available for Release from Fuel Gap in Accident-Initiated Iodine Spike Activity Available Activity Activity Activity Fuel in Available Gap Available In Available in Defect Defective Half Life Number Nuclide In Core (Ci) Fraction Gap (Ci) Gap (Bq) level (%) Fuel (Bq) (sec) of atoms Mass (gm) | |||
I-127 4.31E+03 0.1 4.31E+02 N/A 0.24 1.03E+00 N/A N/A 1.03E+00 (gm) (gm) (gm) | |||
I-129 2.57E+00 0.1 2.57E-01 9.51E+09 0.24 2.28E+07 4.95E+14 1.63E+22 3.49E+00 I-130 1.98E+06 0.1 1.98E+05 7.33E+15 0.24 1.76E+13 4.45E+04 1.13E+18 2.44E-04 I-131 1.01E+08 0.12 1.21E+07 4.48E+17 0.24 1.08E+15 6.95E+05 1.08E+21 2.35E-01 I-132 1.49E+08 0.1 1.49E+07 5.51E+17 0.24 1.32E+15 8.28E+03 1.58E+19 3.46E-03 I-133 2.10E+08 0.1 2.10E+07 7.77E+17 0.24 1.86E+15 7.49E+04 2.02E+20 4.45E-02 I-134 2.36E+08 0.1 2.36E+07 8.73E+17 0.24 2.10E+15 3.16E+03 9.55E+18 2.13E-03 I-135 2.00E+08 0.1 2.00E+07 7.40E+17 0.24 1.78E+15 2.38E+04 6.10E+19 1.37E-02 Total: 4.82E+00 | |||
Westinghouse Non-Proprietary Class 3 Page 98 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Since the atomic masses of cesium and iodine are similar and 95% of the iodine is in particulate form (i.e., CsI), then the total cesium which would be modeled by an accident-initiated iodine spike is 4.58 gm (4.82 gm of iodine | |||
* 0.95 fraction of CsI to total iodine), which is less than the excess radiocesium modeled in the RCS. | |||
Therefore, the dose impact of modeling the design fuel defect level cesium activity in the RCS instead of scaling to the fuel defect level activity bounds the impact of modeling an increase in radiocesium resulting from iodine spiking. | |||
It should be noted that the core composition of cesium isotopes includes approximately 40% | |||
Cs-133 which is a stable nuclide and has no dose impact. Assuming that the core composition of cesium is similar to that released from the fuel gap to the RCS as CsI, then the total radiocesium which would be released during an iodine spike is approximately 60% of that in the calculations above. Therefore, the calculations above are conservative. | |||
ARCB-RAI-28 The column labeled "Comments" of Enclosure VI, Table F for Regulatory Position 5.6 and Table G for Regulatory Position 7.4 states that: "The transport model described in Regulatory Positions 5.5 and 5.6 ... was considered as appropriate [emphasis added] ..." | |||
Please clarify if the models described in these Regulatory Positions were used. If only those considered appropriate were used, please state which were not considered appropriate and justify why they are not appropriate. | |||
===Response=== | |||
The transport model described in Regulatory Positions 5.5 and 5.6 in Appendix E of Regulatory Guide 1.183 were used for the Locked Rotor and Rod Ejection analyses with the following exceptions: | |||
Position 5.5.1: For the Rod Ejection and Locked Rotor radiological consequences analyses, steam generator dryout does not occur due to the secondary system remaining intact so no consideration for flashing of primary-to-secondary leakage was modeled. | |||
Position 5.5.2: As there is no flashing of primary-to-secondary leakage, no credit was taken in the Rod Ejection and Locked Rotor radiological consequences analyses for scrubbing of flashed leakage in the steam generators. | |||
Position 5.6: Comments provided in Table D of Enclosure VI (now Enclosure IV of this LAR) for Regulatory Position 5.6 state that the issue for tube uncovery was addressed in WCAP-13247 [The Topical Report on the Methodology for the Resolution of the Steam Generator Uncovery Issue is included as part of the response to ARCB-RAI-36 for non-SGTR events (Rod Ejection and Locked Rotor included).] It was concluded in WCAP-13247 and confirmed in the NRC response to WCAP-13247 that the effect of tube uncovery would be essentially negligible and the issue could be closed without any further investigation or generic restrictions. This position was accepted by the NRC as | |||
Westinghouse Non-Proprietary Class 3 Page 99 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment noted in Table D. Therefore, the Rod Ejection and Locked Rotor radiological consequences analyses do not model tube uncovery. | |||
ARCB-RAI-29: | |||
Enclosure VI, page 4-54 states, in part, that The minimum SG [steam generator] water mass is increased after 2 hours to take credit for operators maintaining level at narrow range just on span. | |||
Please explain if this this assumption was credited previously. If not, please provide a justification for this assumption. | |||
===Response=== | |||
This assumption was not credited in the previous WCGS analyses of record. This modeling is justified by the WCGS Emergency Management Guidelines (EMGs), which instruct the operators to maintain the level in the steam generators between 6% and 50% of narrow range span (NRS). The minimum mass modeled in the analysis is associated with a NRS SG water level of 0% corresponding to hot zero power conditions. This is conservative given the EMGs instruct the operators to maintain level of at least 6% of NRS. | |||
The following wording is added to Enclosure IV, Sections 4.3.4.2.2, 4.3.5.2.2, and 4.3.6.2.2.2: | |||
as justified by the Emergency Management Guidelines (EMGs). The mass used in the analysis after 2 hours was calculated at 0% narrow range to appropriately bound the just on span level of 6% narrow range. | |||
Westinghouse Non-Proprietary Class 3 Page 100 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-30 Regulatory Position 5.1.2 of RG 1.183 states, in part, that The single active component failure that results in the most limiting radiological consequences should be assumed. | |||
Please provide the most limiting single active failure for each design basis accident. | |||
===Response=== | |||
Consistent with the current control room radiological consequences calculation models, as discussed on page 15A-8 of the USAR, a failure of one of the filtration fans is assumed at the start of emergency mode of operation and a larger unfiltered inflow to the control room is assumed since only half of the makeup flow to the control room can pass through a filter. After a defined time of 90 minutes, operator action isolates the failed train and reduces the unfiltered inflow to the control room. The following accidents modeled switchover to the CREVS and included the failure of one of the filtration fans: | |||
LOCA MSLB SGTR Rod Ejection Locked Rotor Fuel Handing Accident For the SGTR analysis, the limiting single failure is a failed open steam generator ARV on the ruptured steam generator, which is modeled in the mass releases calculated in Section 2.7.2 of WCAP-17658-NP, Revision 1 (Enclosure I of this LAR). The single failure of a failed open steam generator ARV is conservatively modeled in addition to the failure of one of the CREVS filtration fans. | |||
For the remaining dose analyses, no CREVS actuation is generated due to the event and thus the failure of one of the CREVS filtration fans is not modeled. These accidents include: | |||
Tank ruptures Letdown line break Loss of AC power These accidents model a release directly to the environment and do not credit filtration from either the EES or from the CREVS. Thus, there is not a single failure that would decrease the effectiveness of the exhaust or intake filtration and therefore an explicit single failure is not modeled. This is consistent with the current licensing basis AOR. | |||
Westinghouse Non-Proprietary Class 3 Page 101 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-31: | |||
Enclosure VI, page 15.7-1 of the UFSAR markups indicates a proposed change in the licensing bases for the waste gas decay tank failure. Previously, the tank was assumed to fail after 40 years, releasing the peak inventory expected in the tank. The proposed change is assumed to be the maximum activity for each radionuclide during the degassing operations and the Krypton-85 inventory is assumed to be the total activity released during the fuel cycle. RG 1.24, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure," March 1972 (ADAMS Accession No. ML083300020), states, in part, that The maximum content [emphasis added] of the decay tank assumed to fail should be used for the purpose of computing the noble gas inventory in the tank a) If previously the tank was assumed to accumulate activity over 40 years, please explain why the activity from one fuel cycle is conservative and consistent with the RG 1.24 above regulatory position of using the maximum content of the decay tank. Please justify the proposed change in assumptions regarding the tank contents and state why they are conservative. | |||
b) Table 4.3-2a contains 3 columns of source terms used to calculate the tank failures described in Enclosure VI, Sections 4.3.10, "Waste Gas Decay Tank Failure (USAR | |||
[UFSAR] Chapter 15.7.1.5)," and 4.3.11, "Liquid Waste Tank Failure (USAR [UFSAR] | |||
Chapter 15.7.2.5)." Please state the assumptions used to calculate these source terms. | |||
===Response=== | |||
a) The methods and assumptions used to determine the proposed gas decay tank source terms are consistent with, and somewhat more conservative than Regulatory Guide 1.24 as discussed below. | |||
As stated in Regulatory Guide 1.24 Section C.1, Regulatory Position, the assumptions related to the release of radioactive gases from the postulated failure of a gaseous waste storage tank should be, | |||
: a. The reactor has been operating at full power with one percent defective fuel and a shutdown to cold condition has been conducted near the end of an equilibrium core cycle. As soon as possible after shutdown, all noble gases have been removed from the primary cooling system and transferred to the gas decay tank that is assumed to fail. | |||
: b. The maximum content of the decay tank assumed to fail should be used for the purpose of computing the noble gas inventory in the tank. Radiological decay may be taken into account in the computation only for the minimum time period required to transfer the gases from the primary system to the decay tank. | |||
Westinghouse Non-Proprietary Class 3 Page 102 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
: c. The failure is assumed to occur immediately upon completion of the waste gas transfer, releasing the entire contents of the tank to the building. The assumption of the release of the noble gas inventory from only a single tank is based on the premise that all gas decay tanks will be isolated from each other whenever they are in use. | |||
: d. All of the noble gases are assumed to leak out of the building at ground level over a two hour time period. | |||
The WCGS design contains 6 decay tanks that are available for containing waste gases during plant operation and, as noted in the Discussion Section B of Regulatory Guide 1.24, The radioactive components are principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of the halogens. With the exception of krypton-85, the longest half-life of the principal noble gas radionuclides present in reactor effluents is 5.27 days (xenon-133). Thus, storage of these gases for a period of 60 days will essentially eliminate by decay all of the radionuclides except krypton-85. | |||
Thus, the krypton-85 activity that is generated during plant operations prior to cold shutdown will have been discharged from the tank(s) following the storage period during which the short lived species are allowed to decay. If the krypton activity is not discharged, it will be distributed among one or all of the 6 normal operation decay tanks. | |||
The design also includes 2 shutdown gas decay tanks. The gaseous inventory in the primary coolant system is transferred to one (or both) of these tanks via the volume control tank (VCT) as soon as possible after shutdown. Since the shutdown tank will contain a relatively large amount of (primarily) short-lived activity, this is the tank that is assumed to fail. Also, for additional conservatism, all of the krypton-85 activity generated during the previous core cycle (from both normal operation and shutdown operations) is assumed to be present in the failed tank prior to the release. | |||
As noted in the above discussion, the methods and assumptions used in defining the activity inventories in the failed gas decay tank are consistent with the assumptions stated in Regulatory Guide 1.24 Section C.1, Regulatory Position; additionally, conservatism has been added by considering all of the activity that is generated during the previous operating cycle, plus the activity that is removed during the shutdown is contained in the failed gas decay tank. The previous assumption of a 40-year accumulation of krypton-85 activity is overly conservative and inconsistent with the design and operation of the waste gas system. | |||
Westinghouse Non-Proprietary Class 3 Page 103 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment b) All of the source terms are based on the conservative assumption that the plant has operated for a complete equilibrium core cycle with fuel cladding defects in the fuel rods that generate one percent of the core thermal power. A power uncertainty (i.e., calorimetric power uncertainty) factor of 1.02 is considered in the calculation of the core activities; an assumption that meets the requirements of Regulatory Guide 1.195 and Regulatory Guide 1.183, i.e. The inventory of fission products should be based on an assumed core power equal to the current licensed rated thermal power times the emergency core cooling system (ECCS) evaluation uncertainty. In addition, all calculated activity values are increased be a factor of [ ]a,c in order to account for potential core design parameter differences in future core designs. This factor is generally referred to as a fuel management multiplier and is intended to provide bounding values relative to future core designs as applied to accident analyses. Further, the primary coolant concentrations are conservatively assumed to be at the peak values calculated over the operating period for the case with no purge of the volume control tank (VCT) during power operation. | |||
The waste gas decay tank sources are based on solving a set of equations that model the shutdown degassing of the primary coolant in which the isotopic inventory of the gas decay tank varies with time. The letdown of the primary coolant is assumed to occur at the maximum letdown flow rate with coolant degassing operations accomplished by burping of the volume control tank vapor space at 3-hour intervals. The maximum activities over the degassing period are conservatively selected in defining the limiting gas decay tank inventory. In addition to the noble gases, radioactive iodine is considered to partition in the VCT vapor space (a partition factor = 100) and to be carried over into the gas decay tank during the degassing operations. | |||
The Kr-85 activity inventory is treated differently in that the total activity released to the coolant during the fuel cycle is assumed to be transferred to the gas decay tank. | |||
Two separate scenarios for the accidental release of activity from liquid waste tanks are analyzed; that is, 1) the failure of a recycle holdup tank from which the release of 100% of the noble gas nuclides and 10% of the iodine activity in the tank is assumed, and 2) the failure of a hypothetical liquid waste tank from which 100% of the iodine activity is released. The recycle holdup tank activity release is used in analyzing the whole body dose impacts of the accident, and the release of activity from the hypothetical waste tank is considered in determining the thyroid dose consequences of a postulated accident. | |||
The source terms for the tanks consider primary coolant activities as an input and apply appropriate reduction factors in order to determine the tank activity inventories. The reduction factors are based on decay and the reduction in inventory due to removal by the demineralizers which are located upstream of the tank. The contribution of the parent nuclide is also considered. | |||
Two sources of activity are considered in the evaluation of the recycle holdup tank sources; i.e., | |||
the Reactor Coolant System (RCS) and the Reactor Coolant Drain Tank (RCDT). The liquid goes through two mixed bed demineralizers between the exit from the RCS and the recycle holdup tank. The intermittent flow through the cation demineralizer located between the two mixed bed demineralizers is conservatively ignored. Also, there is one demineralizer between | |||
Westinghouse Non-Proprietary Class 3 Page 104 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment the outlet of the RCDT and the recycle holdup tank. Thus, the associated Decontamination Factors (DFs) considered in the analysis are: | |||
For the noble gases, no reduction by the demineralizers is considered, i.e., the DF = 1 For Br and Iodine, a factor of 10 reduction across each demineralizer is considered, i.e., | |||
For the RCS to recycle holdup tank: the DF = 10 x 10 = 100 For the RCDT: the DF = 10 The concept of a hypothetical liquid waste tank was originally developed by WCNOC in order to replace sources associated with the evaporator bottoms tank, since the WCGS does not use the evaporators and the evaporator bottoms tank was no longer considered to be the bounding case for the thyroid dose analysis. Although not associated with an existing piece of equipment, the continued use of sources associated with a hypothetical liquid waste tank is considered to be appropriate. The hypothetical waste tank receives fluid from three separate tanks: | |||
A recycle holdup tank (RHUT), | |||
A waste holdup tank (WHUT), and A floor drain tank (FDT). | |||
It is conservatively assumed that the input to each of the tanks is based on continuous full power operation with equilibrium iodine activity levels in the primary coolant. Based on this activity concentration, it is assumed that each of the feed tanks is filled and drained such that a constant volume in the feed tank is maintained and the activity concentration reaches an equilibrium level. Thus, the feed rate into each tank equals the discharge rate from the tank into the hypothetical tank and the (equilibrium) activity concentration in the feed tank comprises the input of activity into the hypothetical waste tank. Finally, equilibrium activity levels are assumed to exist in the hypothetical waste tank based on decay alone. | |||
ARCB-RAI-32 Enclosure VI markups of UFSAR page 15.3-11 change the sentence from "Steam generator tube leakage is assumed to continue until the pressures in the reactor coolant and secondary systems are equalized" to ''Steam generator tube leakage is assumed to continue until the residual heat removal system can match decay heat and release from the secondary system are terminated." Please justify this proposed change. | |||
===Response=== | |||
The time to terminate steam generator tube leakage to the intact SGs (i.e., SGs neither faulted nor ruptured) was set to the time at which the steam releases were terminated (12 hours, which is the time that the residual heat removal systems match decay heat and eliminate the need for steaming from the secondary system for all analyses involving steam releases). When the | |||
Westinghouse Non-Proprietary Class 3 Page 105 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment steam releases are terminated, the activity leakage into the secondary system no longer has an impact on the resulting doses because the activity released to the environment is terminated. | |||
ARCB-RAI-33 Enclosure VI markups of UFSAR page 6.5A-3 add an Insert N which defines the term kg or the "gas-phase mass-transfer coefficient." The kg used is taken from Reference 2 or Brookhaven National Laboratory (BNL) Technical Report A-3788, "Fission Product Removal Effectiveness of Chemical Additives in PWR Containment Sprays," August 1986, rather than using the reference (Reference 20, entitled: "The Terminal Speed of Single Drops or Bubbles in an Infinite Medium," International Journal of Multiphase Flow, pages 491-511 (1974)) cited in SRP 6.5.2, "Containment Spray as a Fission Product Cleanup System." | |||
Justify why Reference 2 is appropriate rather than the study cited in the SRP and state why it was used. Please provide a copy of the BNL report. | |||
===Response=== | |||
The gas-phase mass-transfer coefficient (kg term) from BNL Technical Report A-3788, dated August 1986 (included in Enclosure VIII of this LAR), was previously used in the Indian Point Unit 2 License Amendment Request (LAR) in the transition to the use of the AST. The Indian Point Unit 2 LAR (reference Proposed Amendment Consisting of Changes to Technical Specification for Containment Air Filtration, Control Room Air Filtration, and Refueling Conditions, ADAMS Accession Number ML993400414) references the Indian Point Unit 2 Licensing Report submitted to the NRC in the October 8, 1999 letter (reference Radiological Consequences of Accidents for the Indian Point Nuclear Generating Station Unit No. 2 Using Source Term Methodology from NUREG-1465, ADAMS Accession Number ML100430988) which gives a kg value of 9.84 ft/min (which is consistent with the value used for the WCGS analysis). It was determined by the NRC staff in the Safety Evaluation (reference Indian Point Nuclear Generating Unit No. 2 - Re: Issuance of Amendment Affecting Containment Air Filtration, Control Room Air Filtration, and Containment Integrity During Fuel Handling Operations (TAC No. MA6955, ADAMS Accession Number ML003727500)) that the licensee used plant-specific parameters and a very conservative value for the gas phase mass transfer coefficient. | |||
Therefore, the kg value from BNL Technical Report A-3788 is also appropriate for use in the WCGS analysis. | |||
The BNL report was used as a reference for the kg value in the proposed revision to SRP Section 6.5.2, Revision 2 (dated April 1987) and was included by the NRC as part of this proposed revision. | |||
From SRP Section 6.5.2, Revision 4: The first-order removal coefficient by spray, s, may be taken to be S=6KgTF/VD where Kg is the gas-phase mass-transfer coefficient and T is the time of fall of the drops, which may be estimated by the ratio of the average fall height to the terminal velocity of the mass-mean drop. | |||
Westinghouse Non-Proprietary Class 3 Page 106 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Earlier revisions of SRP Section 6.5.2 (Revisions 2 and 3) cite Reference 20 at the end of the above sentence. This statement indicates the Reference 20 document is only meant for the terminal velocity calculations that result in the fall time of the drops. Reference 20 is not cited because it does not contain a recommended kg value. | |||
ARCB-RAI-34 Enclosure VI, page 8-14 states in the comments for NRC Regulatory Issue Summary 2006-04, Issue 6 that there were no changes to the plant configuration. The changes to TS 3.9.4 allow an open containment when moving fuel that is not recently irradiated. Consistent with Regulatory Issue Summary 2006-04, please confirm that all pathways to the environment created by the proposed changes are considered and analyzed in the fuel handling accident analysis. | |||
===Response=== | |||
In the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Rev. 2, are being withdrawn. As such, the changes to TS 3.9.4 are being withdrawn and the Limiting Condition for Operation (LCO) 3.9.4 requirements remain applicable during CORE ALTERATIONS and during movement of irradiated fuel assemblies within containment. | |||
The current LCO 3.9.4 allows the equipment hatch to be open and capable of being closed, one personnel air lock capable of being closed, and penetration flow paths that have direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls. These allowances were approved in Amendment No. 146 (Reference 10), Amendment No. 95 (Reference 11), and Amendment No. 135 (Reference 12), respectively. | |||
Consistent with the current analysis of record, the radiological consequences AST analysis for a fuel handling accident inside containment did not credit isolation of containment. Administrative controls (previously approved by the NRC) provide reasonable assurance that containment closure as a defense-in-depth measure can be reestablished quickly to limit releases much lower than assumed in the dose calculations. | |||
ARCB-RAI-35 Enclosure VI, markups to UFSAR Table 15.6-4 remove the words with forced overfill from the steam generator tube rupture. Please explain why this forced overfill was part of the licensing bases and what has changed to allow the change to no longer consider forced overfill. | |||
===Response=== | |||
During the NRC review of the Final Safety Analysis Report (FSAR) for issuance of the operating license, Section 15.4.4, Steam Generator Tube Rupture, of NUREG-0881, Supplement 5, (Reference 13) Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No. 1, specified that the staff will condition the operating license to require satisfactory resolution of this issue before startup following the first refueling outage. This | |||
Westinghouse Non-Proprietary Class 3 Page 107 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment became License Condition 2.C.(11), Steam Generator Tube Rupture (Section 15.4.4, SSER #5) which required Kansas Gas and Electric to submit for NRC review and approval an analysis that demonstrated that the steam generator tube rupture (SGTR) analysis presented in the FSAR is the most severe case with respect to release of fission products and calculated doses. Letter SLNRC 86-1 (Reference 14) dated January 8, 1986, provided the required SGTR analysis. On November 12, 1986, the NRC issued a Request for Additional Information (RAI) (Reference 15) that raised concerns with the margin of 271 ft3 (~5%) to steam generator overfill, which was originally calculated in the required SGTR analysis submitted by letter SLNRC 86-1. The RAI question stated, in part: | |||
The staff believes that, based on this response, there is sufficient uncertainty in the break flow calculations, as well as in operator action times (see Enclosure 2) and interval between safety injection (SI) termination and break flow termination (see question 2 below,) to warrant assuming that safety valve (SV) liquid relief can occur. . Therefore please perform an analysis which assumes a design basis SGTR with loss of offsite power (LOOP), closure of the ruptured SG main steam isolation valve (MSIV), and SG overfill resulting in liquid relief through one SV. | |||
In response to the RAI, WCNOC submitted letter WM 87-0145 (Reference 16) dated May 15, 1987, providing the results of a plant specific analysis that made changes to the modeling and assumptions for the SGTR analysis that would result in steam generator overfill and subsequent liquid relief through one safety valve. Subsequently, the NRC issued a safety evaluation (Reference 17) on May 7, 1991 that concluded that steam generator overfill does occur during a design basis SGTR accident and found the WCNOC SGTR accident analysis to be acceptable. | |||
As part of the transition to the Westinghouse core design and safety analysis methodologies, the SGTR event was analyzed to demonstrate margin to overfill consistent with the Westinghouse standard analysis methodology. This analysis is presented in Section 2.7.2 of Enclosure I of this amendment request. The margin to overfill analysis is consistent with the methodology developed and approved in WCAP-10698-P-A (Reference 18) by Westinghouse. | |||
During the NRC review of the WCAP-10698-P-A methodology, the NRC indicated that an evaluation of the consequences of steam generator overfill would be desirable to demonstrate that the consequences of overfill would be acceptable. In response to this request, WCAP-11002 was submitted to the NRC (Reference 19). The NRC approval of WCAP-10698-P-A included an evaluation of WCAP-11002. The NRC approved methodology in WCAP-10698-P-A does not define a minimum margin to overfill requirement and does not require consideration of a forced overfill scenario. | |||
Westinghouse Non-Proprietary Class 3 Page 108 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-36 The column labeled Comments of Enclosure VI, Table D for Regulatory Position 5.6 discusses WCAP-13247 and an NRC staff response dated March 10, 1993. The NRC staff is unable to locate these documents. Please provide these two documents. | |||
===Response=== | |||
WCAP-13247, Report on the Methodology for the Resolution of the Steam Generator Tube Uncovery Issue, March 1992, was transmitted to the NRC by Westinghouse Owners Group letter OG-92-25 on March 31, 1992, with the accompanying affidavit CAW-92-287. | |||
WCAP-13247, OG-92-25, CAW-92-287, and the NRC response to the Westinghouse Owners Group dated March 10, 1993 (which was an attachment to the Westinghouse Owners Group letter WOG-93-066) are included in Enclosure VIII of this LAR. | |||
ARCB-RAI-37: | |||
Please provide a complete description of how the dose equivalent iodine-131 (DE) and the dose equivalent Xenon-133 are calculated. | |||
===Response=== | |||
The RCS specific activity concentrations for each isotope is based on the assumed fuel defect level (1%), which is multiplied by the ratio of its dose conversion factor to the I-131 dose conversion factor for iodines and Xe-133 dose conversion factor for noble gases to obtain its I-131 or Xe-133 equivalence. The dose conversion factors used are the thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11 for iodines and the effective dose conversion factors from Table III.1 of EPA Federal Guidance Report No. 12, as identified in the revised Definitions for DOSE EQUIVALENT I-131 and revised DOSE EQUIVALENT XE-133 that are included in the Technical Specification markups contained in Enclosure IV, Section 9. The total of the isotopic dose equivalencies is the total I-131 or Xe-133 equivalence corresponding to the design fuel defect level of 1%. Each of the isotopic concentration values at the design defect level is then divided by the total I-131 or Xe-133 dose equivalence to obtain the isotopic concentrations at the applicable Technical Specification 3.4.16, RCS Specific Activity, limits of | |||
<1.0 µCi/gm for DE I-131 and <500 µCi/gm for DE Xe-133. | |||
Note that I-130, Kr-85, and Xe-131m are not listed in the Technical Specification Definitions for DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133. These nuclides do not contribute to the total of the isotopic dose equivalencies; however, their fuel defect level activities are divided by the total I-131 of Xe-133 dose equivalencies and scaled to the Technical Specification limits for DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133. | |||
Westinghouse Non-Proprietary Class 3 Page 109 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-38 SRP 16.0 states, in part, that In TS change requests for facilities with TS based on previous STS [Standard Technical Specifications], licensees should comply with comparable provisions in these STS NUREGs to the extent possible or justify deviations from the STS. | |||
Several of the proposed changes to the TSs and TS bases (an example is provided below) do not align with the STS. Please provide a justification for deviations from the STS. | |||
For example, proposed page B 3.6.3-2 removes detail from the bases of TS 3.6.3, Containment Isolation Valves, which is inconsistent with the STS, NUREG 1431, and Revision 4. The detail provides the limiting total response time for the isolation of containment. Please provide the value used in the DBA analyses assumed for containment isolation. | |||
===Response=== | |||
As noted in the above question, SRP Section 16.0, Technical Specifications, Rev. 3, Subsection II., Acceptance Criteria, states, in part: | |||
In TS change requests for facilities with TS based on previous [emphasis added] STS, licensees should comply with comparable provisions in these STS NUREGs to the extent possible or justify deviations from the STS. | |||
This subsection further states: | |||
Acceptable justifications for deviation would include retention of existing TS requirements, non-adoption of Standard Technical Specifications (STS) requirements not represented in existing TS (e.g., an LCO in STS but not in existing TS), editorial preference, facility design, and a technically justified alternative presentation equivalent to the STS intent. In some cases, comparison to the previous STS may help evaluate the proposed changes by clarifying the TS intent. The previous [emphasis added] STS NUREGs are as follows: | |||
NUREG-0103, STS, Babcock and Wilcox Plants NUREG-0452, STS, Westinghouse Plants NUREG-0212, STS, Combustion Engineering Plants NUREG-0123, STS, General Electric Plants | |||
Westinghouse Non-Proprietary Class 3 Page 110 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment From the above, the quoted subsection specifies the TS change requests based on previous STS (i.e., NUREG-0452, not NUREG-1431) should comply with comparable provisions in these STS NUREGs (i.e., NUREG-1431) to the extent possible or justify deviations from the STS. The current WCGS TSs are based on NUREG-1431, Revision 1. WCNOC submitted a license amendment request on May 15, 1997 (Reference 6) that provided a conversion of the WCGS TSs that were based on NUREG-0452 to TSs that were based on NUREG-1431, Revision 1. | |||
The improved TSs were issued with the issuance of Amendment No. 123 on March 31, 1999 (Reference 7). The process for converting to the improved TSs requires justifications for deviations from the STS. | |||
Subsection III of SRP 16.0, states, in part: | |||
The TS bases are not parts of the Operating License (OL) or Current Operating License (COL); however, the reference TS bases have a wealth of information on safety limit or LCO purposes and action and surveillance requirements. Particularly important is the bases description of how the system or parameter addressed by the LCO satisfies 10 CFR 50.36(c)(2)(ii). Regardless of any technical differences with the reference TS bases, the TS review should verify whether the plant-specific TS bases are consistent with the FSAR plant-specific accident analysis and system description. In addition, the TS bases should describe the basis for each TS requirement accurately. | |||
Nuclear Reactor Regulation (NRR) Office Instruction No. LIC-100, Revision 1, Control of Licensing Bases for Operating Reactors, Section 3.2, Technical Specification Bases Section, states (page 3.10), in part: | |||
If the licensee and staff decide to process TS Bases pages, even though NRC review and approval is not required [emphasis added], the staff may issue the revised pages via a letter to the licensee (or as part of an amendment). | |||
Section 2 of Enclosure VI of Reference 1 provided a description of the changes to the TSs and an associated justification for the change. The changes were further supported by the technical analysis in Section 4 of Enclosure VI of Reference 1. | |||
Regarding the specific changes to TS Bases page B 3.6.3-2, these changes were made for consistency with the AST analyses and to correct a legacy issue associated with the current analyses of record for WCGS. The LOCA and rod ejection dose consequence analyses, which are the only two analyses that model isolation of containment, assumed the containment was isolated at the beginning of event, with the exception of the mini-purge releases for the first 10 seconds of the LOCA event. This is consistent with the current analyses of record for WCGS. | |||
The basis for deleting the wording from the Bases for TS 3.6.3 is that it refers to the containment purge and exhaust valves and not the mini-purge valves. | |||
Westinghouse Non-Proprietary Class 3 Page 111 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 4 RAIs SRXB-RAI-1: | |||
In the Summary of Analysis Codes Utilized in Postulated Accident Analyses, there are two Non-Loss-of-Coolant Accident (LOCA) Safety Analysis Codes listed for the Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, RETRAN, and LOFTRAN. Please provide the information that is obtained from each code and used as the basis for the updated analysis. | |||
===Response=== | |||
The departure from nucleate boiling (DNB) analysis of the rod withdrawal at power (RWAP) event is performed using the RETRAN code, consistent with the approved methodology in WCAP-14882-P-A, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses. RETRAN is also the non-LOCA code used to perform DNB analyses for other Updated Safety Analysis Report (USAR) Chapter 15 events. | |||
The RWAP DNB cases performed consider 10%, 60%, and 100% power cases, at various reactivity insertion rates for minimum and maximum reactivity feedback conditions. Since the Revised Thermal Design Procedure (RTDP) DNB methods were used, initial condition uncertainties are accounted for in the DNB ratio (DNBR) safety analysis limit. As such, the initial conditions are set to their nominal values. A relatively small number of cases are performed in order to determine the most limiting set of conditions (e.g., a sensitivity study to determine conservative treatment of initial condition uncertainties is not required). | |||
For RWAP cases analyzed to address RCS overpressurization concerns, the limiting set of conditions is established through varying a number of parameters (power, conservative treatment of initial condition uncertainties, reactivity insertion rates, etc.) resulting in a large number of cases to be considered. As discussed in WCAP-14882-P-A, the RETRAN and LOFTRAN codes are essentially equivalent. The LOFTRAN code is more suited for running a large number of cases than RETRAN; as such, it is the preferred transient analysis code for addressing this RCS pressure acceptance criterion. The LOFTRAN code (WCAP-7907-P-A, LOFTRAN Code Description) has historically been the approved transient analysis code for non-LOCA analyses performed by Westinghouse and has previously been used to perform RWAP RCS overpressurization analyses for plants that otherwise utilize the RETRAN code for the DNB analyses. As a recent precedent, this approach was utilized for the Turkey Point Extended Power Uprate (Licensing Report Table 2.8.5.0-7: ADAMS Ascension Number ML103560177, and Safety Evaluation Report: ADAMS Ascension Number ML11293A365). | |||
Enclosure I, Section 2.5.2.1.3 of this LAR has been updated to include the above discussion. | |||
Westinghouse Non-Proprietary Class 3 Page 112 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
==References:== | |||
: 1. WCNOC Letter ET 13-0023, License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, August 13, 2013. ADAMS Accession No. ML13247A075. | |||
: 2. Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, Wolf Creek Generating Station - | |||
Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), April 3, 2014. ADAMS Accession No. ML14083A400. | |||
: 3. WCNOC Letter ET 14-0017, Withdrawal of License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, June 18, 2014. ADAMS Accession No. ML14175A119. | |||
: 4. Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, Wolf Creek Generating Station - | |||
Request for Additional Information RE: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), April 30, 2014. ADAMS Accession No. ML14111A100. | |||
: 5. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Wolf Creek Generating Station - Issuance of Amendment Re: Elimination of Requirements for Post Accident Sampling Systems (TAC NO. MB0678), March 2, 2001. ADAMS Accession No. ML010670233. | |||
: 6. WCNOC letter ET 97-0050, Technical Specification Conversion Application, May 15, 1997. ADAMS Accession No. 9705210318 (Public Legacy Library). | |||
: 7. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Conversion to Improved Technical Specification for Wolf Creek Generating Station - Amendment No. 123 to Facility Operating License No. NPF-42 (TAC NO. M98738), March 31, 1999. ADAMS Accession No. ML022050061. | |||
: 8. Letter from A. J. Mendiola, USNRC, to Technical Specifications Task Force, Potential Issues with Plant-Specific Adoption of Travelers TSTF-51, Revision 2, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations, TSTF-286, Revision 2, Operations Involving Positive Reactivity Additions, and TSTF-471, Revision 1, Eliminate Use of Term Core Alterations in Actions and Notes, November 7, 2013. ADAMS Accession No. ML13246A358. | |||
: 9. Not used. | |||
: 10. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Wolf Creek Generating Station - Issuance of Amendment Re: Equipment Hatch Open During Refueling (TAC NO. | |||
MB2599), July 30, 2002. ADAMS Accession No. ML022120428. | |||
Westinghouse Non-Proprietary Class 3 Page 113 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
: 11. Letter from J. C. Stone, USNRC, to N. S. Carns, WCNOC, Wolf Creek Generating Station | |||
- Amendment No. 95 to Facility Operating License No. NPF-42 (TAC NO. M94113), | |||
February 28, 1996. ADAMS Accession No. ML022040718. | |||
: 12. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Wolf Creek Generating Station - Issuance of Amendment Re: Use of Administrative Controls for Open Containment Penetrations During Refueling (TAC NO. MA9293), September 12, 2000. | |||
ADAMS Accession No. ML003750021. | |||
: 13. NUREG-0881, Supplement 5, Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No. 1, March 1985. | |||
: 14. SNUPPS letter SLNRC 86-1, Steam Generator Tube Rupture Analysis - SNUPPS, January 8, 1986. | |||
: 15. Letter from P. W. OConnor, USNRC, to G. L. Koester, Kansas Gas & Electric Company, Request for Additional Information Related to the SNUPPS Steam Generator Tube Rupture Analysis, November 12, 1986. | |||
: 16. WCNOC letter WM 87-0145, Response to RAI Regarding SGTR Analysis, May 15, 1987. | |||
: 17. Letter from D. V. Pickett, USNRC, to B. D. Withers, WCNOC, Safety Evaluation Report for the Wolf Creek Generating Station Steam Generator Tube Rupture Analysis (TAC NO. | |||
57363), May 7, 1991. | |||
: 18. WCAP-10698-P-A, SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, August 1987. | |||
: 19. WCAP-11002 (Proprietary) and WCAP-11003 (Non-Proprietary), Evaluation of Steam Generator Overfill Due to a SGTR Accident, February 1986. | |||
: 20. WCNOC letter ET 14-0003, Response to Request for Additional Information Regarding License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, January 28, 2014. ADAMS Accession No. ML14035A224. | |||
: 21. WCNOC letter WO 14-0031, Response to Request for Additional Information Regarding License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, March 20, 2014. ADAMS Accession No. ML14091A245. | |||
: 22. WCNOC letter WO 14-0032, Supplemental Information for Response to Request for Additional Information Regarding License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, March 26, 2014. ADAMS Accession No. ML14091A261. | |||
Westinghouse Non-Proprietary Class 3 Page 114 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment | |||
: 23. Letter from R. F. Kuntz, USNRC, to C. M. Crane, Exelon Generation Company, Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 - Issuance of Amendment Re: Alternative Source Term (TAC NOS. MC6221, MC6222, MC6223, and MC6224), September 8, 2006. ADAMS Accession No. ML062340420. | |||
: 24. Letter from C. F. Lyon, USNRC, to M. W. Sunseri, WCNOC, Wolf Creek Generating Station - Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), December 13, 2013. ADAMS Accession No. ML13345B335. | |||
: 25. Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, Wolf Creek Generating Station - | |||
Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), March 5, 2014. ADAMS Accession No. ML14058A088. | |||
: 26. Email from Fred Lyon, USNRC, to Steve G Wideman, WCNOC, FW: Wolf Creek Generating Station - Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC No. MF2574), December 23, 2013. ADAMS Accession No. ML13357A250. | |||
: 27. Letter from C. F. Lyon, USNRC, to M. W. Sunseri, WCNOC, WOLF CREEK GENERATING STATION- REQUEST FOR ADDITIONAL INFORMATION RE: | |||
TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSIS (TAC NO. | |||
MF2574), January 28, 2014. ADAMS Accession No. ML14027162.}} | |||
Latest revision as of 19:12, 4 February 2020
| ML17054C229 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/17/2017 |
| From: | Wolf Creek |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17054C103 | List: |
| References | |
| ET 17-0001 LTR-OPB-PMO-SP-MEP-16-017, Rev. 2 - NP | |
| Download: ML17054C229 (114) | |
Text
Westinghouse Non-Proprietary Class 3 Page 1 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Responses to NRC RAIs on August 2013 Methodology Transition LAR Submittal (Non-Proprietary)
November 2016
© 2016 Westinghouse Electric Company LLC All Rights Reserved
Westinghouse Non-Proprietary Class 3 Page 2 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment WCNOC previously submitted a license amendment request (LAR) on August 13, 2013 (Reference 1) requesting approval of changes to the Wolf Creek Generating Station (WCGS)
Technical Specifications (TSs). That application proposed the transition to Westinghouse core design and safety analysis methodologies, full scope implementation of Alternative Source Term (AST), and implementation of instrumentation setpoint and control uncertainty calculations based on the current Westinghouse Setpoint Methodology (including adoption of Option A of Technical Specification Task Force (TSTF) TSTF-493-A, Revision 4). WCNOC subsequently withdrew the LAR on June 18, 2014 (Reference 3) based on deficiencies discovered by WCNOC.
References 2, 4, 24 and 25 are Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAI) related to Reference 1. WCNOC letters ET 14-0003 (Reference 20) and WO 14-0032 (Reference 22) provided responses to the RAIs received in Reference 24. WCNOC letter WO 14-0031 (Reference 21) provided responses to the RAIs received in Reference 25.
WCNOC withdrew the LAR prior to providing a response to References 2 and 4. WCNOC has subsequently developed responses to the RAIs received in References 2 and 4. The responses are provided within this enclosure and the applicable information has been incorporated into the current LAR.
Following the withdrawal of the LAR, two changes were made that affected the responses contained within References 20, 21, and 22. First, WCGS implemented the AST sump pH analysis (refer to Section 4.4 of Enclosure IV) as its licensing basis analysis under the 50.59 process. Second, it was determined that a modification would be implemented to support full implementation of AST. Specifically, the operator action to isolate the failed CREVS train (Discussed in Section 15A.3.1 of the USAR) has historically required local operator action.
However, in order to reduce the difficulty of completing this time critical action and to eliminate dose exposure to a local operator, a design modification will be implemented to supply the CREVS control room isolation dampers with Class IE 120 VAC battery power. This modification will be implemented under the 50.59 process. Based upon these two changes, responses previously transmitted in References 20, 21, and 22 have been updated and are included within this enclosure (in addition to the unchanged responses previously provided). Specifically, the changes to the affected responses (EMCB-RAI-1, ESGB-RAI-1, and EEEB-RAI-2) are shown in bold.
The responses to the RAIs are applicable to the current LAR and are therefore included in this Enclosure. Enclosure VIII provides supplemental documents that were requested by ARCB-RAI-33 and ARCB-RAI-36. The specific NRC request is provided in italics.
Note that Reference 27 also provided NRC RAIs related to Reference 1. These RAIs are associated with the original proposed addition of TSTF-493 and, as such, no longer require a response.
Westinghouse Non-Proprietary Class 3 Page 3 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 24 RAIs as clarified by Reference 26 EMCB-RAI-1 Implementation of an Alternate Source Term (AST) in accordance with Section 50.67, Accident source term, of Title 10 of the Code of Federal Regulations (10 CFR) could affect structures, systems, and components (SSCs) which were not previously evaluated for the consequences of a design-basis accident (DBA) and, as such, may not be seismically qualified. Appendix A to 10 CFR Part 100, Reactor Site Criteria, requires that SSCs necessary to assure the capability of the plant to mitigate the consequences of accidents, which could result in exposures comparable to the guideline exposures provided in 10 CFR Part 100, be designed to remain functional during and after a safe shutdown earthquake. In accordance with the 10 CFR Part 100 requirements, please identify SSCs which may be affected by the implementation of the proposed AST and address the following:
a) Indicate whether any non-safety-related SSCs are being credited in the proposed AST license amendment.
b) For any nonsafety-related SSCs credited in the AST, confirm that the SSCs have been seismically qualified in accordance with the plant licensing basis.
c) Indicate whether the SSCs are new or existing.
d) Describe the location of the SSCs and the seismic qualification method employed to demonstrate the seismic ruggedness of these SSCs, such as the plant licensing basis or an NRC-endorsed industry standard.
e) Summarize the results of the seismic qualification of the equipment, indicating whether any modifications or re-design will be necessary in support of the AST.
Response
The proposed license amendment request included the adoption of Alternative Source Term (AST) radiological analysis methodology consistent with 10 CFR 50.67, Accident source term. The design basis radiological consequence analyses performed for AST did not credit any non-safety related structures, systems, and components (SSCs). This is consistent with the current licensing basis analysis of record that does not credit any non-safety related SSCs. The design basis radiological consequences analyses performed for AST do not require the installation of any new SSCs (the damper modification being implemented will only change the damper power supply; it will not add new components). New cable trays or conduit may need to be installed to support the one modification being implemented, depending on cable tray and conduit availability. If the modification requires installation of a new cable tray or conduit for the power cables, the associated seismic evaluations will be performed in accordance with Wolf Creek methods used for existing cable trays and conduit.
Westinghouse Non-Proprietary Class 3 Page 4 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ESGB-RAI-1 Describe the analysis methodology used to determine the pH in the sump water during the period of 30 days post-LOCA. Include detailed calculations of time dependent pH values in the sump during a 30 day period post-LOCA to demonstrate that the pH remains basic throughout this time period.
Response
As discussed in Section 4.4.2 of Enclosure VI to Reference 1, the containment sump pH calculation did not include consideration of acid generation (nitric acid produced by the irradiation of water and air or hydrochloric acid produced by the radiolysis of chlorine bearing materials) as they were considered secondary effects. In Section 4.4.2 of Enclosure VI to Reference 1, WCNOC indicated that it was expected that the effect of acid generation on sump pH would decrease the pH value less than 0.1 pH based on a comparison of Byron Station information and the conclusion reached by the NRC in the safety evaluation for Amendment No.
147 for the Byron Station and Amendment No. 140 for the Braidwood Station (Reference 23).
WCNOC letter WO 14-0032 (Reference 22) provided the below details and results of the calculation.
Methodology In order to calculate the minimum sump pH, the maximum amount of boric acid from the various sources of borated water that enter the containment sump and the strong acids that are generated from the effects of radiolysis post-LOCA combined with a minimum amount of caustic from the spray additive tank (SAT) will yield a minimum pH value. The concentrations of these substances are used to compute the value of the sump pH as a function of time using verified titration curve data for aqueous solutions of boric acid and sodium hydroxide.
The current licensing basis (CLB) for the WCGS sump pH calculation (EN-03-W, Rev. 2, 18 Months Fuel Cycle, Cycle 4 Specific Boron - pH Calculations for TSA 20038-001, Rev. 3) forms the basis of existing Updated Safety Analysis Report (USAR) Figure 6.5-5. Two scenarios are analyzed: both containment spray trains operating with one NaOH eductor in service, and both containment spray trains operating with both NaOH eductors in service.
Westinghouse Non-Proprietary Class 3 Page 5 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment No computer codes were used in the CLB calculation. To determine the pH in this calculation, NaOH and H3BO3 molarities were first calculated and then used with Oak Ridge National Laboratory (ORNL) titration curve data to determine the pH of the spray and sump solutions during the injection and recirculation phases of Emergency Core Cooling System (ECCS) operation. The CLB calculation did not consider the effects of the generation of strong acids inside containment post-LOCA.
Westinghouse has subsequently performed an independent calculation (CN-SEE-I-13-10, Rev. 1, Wolf Creek Sump pH,) to update the conclusions of the CLB calculation for the transition to the Westinghouse Core Design and Safety Analysis methodologies. The Westinghouse calculation used pH data for boric acid/sodium hydroxide solutions from verified titration curve data for aqueous solutions of boric acid and sodium hydroxide. A revision of this calculation was performed in response to this RAI question (ESGB-RAI-1) and the effect of strong acid generation on the post-LOCA sump pH was considered. In accordance with the guidance of NUREG/CR-5950, Iodine Evolution and pH Control, the mass of nitric acid that is generated by the radiolysis of air and water inside containment and the mass of hydrochloric acid that is generated by the radiolysis of Hypalon electrical cable insulation inside containment were calculated. The amount of hydriodic acid that is generated from iodine released inside containment was judged to be negligible and was not considered in the sump pH calculation.
The masses (moles) of the strong acids that are generated inside containment post-LOCA were assumed to instantaneously be neutralized by a molar equivalent of NaOH, which reduced the net mass of NaOH that was available to neutralize the boric acid injected from the Reactor Coolant System (RCS), accumulators and refueling water storage tank (RWST). The net effect of the neutralization of all acid species in the sump by the NaOH from the SAT determined the equilibrium sump pH.
The CLB has been updated since the time the RAIs were sent and thus the current USAR figure is aligned with the AST analysis.
Assumptions and Inputs Among the various assumptions that were made in the sump pH calculation is that of effective mixing of the water inventory in the sump. This is a valid assumption because of the long term operation of the Containment Spray System and the uniform dispersion of spray over the containment cross section. With respect to the generation of strong acids inside containment, the gaseous hydrochloric acid produced by electric cable insulation radiolysis was conservatively assumed to be instantly dissolved in the sump water, and the hydrochloric acid generated from the radiolysis of cable insulation and nitric acid generated by the irradiation of containment sump water and air was assumed to fully dissociate in the sump water. At the completion of the injection and mixing of all chemical species (acids and sodium hydroxide) inside containment post-LOCA, it was assumed that there would be no further change in the
Westinghouse Non-Proprietary Class 3 Page 6 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment sump liquid inventory and, consequently, that there would be no further change in the calculated sump pH endpoint over the 30 day post-LOCA period.
The minimum long-term sump pH in the revised Westinghouse calculation uses the following conservative bases:
- 7. Maximum RWST and safety injection (SI) accumulator Technical Specification 3.5.1, Accumulators, and 3.5.4, Refueling Water Storage Tank (RWST), boron concentration of 2500 ppm.
- 8. A conservatively high RCS boron concentration of 1900 ppm in the CLB, and 1980 ppm in the Westinghouse calculation. Given that the RCS is only approximately 15% of the sump fluid, this difference in boron concentration is a small effect.
- 9. The SAT is assumed to contain the Technical Specification 3.6.7, Spray Additive System, minimum 28% by weight NaOH solution.
- 10. The mass of Hypalon electrical cable insulation inside containment is 50,000 lbm.
- 11. The 30-day integrated containment doses inside containment (beta and gamma sources) to which the Hypalon insulation is exposed post-LOCA were derived from the plant-specific report for the Environmental Qualification of Safety-Related Electrical Equipment.
- 12. The maximum sump liquid mass is approximately 4.1x106 lbm, which corresponds to a liquid volume of 1.87x106 liters.
While the Technical Specification 3.6.7 SAT minimum contained volume is 4340 gallons, the CLB analysis assumed a conservative minimum delivered volume of only 2960 gallons. The Westinghouse calculation used an adjusted delivered volume of 2752 gallons, which accounted for the neutralization of the strong acid generated inside containment post-LOCA.
Results and Conclusions The Westinghouse calculation determined a long-term sump pH of 8.7, while the CLB calculation determined a minimum value of 8.6. The slightly higher pH in the Westinghouse calculation is attributed to the use of a different boron/NaOH/pH correlation. In either case, the results show that the sump pH remains above the CLB value of 8.5 and well above the NRC-required value of 7.0 after the minimum amount of NaOH is injected into the sump.
In the CLB calculation, the limiting system alignment of both containment spray trains operating with one NaOH eductor in service results in a sump pH of 7 in approximately 15 minutes, with the final long-term pH of 8.6 in approximately 80 minutes. The less limiting case of both Containment Spray trains operating with both NaOH supplies in service results in a sump pH of 7 in approximately 11 minutes, with the final long-term pH of 8.6 in approximately 45 minutes.
Westinghouse Non-Proprietary Class 3 Page 7 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment For the Westinghouse calculation, the limiting system alignment of both containment spray trains operating with one NaOH eductor in service results in a sump pH of 7 in approximately 11 minutes, with the final long-term pH of 8.7 in approximately 70 minutes. The less limiting case of both containment spray trains operating with both NaOH supplies in service results in a sump pH of 7 in approximately 9 minutes, with the final long-term pH of 8.7 in approximately 35 minutes.
The figure below from the Wolf Creek USAR shows that the acceptance criteria of sump pH >
8.5 (minimum) is met for one or two containment spray eductors in service, considering also the generation of strong acids inside containment post-LOCA, consistent with the guidance of NUREG/CR-5950.
Westinghouse Non-Proprietary Class 3 Page 8 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 25 RAIs EEEB-RAI-1 Please confirm whether any loads are being added to the WCGS emergency diesel generators (EDGs). If so, describe their impact on the capability and capacity of the EDGs. Also, describe changes, if any, being made to the EDG loading sequence to support this license amendment request (LAR).
Response
The proposed license amendment request included the adoption of Alternative Source Term (AST) radiological analysis methodology consistent with 10 CFR 50.67, Accident source term.
The design basis radiological consequence analyses performed for AST did not result in the addition of loads to the onsite standby power source (train A and train B diesel generators (DGs)) and did not result in changes to the DG loading sequence.
EEEB-RAI-2 Please confirm whether any components are being added to the Environmental Qualification (EQ) equipment list to comply with Title 10 of the Code of Federal Regulations, Section 50.49 (10 CFR 50.49) due to this LAR. If components are being added, describe the equipment qualification for the environmental conditions to which the components are expected to be exposed.
Response
The design basis radiological consequences analyses performed for AST do not require the installation of any new SSCs. The one modification being implemented to support full implementation of AST will not add any components to the EQ list (the damper modification being implemented will only change the damper power supply; it will not add new components). As such, no components are added to the Environmental Qualification equipment list to comply with 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants.
Westinghouse Non-Proprietary Class 3 Page 9 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 2 RAIs ARCB-RAI-1 Please justify all changes from the current licensing basis (see Issue 1 of NRC Regulatory Issue Summary 2006-04, Experience with Implementation of Alternative Source Terms, dated March 7, 2006 (ADAMS Accession No. ML053460347), for more detail). No justification is needed for changes that are consistent with Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ADAMS Accession No. ML003716792), or are provided in the submittal dated August 13, 2013 (ADAMS Accession No. ML13247A076), unless requested by these RAIs.
Response
NRC Regulatory Issue Summary 2006-004, Experience with Implementation of Alternative Source Terms, indicates that the AST amendment request should (1) provide justification for each individual proposed change to the technical specifications (TS), (2) identify and justify each change to the licensing basis accident analyses, and (3) contain enough details (e.g., assumptions, computer analyses input and output) to allow the NRC staff to confirm the dose analyses results in independent calculations. Section 1 of Enclosure VI of Reference 1 identified the changes to the WCGS licensing basis as:
The Control Room Habitability Envelope (CRHE) unfiltered inleakage is revised from 20 scfm to 50 scfm.
The Control Building unfiltered inleakage is revised from 300 scfm to 400 scfm.
Revise the USAR Chapter 15 dose analyses in accordance with the guidance in Regulatory Guide 1.183.
Revise the Technical Specification (TS) to address the update of the accident source term and associated design basis accidents utilizing the guidance provided in Regulatory Guide 1.183 and the associated control room dose limits of General Design Criterion (GDC) 19, and offsite dose limits of 10 CFR 50.67.
In addition, a detailed comparison of the analyses parameters for the AST versus the current licensing basis is included in Enclosure IV of this LAR, Section 4.3, Tables 4.3-5 through 4.3-16.
These tables have been included below to facilitate the review. For the resubmittal of the LAR, the justification for the proposed changes to the TSs and the control room and control building unfiltered inleakages is justified in Section 2 of Enclosure IV.
Westinghouse Non-Proprietary Class 3 Page 10 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Doses Analyses Parameters Summary Tables (excerpt from Enclosure IV)
Table 4.3-5 Control Room and Control Building Parameters AST CLB Reason for Change Control room volume (ft3) 100,000 100,000 No change Control building volume (ft3) 239,000 239,000 No change Normal ventilation flow rates (cfm)
Unfiltered makeup flow rate from 13,050 13,050 No change environment to control building Unfiltered makeup flow rate from 1950 1950 No change environment to control room Unfiltered inleakage to control room 50 10 Increased to allow for additional surveillance testing margin Emergency mode of operation flow rates prior to operator action (cfm)
Filtered makeup flow rate from 1350 1350 No change environment to control building Filtered makeup flow rate from control 550 550 No change building to control room Unfiltered makeup flow rate from 400 300 Increased to allow for environment to control building additional surveillance testing margin Unfiltered makeup flow rate from control 550 550 No change building to control room Unfiltered inleakage to control room 50 20 Increased to allow for additional surveillance testing margin Filtered control room recirculation flow 1250 1250 No change Emergency mode of operation flow rates following operator action (cfm)
Filtered makeup flow rate from 675 675 No change environment to control building Filtered makeup flow rate from control 550 550 No change building to control room Unfiltered makeup flow rate from 400 300 Increased to allow for environment to control building additional surveillance testing margin
Westinghouse Non-Proprietary Class 3 Page 11 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-5 Control Room and Control Building Parameters (cont.)
AST CLB Reason for Change Unfiltered makeup flow rate from control 0 0 No change building to control room Unfiltered inleakage to control room 50 20 Increased to allow for additional surveillance testing margin Filtered control room recirculation flow 1250 1250 No change Operator action time to terminate failed train 90 90 No change of filtered makeup flow from start of event (minutes)
Filter efficiencies (%)
Elemental iodine 95 95 No change Organic iodine 95 95 No change Particulates 95 95 No change Isolation setpoint for control room air supply 2.12E-03 N/A Change due to modeling of CR for all radiation monitors (GKRE0004 and accidents GKRE0005) (µCi/cc Xe-133) (not AST specific)
Delay to switch to emergency mode of 60 N/A Change due to operation following receipt of isolation signal modeling of CR for all (seconds) accidents (not AST specific)
Control room breathing rate for duration of the 3.5E-04 3.47E-04 Regulatory Guide event (m3/sec) update (not AST specific)
Control room occupancy factors 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 1.0 No change 1 - 4 days 0.6 0.6 No change 4 - 30 days 0.4 0.4 No change
Westinghouse Non-Proprietary Class 3 Page 12 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis AST CLB Reason for Change RCS activity See Table 4.3-1a See Table 4.3-1b Updated calculations Initial secondary system activity See Table 4.3-1a See Table 4.3-1b Updated calculations Pre-accident iodine spike factor 60 60 No change Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 75 Modeling update (not AST specific)
Letdown flow decontamination 100 N/A Modeling update
(%) (not AST specific)
RCS leakage (gpm) 11 1 Modeling update (not AST specific)
Spike factor 500 500 No change Duration of accident-initiated 8 8 No change iodine spike (hr)
RCS mass, maximum (lbm) 8.42E+05 4.94E+05 Modeling update (not AST specific)
Equilibrium appearance rates (Ci/min)
I-130 9.87E-03 N/A Modeling update (not AST specific)
I-131 4.39E-01 N/A Modeling update (not AST specific)
I-132 1.98E+00 N/A Modeling update (not AST specific)
I-133 8.93E-01 N/A Modeling update (not AST specific)
I-134 9.65E-01 N/A Modeling update (not AST specific)
I-135 8.17E-01 N/A Modeling update (not AST specific)
Westinghouse Non-Proprietary Class 3 Page 13 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis (cont.)
AST CLB Reason for Change Iodine chemical form of releases (%)
Elemental 97 N/A Regulatory Guide Update (not AST specific)
Organic 3 N/A Regulatory Guide Update (not AST specific)
Particulate 0 N/A Regulatory Guide Update (not AST specific)
Approximate timing of events Safety injection (SI) signal 30 N/A Modeling update (sec) (not AST specific)
Control room isolation 90 N/A Modeling update (including delay) (sec) (not AST specific)
Faulted SG releases all initial 2 N/A Modeling update activity (min) (not AST specific)
RHR cooling takes over 12 N/A Modeling update (releases from intact SGs (not AST specific) terminated) (hr)
RCS cooled below 212°F 34 N/A Modeling update (releases from faulted SG (not AST specific) terminated) (hr)
Mass transfer data Initial faulted SG release 165,000 164,500 Updated calculations (in first 2 minutes) (lbm)
Total primary-to-secondary leakage Leakage through faulted 1 1 No change SG to atmosphere (gpm)
Leakage into intact SGs 450 N/A Modeling update (gpd, total) (not AST specific)
Steam Released from Intact SGs to Atmosphere 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 419,340 404,452 Updated calculations 2 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (lbm) 1,310,269 945,973 Updated calculations
Westinghouse Non-Proprietary Class 3 Page 14 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis (cont.)
AST CLB Reason for Change RCS Mass, Minimum (lbm) 3.99E+05 4.94E+05 Input change (minimum CLB does not vs nominal) use maximum and minimum Faulted SG Mass, Maximum (lbm) 1.65E+05 164,500 Updated calculations Intact SGs Mass, Minimum 2.47E+05 286,500 Updated calculations (lbm, Total)
SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Control room atmospheric dispersion factors (sec/m3)
Intact SGs 0 - 0.025 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> 2.55E-02 N/A Not modeled in current AORs 0.025 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.39E-04 N/A Not modeled in current AORs Faulted SG 0 - 0.025 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> 2.11E-03 N/A Not modeled in current AORs 0.025 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.12E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.38E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.79E-04 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.14E-04 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 15 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-6 Assumptions Used for Main Steamline Break Analysis (cont.)
AST CLB Reason for Change 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 8.94E-05 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3)
Intact SGs 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 4.89E-05 N/A Not modeled in current AORs Faulted SGs 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.80E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.80E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 6.44E-05 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 4.42E-05 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 3.22E-05 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 16 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-7 Assumptions Used for Loss of Non-Emergency AC Power Analysis Reason for AST CLB Change RCS activity See Table 4.3-1a See Table 4.3-1b Updated calculations Initial secondary system activity See Table 4.3-1a See Table 4.3-1b Updated calculations Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 75 Modeling update (not AST specific)
Letdown flow decontamination (%) 100 N/A Modeling update (not AST specific)
RCS leakage (gpm) 11 1 Modeling update (not AST specific)
Spike factor 500 N/A Modeling update (not AST specific)
Duration of accident-initiated iodine 8 N/A Modeling update spike (hr) (not AST specific)
RCS mass, maximum (lbm) 8.42E+05 4.94E+05 Input change CLB does not use (maximum vs max and min nominal)
Equilibrium appearance rates (Ci/min)
I-130 9.87E-03 N/A Modeling update (not AST specific)
I-131 4.39E-01 N/A Modeling update (not AST specific)
I-132 1.98E+00 N/A Modeling update (not AST specific)
I-133 8.93E-01 N/A Modeling update (not AST specific)
I-134 9.65E-01 N/A Modeling update (not AST specific)
I-135 8.17E-01 N/A Modeling update (not AST specific)
Westinghouse Non-Proprietary Class 3 Page 17 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-7 Assumptions Used for Loss of Non-Emergency AC Power Analysis (cont.)
Reason for AST CLB Change Iodine chemical form of releases (%)
Elemental 97 N/A Regulatory Guide Update (not AST specific)
Organic 3 N/A Regulatory Guide Update (not AST specific)
Particulate 0 N/A Regulatory Guide Update (not AST specific)
Time RHR cooling matched decay heat 12 8 Updated (SG releases terminated) (hr) calculations Mass transfer data Total primary-to-secondary leakage 1 1 No change (gpm)
Steam released from SGs to atmosphere 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 419,846 549,000 Updated calculations 2 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (lbm) 1,352,918 1,030,000 Updated calculations RCS mass, minimum (lbm) 3.99E+05 4.94E+05 Input change CLB does not (minimum vs use max and min nominal)
Plant total SG mass, minimum (lbm)
Until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 3.30E+05 382,000 Updated calculations After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 4.85E+05 382,000 Modeling update (not AST specific)
SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Control room isolation None N/A No change
Westinghouse Non-Proprietary Class 3 Page 18 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-7 Assumptions Used for Loss of Non-Emergency AC Power Analysis (cont.)
Reason for AST CLB Change Control room atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.55E-2 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.84E-2 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.46E-3 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 4.72E-3 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 3.43E-3 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 4.89E-05 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 19 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-8 Assumptions Used for Locked Rotor Analysis AST CLB Reason for Change Core activity See Table 4.3-1a See Table 4.3-1b Updated calculations Failed fuel (% of Core) 5 5 No change Melted fuel (% of Core) 0 0 No change Peaking factor 1.65 1.65 No change Gap fractions I-131 0.08 0.12 Modeling change for AST Kr-85 0.10 0.30 Modeling change for AST Other iodines and noble gases 0.05 0.10 Modeling change for AST Iodine chemical form of releases (%)
Elemental 97 N/A Regulatory Guide Update (not AST specific)
Organic 3 N/A Regulatory Guide Update (not AST specific)
Particulate 0 N/A Regulatory Guide Update (not AST specific)
Time RHR cooling matched decay 12 8 Updated calculations heat (SG releases terminated) (hr)
Mass transfer data Total primary-to-secondary 1 1 No change leakage (gpm)
Steam released from SGs to atmosphere 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 419,846 5.49E+05 Updated calculations 2 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (lbm) 1,352,918 1.03E+06 Updated calculations RCS mass, minimum (lbm) 3.99E+05 4.94E+05 Input change (minimum vs nominal)
Plant total SG mass, minimum (lbm)
Until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 3.30E+05 3.82E+05 Updated calculations After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 4.85E+05 3.82E+05 Updated calculations
Westinghouse Non-Proprietary Class 3 Page 20 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-8 Assumptions Used for Locked Rotor Analysis (cont.)
AST CLB Reason for Change SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Time of Control room isolation 120 N/A No change (including delays) (sec)
Control room atmospheric dispersion factors (sec/m3)
HVAC Flows Except for Control Room Unfiltered Inleakage 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 2.55E-02 N/A Not modeled in current AORs 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.39E-04 N/A Not modeled in current AORs Control Room Unfiltered Inleakage 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.55E-02 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.84E-02 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.46E-03 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 4.72E-03 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 3.43E-03 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 21 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-8 Assumptions Used for Locked Rotor Analysis (cont.)
AST CLB Reason for Change TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 4.89E-05 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 22 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis Reason for AST CLB Change Core activity See Table 4.3-1a See Table 4.3-1b Updated calculations Failed fuel (% of core) 10 10 No change Melted fuel (% of core) 0.25 0.25 No change Peaking factor 1.65 1.65 No change Gap fractions Iodines and noble gases 0.10 0.10 No change Alkali metals 0.12 N/A Modeling change for AST Containment Leakage Activity released to containment from failed fuel (%)
Iodines and noble gases 10 10 No change Alkali metals 12 N/A Modeling change for AST Activity released to containment from melted fuel (%)
Iodines and alkali metals 50 50 No change Noble gas 100 100 No change Iodine chemical form of releases (%)
Elemental 4.85 91 Modeling change for AST Particulate 95 5 Modeling change for AST Organic 0.15 4 Modeling change for AST Containment leak rates (weight %/day) 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.2 0.2 No change 1 - 30 days 0.1 0.1 No change 3
Containment volume (ft ) 2.5E+06 2.5E+06 No change Removal of airborne activity in None None No change containment (other than leakage or decay)
Westinghouse Non-Proprietary Class 3 Page 23 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.)
Reason for AST CLB Change SI signal (sec) 150 N/A Modeling update (not AST specific)
Time of control room isolation (including 210 N/A Modeling update delays) (sec) (not AST specific)
Control room atmospheric dispersion factors (sec/m3) 0 - 0.0583 hours0.00675 days <br />0.162 hours <br />9.63955e-4 weeks <br />2.218315e-4 months <br /> 2.11E-03 N/A Not modeled in current AORs 0.0583 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.12E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.38E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.81E-04 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.29E-04 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 9.65E-05 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.91E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.66E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.62E-05 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 7.05E-05 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 5.52E-05 N/A Not modeled in current AORs Primary-to-Secondary Leakage Activity released to RCS from failed fuel
(%)
Iodines and noble gases 10 10 No change Alkali metals 12 N/A Modeling change for AST
Westinghouse Non-Proprietary Class 3 Page 24 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.)
Reason for AST CLB Change Activity released to RCS from melted fuel (%)
Iodines and alkali metals 50 50 No change Noble gas 100 100 No change Iodine chemical form of releases (%)
Elemental 97 N/A Regulatory Guide Update (not AST specific)
Organic 3 N/A Regulatory Guide Update (not AST specific)
Particulate 0 N/A Regulatory Guide Update (not AST specific)
Time RHR cooling matched decay heat 12 8 Updated (SG releases terminated) (hr) calculations Mass transfer data Total primary-to-secondary leakage 1 1 No change (gpm)
Steam released from SGs to atmosphere 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 419,846 48,600 (140 sec) Updated calculations 2 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (lbm) 1,352,918 N/A Updated calculations RCS mass, minimum (lbm) 3.99E+05 4.94E+05 Input change (minimum vs nominal)
Plant total SG mass, minimum (lbm)
Until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 3.30E+05 4.16E+05 Updated calculations After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (lbm) 4.85E+05 4.16E+05 Updated calculations
Westinghouse Non-Proprietary Class 3 Page 25 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.)
Reason for AST CLB Change SG iodine water/steam partition 100 100 No change coefficient Moisture carryover (%) 0.25 0.25 No change Time of Control room isolation (including 120 N/A No change delays) (sec)
Control room atmospheric dispersion factors (sec/m3)
HVAC Flows Except for Control Room Unfiltered Inleakage 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 2.55E-02 N/A Not modeled in current AORs 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.39E-04 N/A Not modeled in current AORs Control Room Unfiltered Inleakage 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.55E-02 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.84E-02 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.46E-03 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 4.72E-03 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 3.43E-03 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.83E-04 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 26 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-9 Assumptions Used for Rod Ejection Analysis (cont.)
Reason for AST CLB Change 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 4.89E-05 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 27 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-10 Assumptions Used for Letdown Line Break Analysis AST CLB Reason for Change RCS activity See Table 4.3-1a See Table 4.3-1b Updated calculations Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 195 Modeling update (not AST specific)
Letdown flow decontamination 100 N/A Modeling update
(%) (not AST specific)
RCS leakage (gpm) 11 N/A Modeling update (not AST specific)
Spike factor 500 N/A Modeling update (not AST specific)
Duration of accident-initiated 8 N/A Modeling update iodine spike (hr) (not AST specific)
Reactor coolant mass, maximum 8.42E+05 4.94E+05 Input change (lbm) (maximum vs nominal)
Equilibrium appearance rates (Ci/min)
I-130 9.87E-03 N/A Modeling update (not AST specific)
I-131 4.39E-01 N/A Modeling update (not AST specific)
I-132 1.98E+00 N/A Modeling update (not AST specific)
I-133 8.93E-01 N/A Modeling update (not AST specific)
I-134 9.65E-01 N/A Modeling update (not AST specific)
I-135 8.17E-01 N/A Modeling update (not AST specific)
Westinghouse Non-Proprietary Class 3 Page 28 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-10 Assumptions Used for Letdown Line Break Analysis (cont.)
AST CLB Reason for Change Iodine chemical form of releases (%)
Elemental 97 N/A Regulatory Guide Update (not AST specific)
Organic 3 N/A Regulatory Guide Update (not AST specific)
Particulate 0 N/A Regulatory Guide Update (not AST specific)
Reactor coolant mass, minimum (lbm) 3.99E+05 N/A Modeling update (not AST specific)
Flow rate out of broken line (gpm) 141 141 No change Iodine and alkali metal airborne 0.18 N/A Modeling update fraction (not AST specific)
Maximum RCS letdown pressure 600 2200 Modeling update (psig) (not AST specific)
Maximum RCS letdown 380 286 Modeling update temperature (°F) (not AST specific)
Time to isolate break flow 30.167 30.167 No change (terminating releases) (min)
Control room isolation None N/A No change Control room atmospheric dispersion N/A factor (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.11E-03 N/A Not modeled in current AORs TSC atmospheric dispersion factor (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.80E-04 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 29 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis Reason for AST CLB Change RCS activity See Table 4.3-1a See Table 4.3-1c Updated calculations Initial secondary system activity See Table 4.3-1a See Table 4.3-1c Updated calculations Pre-accident iodine spike factor 60 60 No change Accident-initiated iodine spike appearance rate calculations Letdown flow, maximum (gpm) 132 120 Modeling update (not AST specific)
Letdown flow decontamination 100 N/A Modeling update
(%) (not AST specific)
RCS leakage (gpm) 11 1 Modeling update (not AST specific)
Spike factor 335 335 No change Reactor coolant mass, maximum 8.42E+05 5.05E+5 Input change (lbm) (maximum vs nominal)
Duration of accident-initiated 8 8 No change iodine spike (hr)
Equilibrium appearance rates (Ci/min)
I-130 9.87E-03 N/A Modeling update (not AST specific)
I-131 4.39E-01 3.54E-1 Modeling update (not AST specific)
I-132 1.98E+00 1.36E+00 Modeling update (not AST specific)
I-133 8.93E-01 7.72E+00 Modeling update (not AST specific)
I-134 9.65E-01 7.02E-01 Modeling update (not AST specific)
I-135 8.17E-01 6.63E-01 Modeling update (not AST specific)
Westinghouse Non-Proprietary Class 3 Page 30 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.)
Reason for AST CLB Change Iodine chemical form of releases (%)
Elemental 97 N/A Regulatory Guide Update (not AST specific)
Organic 3 N/A Regulatory Guide Update (not AST specific)
Particulate 0 N/A Regulatory Guide Update (not AST specific)
Approximate timing of events (sec) See Table 2.7.3-2 N/A Modeling update in Enclosure I of (not AST specific) this LAR (For dose input, exclude the 100 seconds of steady-state operation.)
Time of control room isolation 120 N/A Modeling update (including delay) (sec) (not AST specific)
Time to complete control room 600 N/A Modeling update isolation with SI signal from event (not AST specific) initiation (sec)
Westinghouse Non-Proprietary Class 3 Page 31 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.)
Reason for AST CLB Change Transient mass transfer data Non-flashed break flow (lbm) 0 - 52 seconds 2227.5 N/A Updated calculations 52 - 1102 seconds 43,129.9 N/A Updated calculations 1102 - 2902 seconds 88,387.2 N/A Updated calculations 2902 - 3502 seconds 32,991.2 N/A Updated calculations 3502 - 3846 seconds 18,224.8 N/A Updated calculations 3846 - 5155 seconds 61,523.0 N/A Updated calculations 5155 - 7527 seconds 41,166.4 N/A Updated calculations Flashed break flow (lbm) 0 - 52 seconds 438.9 N/A Updated calculations 52 - 1102 seconds 2,901.8 N/A Updated calculations 1102 - 2902 seconds 13,432.1 N/A Update calculations 2902 - 3502 seconds 2,635.6 N/A Update calculations 3502 - 3846 seconds 606.1 N/A Update calculations
Westinghouse Non-Proprietary Class 3 Page 32 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.)
Reason for AST CLB Change Steam released from ruptured SG (lbm) 0 - 52 seconds 188,100 N/A Update calculations 52 - 1102 seconds 27,469.2 N/A Update calculations 1102 - 2902 seconds 149,850.8 N/A Update calculations 2902 - 7527 seconds 0 N/A Update calculations 7527 - 43,200 seconds 2530 N/A Update calculations Steam released from intact SGs (lbm) 0 - 52 seconds 562,650 N/A Update calculations 52 - 1102 seconds 69,877.5 N/A Update calculations 1102 - 3502 seconds 0 N/A Update calculations 3502 - 3846 seconds 94,307.4 N/A Update calculations 3846 - 5155 seconds 130,799.9 N/A Update calculations 5155 - 7527 seconds 98,156.3 N/A Update calculations 7527 - 43,200 seconds 1,645,930 N/A Update calculations Reactor coolant mass, minimum (lbm) 3.99E+05 5.05E+05 CLB Input change does not use max (minimum vs and min nominal)
Ruptured SG mass, minimum (lbm) 7.00E+04 N/A Updated calculations Intact SGs mass, minimum (lbm, 1.95E+05 N/A Updated total) calculations Condenser iodine and alkali metal 100 N/A Modeling update removal factor (not AST specific)
SG iodine water/steam partition 100 N/A Regulatory Guide coefficient Update (not AST specific)
Moisture carryover (%) 0.25 0.25 No change
Westinghouse Non-Proprietary Class 3 Page 33 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-11 Assumptions Used for SGTR Dose Analysis (cont.)
Reason for AST CLB Change Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 2.55E-02 N/A Not modeled in current AORs 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />(1) 1.04E-03 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.46E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.03E-04 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.90E-04 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.39E-04 N/A Not modeled in current AORs TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.83E-04 N/A Not modeled in current AORs 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.58E-04 N/A Not modeled in current AORs 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.63E-05 N/A Not modeled in current AORs 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 6.45E-05 N/A Not modeled in current AORs 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 4.89E-05 N/A Not modeled in current AORs Note: (1) The control room unfiltered inleakage continues to be associated with the normal mode /Q of 2.55E-02 sec/m3 until completion of control room isolation from a safety injection signal at 10 minutes.
Westinghouse Non-Proprietary Class 3 Page 34 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis Reason for AST CLB Change Core activity See Table 4.3-1a See Table 4.3-1b Updated (containment leakage, calculations ECCS leakage, and RWST back-leakage)
RCS activity See Table 4.3-1a See Table 4.3-1b Updated (containment purge) calculations Fuel release fractions 100% of the core activity is Modeling change and timing released immediately for AST following event initiation Nuclide Group Gap Early Gap Early Release In-Vessel Release In-Vessel Phase Phase Phase Phase Fraction Fraction Fraction Fraction Noble gases 0.05 0.95 N/A N/A Modeling change for AST Iodines 0.05 0.35 N/A N/A Modeling change for AST Alkali metals 0.05 0.25 N/A N/A Modeling change for AST Tellurium metals 0.00 0.05 N/A N/A Modeling change for AST Barium and 0.00 0.02 N/A N/A Modeling change strontium for AST Noble metals 0.00 0.0025 N/A N/A Modeling change for AST Cerium 0.00 0.0005 N/A N/A Modeling change for AST Lanthanides 0.00 0.0002 N/A N/A Modeling change for AST
Westinghouse Non-Proprietary Class 3 Page 35 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.)
Reason for AST CLB Change Duration of phases Phase Onset Duration Onset Duration Gap release 30 sec 0.49167 hr N/A N/A Modeling change for AST Early in-vessel 0.5 hr 1.3 hr N/A N/A Modeling change for AST SI signal (sec) 0 N/A Modeling update (not AST specific)
Time of control room 120 0 Modeling update isolation (including (not AST delays) (sec) specific)
Containment Leakage Iodine chemical form of releases (%)
Elemental 4.85 91 Modeling change for AST Organic 0.15 4 Modeling change for AST Particulate 95 5 Modeling change for AST Containment volume, 2.7E+06 2.5E+06 Since a maximum (ft3) maximum volume is conservative, the larger value listed in USAR Section 6.2.1.5.3 was modeled
% Sprayed 85 85 No change
% Unsprayed 15 15 No change
Westinghouse Non-Proprietary Class 3 Page 36 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.)
AST CLB Reason for Change Mixing between sprayed and unsprayed 6.94E+04 8.50E+04 The CLB value corresponds containment volumes (cfm) to two hydrogen mixing fans operating in slow speed.
While a mixing rate of 85,000 cfm will still be present following the event, the value has been conservatively reduced to the flow rate from one containment cooler fan Start of fan cooler mixing (min) 2 N/A Modeling update (not AST specific)
Containment leak rates (weight %/day) 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.2 0.2 No change 1 - 30 days 0.1 0.1 No change Spray timing Initiation (min) 2 0 The time to credit containment spray was conservatively increased to account for the delay to reaching full flow conditions Termination (hr) 5 9.55 Conservative time used Spray removal coefficients Organic iodine spray removal 0.0 0.0 No change coefficient (hr -1)
Elemental iodine spray removal coefficient calculations Spray removal coefficient (hr-1), 10 10 No change DF < 200 Gas phase mass transfer 3 N/A Modeling update (not AST coefficient (m/min) specific)
Time of fall of the spray drops 0.146 N/A Updated calculations (min)
Volume flow rate of sprays 658.66 711.13 Conservative low flow (m3/hr) modeled.
Westinghouse Non-Proprietary Class 3 Page 37 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.)
AST CLB Reason for Change Containment sprayed volume 6.5E+04 6.0E+04 The maximum containment (m3) volume increased from 2.5E+06 ft3 to 2.7E+06 ft3.
Since 85% of the containment is sprayed, the sprayed volume increased proportionally with the increase in total volume.
Mass-mean diameter of the 0.00116 N/A Updated calculations spray drops (m)
Particulate spray removal coefficient calculations Spray removal coefficient (hr-1), 5 0.45 Updated calculations DF < 50 (Used DFs up to 100)
Spray removal coefficient (hr-1), 0.5 N/A Updated calculations DF > 50 Drop fall height (m) 35.966 36.017 Conservative rounding Volume flow rate of sprays 658.66 711.13 Conservative low flow (m3/hr) modeled.
Containment sprayed volume 6.5E+04 6.0E+04 WCGS input change The (m3) maximum containment volume increased from 2.5E+06 ft3 to 2.7E+06 ft3.
Since 85% of the containment is sprayed, the sprayed volume increased proportionally with the increase in total volume.
Ratio of dimensionless collection efficiency to average spray drop diameter Prior to DF of 50 (m-1) 10 N/A Modeling update (not AST specific)
After DF of 50 (m-1) 1 N/A Modeling update (not AST specific)
Westinghouse Non-Proprietary Class 3 Page 38 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.)
AST CLB Reason for Change Particulate sedimentation removal 0.1 N/A Modeling update (not AST coefficient (hr-1), DF < 1000 specific) pH of sump 7.0 8.5 The pH value was changed to 7.0 to be consistent with the iodine retention assumptions contained within the analysis.
Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 2.11E-03 5.30E-04 Updated calculations 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.12E-04 5.30E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.38E-04 5.30E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.81E-04 3.6E-04 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.29E-04 6.60E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 9.65E-05 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.91E-04 2.2E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.66E-04 2.2E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 9.62E-05 1.17E-04 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 7.05E-05 2.04E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 5.52E-05 0 Updated calculations ECCS Leakage Iodine chemical form of releases (%)
Elemental 97 N/A Regulatory Guide Update Organic 3 N/A Regulatory Guide Update Particulate 0 N/A Regulatory Guide Update Sump volume (gal) 4.60E+05 4.60E+05 No change Time to initiate ECCS recirculation (min) 0 28.2 Modeling update (not AST specific)
ECCS leakage to auxiliary building (gpm) 2 2 No change Iodine airborne fraction 0.10 0.10 No change
Westinghouse Non-Proprietary Class 3 Page 39 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.)
AST CLB Reason for Change Auxiliary building exhaust filter efficiency 90 90 No change (all forms of iodine) (%)
Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 2.11E-03 1.10E-04 Updated calculations 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.12E-04 1.10E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.38E-04 1.10E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.79E-04 6.80E-05 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.14E-04 1.70E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 8.94E-05 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.80E-04 2.2E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.80E-04 2.2E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 6.44E-05 1.17E-04 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 4.42E-05 2.04E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 3.22E-05 0 Updated calculations RWST Back-Leakage RWST initial gas volume, minimum (gal) 3.54E+05 N/A Modeling update Time to initiate ECCS recirculation (min) 0 28.2 Modeling update ECCS leakage to RWST (gpm) 3.8 3.8 No change Iodine airborne fraction 0.10 0.10 No change Release from RWST gas space (gpm) 3.8 3.8 No change Iodine chemical form of releases (%)
Elemental 97 91 Regulatory Guide Update Organic 3 4 Regulatory Guide Update Particulate 0 5 Regulatory Guide Update
Westinghouse Non-Proprietary Class 3 Page 40 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.)
AST CLB Reason for Change Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 1.03E-03 1.10E-04 Updated calculations 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.80E-04 1.10E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.19E-04 1.10E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.27E-04 6.80E-05 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.96E-04 1.70E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.53E-04 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.87E-04 2.2E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.25E-04 2.2E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.51E-05 1.17E-04 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 3.33E-05 2.04E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 2.61E-05 0 Updated calculations Containment Purge RCS activity released (%) 100 100 No change Iodine chemical form of releases (%)
Elemental 97 91 Regulatory Guide Update Organic 3 4 Regulatory Guide Update Particulate 0 5 Regulatory Guide Update RCS mass, maximum (lbm) 8.42E+05 4.94E+05 Input change (maximum vs CLB does nominal) not use maximum and minimum Containment volume, minimum (ft3) 2.5E+06 2.5E+06 No change Maximum purge flow rate, unfiltered (cfm) 4,680 4,680 No change Duration of purge release (sec) 10 8 Updated input
Westinghouse Non-Proprietary Class 3 Page 41 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-12 Assumptions Used for LOCA Analysis (cont.)
AST CLB Reason for Change Control room atmospheric dispersion factors (sec/m3) 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 2.11E-03 1.10E-04 Updated calculations 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.12E-04 1.10E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.38E-04 1.10E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.79E-04 6.80E-05 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.14E-04 6.60E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 8.94E-05 0 Updated calculations TSC atmospheric dispersion factors (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.80E-04 2.2E-04 Updated calculations 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.80E-04 2.2E-04 Updated calculations 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 6.44E-05 1.17E-04 Updated calculations 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 4.42E-05 2.04E-05 Updated calculations 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 3.22E-05 0 Updated calculations
Westinghouse Non-Proprietary Class 3 Page 42 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-13 Assumptions Used for Waste Gas Decay Tank Failure Analysis AST CLB Reason for Change Activity in ruptured tank See See Updated calculations Table 4.3-2a Table 4.3-2b Iodine chemical form of releases (%)
Elemental 100 N/A Modeling update (not AST specific)
Duration of release (hr) 2 2 No change Control room isolation None N/A No change Control room atmospheric dispersion factor (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7.17E-04 5.30E-04 Updated calculations TSC atmospheric dispersion factor (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.97E-04 N/A Not modeled in current AORs Table 4.3-14 Assumptions Used for Liquid Waste Tank Failure Analysis AST CLB Reason for Change Activity in ruptured tank Recycle holdup tank See See Updated calculations Table 4.3-2a Table 4.3-2b Hypothetical tank maximizing See Table N/A Updated calculations iodine 4.3-2a Iodine chemical form of releases (%)
Elemental 100 N/A Modeling update (not AST specific)
Duration of release (hr) 2 2 No change Control room isolation None N/A No change Control room atmospheric dispersion factor (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 7.17E-04 N/A Not modeled in current AORs TSC atmospheric dispersion factor (sec/m3) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.97E-04 N/A Not modeled in current AORs
Westinghouse Non-Proprietary Class 3 Page 43 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-15 Assumptions Used for Fuel Handling Accident Analysis AST CLB Reason for Change Core activity for one assembly at minimum time prior to fuel movement (Ci)
Kr-85m 1.10E+00 See Table 4.3-4c Updated calculations Kr-85 5.69E+03 See Table 4.3-4c Updated calculations Xe-131m 5.38E+03 See Table 4.3-4c Updated calculations Xe-133m 1.76E+04 See Table 4.3-4c Updated calculations Xe-133 8.19E+05 See Table 4.3-4c Updated calculations Xe-135m 5.58E+01 See Table 4.3-4c Updated calculations Xe-135 8.14E+03 See Table 4.3-4c Updated calculations I-130 1.45E+02 See Table 4.3-4c Updated calculations I-131 4.12E+05 See Table 4.3-4c Updated calculations I-132 3.95E+05 See Table 4.3-4c Updated calculations I-133 8.90E+04 See Table 4.3-4c Updated calculations I-135 3.42E+02 See Table 4.3-4c Updated calculations Number of fuel assemblies damaged 1.2 1.2 No change Peaking factor 1.65 1.65 No change Time of Control room isolation 120 N/A No change (including delays) (sec)
Decay time prior to fuel movement, 76 76 No change minimum (hr)
Gap fractions I-131 0.12 0.12 No change Kr-85 0.30 0.30 No change Other iodines and noble gases 0.10 0.10 No change Iodine chemical form in gap (%)
Elemental 99.85 N/A Modeling change for AST Organic 0.15 N/A Modeling change for AST Fuel pool water depth, minimum (ft) 23 23 No change Fuel rod internal pressure, maximum 1500 1200 Modeling update (not (psig) AST specific)
Westinghouse Non-Proprietary Class 3 Page 44 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-15 Assumptions Used for Fuel Handling Accident Analysis (cont.)
AST CLB Reason for Change Overall pool iodine DF 200 100 Regulatory Guide Update Iodine airborne fractions (%)
Elemental 70 N/A Modeling change for AST Organic 30 N/A Modeling change for AST Duration of release (hr) 2 2 No change Removal of airborne activity in No change containment/fuel building (other than None None decay)
Control room atmospheric dispersion factor (sec/m3)
Auxiliary Building Release HVAC Flows Except for Control Room Unfiltered Inleakage 0 - 0.0333 hours0.00385 days <br />0.0925 hours <br />5.505952e-4 weeks <br />1.267065e-4 months <br /> 2.11E-03 5.30E-04 Updated calculations 0.0333 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.12E-04 5.30E-04 Updated calculations Control Room Unfiltered Inleakage 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.11E-03 5.30E-04 Updated calculations Containment Release(1) 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.38E-03 N/A Not modeled in current AORs TSC atmospheric dispersion factor (sec/m3)2 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.91E-04 N/A Not modeled in current AORs Note: (1) The containment release /Q is lower that the auxiliary building release /Q; therefore the auxiliary building release /Q is used in the calculation of the doses. The containment release /Q is only used to determine the time the control room isolation setpoint is reached since it is lower than the containment leakage /Q and results in a conservative isolation time.
Westinghouse Non-Proprietary Class 3 Page 45 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.3-16 Technical Support Center Parameters AST CLB Reason for Change TSC volume (ft3) 44,000 52,800 The CLB value is based on a simple bounding one volume calculation (length x width x height). The more accurate AST value was calculated by determining the volume of all individual TSC rooms and then summing them.
Normal ventilation flow rates (cfm)
Unfiltered makeup flow rate 550 550 No change Unfiltered inleakage 20 20 No change Emergency mode flow rates (cfm)
Filtered makeup flow rate 550 550 No change Unfiltered makeup flow rate 0 N/A No change Unfiltered inleakage 20 20 No change Filtered recirculation flow 450 450 No change Filter efficiencies (%)
Elemental iodine 95 90 Test procedure revised to increase acceptance criterion to 95%
Organic iodine 95 90 Test procedure revised to increase acceptance criterion to 95%
Particulates 95 90 Test procedure revised to increase acceptance criterion to 95%
Delay to switch to emergency mode 60 0 A delay of 60 minutes was of operation after event initiation added to ensure the analysis (minutes) bounds the plant specific time to switch to the emergency mode of operation after event initiation TSC breathing rate for duration of 3.5E-04 3.5E-04 No change the event (m3/sec)
TSC occupancy factors 0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 1.0 No change 1 - 4 days 0.6 0.6 No change 4 - 30 days 0.4 0.4 No change
Westinghouse Non-Proprietary Class 3 Page 46 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-2:
Enclosure VI, page 4-38 states, in part, that An additional fuel management multiplier is applied to the calculated core activity to account for anticipated variations in fuel cycle design.
Please provide the value of the multiplier, how it was derived, and a justification for its use.
Response
The analyses presented in Section 4.2 of Enclosure VI include a fuel management multiplier on the core inventory to accommodate anticipated changes in the values of fuel enrichment and fuel burnup, since the fuel cycle designs vary from cycle to cycle.
The fuel management multiplier is equal to [ ]a,c. The fuel management multiplier increases the core activity calculated for the cycle 19 fuel cycle design, which approximates an equilibrium fuel cycle, to account for any differences in cycle 20 and future fuel cycles; therefore, it adds margin to the calculated doses. The addition of this margin is not required by Regulatory Guide 1.183 and provides an additional conservatism in the dose calculations.
The multiplier is derived from a number of historical sensitivity analyses on core activity and post-accident dose consequence analyses. Variations in fuel enrichment, core loading (i.e. fuel mass), and cycle burnup were considered in the sensitivities analyses. Based on these sensitivity analyses, a value of [ ] a,c was used for the WCGS.
ARCB-RAI-3 Enclosure VI, page 1-1 proposes an exception to the full scope implementation of the AST methodology. The exception to retain the current licensing basis (using Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, dated March 23, 1962 (not publicly available, proprietary information), for NUREG-0737, Clarification of TMI Action Plan Requirements, July 2000 (ADAMS Accession No. ML051400209), evaluations other than Control Room Habitability Envelope (CRHE) and Technical Support Center (TSC) doses) is based upon NRC Regulatory Guide (RG) 1.183, Section 1.3.5, Equipment Environmental Qualification and Section 6, Assumptions for Evaluating the Radiation Doses for Equipment Qualification. These RG 1.183 sections are only for equipment qualification and are not for NUREG-0737 evaluations.
RG 1.183, Regulatory Position 4.3, under Other Dose Consequences, states, in part, that
Westinghouse Non-Proprietary Class 3 Page 47 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737 .... Design envelope source terms provided in NUREG-0737 should be updated for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE
[total effective dose equivalent].
b) Please confirm if the current licensing basis calculations are used to determine post-accident vital area access unaffected or bounding (using the TID-14844 source term) compared to those using the AST.
b) Please explain if any new operator actions have been credited to support the AST methodology.
c) Please explain if the current licensing basis calculations are used to determine the doses for operation of the post-accident sampling system or the containment high range radiation monitors used to monitor post-accident primary containment radiation levels (using the TID-14844 source term) unaffected or bounding compared to those using the AST.
d) Please provide a detailed justification for the proposed exemption or follow Regulatory Position 4.3 of RG 1.183.
Response
a) Regulatory Position 1.3.2 of Regulatory Guide 1.183 specifies that an evaluation of AST impact was performed on three representative operating reactors and concluded that the Technical Information Document (TID)-14844 based methodology and assumptions generally bound the AST methodology. Regulatory Position 1.3.2, states, in part:
This evaluation determined that radiological analysis results based on TID-14844 source term assumptionsand the whole body and thyroid methodology generally bound the results from the analyses based on AST and TEDE methodology.
Licensees may use the applicable conclusions of this evaluation in addressing the impact of the AST on design basis radiological analyses.
Full implementation of AST at WCGS does not impact the assumptions or inputs to the current TID based analyses. As such, AST implementation would not impact the conclusions from Regulatory Position 1.3.2, the TID based analyses and results, or the ability to implement AST in other radiological analyses per the guidance in Regulatory Position 1.3.4.
b) No new operator actions have been credited to support the AST methodology.
Regarding the time critical action to isolate the failed CREVS train, the overall action
Westinghouse Non-Proprietary Class 3 Page 48 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment described in USAR Section 15A.3.1 remains unchanged, i.e., operator action is required to ensure no bypass pathways exist for unfiltered air to enter the control room following the failure of a filtration fan. However, the damper modification being implemented will eliminate the possibility of having to locally isolate the failed CREVS train. Specifically, the modification will result in the system automatically isolating the CREVS by normal system operation when the control room AC fan unit is turned off, which can be performed remotely from the control room following the identification of the failed filtration fan.
c) The requirements for the post-accident sampling system were eliminated in Amendment No. 137 (Reference 5) and thus is unaffected by the change to AST methodology. With regards to the containment high range radiation monitors, the controlling isotope for the TID-14844 source term, Xe133, remains the controlling isotope for the AST source term. Thus, the containment high range radiation monitors are also unaffected.
d) WCNOC is only proposing exceptions to the full scope implementation of the AST for equipment qualification and NUREG-0737 evaluations other than Control Room Habitability Envelope (CRHE) and Technical Support Center (TSC) doses. With regards to NUREG-0737 evaluations, full implementation of AST at WCGS does not impact the assumptions or inputs to the current TID based analyses, as documented in the part a) response to this request. Thus, AST implementation would not impact their conclusions from Regulatory Position 1.3.2, the TID based analyses, their results, or the ability to implement AST in other radiological analyses per the guidance in Regulatory Position 1.3.4. Once AST is part of the sites licensing basis this regulatory position specifies that all characteristics of the AST and the TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an individual as-needed basis. WCNOC plans to incorporate AST into these remaining analyses on an as needed basis as described in Regulatory Position 1.3.4.
ARCB-RAI-4:
Enclosure VI, page 5-2 states, in part, that Core design parameters (enrichment, burnup, and MTU loading) are based on cycle 19 core design.
c) Please describe how many batches or core regions were assumed to determine the source term.
d) Please explain what period of irradiation (burnup) and specific power (i.e., megawatts days per metric ton of uranium (MWD/MTU)) was assumed for each region.
c) Please explain if the activity from each batch was taken at the end of life for each cycle.
Westinghouse Non-Proprietary Class 3 Page 49 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment d) Please provide the number of dose significant isotopes form the source term input for the RADTRAD dose evaluations.
e) For the purpose of the design basis, please explain the maximum enrichment assumed and if the assumed period of irradiation allows for the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values.
Response
a) The core is modeled in seven fuel regions (batches).
b) The length of cycle 19 was modelled as 511 effective full-power days (EFPD), which bounds the actual cycle length of 505 EFPD. Prior cycles were assumed to operate for 455 EFPD consistent with the WCGS fuel cycle operating experience. The region-specific data is tabulated below for cycle 19.
Wt% MWt/FA MWt/FA Region # of FAs U-235 (pre- cycle 19) (cycle 19) a,c
Westinghouse Non-Proprietary Class 3 Page 50 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment c) The activity reported in Enclosure IV, Table 4.2-1 is taken at the end of cycle 19 for each fuel region. (511 EFPD) d) The LOCA dose consequence analysis modeled the 63 nuclides presented in Enclosure IV, Table 4.3-1a. The main steamline break, loss of non-emergency AC power, locked rotor, rod ejection, letdown line break, and steam generator tube rupture dose consequence analyses consider the 21 noble gas, iodine, and alkali metal nuclides listed in Enclosure IV, Table 4.3-1a. The fuel handling accident dose consequence analysis (both inside containment and in the fuel building) considers the 12 noble gas and iodine nuclides listed in Enclosure IV, Table 4.3-15. The tank rupture dose consequence events (waste gas decay tank failure and liquid waste tank failure) consider the 16 noble gas and iodine nuclides listed in Enclosure IV, Table 4.3-2a.
e) The maximum enrichment considered for a single fuel assembly is [ ]a,c wt% U-235, consistent with the cycle 19 fuel cycle design.
In general, radionuclides with short half-lives are present in quantities that are proportional to core power. Those are conservatively treated by including the emergency core cooling system (ECCS) evaluation uncertainty in core thermal power. Quantities of longer-lived radionuclides will build up during the fuel cycle. Those are conservatively treated by modelling the longest anticipated cycle 19 length (511 EFPD). The end of life values are reported in Enclosure IV, Table 4.2-1.
This is consistent with Regulatory Guide 1.183 Section C.3.1, which states:
The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values.9 9
Note that for some radionuclides, such as Cs-137, equilibrium will not be reached prior to fuel offload. Thus, the maximum inventory at the end of life should be used.
ARCB-RAI-5 Enclosure VI, page 15B-3 states that the analysis conforms to Regulatory Position 3.2 of RG 1.183. Please confirm that reactor fuel will have a peak burnup of less than 62,000 MWD/MTU of uranium and a maximum linear heat generation rate of 6.3 kilowatts per foot (kW/ft.) or less for peak rod average power for burnups exceeding 54,000 MWD/MTU. If not, please justify how the analysis conforms to Regulatory Position 3.2.
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Response
The fuel handling accident (FHA) analysis conservatively assumed that 100% of the fuel will not meet the Footnote 11 from Regulatory Position 3.2 of Regulatory Guide 1.183. Footnote 11 contains burnup and linear heat generation rate limits for the gap fractions contained in Table 3 of Regulatory Guide 1.183 as well as an alternative method for determining gap fractions provided they bound the limiting projected plant-specific power history for the specific load. The analysis conforms with the position such that alternative gap fractions were used which were appropriate for the assumption of not meeting the Footnote 11 limits. These gap fractions are obtained from Regulatory Guide 1.25, as modified by NUREG/CR-5009, which provides higher, i.e., more conservative gap fractions than Regulatory Guide 1.183, which are not constrained by the Footnote 11 burnup limits as they can be applied to higher burnup to bound power history.
ARCB-RAI-6:
Enclosure VI, page 4-63 states, in part, that Sedimentation is credited in the portion of containment that is not impacted by spray removal and in the sprayed portion when sprays are not on at a rate of 0.1 hr -1 until a DF [decontamination factor] of 1000 is reached at 23.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. After this time sedimentation removal is terminated.
Removal of aerosol by sprays and natural deposition are competing processes. Please justify crediting both spray removal and the proposed 0.1 hr -1 sedimentation rate. Describe how the natural deposition model accounts for removal due to the spray model used. If any further credit for a reduction in aerosols is taken for any pathway, please provide a justification for that credit while considering the impact of any other removal mechanism credited. Please confirm if the assumed sedimentation rate impacts the final dose.
Response
The sprayed and unsprayed regions of containment were modeled as separate compartments. Activity removal via natural deposition was only modeled in the unsprayed region of containment and in the sprayed region of containment when sprays were not credited for removal of aerosols (e.g., after containment spray termination). Activity removal via the sprays and activity removal via natural deposition were not modeled concurrently in the same compartment of containment. Therefore, there was no double-accounting for the removal of aerosols in either the sprayed or unsprayed regions of containment. The user-defined coefficients in the RADTRAD model for sprays and natural deposition were used.
There is no other credit for reduction of aerosols in any pathway except for removal by filters in the control building/control room ventilation systems. This removal does not
Westinghouse Non-Proprietary Class 3 Page 52 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment compete with any other removal process and is justified due to the availability of the filters in these systems.
The assumed sedimentation rate does impact the final dose results.
ARCB-RAI-7:
Enclosure VI, page 4-63 states, in part, that The resulting removal coefficient for elemental iodine is 22.9* hr - 1 SRP [Standard Review Plan] 6.5.2 allows for elemental iodine removal credit of up to 20 hr-1 during injection spray; however, to avoid sensitivities with spray switchover times from injection to recirculation and to conservatively address iodine loading in the spray fluid during recirculation, the removal is limited to 10 hr - 1 for either spray mode.
Please explain in more detail why the removal coefficient is determined to be limited to 10 hr-1 and what is meant by "sensitivities with spray switchover times from injection to recirculation and to conservatively address iodine loading in the spray fluid during recirculation."
Response
In accordance with Standard Review Plan (SRP) Section 6.5.2, the spray removal coefficient, if calculated to be greater than 20 hr-1, is to be limited to 20 hr-1. For the analysis performed for the WCGS, the removal coefficient is calculated to be greater than 20 hr-1; however, the removal coefficient is reduced further to 10 hr-1 to conservatively address iodine loading in the spray fluid during recirculation. During recirculation, the spray solution will gradually become loaded with elemental iodine, which will limit the capacity of the spray to remove airborne iodine. The spray removal coefficient would be inversely proportional to the decontamination factor achieved for elemental iodine. Thus, when recirculation spray is first credited, there is a low enough level of elemental iodine in the sump solution that it is appropriate to use the removal coefficient of 20 hr-1. However, when the decontamination factor approaches its defined limit of 200, the removal coefficient would be only a small fraction of its original value. The impact of this varying nature of the removal coefficient is approximated by setting the removal coefficient to one half of the calculated value. However, for conservatism and simplification, a value of 10 hr-1 (half of the 20 hr-1 limits specified in SRP 6.5.2) was used.
The spray removal coefficient is a function of the spray flow rate. The analysis used a spray removal coefficient based on a conservatively low spray flow rate (658.66 m3/hr) to bound both modes (injection and recirculation, with spray flow rates of 665.86 m3/hr and 780.00 m3/hr, respectively) of spray operation. Therefore, sensitivities on timing on the spray switchover times are not necessary.
Westinghouse Non-Proprietary Class 3 Page 53 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The lower spray flow rate and the factor of 2 reduction in the removal coefficient applied for the entire duration of spray operation bounds the impacts on the analysis results due to potential changes in switchover timing.
The following wording is added to Enclosure IV, Section 4.3.9.2.2.1:
During recirculation, the spray solution will gradually become loaded with elemental iodine, which will limit the capacity of the spray to remove airborne iodine. The spray removal coefficient would be inversely proportional to the decontamination factor (DF) achieved for elemental iodine. Thus, when recirculation spray is first credited, there is a low enough level of elemental iodine in the sump solution that it is appropriate to use the removal coefficient of 20 hr-1. However, when the DF approaches its defined limit of 200, the removal coefficient would be only a small fraction of its original value. The impact of this varying nature of the removal coefficient is approximated by setting the removal coefficient to one half of the calculated value.
Also, the spray removal coefficient is a function of the spray flow rate. The analysis used a spray removal coefficient based on a conservatively low spray flow rate to bound both modes (injection and recirculation) of spray operation.
ARCB-RAI-8:
Enclosure VI, page 4-64 states, in part, that An adjustment is made to account for a reduction in the RWST [refueling water storage tank] gas volume available for dilution as the leakage into the RWST increase the water level.
Please provide details regarding this adjustment so that the NRC staff can independently confirm the doses from the RWST back-leakage.
Response
As leakage enters the RWST at a rate of 3.8 gpm (assumed to be all liquid), the gas volume that can be credited for dilution (initially 3.54E+05 gallons) is reduced, resulting in an increased concentration of activity in the RWST gas volume. The leakage is modeled for the duration of the analysis, resulting in a gas volume of 189,840 gallons at 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />. Because RADTRAD maintains a constant compartment volume, the effects of the increased concentration are accounted for via an increased flow rate from the RWST. The flow rate is increased in proportion to the expected increase in concentration resulting from the reduced RWST gas volume.
The following wording is added to Enclosure IV, Section 4.3.9.2.2.3:
The RWST gas volume is decreased by the rate of leakage into the RWST.
Westinghouse Non-Proprietary Class 3 Page 54 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-9:
Please provide the doses from each pathway analyzed for the loss-of-coolant accident (for the exclusion area boundary (EAB) (worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), low population zone (LPZ), and control room (at 30 days)). Please explain whether the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB dose is determined using the sum of the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose for each pathway to the EAB or by first summing all the time dependent dose pathways and then determining the worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose.
Response
The following summarizes the doses for each pathway analyzed for the LOCA dose consequence analysis:
Exclusion Area Boundary (EAB) (0.5 hr to 2.5 hr):
Containment Leakage 4.73E+0 rem TEDE ECCS Leakage 4.63E-1 rem TEDE RWST Back-Leakage 4.36E-3 rem TEDE Containment Purge 0.00E+0 rem TEDE Low Population Zone (LPZ):
Containment Leakage 2.00E+0 rem TEDE ECCS Leakage 1.26E+0 rem TEDE RWST Back-Leakage 1.57E+0 rem TEDE Containment Purge 8.24E-4 rem TEDE Control Room:
Containment Leakage 1.01E+0 rem TEDE ECCS Leakage 8.48E-1 rem TEDE RWST Back-Leakage 2.69E+0 rem TEDE Containment Purge 6.52E-2 rem TEDE External Sources 1.53E-1 rem TEDE The worst 2-hour EAB dose was determined by summing all the time-dependent dose pathways and then determining the worst 2-hour dose. Note that releases via the mini-purge terminate prior to 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and thus do not contribute to the limiting 2-hour EAB dose.
This information is included in Enclosure IV, Section 4.3.9.4.
ARCB-RAI-10:
Enclosure VI, page 4-38 states, in part, that The FIPCO-V computer code calculates the buildup of fission product activities in plant systems and components, including the reactor coolant system, chemical and volume control
Westinghouse Non-Proprietary Class 3 Page 55 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment system demineralizer resins, VCT [volume control tank] liquid and vapor phases, and waste decay tank (WGDT).
Please provide details regarding the input and methodologies used in the FIPCO-V code for the staff to replicate the calculations performed by the FIPCO-V code.
Response
The equations which are used to determine the concentrations of fission products at any time are developed from the general balance equation:
Net Rate of = Rate of - Rate of Accumulation Formation Loss Reactor Core Fission Products The buildup of fission products in the fuel assemblies in the core is described by the following equations:
[
]a,c
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[
]a,c
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Westinghouse Non-Proprietary Class 3 Page 63 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The inputs used in the FIPCO calculations are provided in the following two tables. The data below are taken from operating data, fuel cycle design data, and plant information.
Key Input Parameters for WCGS RCS Activity Analyses Parameter Value Notes a,c Core thermal power, MWt Fuel Management multiplier RCS boron concentrations Fuel defect level, %
RCS volume, ft3 RCS expansion factor Steam generator (S/G) tube plugging level, %
Normal pressurizer liquid level @
full power, %
Core Tavg, oF Letdown flow rate, gpm Volume control tank (VCT) volume, gallons/ft3 VCT liquid volume, %
VCT conditions:
Temperature, oF Pressure, psig VCT continuous purge rate, scfm .
Gas decay tank volume, ft3 Number of gas decay tanks .
Effective deborating demineralizer cut-in concentration, ppm Mixed bed resin volume, ft3
Westinghouse Non-Proprietary Class 3 Page 64 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Critical Boron Concentrations Burnup B Conc. Burnup B Conc. Burnup B Conc.
(MWd/MTU) (ppm) (MWd/MTU) (ppm) (MWd/MTU) (ppm) a,c ARCB-RAI-11:
For the containment purge pathway analyzed for the loss-of-coolant accident (LOCA),
please describe what is the assumed form of radioiodine released from the reactor coolant system prior to isolation of the containment purge. Please justify your answer.
Response
The forms of radioiodine assumed for the containment purge pathway is 97% elemental and 3%
organic. This is an assumed split consistent with other releases specified in Regulatory Guide 1.183. There is no guidance on forms of radioiodine for purge releases in Regulatory Guide 1.183 and the split has no impact on the analysis results since no spray removal is credited for the RCS activity in containment. The Control Room Emergency Ventilation System (which includes the Control Building) filters are modeled with the same efficiency for all forms of iodine.
The following wording is added to Enclosure IV, Section 4.3.9.2.2.4:
Note that the assumed iodine chemical fractions do not impact the analysis results since spray removal is not credited for the RCS activity in containment.
The following wording is added to Enclosure IV, Section 4.3.9.2.3:
Note that the assumed iodine chemical fractions for the containment purge pathway do not impact the analysis results since the emergency mode ventilation system filters are modeled with the same efficiency for all forms of iodine.
Westinghouse Non-Proprietary Class 3 Page 65 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-12 NRC Information Notice (IN) 2012-01: Seismic Considerations- Principally Issues Involving Tanks, dated January 26, 2012 (ADAMS Accession No. ML11292A175), provides examples and references to events in which licensees failed to recognize various seismic considerations and system alignment issues that could impact safety. The NRC staff has identified recent concerns about licensees who failed to recognize that aligning non-seismic piping to the RWST would require TS limiting condition for operation (LCO) action statement entry, system modifications, or license amendments.
RG 1.183, Regulatory Position 5.1.2, Credit for Engineered Safeguard Features, states:
Credit may be taken for accident mitigation features that are classified as safety related, are required to be operable by technical specifications, are powered by emergency power sources, and are either automatically actuated or, in limited cases, have actuation requirements explicitly addressed in emergency operating procedures. The single active component failure that results in the most limiting radiological consequences should be assumed. Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences.
During operations at WCGS, please describe if there is any non-seismic piping aligned to the RWST or systems which recirculate post-LOCA sump fluid. Also, provide details of how the AST analysis is modeled. Please provide enough detail so the NRC staff can independently model this configuration to assess its impact on design basis accident (DBA) doses.
Response
WCNOC review of NRC Information Notice 2012-01 and associated operating experience is documented in the corrective action program. The Refueling Water Storage Tank (RWST) is described in USAR Section 6.3 as part of the Emergency Core Cooling System (ECCS) and its ECCS function is to contain an inventory of 394,000 gallons of water to provide for ECCS mitigation purposes. The RWST is required to perform its function during the short term (less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) following a LOCA, MSLB, SGTR or any other accident requiring the ECCS following receipt of an safety injection (SI) signal. The ECCS consists of three separate subsystems: centrifugal charging (high head), SI (intermediate head), and residual heat removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains.
A review of the interface piping connections with the RWST was performed. There are eight interface piping connections to the seismic category 1 piping of the RWST excluding the piping connections from RWST supply header to the suctions of the ECCS pumps and Containment Spray (CS) System pumps. The interface piping connections from RWST supply header to the suctions of the ECCS pumps and CS System pumps are seismic category 1 piping connections.
Westinghouse Non-Proprietary Class 3 Page 66 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The remaining eight interface piping connections on the RWST were reviewed for seismic to non-seismic piping alignment concerns as follows:
- 1. RWST to Spent Fuel Pool Cooling and Cleanup System Flow Path The Spent Fuel Cooling and Cleanup System consists of three subsystems: fuel pool cooling subsystem, fuel pool cleanup subsystem and fuel pool surface skimmer subsystem. The fuel pool cleanup subsystem is non-safety related and has no safety design basis. The fuel pool cleanup subsystem limits the fission and corrosion product concentrations in the refueling pool water, the transfer canal water, and the fuel storage pool water to permit operator access to the fuel storage area and for fuel handling operations.
WCGS has two safety related air operated isolation valves (BNHV8800A/B) in series on the safety related piping from the RWST before it is connected to the non-safety related piping of the fuel pool cleanup subsystem. By design, valves BNHV8800A/B receive a safety injection signal to initiate automatic closure. The valves receive independent and redundant auto closure signals (i.e., each valve is actuated from a different SI actuation train) ensuring these safety related isolation valves are closed when the RWST may be required to support the ECCS. The fail-safe position of these valves is closed (the air solenoid fails open on a loss of power). The design features of the interconnecting piping between the RWST and the fuel pool cleanup subsystem meet GDC 35, Emergency core cooling. This criterion is satisfied by these two valves in that the system safety function can be accomplished with the failure of one valve as the other valve will fulfill the isolation function. Following a design basis accident that requires the ECCS mitigation function, a safety injection signal is generated and results in the closure of BNHV8800A and BNHV8800B, if the RWST was in the purification alignment mode.
- 2. Safety Injection (SI) Pump Recirculation Flow Path to RWST The SI pumps are provided with a recirculation flow path to the top of the RWST to support pump operation. The SI pumps recirculation flow path to the RWST is not in use during normal plant operations as the SI pumps are in standby for ECCS mitigation. The SI pumps recirculation flow path is not used to bring the plant to a cold shutdown condition. The SI pumps recirculation flow path is typically used to support pump operation for pump testing during normal plant operations. The SI pump recirculation flow path to the RWST is seismically qualified piping (seismic category 1 and II/I Special Scope). II/I Special Scope (SS), is non-safety related piping that runs above safety related piping that has been supported in this area seismically so as not to degrade the function of the safety related piping below it. Piping with this qualification is designed and constructed to ensure that its failure will not reduce the functioning of a safety related component to an unacceptable safety level.
- 3. Containment Spray (CS) Pump Test Flow Path to RWST
Westinghouse Non-Proprietary Class 3 Page 67 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment The CS pumps are provided with a flow path to the top of the RWST to provide a recirculation flow path to test each pump or to support system testing. The CS pumps recirculation flow path to the RWST is seismically qualified piping (seismic category 1 and II/I SS).
- 4. Charging and Volume Control System (CVCS) Boric Acid Blending Tee Flow Path to RWST The CVCS boric acid blending tee flow path to the top of the RWST provides for the addition of blended boric acid solution to the RWST or a path for testing the boric acid transfer pumps. The CVCS boric acid blending tee flow path to the RWST is seismically qualified piping (seismic category 1 and II/I SS).
- 5. SI System Test Flow Path to RWST The SI System test flow path to the top of the RWST allows for the testing of ECCS check valves, supports depressurization of ECCS piping, and provides dampening of the RHR and SI pump start pressure transients. The non-safety related piping in the SI System test flow path is 3/4 inch piping. Due to this small line size, a break or crack in this size of line is not postulated and isolation is provided by the containment isolation valves.
- 6. RHR Pump Flow Path to RWST The RHR Pump flow path to the RWST is only used during refueling outages to transfer/drain the refueling pool water to the RWST. An isolation valve (BN8717) is normally locked closed and is opened under administrative controls when the plant is in a refueling outage. The RHR pump flow path to the RWST is seismic category 1 qualified piping.
- 7. RWST Drain Flow Path to Waste Holdup Tank The RWST can be drained via a flow path to the Waste Holdup Tank in the Radwaste Building.
This drain path is seismic category 1 qualified piping out to and including its isolation valve BNV0017 that is locked closed. Downstream of the isolation valve, the piping is classified as non-safety related. When the RWST is required to be drained for maintenance or internal inspection during refueling outages, this drain flow path is utilized.
- 8. RWST Overflow Flow Path to Waste Holdup Tank The RWST is provided with an overflow path at the top of the RWST that is also connected to the Waste Holdup Tank. The overflow piping is seismic category 1 qualified near the RWST and connects with non-safety related piping going to the Waste Holdup Tanks. There are no isolation valve(s) in the overflow flow path as it is designed to provide overflow protection for the RWST.
The piping aligned to systems which recirculate post-LOCA sump fluid is seismically qualified piping.
Westinghouse Non-Proprietary Class 3 Page 68 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Relevant Details of AST Analysis Model:
The AST analysis models RWST back-leakage at a rate of 3.8 gpm from the sump. The activity is modeled to be delivered directly to the gas filled portion of the RWST; however, only 10 percent of the activity becomes airborne and is available for release to the environment. The release rate from the RWST to the environment is based on the volume displacement from the incoming leakage. An adjustment to the release rate is made to account for a reduction in the RWST gas volume available for dilution as the leakage into the RWST increases the water level.
The adjusted flow rate is based on increasing the modeled flow in order to exhaust the appropriate activity concentration. This is necessary as RADTRAD maintains a constant compartment volume and transfers are concentration-based.
The RWST modeling description is included in Enclosure IV, Section 4.3.9.2.2.3 and also included as a response to ARCB-RAI-8 in Enclosures VI and VII.
ARCB-RAI-13 According to Enclosure VI, Table 4.3-5, Control Room and Control Room Building Parameters, the delay due to switch to emergency mode operation following receipt of isolation signal in the current licensing basis is stated as N/A or not applicable. The submittal proposes to change this delay to 60 seconds. An isolation set point for the R-23 detector is also provided.
Tables are provided for each accident with analysis parameters and assumptions (i.e., Tables 4.3-6 through Table 4.3-16.). These tables include either the time delay to switch to emergency mode of operation after event initiation, or the time delay to switch to emergency mode operation following receipt of isolation signal.
a) For each change to the current licensing bases, please provide details about how the total time to isolate was calculated.
b) Please state and justify the isolation signal assumed to initiate the switch to emergency mode.
c) Please state whether the time to isolate includes the worst case single failure, control room isolation signal time, emergency diesel generator startup time, time to load the electrical bus, and damper closure time. If not, please justify why these assumptions and delays are not considered.
d) Please describe how the R-23 detector is used to isolate the control room and for which accidents it is credited. Provide enough detail so that the NRC staff can independently calculate the isolation time credited.
e) Please state if the R-23 detector is a general area monitor or is it located in the control room HVAC ductwork.
Westinghouse Non-Proprietary Class 3 Page 69 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment f) Please state if the R-23 detector and initiation signal comply with Regulatory Position 5.1.2, Credit for Engineered Safeguard Features of RG 1.183 and explain how compliance with Regulatory Position 5.1.2 is met.
Response
The R-23 detector discussed above is hereafter referred to as the control room air supply radiation monitors (GKRE0004 and GKRE0005).
a) The total time to isolate was calculated by determining a time by which the isolation signal(s) would have been generated and then adding a delay to account for the Control Room Ventilation Isolation Signal (CRVIS) to actuate the Control Room Emergency Ventilation System (CREVS). Isolation signals include safety injection or high radiation in the control room air supply radiation monitors. To ensure that the time delay modeled in the analyses is bounding, new TS Surveillance Requirement (SR) 3.3.7.6 has been added for the performance of the required response time verification.
For the loss of coolant accident (LOCA) radiological consequences analysis, the safety injection isolation signal is assumed to occur immediately after event initiation. For conservatism, the switchover to emergency mode ventilation was not assumed to occur until 2 minutes after the onset of the event (including the 1 minute delay for switchover). This is presented in Enclosure IV of this LAR, Section 4.3.9.2.3.
For the main steam line break (MSLB) radiological consequences analysis, the safety injection isolation signal is assumed to occur immediately after event initiation. For conservatism, the switchover to emergency mode ventilation was not assumed to occur until 90 seconds after the onset of the event (including the 1 minute delay for switchover). This is presented in Enclosure IV of this LAR, Section 4.3.3.2.3.
For the rod ejection containment leakage case, the safety injection isolation signal is assumed to occur within 150 seconds of event initiation. The switchover to emergency mode ventilation was not assumed to occur until 210 seconds after the onset of the event (including the 1 minute delay for switchover). This is presented in Enclosure IV of this LAR, Section 4.3.6.2.3.
For the steam generator tube rupture (SGTR), rod ejection primary-to-secondary case, locked rotor, and fuel handling accident (FHA) radiological consequences analysis, the control room air supply radiation monitors setpoint is shown to be reached for each accident immediately after the onset of the event. This is based on the amount of Xe-133 activity initially in the primary system, the environmental release rate modeled for each analysis, and the atmospheric dispersion factor modeled for the control room.
Since the control room air supply radiation monitors are located in the control room HVAC ductwork, the concentration of activity entering the control room is the concentration seen by the monitors. The delay from the control room air supply
Westinghouse Non-Proprietary Class 3 Page 70 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment radiation monitors setpoint being reached to the switchover to emergency mode ventilation is 60 seconds. Each analysis conservatively assumed that switchover did not occur for at least 120 seconds. This is presented in Enclosure IV of this LAR, Sections 4.3.8.2.3, 4.3.6.2.3, 4.3.5.2.3, and 4.3.12.2.3.
b) For the LOCA, rod ejection containment leakage case, and MSLB analyses, a safety injection signal is credited. For LOCA and rod ejection containment leakage, the safety injection signal is generated by the reduction in RCS pressure as a result of the respective breaks. For MSLB, the safety injection signal is generated by the low steamline pressure as a result of the break. This is presented in Enclosure IV of this LAR, Sections 4.3.9.2.3, 4.3.6.2.3, and 4.3.3.2.3.
c) The CRVIS time, emergency diesel generator startup time, time to load the electrical bus and damper closure times have been considered and are bounded by the 60 second delay time assumed for each accident scenario as described in response to item a) above.
The worst case single failure of a control room filtration fan is considered and results in a pathway for contaminated air from the control building to bypass the control room filtration system filter adsorber unit and thereby enter the control room unfiltered. The mitigative action for this single failure is an operator action to isolate this flow path by closing the appropriate train-specific damper. This single failure is the same as that assumed in the current licensing basis analysis of record (AOR), where the operator action is assumed to take place 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after event initiation. Since the two CREVS trains are independent, this single failure only affects one train of the CREVS.
d) The control room air supply radiation monitors are credited to isolate the control room in the SGTR, rod ejection primary-to-secondary leakage, locked rotor, and FHA analyses. The activity concentration seen by the control room air supply radiation monitors setpoint is determined by the following equation:
Activity Concentration at Control Room Air Supply Radiation Monitors = Activity Conc
- Environment Release Rate
- CR /Q The following calculations determine the time to reach the setpoint for each accident.
Note that the analyses modeled a control room radiation monitor setpoint of 2.12E-3 Ci/m3 Xe-133.
Westinghouse Non-Proprietary Class 3 Page 71 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment SGTR Since the analysis assumes no delay in the release of activity and no delays in the transport of activity through the primary and secondary systems or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 4.19E4 Ci and the RCS mass is 3.99E5 lbm. The initial break flow rate is 3085 lbm/min and the 0-2 hour CR /Q for steaming releases from the SG for normal operation is 2.55E-2 sec/m3.
The activity concentration at the detector is:
(4.19E4 Ci / 3.99E5 lbm)(3085 lbm/min
- 1 min/60 sec)(2.55E-2 sec/m3) = 1.4E-1 Ci/m3 The control room radiation monitor is shown above to exceed its setpoint immediately.
This comparison is presented in Enclosure IV of this LAR, Section 4.3.8.2.3.
Note that Enclosure IV, Section 4.3.2.1 describes that isolation based solely on the control room radiation monitor does not fully switch the HVAC systems such that the control room unfiltered inleakage continues to be associated with the normal mode X/Qs. The SGTR analysis credits the safety injection signal due to low pressurizer pressure to complete isolation of the control room. From Enclosure I, Section 2.7.3, the safety injection signal occurs less than 6 minutes after the onset of the event. For conservatism, the isolation of the control room is completed at 10 minutes after the onset of the event.
Rod Ejection Primary-to-Secondary Leakage Since the analysis assumes no delay in the release of activity from the damaged fuel and no delays in the transport of activity through the primary and secondary systems or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 4.15E6 Ci and the RCS mass is 3.99E5 lbm. The primary to secondary leak rate is 8.342 lbm/min (applying a conservative analysis density of 62.4 lbm/ft3) and the 0-2 hour CR /Q for steaming releases from the SG for normal operation is 2.55E-2 sec/m3.
The activity concentration at the detector is:
(4.15E6 Ci / 3.99E5 lbm)(8.342 lbm/min
- 1 min/60 sec)(2.55E-2 sec/m3) = 3.7E-2 Ci/m3 The control room air supply radiation monitors are shown above to exceed the setpoint immediately.
This comparison is presented in Enclosure IV of this LAR, Section 4.3.6.2.3.
Westinghouse Non-Proprietary Class 3 Page 72 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Locked Rotor Since the analysis assumes no delay in the release of activity from the damaged fuel and no delays in the transport of activity through the primary and secondary systems or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 8.30E5 Ci and the RCS mass is 3.99E5 lbm. The primary to secondary leak rate is 8.342 lbm/min (applying a conservative analysis density of 62.4 lbm/ft3) and the 0-2 hour CR /Q for steaming releases from the SG for normal operation is 2.55E-2 sec/m3.
The activity concentration at the detector is:
(8.30E5 Ci / 3.99E5 lbm)(8.342 lbm/min
- 1 min/60 sec)(2.55E-2 sec/m3) = 7.4E-3 Ci/m3 The control room air supply radiation monitors are shown above to exceed the setpoint immediately.
This comparison is presented in Enclosure IV of this LAR, Section 4.3.5.2.3.
FHA Since the analysis assumes no delay in the release of activity from the damaged fuel and no delays in the transport of activity through the spent fuel pool or in transport from the release point to the air intake, an instantaneous generation of the high radiation signal could be assumed. The total Xe-133 activity is 1.62E5 Ci. Given the linear release over a 2-hour period, the release rate is 8.1E4 Ci/hr or 22.5 Ci/sec. The 0-2 hour CR /Q for releases is 1.38E-3 sec/m3. (Note that this is the /Q associated with releases from a FHA inside containment. It is selected for this check as the lower
/Q is conservative in determining when the setpoint is reached.) Therefore the concentration of Xe-133 at the control room air supply radiation monitors is:
(22.5 Ci/sec)(1.38E-3 sec/m3) = 3.11E-2 Ci/m3 The control room air supply radiation monitors are shown above to exceed the setpoint immediately.
This comparison is presented in Enclosure IV of this LAR, Section 4.3.12.2.3.
Westinghouse Non-Proprietary Class 3 Page 73 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment e) The control room air supply radiation monitors are located in the control room HVAC ductwork.
f) Yes, the control room air supply radiation monitors comply with Regulatory Position 5.1.2 of Regulatory Guide 1.183. Specifically, compliance is demonstrated by the following:
- 1) The two redundant radiation monitors and associated actuation instrumentation are safety-related.
- 2) The radiation monitors are required to be operable by TS limiting condition for operation (LCO) 3.3.7, CREVS Actuation Instrumentation.
- 3) The radiation monitors are powered by Class 1E electrical equipment.
- 4) The control room air supply radiation monitors provide input to initiate a CRVIS.
Proper actuation of the CREVS is verified in the emergency operating procedures.
ARCB-RAI-14 Please provide a diagram describing the model used for modeling the control building and control room for each design basis accident. Please provide the unfiltered in-leakage into the control building and justify the value. Please explain if this value has been confirmed by testing and if it will be confirmed periodically as part of a TS surveillance program. (Note this issue was previously identified in NRC Regulatory Issue Summary 2006-04, Issue Number 3.)
Response
The figure displayed below (obtained from Section 15A.3 of the Wolf Creek USAR), provides a diagram of the flow rates modeled for the control building and control room. The Normal Mode HVAC flow rates are modeled prior to CREVS actuation. Following the CREVS actuation, the Emergency Mode HVAC Prior to Operator Action provides the flow rates considering a single failure of a control room filtration fan, which is consistent with the analysis of record single failure. The Emergency Mode HVAC after Operator Action flow rates are modeled after the failed train has been isolated by closing the train-specific damper.
Westinghouse Non-Proprietary Class 3 Page 74 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Enclosure IV, Table 4.3-5 contains the flow rates input values used in the analysis. A value of 400 cfm was modeled as the inleakage into the control building. The inleakage value was increased from the analysis of record value of 300 cfm to allow additional testing margin. An acceptance criterion of 300 cfm was confirmed in October 2010 during performance of TS SR 3.7.10.4. Although the analysis acceptable inleakage increased, the previous TS surveillance test performance in accordance with the Control Room Envelope Habitability Program (TS 5.5.18) demonstrates that the control building inleakage is less than 400 cfm. The Frequency of SR 3.7.10.4 is also in accordance with the Control Room Envelope Habitability Program. This Frequency is procedurally controlled to 6 years.
ARCB-RAI-15 In the proposed TS Bases B 3.3.8, Emergency Exhaust System (EES) Actuation Instrumentation, which ensures that radioactive materials in the fuel building atmosphere are filtered and absorbed prior to exhausting to the environment, the EES is not credited after 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> of decay from the fuel. Without the EES credited, the leakage (source) could occur anywhere there is a penetration or hole in the fuel building (rather than through the exhaust of the EES). Please explain how the worst-case source-receptor pairings are determined for the calculation of atmospheric dispersion factors when the EES is not credited.
Response
In Reference 1, WCNOC proposed to adopt TSTF-51-A, Revision 2, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations. The adoption of TSTF A revised TS 3.3.8 by adding recently to the Applicability. The revised Applicability stated:
During movement of recently irradiated fuel assemblies in the fuel building. In the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Revision 2, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations. are being
Westinghouse Non-Proprietary Class 3 Page 75 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment withdrawn. Subsequently, WCNOC withdrew the license amendment request (Reference 1) in WCNOC letter dated June 18, 2014 (Reference 3). The resubmittal of the license amendment request excluded the adoption of TSTF-51-A, Revision 2. With the removal of changes proposed by TSTF-51-A, the Emergency Exhaust System (EES) actuation instrumentation is required to be OPERABLE during movement of irradiated fuel assemblies in the fuel building.
As such, if a FHA were to occur, upon receipt of a fuel building ventilation isolation signal generated by the gaseous radioactivity monitors in the fuel building exhaust line, normal air discharges from the building are terminated, the fuel building is isolated, and the stream of ventilation air discharges through the EES filter trains.
However, as described in Enclosure IV of this LAR, Section 4.3.12, while the EES is credited for the event and thus the release location will be the unit vent for a fuel handling accident within the fuel building, filtration from the EES was conservatively not credited for the FHA analysis as the dose limits were met without crediting EES filtration.
Updates to the TS Bases are presented in Enclosure IV of this LAR, page B 3.3.8-1 of Section 11.
ARCB-RAI-16 WCNOC assumes the control room does not isolate after a fuel handling accident. The normal unfiltered outside air makeup flow to the control building and the control room is 13,050 cubic feet per minute (cfm) and 1950 cfm, respectively. The unfiltered in-leakage to the control room is assumed to be 50 cfm.
a) Please justify the use of 50 cfm for unfiltered in-leakage in the configuration when the control room ventilation system is not isolated. Please state if this value has been confirmed by testing and will it be confirmed periodically as part of a TS surveillance program. (Note this issue was previously identified in NRC Regulatory Issue Summary 2006-04, Issue Number 3.)
b) In this case only, the normal makeup for the control building and control room is used to mitigate the consequences of the fuel handling accident. Please state if the normal control room heating and ventilation systems is credited meet the qualities, attributes, and performance criteria as described in RG 1.183, Regulatory Positions 4.2.4 and 5.1.2. (Note this issue was previously identified in NRC Regulatory Issue Summary 2006-04, Issue Number 3.) If so, please justify how the credited ventilation system complies with Regulatory Positions 4.2.4 and 5.1.2. If not, justify the proposed alternative assumptions used.
c) Please state if there is a surveillance for these normal makeup flow rates in the TSs. If not, please explain if a sensitivity analysis has been performed to determine the limiting dose based upon the range of possible makeup flow rates. Please provide the results (dose versus makeup flow rates) of any sensitivity analyses performed.
Westinghouse Non-Proprietary Class 3 Page 76 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment d) For other accidents that do not assume control room isolation is actuated (i.e., locked reactor coolant pump rotor, loss of alternating current power, letdown line break and tank ruptures) please explain if the assumptions used to model the control building or control room are the same as those used for the fuel handling accident. In your submittal for these accidents, where the control room isolation is not credited, the control room ventilation system is assumed to remain in the normal mode of operation.
Please state whether this assumption yields more conservative doses than if the emergency mode is assumed to be actuated and justify your answer. For these accidents (other than the fuel handling accident) that do not credit control room isolation, please explain if a sensitivity analysis has been performed to determine the limiting dose based upon a range of possible makeup flows. If so, please provide the results of the analysis (dose vs. makeup flow). If not, justify why the assumed makeup flows are limiting.
Response
In the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Rev. 2, are being withdrawn. As such, the normal HVAC lineup is not credited to mitigate the effects of a fuel handling accident. Specific responses are provided below:
a) The 50 cfm of unfiltered inleakage in the configuration when the control room ventilation system is in the normal mode of operation is consistent with the unfiltered inleakage in the configuration when the CREVS is in service. The 50 cfm is conservatively added to the 1950 cfm air makeup flow to the control room to yield a total flow rate of 2000 cfm of unfiltered atmosphere into the control room. Since the normal makeup flow rates are not credited to mitigate the consequence of a fuel handling accident there is not a TS Surveillance Requirement to measure the normal makeup flow rates and inleakage.
b) As stated in the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Rev. 2, are being withdrawn. As such, the normal makeup for the control building and control room is not credited to mitigate the consequences of a fuel handling accident.
c) The normal makeup flow rates are no longer credited to mitigate the consequence of a fuel handling accident. Thus, there is not a TS Surveillance Requirement for the normal makeup flow rates. A sensitivity analysis was not performed for the normal flow rates.
However, the actual normal makeup flow rates were measured and compared to the values modeled in the dose analyses. The normal makeup flow rates modeled in the dose analyses are greater than the measured plant flow rates for the control building and control room by more than 10%.
Westinghouse Non-Proprietary Class 3 Page 77 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment d) The following analyses only credit normal makeup flow rates to mitigate the consequences of the accident: loss of alternating current power, letdown line break, and tank ruptures. The assumptions used in these analyses are consistent with the assumptions for the normal mode of operation used in each accident analyzed for the AST implementation, as described in Section 4.3.2.1 and Table 4.3.5 of Enclosure IV of this LAR. Remaining in normal mode of operation is conservative for the accidents since the normal mode flow rates are significantly higher than the emergency mode flows and no filtration is credited for activity entering the control building or control room during normal mode. In addition, filtered recirculation is credited for activity removal in the control room during emergency mode of operation. Thus, the activity in the control room is greater for normal mode of operation.
The following wording is added to Enclosure IV of this LAR, Sections 4.3.4.2.3, 4.3.7.2.3, 4.3.10.2.3 and 4.3.11.2.3: This modeling is conservative since activity reduction due to filtration of inflow and filtered recirculation is not credited.
A sensitivity analysis was not performed for the normal flow rates. However, the actual normal makeup flow rates were measured and compared to the values modeled in the dose analyses. The normal makeup flow rates modeled in the dose analyses are greater than the measured plant flow rates for the control building and control room by more than 10%.
ARCB-RAI-17:
The current licensing basis for the radioactive waste gas decay tank failure, from Updated Final Safety Analysis Report (UFSAR) Section 15.7.1.2, states that the tank is assumed to fail after 40 years, releasing the peak inventory expected in the tank. The proposed change requests a change to this assumption. Please justify this change and explain why it is conservative.
Response
A detailed discussion of the waste gas decay tank failure calculation and justification for the assumption is provided in the response to ARCB-RAI-31 below.
Westinghouse Non-Proprietary Class 3 Page 78 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-18 Page 15.7-13 of the UFSAR "markup" states that the gap fractions are obtained for high burnup fuel from Regulatory Guide (RG) 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Safety Guide 25)," dated March 1972 (ADAMS Accession No. ML083300022), as modified by NUREG/CR-5009, "Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors," dated February 1988 (not publicly available), to support the conservative assumption that 100 percent of the rods do not meet the burnup and kilowatts per foot (kW/ft) limits set forth in Footnote 11 of RG 1.183. Enclosure VI, page 4-69 of the submittal elaborates on the use of NUREG/CR-5009.
a) Please justify the use and applicability of NUREG/CR-5009 instead of industry standards such as ANSI/ANS-5.4-2011, "Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel."
b) Please justify the statement made on page 3A-12 of the UFSAR markup that "Use of this regulatory guide [RG 1.25] has been replaced by Regulatory Guide 1.183 for alternative source term application," in light of the statement that the gap fractions are obtained from RG 1.25.
Response
a) The following table compares the nuclide gap fractions used in the WCGS Fuel Handling Accident analysis (which are obtained from RG 1.25, as modified by NUREG/CR-5009) versus those specified in Table 2.9 of PNNL-18212, Revision 1 (Update of Gap Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard, ADAMS Accession Number ML112070118), which calculated the gap release fractions utilizing ANSI/ANS-5.4-2011.
WCGS Analysis Calculated Gap Fractions Nuclide Gap Fractions Utilizing ANSI/ANS-5.4-2011 Kr-85 0.30 0.38 I-131 0.12 0.08 I-132 0.10 0.09 Other Noble Gases 0.10 0.08 Other Halogens 0.10 0.05
Westinghouse Non-Proprietary Class 3 Page 79 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment As shown in the table above, the gap fractions used in the WCGS analysis are conservatively higher than those specified in ANSI/ANS-5.4-2011, with the exception of Kr-85. Kr-85 is not a significant dose contributor when compared to the other nuclides being modeled. Kr-85 has an effective dose equivalent (EDE) dose conversion factor (DCF) of 1.19E-16 Sv-m3/Bq-sec and a released activity of 3.38E+03 Ci, whereas Xe-133 (a major dose contributor) has an EDE DCF of 1.56E-15 Sv-m3/Bq-sec and a released activity of 1.62E+05 Ci.
In the fuel handling accident radiological consequences analysis both release scenarios are considered, namely inside containment and inside the fuel building. For both release scenarios, Kr-85 contributes less than 0.03% of the total dose at each location (exclusion area boundary, low population zone, control room, and technical support center). The dose increase resulting from an increase of ~27% of Kr-85 activity (gap fraction of 0.30 vs 0.38), would be more than offset by decreasing the activity releases for other nuclides when modeling the ANSI/ANS-5.4-2011 gap fractions.
Therefore, the gap fractions used for WCGS, which are based on RG 1.25, as modified by NUREG/CR-5009, are conservative in comparison to the gap fractions using ANSI/ANS-5.4-2011.
b) The USAR text is revised to state that the use of RG 1.25 is being specified in Table 15B-1 [RG 1.183 Conformance Table].
ARCB-RAI-19 The current single failure taken for the fuel handling accident is the failure of the humidly control system for the engineered safety feature (ESF) emergency filtration system as stated on page 15.7-13 of the UFSAR. Please explain the worst case single failure that is assumed for the fuel handling accident in the proposed analysis. Please justify your answer.
Response
The current single failure of the humidity control system resulted in a reduced iodine removal efficiency. The fuel handing accident analysis, documented in Section 4.3.12 of Enclosure IV of this LAR, modeled a direct release of activity to the environment with no credit for the filtration of the EES. Therefore, by conservatively not crediting the EES filtration, the analysis results documented in Section 4.3.12 of Enclosure IV of this LAR are more limiting than if the previous single failure of the humidity control system had been retained and a reduced iodine removal efficiency was credited.
Consistent with the current control room radiological consequences calculation models, as discussed on page 15A-8 of the USAR, a failure of one of the filtration fans is assumed at the start of emergency mode of operation and a larger unfiltered inflow to the control room is assumed since only half of the makeup flow to the control room can pass through a filter. After a
Westinghouse Non-Proprietary Class 3 Page 80 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment defined time of 90 minutes, operator action isolates the failed train and reduces the unfiltered inflow to the control room.
ARCB-RAI-20 Enclosure VI, page 4-69 of the submittal states, in part, that Although not explicitly discussed, the specified overall DF [decontamination factor] also applies to rod internal pressures up to 1500 psig [pounds per square inch gauge].
The DF of 200 provided in RG 1.183 is based upon Reference B-1 ("Evaluation of Fission Product Release and Transport," dated October 5, 1971 (ADAMS Legacy Accession No.
8402080322)) of RG 1.183. The data upon which the pool DF of 200 is based was developed in 1971 and was based on the Westinghouse fuel marketed at the time (the assumed internal fuel pressure of 1200 psig was used). Since higher pressures correlate to lower DFs, the NRC staff is concerned that a DF of 200 might not be sufficiently conservative for pressures higher than 1200 psig.
Please provide the data for current fuel types used at WCNOC that justify a DF of 200 for fuel pressures up to 1500 psig. Also, please provide a detailed justification for using a DF of 200 for pressures up to 1500 psig.
Response
The current fuel type for WCNOC (17x17 RFA-2) is generically addressed for a DF of 200 at higher rod internal pressures by WCAP-16072-P-A (Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, ADAMS Accession Number ML042510053). WCAP-16072-P-A was submitted to the NRC for review and approval, and included an evaluation of iodine decontamination factors for fuel rod pressures up to 1500 psig. In the Final Safety Evaluation for WCAP-16072-P-A (ADAMS Accession Number ML041270102), the NRC stated that, the staff has determined that there is reasonable assurance that fuel rod design pressures of up to 1500 psig will not invalidate analysis assumptions related to iodine decontamination. The staff has also determined that this conclusion remains valid for the decontamination factor of 200 provided in RG 1.183 and RG 1.195, which supersede SG 25 for alternative source terms and TID14844 source terms, respectively.
WCAP-16072-P-A was prepared for CE NSSS plant fuel designs. However, the justification for the continued applicability of a DF of 200 was based on evaluations performed in WCAP-7518-L Radiological Consequences of a Fuel Handling Accident which did not distinguish between fuel types. WCAP-7518-L provides a method for calculating an iodine DF given a bubble rise time which is dependent on rod internal pressures. This method is not fuel type specific and was used and discussed in WCAP-16072-P-A to justify the continued applicability of a DF value of
Westinghouse Non-Proprietary Class 3 Page 81 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment 200. Therefore, the justification provided in WCAP-16072-P-A is applicable to all Westinghouse fuel types.
It should be noted that Regulatory Guide 1.183 refers to NRC Staff Technical Paper Evaluation of Fission Product Release and Transport for a Fuel Handling Accident (ADAMS Accession Number 8402080322). WCAP-7518-L is discussed in the NRC Staff Technical Paper as being submitted to the NRC to present new experimental data pertaining to the consequences of a refueling accident.
Also, the NRC has previously approved the use of a DF of 200 for fuel pressures up to 1500 psig (reference Indian Point Nuclear Unit No. 3 - Issuance of Amendment Re: Full Scope Adoption of Alternative Source Term, ADAMS Accession Number ML050750431). In the NRC approval of this amendment, the NRC noted that the DF for elemental iodine would remain above 400 for fuel pressures at 1500 psig. This is greater than the DF of 285 for elemental iodine used in the analysis to obtain an overall iodine DF of 200. Therefore, the use of a DF of 200 remains conservative and is appropriate for use.
ARCB-RAI-21 , page 4-69 of the submittal states: The decay time used in determining the inventory of the damaged rods is 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br />. Thus, the analysis supports the TS limit [emphasis added] of 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> decay time prior to fuel movement. The staff agrees that the decay time appears to meet Criterion 2 of 10 CFR 50.36, Technical Specifications, but could not locate the TS limit of 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br />. Please state where in the TSs this limit exists.
Response
The sentence on page 4-69 of Enclosure VI of Reference 1 is in error. The 76 hour8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> decay time prior to fuel movement limit is not specified in the Technical Specifications. In the conversion to the improved Technical Specifications (Reference 6), discussion of change (DOC) 3-01-R proposed the relocation of specification 3/4.9.3, Decay Time, to a licensee controlled document. The NRC approved the improved Technical Specifications in License Amendment No. 123 (Reference 7). The safety evaluation for License Amendment No. 123 states, in part:
The requirements in CTS 3/4.9.3 on the decay time that the reactor core must be subcritical before there is movement of irradiated fuel in the reactor core are being relocated to the TRM. This LCO requires the reactor to be subcritical for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to allow the radioactive decay of the short-lived fission products. The screening criteria for including the requirements in the ITS have been satisfied for Criterion 2 since decay time is consistent with the assumptions used in an accident analysis; however, the activities necessary to be performed at WCGS before commencing movement of irradiated fuel
Westinghouse Non-Proprietary Class 3 Page 82 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ensure that 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of subcriticality will elapse before there is movement of irradiated fuel in the core. Therefore, because the CTS is not required to assure that 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> have elapsed prior to fuel movement, the decay time LCO and SRs in the CTS may be relocated to the TRM, a licensee-controlled document outside TS. The TRM is included by reference in the USAR and is an acceptable licensee-controlled document; therefore, the relocation is acceptable.
The two relocated specifications from the CTS discussed above are not required to be in the ITS because they do not fall within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii). They are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety. In addition, the NRC staff finds that sufficient regulatory controls exist under 10 CFR 50.59 to maintain the effect of the provisions in these specifications. The NRC staff has concluded that appropriate controls have been established for all of the current specifications that are being moved to the USAR.
The decay time requirement was relocated to USAR Section 9.1.4.2.3. Additionally, the TS 3.9.7 Bases, Refueling Pool Water Level, discussed the minimum decay time prior to fuel handling.
ARCB-RAI-22:
Please confirm that with the exception of different release points, the assumptions and inputs are identical for the fuel handling accident within the containment and the fuel handling accident outside the containment.
Response
The assumptions and inputs for the fuel handling accident within containment and outside of containment are identical with the exception of control room and technical support center atmospheric dispersion factors associated with the different release points.
ARCB-RAI-23:
a) Please confirm that the most limiting combination of release point and receptor for the control room and technical support center (TSC) were used to determine atmospheric dispersion factors for each accident.
b) Please explain why the control room atmospheric dispersion factors provided in the tables for individual accidents (i.e., Tables 4.3-11 and 4.3-15) do not correlate to values provide in Table 4.1.2-3.
c) Please state and justify the release points that correlate to the atmospheric dispersion factors for each design basis accident.
Westinghouse Non-Proprietary Class 3 Page 83 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment
Response
a) Based on the NRC question, WCNOC re-reviewed the release point and receptor combinations and determined more limiting pairs should be used. Two additional release points were determined to need atmospheric dispersion factor calculations: (1) release from a reactor building wall diffuse area source and (2) release from the radwaste building.
In addition, it was also discovered during the review (of release point and receptor combinations) that atmospheric dispersion factors needed to be determined for one additional receptor: the control room normal HVAC intake.
Due to the new atmospheric dispersion factor calculations, four of the presented accident and transport path combinations were determined to not be using the most limiting combination of release point and receptor. The affected accident and transport path combinations are: (1) rod ejection accident containment leakage transport path to the control room emergency intake vent, (2) loss of coolant accident containment leakage transport path to the control room emergency intake vent, (3) waste gas decay tank failure transport path to the control room emergency intake vent and TSC intake vent, and (4) liquid waste tank failure transport path to the control room emergency intake vent and TSC intake vent. These accident and transport path combinations have been updated and are included in the response to item c) below, and also contained in Enclosure IV of this LAR, Section 4.1.2.4 and Table 4.1.2-5.
b) The updated (refer to response a) above) Table 4.1.2-3 and updated individual accident tables (Table 4.3-6 through 4.3-15) have been included in Enclosure IV of this LAR, and have been confirmed to contain the same information. The updated version of the Table 4.1.2-3, which has been renumbered to 4.1.2-3(a), is included below for reference.
Westinghouse Non-Proprietary Class 3 Page 84 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.1.2-3(a) Calculated /Q (sec/m3) for the Emergency Control Room Intake Vent Equipment Hatch to Emergency Control Room Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5.44E-04 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.35E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.62E-04 1 to 4 days 1.22E-04 4 to 30 days 8.70E-05 Unit Vent Stack to Emergency Control Room Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.12E-04 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.38E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.79E-04 1 to 4 days 1.14E-04 4 to 30 days 8.94E-05 MSSVs/ARVs to Emergency Control Room Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.04E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.46E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.03E-04 1 to 4 days 1.90E-04 4 to 30 days 1.39E-04 RWST Vent to Emergency Control Room Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.80E-04 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.19E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.27E-04 1 to 4 days 1.96E-04 4 to 30 days 1.53E-04 TDAFW Exhaust to Emergency Control Room Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5.17E-04 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.99E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.60E-04 1 to 4 days 1.00E-04 4 to 30 days 7.21E-05 Radwaste Building to Emergency Control Room Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.92E-04 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.06E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.39E-04 1 to 4 days 2.00E-04 4 to 30 days 1.47E-04 Reactor Building Wall to Emergency Control Room Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 6.02E-04 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 4.35E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.81E-04 1 to 4 days 1.29E-04 4 to 30 days 9.65E-05
Westinghouse Non-Proprietary Class 3 Page 85 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment c) Enclosure IV of this LAR, Table 4.1.2-5, included for reference below, provides the release points considered in the calculation of the /Q values for each transport path of each accident. The table is structured according to the accident type and associated transport paths. The accident types include (1) main steam line break (two transport paths), (2) loss of non-emergency AC power (one transport path), (3) locked rotor (one transport path), (4) rod ejection (two transport paths), (5) letdown line break (one transport path), (6) steam generator tube rupture (one transport path), (7) loss of coolant accident (four transport paths), (8) waste gas decay tank failure (one transport path), (9) liquid waste tank failure (one transport path), and (10) fuel handling accident (two transport paths). Some accidents have multiple transport paths. The /Q values assigned to a given accident depend on the transport paths associated with that accident. Accidents with the same transport path are assigned the same /Q values.
As explained in Enclosure IV of this LAR, Section 4.1.2.4, for each accident, transport paths and release points are identified. Transport paths refer to paths such as primary-to-secondary leakage, primary-to-secondary break, containment leakage, ECCS leakage, RWST back-leakage, containment purge, direct release to auxiliary building, and direct release to atmosphere. The release points refer to the release points whose /Q values for CR and TSC intakes have been determined. These release points include the reactor building equipment hatch, reactor building wall diffuse source, unit vent stack, TDAFW exhaust vent, MSSV/ARV vents, RWST vent, and radwaste building. If a transport path has a single release point, the /Q associated with that release point is the limiting value. Some transport paths may have multiple release points. For such cases, the limiting /Q values are assumed to be the maximum of all release points associated with that transport path and release period. A site diagram with the considered release points and receptor points labeled is shown in Figure 4.1.2-1 below, and contained in Enclosure IV of this LAR, Section 4.1.2.
The containment leakage path considers three release points (equipment hatch, unit vent stack, and diffuse reactor building wall). For the containment leakage path, the limiting /Q values applied to the emergency control room intake are from the unit vent stack for a period from 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and from the reactor building wall for a period from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 30 days. The limiting /Q values for the normal control room intake are from the unit vent stack for all periods. The limiting
/Q values for the technical support intake are from the equipment hatch for all periods.
The primary-to-secondary leakage to the intact SGs is another example of a transport path with multiple release points. The path through the intact SGs leads to the release points at the MSSV/ARV vents and the TDAFW exhaust vent. For this case, the limiting /Q values are assigned to the larger values of the two release points.
Westinghouse Non-Proprietary Class 3 Page 86 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment For the main steam line break accident, the primary-to-secondary leakage to the intact SGs and the faulted SG are the two transport paths that lead to different release points to the environment. The path through the intact SGs leads to the release points at the MSSV/ARV vents and the TDAFW exhaust vent. The limiting /Q values for the intact SG release path are assumed the larger values of the two release points. However, the path through the faulted SG releases radionuclides into the auxiliary building if a steam line break occurs outside containment. The main steam line break is assumed to occur outside of containment because a break outside of the containment will bound any break inside containment. The radionuclides released from the break are drawn by the building ventilation system to the unit vent stack and eventually exhausted to the environment through the unit vent stack. Hence, the /Q values for the broken steam line release path are the values for the unit vent.
For the loss of non-emergency AC power accident, the primary-to-secondary leakage to intact SGs transport path leads to two release points at the MSSV/ARV vents and the TDAFW exhaust vent. The /Q values for this release path are assumed as the larger value of the two release points.
For the locked rotor accident, the primary-to-secondary leakage to SGs transport path leads to two release points at the MSSV/ARV vents and the TDAFW exhaust vent. The /Q values for this release path are assumed as the larger value of the two release points.
For the rod ejection accident there are two transport paths, containment leakage to the environment, and primary to secondary leakage. The /Q values for the containment leakage release path are assumed as the maximum value of the three release points: (1) equipment hatch, (2) reactor building wall, and (3) unit vent stack. The /Q values for the primary to secondary leakage release path are assumed as the larger value of the MSSV/ARV vents and the TDAFW exhaust vent release points.
For the letdown line break accident, the line break is assumed to occur outside of containment and radionuclides are directly released into the auxiliary building. The /Q values for this case are those of the unit vent stack.
For the steam generator tube rupture accident, the primary-to-secondary break to the ruptured SG is the major transport path to the environment. The /Q values for the ruptured SG release path are assumed the larger value of the MSSV/ARV vents and the TDAFW exhaust vent.
For the loss of coolant accident, there are four transport paths. The first path is the containment leakage transport path to the environment. The /Q values for this transport path are assumed as the maximum value of the three release points: (1) equipment hatch, (2) reactor building wall, and (3) unit vent stack. The second transport path is the ECCS leakage path. The /Q values for this path are those of the unit vent stack. The third transport path is the RWST back-leakage path. The /Q values for this path are of the RWST vent. The fourth transport path is the containment purge path. The containment purge system vents the containment atmosphere
Westinghouse Non-Proprietary Class 3 Page 87 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment through the containment exhaust penetrations up to the unit vent stack. Therefore, the /Q values for this path are those of the unit vent stack.
For the waste gas decay tank failure, direct release from the radwaste building is the only credited transport path to the environment. The /Q values for this case are those of the radwaste building.
For the liquid waste tank failure, direct release from the radwaste building is the only credited transport path to the environment. The /Q values for this case are those of the radwaste building.
For the fuel handling accident, two possible accident types are considered: (1) in containment and (2) in fuel building. For a fuel handling accident occurring in containment, the transport path is through the open equipment hatch. For a fuel handling accident occurring in the fuel building, the radionuclides released are drawn by the building ventilation system to the unit vent stack and eventually exhausted to the environment through the unit vent stack.
Westinghouse Non-Proprietary Class 3 Page 88 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment
Westinghouse Non-Proprietary Class 3 Page 89 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Figure 4.1.2-1. Diagram of Source and Receptor Locations for Wolf Creek Table 4.1.2-5 Accident Release Sources Accident Transport Path Release Points Limiting /Q Main Steamline Primary to Secondary Leakage to Intact MSSVs/ARVs Maximum of all Break SG Associated TDAFW Exhaust Release Points Primary to Secondary Leakage to Unit Vent Unit Vent Faulted SG for a Break Outside of Containment Primary to Secondary Leakage to Equipment Hatch Maximum of all Faulted SG for a Break Inside of Associated Unit Vent Containment Release Points (This transport path is not used Reactor Building because the break outside of Wall (Diffuse containment is bounding) Source)
Loss of Primary to Secondary Leakage MSSVs/ARVs Maximum of all Non-Emergency Associated TDAFW Exhaust AC Power Release Points Locked Rotor Primary to Secondary Leakage MSSVs/ARVs Maximum of all Associated TDAFW Exhaust Release Points Rod Ejection Containment Leakage Equipment Hatch Maximum of all Associated Unit Vent Release Points Reactor Building Wall (Diffuse Source)
Primary to Secondary Leakage MSSVs/ARVs Maximum of all Associated TDAFW Exhaust Release Points Letdown Line Release Outside of Containment Unit Vent Unit Vent Break Steam Generator Primary to Secondary Break MSSVs/ARVs Maximum of all Tube Rupture Associated TDAFW Exhaust Release Points
Westinghouse Non-Proprietary Class 3 Page 90 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4.1.2-5 Accident Release Sources (cont.)
Accident Transport Path Release Points Limiting /Q Loss of Coolant Containment Leakage Equipment Hatch Maximum of all Accident Associated Unit Vent Release Points Reactor Building Wall (Diffuse Source)
ECCS Leakage Unit Vent Unit Vent RWST Back-Leakage RWST Vent RWST Vent Containment Purge Unit Vent Unit Vent Waste Gas Release to Atmosphere Radwaste Building Radwaste Decay Tank Building Failure Liquid Waste Release to Atmosphere Radwaste Building Radwaste Tank Failure Building Fuel Handling Release In Containment Equipment Hatch Equipment Hatch Accident Release In Fuel Building Unit Vent Unit Vent ARCB-RAI-24 In a letter dated November 7, 2013 (ADAMS Accession No. ML13246A358), the NRC informed the Technical Specifications Task Force of concerns that the NRC staff had recently identified during a review of plant-specific license amendments requesting adoption of three travelers including traveler TSTF-51, Revision 2, Revise Containment Requirements during Handling Irradiated Fuel and Core Alterations.
Enclosure VI, page 2-4 of the submittal discusses the proposed TSTF-51 changes. TSTF-51 states, in part, that The addition of the term recently associated with handling irradiated fuel in all of the containment function Technical Specification requirements is only applicable to those licensees who have demonstrated by analysis that after sufficient radioactive decay has occurred, off-site doses resulting from a fuel handling accident remain below the Standard Review Plan limits (well within 10CFR100). [or 10 CFR 50.67]
Westinghouse Non-Proprietary Class 3 Page 91 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment NUREG-0800, Standard Review Plan (SRP) 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, July 2000 (ADAMS Accession No. ML003734190), states, in part, that The models, assumptions, and parameter inputs used by the licensee should be reviewed to ensure that the conservative design basis assumptions outlined in RG-1.183 have been incorporated.
Appendix B of RG 1.183, Regulatory Position 1.1 states:
The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or weight of a dropped fuel assembly ...
With regard to the WCGS submittal to adopt TSTF-51, please provide plant-specific information to verify that the limiting cases have been considered, e.g., a fuel handling accident analysis that evaluates the dropping of loads allowed over irradiated fuel assemblies (i.e., new fuel assembly, sources, or reactivity control components) onto irradiated fuel assemblies prior to and after the proposed 76-hour decay time. Such an analysis should only credit those safety systems required to be operable as required by TS. There must be reasonable assurance that the fuel handling accident analysis doses remain within regulatory limits when references to Core Alterations are removed from TSs and Engineered Safety Features are no longer required during movement of loads such as new fuel assemblies, sources or reactivity control components.
Response
Reference 1 included the adoption of TSTF-51-A, Revision 2. WCNOC became aware of the NRC concerns with TSTF-51 in early October 2013 after Reference 1 had been submitted to the NRC. With the subsequent issuance of a letter dated November 7, 2013 (Reference 8),
WCNOC communicated with the NRC Project Manager on November 14, 2013 by electronic mail that consideration would be given to withdrawing the changes proposed by TSTF-51-A depending on the significance of the impact to Reference 1. WCNOC withdrew the license amendment request (Reference 1) by WCNOC letter dated June 18, 2014 (Reference 3). The resubmittal of the license amendment request excludes the adoption to TSTF-51, Revision 2.
Westinghouse Non-Proprietary Class 3 Page 92 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-25 Several calculations assume an iodine partition in the steam generators of 100 which is applied to the releases resulting from steaming of secondary side fluid (i.e., Enclosure VI, page 4-59). Please confirm that the partition factor is applied only to the elemental iodine or justify how organic iodine is partitioned.
Response
Regulatory Position 5.5.4 in Appendix E of Regulatory Guide 1.183, which specifies the iodine partition of 100 in the steam generators, does not distinguish between elemental and organic iodine. As a result, the partition factor of 100 was applied to both forms of iodine being released from the steam generators for all applicable analyses. The NRC approved this modeling of partitioning on iodine releases from the steam generators in previous AST submittals (reference Point Beach Nuclear Plant (PBNP), Units 1 and 2 -Issuance of License Amendments Regarding Use of Alternate Source Term (TAC Nos. ME0219 and ME0220), ADAMS Accession Number ML110240054).
ARCB-RAI-26 The column labeled "Comments" of Enclosure VI, Table A for Regulatory Position 4.1.1, states that progeny was not included in the dose calculations consistent with the two previously approved submittals. The NRC staffs review of the subject safety evaluations of these two submittals did not find explicit approval or a review of excluding the effects of progeny. Since this appears to conflict with Regulatory Position 4.1.1 and could potentially yield a non-conservative estimate of doses, please either provide justification for not including the progeny, or include progeny in these calculations.
Response
The modeling of progeny has been included in the updated results presented in Enclosure IV of this LAR, Section 4.3. The comments in Enclosure IV of this LAR, Table A for Regulatory Position 4.1.1 have been updated to the following response:
The dose calculations determine the TEDE and consider all radionuclides, including progeny from the decay of parent radionuclides, that are significant with regard to dose consequences.
Westinghouse Non-Proprietary Class 3 Page 93 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-27 For those accidents that model accident and pre-existing spikes, RG 1.183 specifies the release from the fuel to be in the form of 95 percent cesium iodide. Please confirm that the spike modeled in the reactor coolant system models the increase in cesium in addition to the iodine modeled.
Response
A spike in cesium associated with the iodine spikes (both pre-accident and accident-initiated) in the WCGS radiological consequences analyses (main steamline break, steam generator tube rupture, loss of AC power, and letdown line break) was not explicitly modeled, consistent with methods used in previously approved AST submittals (e.g., Point Beach Units 1 and 2 AST Approval, April 14, 2011 under ADAMS Accession Number ML110240054). However the initial cesium activity modeled in the RCS bounds the cesium which would be released to the RCS in association with iodine spiking.
The cesium activity initially modeled in the RCS for the above accident analyses was that associated with the design basis fuel defect level of 1%. The initial RCS iodine activity was scaled down to the Technical Specification 3.4.16, RCS Specific Activity, limit of 1.0 µCi/gm dose equivalent (DE) I-131 as identified in Enclosure IV of this LAR, Section 4.3 and explained further in the response to ARCB-RAI 37. As shown in the response to ARCB-RAI 37, the total isotopic dose equivalence for the iodines considered is 4.22, which means that the 1% fuel defect iodine activity is reduced by a factor of 4.22 to be at the Technical Specification 3.4.16 limit of 1.0 µCi/gm DE I-131. Therefore, the fuel defect level corresponding to the Technical Specification limit is approximately 0.24% fuel defects.
Tables 1 and 2 convert the fuel defect level activities to corresponding nuclide masses.
The RCS mass used in the calculation is the mass used in the dose calculations (1.81E+08 gm). Note that the calculations were performed in a spreadsheet and additional decimal places were not included in the tables for ease of presentation.
A sample calculation for Cs-134 is as follows:
RCS Activity (Ci) = 1% Fuel Defect Activity
- RCS Mass
- 1 Ci / 1E6 µCi
= 4.82E+00 µCi/gm
- 1.81E+08 gm
- 3.7E+10 Bq/Ci
= 8.72E+02 Ci
- 3.7E+10 Bq/Ci = 3.23E+13 Bq Number of atoms = RCS Activity (Bq)
- Half-life (sec) / ln(2)
= 3.23E+13 Bq
- 6.50E+07 sec / ln(2) = 3.03E+21 atoms Mass (gm) = Number of atoms
- Atomic Mass Number (gm/mole) / 6.022E+23 (atoms/mole)
= 3.03E+21 atoms
- 134 gm/mole / 6.022E+23 atoms/mole = 6.74E-01 gm
Westinghouse Non-Proprietary Class 3 Page 94 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 1 Cesium Mass at 1% Fuel Defect Level 1% Fuel Defect Activity RCS RCS Activity Half-life Number of Nuclide (µCi/gm) Activity (Ci) (Bq) (sec) atoms Mass (gm)
Cs-134 4.82E+00 8.72E+02 3.23E+13 6.50E+07 3.03E+21 6.74E-01 Cs-136 4.35E+00 7.87E+02 2.91E+13 1.13E+06 4.75E+19 1.07E-02 Cs-137 2.68E+00 4.85E+02 1.79E+13 9.46E+08 2.45E+22 5.57E+00 Cs-138 1.16E+00 2.10E+02 7.77E+12 1.93E+03 2.16E+16 4.96E-06 Total: 6.26E+00 Table 2 Cesium Mass at Technical Specification Limit Fuel Defect Level 0.24% Fuel Defect Activity RCS RCS Half-life Number of Nuclide (µCi/gm) Activity (Ci) Activity (Bq) (sec) atoms Mass (gm)
Cs-134 1.16E+00 2.09E+02 7.75E+12 6.50E+07 7.26E+20 1.62E-01 Cs-136 1.04E+00 1.89E+02 6.99E+12 1.13E+06 1.14E+19 2.57E-03 Cs-137 6.43E-01 1.16E+02 4.31E+12 9.46E+08 5.88E+21 1.34E+00 Cs-138 2.78E-01 5.04E+01 1.86E+12 1.93E+03 5.19E+15 1.19E-06 Total: 1.50E+00 As shown, the amount of excess cesium mass in modeling 1% fuel defects as opposed to the fuel defects of 0.24% is 4.76 gm. The amount of excess was determined for the radiocesium nuclides considered in the dose analyses (Cs-134, Cs-136, Cs-137, and Cs-138).
For the pre-accident spike scenario, the total iodine mass modeled in the RCS at the spike limit of 60 µCi/gm DE I-131 is calculated. The mass contributions of both I-127 and I-129, as well as the mass contributions of iodine modeled in the dose analyses (I-130 through I-135), are calculated. Since I-127 is stable, its contribution to 1% fuel defects is given in units of grams of I-127 per gram of RCS coolant.
Table 3 converts the fuel defect level activities to corresponding nuclide masses. The pre-accident spike activity level used is 60 µCi/gm DE I-131 and the RCS mass used in the calculation is the mass used in the dose calculations (1.81E8 gm). The calculations were performed in a spreadsheet and additional decimal places were not included in the tables for ease of presentation.
Westinghouse Non-Proprietary Class 3 Page 95 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 3 Iodine Mass in RCS for Pre-Accident Iodine Spike Isotopic Activity at Activity Scaled Isotopic Isotopic Isotopic 1.0% Fuel to 1.0 µCi/gm Activity at Activity at Activity at Defect Level DE I-131 1.0 µCi/gm DE 60 µCi/gm 60 µCi/gm DE I- Half-life Number of Nuclide (µCi/gm) (µCi/gm) I-131 (Ci) DE I-131 (Ci) 131 (Bq) (sec) atoms Mass (gm)
I-127 1.24E-10 2.94E-11 5.32E-03 (gm) 3.19E-01 (gm) N/A N/A N/A 3.19E-01 (gm/gm) (gm/gm)
I-129 7.17E-08 1.70E-08 3.07E-06 1.84E-04 6.82E+06 4.95E+14 4.87E+21 1.04E+00 I-130 4.65E-02 1.10E-02 1.99E+00 1.20E+02 4.43E+12 4.45E+04 2.84E+17 6.13E-05 I-131 3.28E+00 7.77E-01 1.41E+02 8.44E+03 3.12E+14 6.95E+05 3.13E+20 6.81E-02 I-132 3.39E+00 8.03E-01 1.45E+02 8.72E+03 3.23E+14 8.28E+03 3.85E+18 8.45E-04 I-133 5.04E+00 1.19E+00 2.16E+02 1.30E+04 4.80E+14 7.49E+04 5.18E+19 1.14E-02 I-134 7.30E-01 1.73E-01 3.13E+01 1.88E+03 6.95E+13 3.16E+03 3.17E+17 7.05E-05 I-135 2.85E+00 6.75E-01 1.22E+02 7.33E+03 2.71E+14 2.38E+04 9.31E+18 2.09E-03 Total 1.45E+00
Westinghouse Non-Proprietary Class 3 Page 96 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Since the atomic masses of cesium and iodine are similar and 95% of the iodine is in particulate form (i.e., CsI), then the total cesium which would be modeled by a pre-accident iodine spike is 1.37 gm (1.45 gm of iodine
- 0.95 fraction of CsI to total iodine), which is less than the excess radiocesium initially modeled in the RCS.
For the accident-initiated spike scenario, the total iodine mass in the gap of the defective fuel rods is calculated. The mass contributions of both I-127 and I-129, as well as the mass contributions of iodine modeled in the dose analyses (I-130 through I-135), are calculated. Since I-127 is stable, its contribution in the core is given in units of grams of I-127.
Table 4 converts the iodine activity available in the core to corresponding nuclide masses that are available for release from the fuel gap in the defective fuel rods. The fuel gap fractions of 12% for I-131 and 10% for the remaining iodines are used and are conservative gap fractions based on those used in the fuel handling accident analysis. Also, the fuel defect level of 0.24%
(described above) is used. Note that the calculations were performed in a spreadsheet and additional decimal places were not included in the tables for ease of presentation.
Westinghouse Non-Proprietary Class 3 Page 97 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Table 4 Iodine Mass Available for Release from Fuel Gap in Accident-Initiated Iodine Spike Activity Available Activity Activity Activity Fuel in Available Gap Available In Available in Defect Defective Half Life Number Nuclide In Core (Ci) Fraction Gap (Ci) Gap (Bq) level (%) Fuel (Bq) (sec) of atoms Mass (gm)
I-127 4.31E+03 0.1 4.31E+02 N/A 0.24 1.03E+00 N/A N/A 1.03E+00 (gm) (gm) (gm)
I-129 2.57E+00 0.1 2.57E-01 9.51E+09 0.24 2.28E+07 4.95E+14 1.63E+22 3.49E+00 I-130 1.98E+06 0.1 1.98E+05 7.33E+15 0.24 1.76E+13 4.45E+04 1.13E+18 2.44E-04 I-131 1.01E+08 0.12 1.21E+07 4.48E+17 0.24 1.08E+15 6.95E+05 1.08E+21 2.35E-01 I-132 1.49E+08 0.1 1.49E+07 5.51E+17 0.24 1.32E+15 8.28E+03 1.58E+19 3.46E-03 I-133 2.10E+08 0.1 2.10E+07 7.77E+17 0.24 1.86E+15 7.49E+04 2.02E+20 4.45E-02 I-134 2.36E+08 0.1 2.36E+07 8.73E+17 0.24 2.10E+15 3.16E+03 9.55E+18 2.13E-03 I-135 2.00E+08 0.1 2.00E+07 7.40E+17 0.24 1.78E+15 2.38E+04 6.10E+19 1.37E-02 Total: 4.82E+00
Westinghouse Non-Proprietary Class 3 Page 98 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Since the atomic masses of cesium and iodine are similar and 95% of the iodine is in particulate form (i.e., CsI), then the total cesium which would be modeled by an accident-initiated iodine spike is 4.58 gm (4.82 gm of iodine
- 0.95 fraction of CsI to total iodine), which is less than the excess radiocesium modeled in the RCS.
Therefore, the dose impact of modeling the design fuel defect level cesium activity in the RCS instead of scaling to the fuel defect level activity bounds the impact of modeling an increase in radiocesium resulting from iodine spiking.
It should be noted that the core composition of cesium isotopes includes approximately 40%
Cs-133 which is a stable nuclide and has no dose impact. Assuming that the core composition of cesium is similar to that released from the fuel gap to the RCS as CsI, then the total radiocesium which would be released during an iodine spike is approximately 60% of that in the calculations above. Therefore, the calculations above are conservative.
ARCB-RAI-28 The column labeled "Comments" of Enclosure VI, Table F for Regulatory Position 5.6 and Table G for Regulatory Position 7.4 states that: "The transport model described in Regulatory Positions 5.5 and 5.6 ... was considered as appropriate [emphasis added] ..."
Please clarify if the models described in these Regulatory Positions were used. If only those considered appropriate were used, please state which were not considered appropriate and justify why they are not appropriate.
Response
The transport model described in Regulatory Positions 5.5 and 5.6 in Appendix E of Regulatory Guide 1.183 were used for the Locked Rotor and Rod Ejection analyses with the following exceptions:
Position 5.5.1: For the Rod Ejection and Locked Rotor radiological consequences analyses, steam generator dryout does not occur due to the secondary system remaining intact so no consideration for flashing of primary-to-secondary leakage was modeled.
Position 5.5.2: As there is no flashing of primary-to-secondary leakage, no credit was taken in the Rod Ejection and Locked Rotor radiological consequences analyses for scrubbing of flashed leakage in the steam generators.
Position 5.6: Comments provided in Table D of Enclosure VI (now Enclosure IV of this LAR) for Regulatory Position 5.6 state that the issue for tube uncovery was addressed in WCAP-13247 [The Topical Report on the Methodology for the Resolution of the Steam Generator Uncovery Issue is included as part of the response to ARCB-RAI-36 for non-SGTR events (Rod Ejection and Locked Rotor included).] It was concluded in WCAP-13247 and confirmed in the NRC response to WCAP-13247 that the effect of tube uncovery would be essentially negligible and the issue could be closed without any further investigation or generic restrictions. This position was accepted by the NRC as
Westinghouse Non-Proprietary Class 3 Page 99 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment noted in Table D. Therefore, the Rod Ejection and Locked Rotor radiological consequences analyses do not model tube uncovery.
ARCB-RAI-29:
Enclosure VI, page 4-54 states, in part, that The minimum SG [steam generator] water mass is increased after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to take credit for operators maintaining level at narrow range just on span.
Please explain if this this assumption was credited previously. If not, please provide a justification for this assumption.
Response
This assumption was not credited in the previous WCGS analyses of record. This modeling is justified by the WCGS Emergency Management Guidelines (EMGs), which instruct the operators to maintain the level in the steam generators between 6% and 50% of narrow range span (NRS). The minimum mass modeled in the analysis is associated with a NRS SG water level of 0% corresponding to hot zero power conditions. This is conservative given the EMGs instruct the operators to maintain level of at least 6% of NRS.
The following wording is added to Enclosure IV, Sections 4.3.4.2.2, 4.3.5.2.2, and 4.3.6.2.2.2:
as justified by the Emergency Management Guidelines (EMGs). The mass used in the analysis after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> was calculated at 0% narrow range to appropriately bound the just on span level of 6% narrow range.
Westinghouse Non-Proprietary Class 3 Page 100 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-30 Regulatory Position 5.1.2 of RG 1.183 states, in part, that The single active component failure that results in the most limiting radiological consequences should be assumed.
Please provide the most limiting single active failure for each design basis accident.
Response
Consistent with the current control room radiological consequences calculation models, as discussed on page 15A-8 of the USAR, a failure of one of the filtration fans is assumed at the start of emergency mode of operation and a larger unfiltered inflow to the control room is assumed since only half of the makeup flow to the control room can pass through a filter. After a defined time of 90 minutes, operator action isolates the failed train and reduces the unfiltered inflow to the control room. The following accidents modeled switchover to the CREVS and included the failure of one of the filtration fans:
LOCA MSLB SGTR Rod Ejection Locked Rotor Fuel Handing Accident For the SGTR analysis, the limiting single failure is a failed open steam generator ARV on the ruptured steam generator, which is modeled in the mass releases calculated in Section 2.7.2 of WCAP-17658-NP, Revision 1 (Enclosure I of this LAR). The single failure of a failed open steam generator ARV is conservatively modeled in addition to the failure of one of the CREVS filtration fans.
For the remaining dose analyses, no CREVS actuation is generated due to the event and thus the failure of one of the CREVS filtration fans is not modeled. These accidents include:
Tank ruptures Letdown line break Loss of AC power These accidents model a release directly to the environment and do not credit filtration from either the EES or from the CREVS. Thus, there is not a single failure that would decrease the effectiveness of the exhaust or intake filtration and therefore an explicit single failure is not modeled. This is consistent with the current licensing basis AOR.
Westinghouse Non-Proprietary Class 3 Page 101 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-31:
Enclosure VI, page 15.7-1 of the UFSAR markups indicates a proposed change in the licensing bases for the waste gas decay tank failure. Previously, the tank was assumed to fail after 40 years, releasing the peak inventory expected in the tank. The proposed change is assumed to be the maximum activity for each radionuclide during the degassing operations and the Krypton-85 inventory is assumed to be the total activity released during the fuel cycle. RG 1.24, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure," March 1972 (ADAMS Accession No. ML083300020), states, in part, that The maximum content [emphasis added] of the decay tank assumed to fail should be used for the purpose of computing the noble gas inventory in the tank a) If previously the tank was assumed to accumulate activity over 40 years, please explain why the activity from one fuel cycle is conservative and consistent with the RG 1.24 above regulatory position of using the maximum content of the decay tank. Please justify the proposed change in assumptions regarding the tank contents and state why they are conservative.
b) Table 4.3-2a contains 3 columns of source terms used to calculate the tank failures described in Enclosure VI, Sections 4.3.10, "Waste Gas Decay Tank Failure (USAR
[UFSAR] Chapter 15.7.1.5)," and 4.3.11, "Liquid Waste Tank Failure (USAR [UFSAR]
Chapter 15.7.2.5)." Please state the assumptions used to calculate these source terms.
Response
a) The methods and assumptions used to determine the proposed gas decay tank source terms are consistent with, and somewhat more conservative than Regulatory Guide 1.24 as discussed below.
As stated in Regulatory Guide 1.24 Section C.1, Regulatory Position, the assumptions related to the release of radioactive gases from the postulated failure of a gaseous waste storage tank should be,
- a. The reactor has been operating at full power with one percent defective fuel and a shutdown to cold condition has been conducted near the end of an equilibrium core cycle. As soon as possible after shutdown, all noble gases have been removed from the primary cooling system and transferred to the gas decay tank that is assumed to fail.
- b. The maximum content of the decay tank assumed to fail should be used for the purpose of computing the noble gas inventory in the tank. Radiological decay may be taken into account in the computation only for the minimum time period required to transfer the gases from the primary system to the decay tank.
Westinghouse Non-Proprietary Class 3 Page 102 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment
- c. The failure is assumed to occur immediately upon completion of the waste gas transfer, releasing the entire contents of the tank to the building. The assumption of the release of the noble gas inventory from only a single tank is based on the premise that all gas decay tanks will be isolated from each other whenever they are in use.
- d. All of the noble gases are assumed to leak out of the building at ground level over a two hour time period.
The WCGS design contains 6 decay tanks that are available for containing waste gases during plant operation and, as noted in the Discussion Section B of Regulatory Guide 1.24, The radioactive components are principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of the halogens. With the exception of krypton-85, the longest half-life of the principal noble gas radionuclides present in reactor effluents is 5.27 days (xenon-133). Thus, storage of these gases for a period of 60 days will essentially eliminate by decay all of the radionuclides except krypton-85.
Thus, the krypton-85 activity that is generated during plant operations prior to cold shutdown will have been discharged from the tank(s) following the storage period during which the short lived species are allowed to decay. If the krypton activity is not discharged, it will be distributed among one or all of the 6 normal operation decay tanks.
The design also includes 2 shutdown gas decay tanks. The gaseous inventory in the primary coolant system is transferred to one (or both) of these tanks via the volume control tank (VCT) as soon as possible after shutdown. Since the shutdown tank will contain a relatively large amount of (primarily) short-lived activity, this is the tank that is assumed to fail. Also, for additional conservatism, all of the krypton-85 activity generated during the previous core cycle (from both normal operation and shutdown operations) is assumed to be present in the failed tank prior to the release.
As noted in the above discussion, the methods and assumptions used in defining the activity inventories in the failed gas decay tank are consistent with the assumptions stated in Regulatory Guide 1.24 Section C.1, Regulatory Position; additionally, conservatism has been added by considering all of the activity that is generated during the previous operating cycle, plus the activity that is removed during the shutdown is contained in the failed gas decay tank. The previous assumption of a 40-year accumulation of krypton-85 activity is overly conservative and inconsistent with the design and operation of the waste gas system.
Westinghouse Non-Proprietary Class 3 Page 103 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment b) All of the source terms are based on the conservative assumption that the plant has operated for a complete equilibrium core cycle with fuel cladding defects in the fuel rods that generate one percent of the core thermal power. A power uncertainty (i.e., calorimetric power uncertainty) factor of 1.02 is considered in the calculation of the core activities; an assumption that meets the requirements of Regulatory Guide 1.195 and Regulatory Guide 1.183, i.e. The inventory of fission products should be based on an assumed core power equal to the current licensed rated thermal power times the emergency core cooling system (ECCS) evaluation uncertainty. In addition, all calculated activity values are increased be a factor of [ ]a,c in order to account for potential core design parameter differences in future core designs. This factor is generally referred to as a fuel management multiplier and is intended to provide bounding values relative to future core designs as applied to accident analyses. Further, the primary coolant concentrations are conservatively assumed to be at the peak values calculated over the operating period for the case with no purge of the volume control tank (VCT) during power operation.
The waste gas decay tank sources are based on solving a set of equations that model the shutdown degassing of the primary coolant in which the isotopic inventory of the gas decay tank varies with time. The letdown of the primary coolant is assumed to occur at the maximum letdown flow rate with coolant degassing operations accomplished by burping of the volume control tank vapor space at 3-hour intervals. The maximum activities over the degassing period are conservatively selected in defining the limiting gas decay tank inventory. In addition to the noble gases, radioactive iodine is considered to partition in the VCT vapor space (a partition factor = 100) and to be carried over into the gas decay tank during the degassing operations.
The Kr-85 activity inventory is treated differently in that the total activity released to the coolant during the fuel cycle is assumed to be transferred to the gas decay tank.
Two separate scenarios for the accidental release of activity from liquid waste tanks are analyzed; that is, 1) the failure of a recycle holdup tank from which the release of 100% of the noble gas nuclides and 10% of the iodine activity in the tank is assumed, and 2) the failure of a hypothetical liquid waste tank from which 100% of the iodine activity is released. The recycle holdup tank activity release is used in analyzing the whole body dose impacts of the accident, and the release of activity from the hypothetical waste tank is considered in determining the thyroid dose consequences of a postulated accident.
The source terms for the tanks consider primary coolant activities as an input and apply appropriate reduction factors in order to determine the tank activity inventories. The reduction factors are based on decay and the reduction in inventory due to removal by the demineralizers which are located upstream of the tank. The contribution of the parent nuclide is also considered.
Two sources of activity are considered in the evaluation of the recycle holdup tank sources; i.e.,
the Reactor Coolant System (RCS) and the Reactor Coolant Drain Tank (RCDT). The liquid goes through two mixed bed demineralizers between the exit from the RCS and the recycle holdup tank. The intermittent flow through the cation demineralizer located between the two mixed bed demineralizers is conservatively ignored. Also, there is one demineralizer between
Westinghouse Non-Proprietary Class 3 Page 104 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment the outlet of the RCDT and the recycle holdup tank. Thus, the associated Decontamination Factors (DFs) considered in the analysis are:
For the noble gases, no reduction by the demineralizers is considered, i.e., the DF = 1 For Br and Iodine, a factor of 10 reduction across each demineralizer is considered, i.e.,
For the RCS to recycle holdup tank: the DF = 10 x 10 = 100 For the RCDT: the DF = 10 The concept of a hypothetical liquid waste tank was originally developed by WCNOC in order to replace sources associated with the evaporator bottoms tank, since the WCGS does not use the evaporators and the evaporator bottoms tank was no longer considered to be the bounding case for the thyroid dose analysis. Although not associated with an existing piece of equipment, the continued use of sources associated with a hypothetical liquid waste tank is considered to be appropriate. The hypothetical waste tank receives fluid from three separate tanks:
A recycle holdup tank (RHUT),
A waste holdup tank (WHUT), and A floor drain tank (FDT).
It is conservatively assumed that the input to each of the tanks is based on continuous full power operation with equilibrium iodine activity levels in the primary coolant. Based on this activity concentration, it is assumed that each of the feed tanks is filled and drained such that a constant volume in the feed tank is maintained and the activity concentration reaches an equilibrium level. Thus, the feed rate into each tank equals the discharge rate from the tank into the hypothetical tank and the (equilibrium) activity concentration in the feed tank comprises the input of activity into the hypothetical waste tank. Finally, equilibrium activity levels are assumed to exist in the hypothetical waste tank based on decay alone.
ARCB-RAI-32 Enclosure VI markups of UFSAR page 15.3-11 change the sentence from "Steam generator tube leakage is assumed to continue until the pressures in the reactor coolant and secondary systems are equalized" to Steam generator tube leakage is assumed to continue until the residual heat removal system can match decay heat and release from the secondary system are terminated." Please justify this proposed change.
Response
The time to terminate steam generator tube leakage to the intact SGs (i.e., SGs neither faulted nor ruptured) was set to the time at which the steam releases were terminated (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, which is the time that the residual heat removal systems match decay heat and eliminate the need for steaming from the secondary system for all analyses involving steam releases). When the
Westinghouse Non-Proprietary Class 3 Page 105 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment steam releases are terminated, the activity leakage into the secondary system no longer has an impact on the resulting doses because the activity released to the environment is terminated.
ARCB-RAI-33 Enclosure VI markups of UFSAR page 6.5A-3 add an Insert N which defines the term kg or the "gas-phase mass-transfer coefficient." The kg used is taken from Reference 2 or Brookhaven National Laboratory (BNL) Technical Report A-3788, "Fission Product Removal Effectiveness of Chemical Additives in PWR Containment Sprays," August 1986, rather than using the reference (Reference 20, entitled: "The Terminal Speed of Single Drops or Bubbles in an Infinite Medium," International Journal of Multiphase Flow, pages 491-511 (1974)) cited in SRP 6.5.2, "Containment Spray as a Fission Product Cleanup System."
Justify why Reference 2 is appropriate rather than the study cited in the SRP and state why it was used. Please provide a copy of the BNL report.
Response
The gas-phase mass-transfer coefficient (kg term) from BNL Technical Report A-3788, dated August 1986 (included in Enclosure VIII of this LAR), was previously used in the Indian Point Unit 2 License Amendment Request (LAR) in the transition to the use of the AST. The Indian Point Unit 2 LAR (reference Proposed Amendment Consisting of Changes to Technical Specification for Containment Air Filtration, Control Room Air Filtration, and Refueling Conditions, ADAMS Accession Number ML993400414) references the Indian Point Unit 2 Licensing Report submitted to the NRC in the October 8, 1999 letter (reference Radiological Consequences of Accidents for the Indian Point Nuclear Generating Station Unit No. 2 Using Source Term Methodology from NUREG-1465, ADAMS Accession Number ML100430988) which gives a kg value of 9.84 ft/min (which is consistent with the value used for the WCGS analysis). It was determined by the NRC staff in the Safety Evaluation (reference Indian Point Nuclear Generating Unit No. 2 - Re: Issuance of Amendment Affecting Containment Air Filtration, Control Room Air Filtration, and Containment Integrity During Fuel Handling Operations (TAC No. MA6955, ADAMS Accession Number ML003727500)) that the licensee used plant-specific parameters and a very conservative value for the gas phase mass transfer coefficient.
Therefore, the kg value from BNL Technical Report A-3788 is also appropriate for use in the WCGS analysis.
The BNL report was used as a reference for the kg value in the proposed revision to SRP Section 6.5.2, Revision 2 (dated April 1987) and was included by the NRC as part of this proposed revision.
From SRP Section 6.5.2, Revision 4: The first-order removal coefficient by spray, s, may be taken to be S=6KgTF/VD where Kg is the gas-phase mass-transfer coefficient and T is the time of fall of the drops, which may be estimated by the ratio of the average fall height to the terminal velocity of the mass-mean drop.
Westinghouse Non-Proprietary Class 3 Page 106 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Earlier revisions of SRP Section 6.5.2 (Revisions 2 and 3) cite Reference 20 at the end of the above sentence. This statement indicates the Reference 20 document is only meant for the terminal velocity calculations that result in the fall time of the drops. Reference 20 is not cited because it does not contain a recommended kg value.
ARCB-RAI-34 Enclosure VI, page 8-14 states in the comments for NRC Regulatory Issue Summary 2006-04, Issue 6 that there were no changes to the plant configuration. The changes to TS 3.9.4 allow an open containment when moving fuel that is not recently irradiated. Consistent with Regulatory Issue Summary 2006-04, please confirm that all pathways to the environment created by the proposed changes are considered and analyzed in the fuel handling accident analysis.
Response
In the response to ARCB-RAI-24, WCNOC identified that the changes proposed in TSTF-51-A, Rev. 2, are being withdrawn. As such, the changes to TS 3.9.4 are being withdrawn and the Limiting Condition for Operation (LCO) 3.9.4 requirements remain applicable during CORE ALTERATIONS and during movement of irradiated fuel assemblies within containment.
The current LCO 3.9.4 allows the equipment hatch to be open and capable of being closed, one personnel air lock capable of being closed, and penetration flow paths that have direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls. These allowances were approved in Amendment No. 146 (Reference 10), Amendment No. 95 (Reference 11), and Amendment No. 135 (Reference 12), respectively.
Consistent with the current analysis of record, the radiological consequences AST analysis for a fuel handling accident inside containment did not credit isolation of containment. Administrative controls (previously approved by the NRC) provide reasonable assurance that containment closure as a defense-in-depth measure can be reestablished quickly to limit releases much lower than assumed in the dose calculations.
ARCB-RAI-35 Enclosure VI, markups to UFSAR Table 15.6-4 remove the words with forced overfill from the steam generator tube rupture. Please explain why this forced overfill was part of the licensing bases and what has changed to allow the change to no longer consider forced overfill.
Response
During the NRC review of the Final Safety Analysis Report (FSAR) for issuance of the operating license, Section 15.4.4, Steam Generator Tube Rupture, of NUREG-0881, Supplement 5, (Reference 13) Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No. 1, specified that the staff will condition the operating license to require satisfactory resolution of this issue before startup following the first refueling outage. This
Westinghouse Non-Proprietary Class 3 Page 107 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment became License Condition 2.C.(11), Steam Generator Tube Rupture (Section 15.4.4, SSER #5) which required Kansas Gas and Electric to submit for NRC review and approval an analysis that demonstrated that the steam generator tube rupture (SGTR) analysis presented in the FSAR is the most severe case with respect to release of fission products and calculated doses. Letter SLNRC 86-1 (Reference 14) dated January 8, 1986, provided the required SGTR analysis. On November 12, 1986, the NRC issued a Request for Additional Information (RAI) (Reference 15) that raised concerns with the margin of 271 ft3 (~5%) to steam generator overfill, which was originally calculated in the required SGTR analysis submitted by letter SLNRC 86-1. The RAI question stated, in part:
The staff believes that, based on this response, there is sufficient uncertainty in the break flow calculations, as well as in operator action times (see Enclosure 2) and interval between safety injection (SI) termination and break flow termination (see question 2 below,) to warrant assuming that safety valve (SV) liquid relief can occur. . Therefore please perform an analysis which assumes a design basis SGTR with loss of offsite power (LOOP), closure of the ruptured SG main steam isolation valve (MSIV), and SG overfill resulting in liquid relief through one SV.
In response to the RAI, WCNOC submitted letter WM 87-0145 (Reference 16) dated May 15, 1987, providing the results of a plant specific analysis that made changes to the modeling and assumptions for the SGTR analysis that would result in steam generator overfill and subsequent liquid relief through one safety valve. Subsequently, the NRC issued a safety evaluation (Reference 17) on May 7, 1991 that concluded that steam generator overfill does occur during a design basis SGTR accident and found the WCNOC SGTR accident analysis to be acceptable.
As part of the transition to the Westinghouse core design and safety analysis methodologies, the SGTR event was analyzed to demonstrate margin to overfill consistent with the Westinghouse standard analysis methodology. This analysis is presented in Section 2.7.2 of Enclosure I of this amendment request. The margin to overfill analysis is consistent with the methodology developed and approved in WCAP-10698-P-A (Reference 18) by Westinghouse.
During the NRC review of the WCAP-10698-P-A methodology, the NRC indicated that an evaluation of the consequences of steam generator overfill would be desirable to demonstrate that the consequences of overfill would be acceptable. In response to this request, WCAP-11002 was submitted to the NRC (Reference 19). The NRC approval of WCAP-10698-P-A included an evaluation of WCAP-11002. The NRC approved methodology in WCAP-10698-P-A does not define a minimum margin to overfill requirement and does not require consideration of a forced overfill scenario.
Westinghouse Non-Proprietary Class 3 Page 108 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-36 The column labeled Comments of Enclosure VI, Table D for Regulatory Position 5.6 discusses WCAP-13247 and an NRC staff response dated March 10, 1993. The NRC staff is unable to locate these documents. Please provide these two documents.
Response
WCAP-13247, Report on the Methodology for the Resolution of the Steam Generator Tube Uncovery Issue, March 1992, was transmitted to the NRC by Westinghouse Owners Group letter OG-92-25 on March 31, 1992, with the accompanying affidavit CAW-92-287.
WCAP-13247, OG-92-25, CAW-92-287, and the NRC response to the Westinghouse Owners Group dated March 10, 1993 (which was an attachment to the Westinghouse Owners Group letter WOG-93-066) are included in Enclosure VIII of this LAR.
ARCB-RAI-37:
Please provide a complete description of how the dose equivalent iodine-131 (DE) and the dose equivalent Xenon-133 are calculated.
Response
The RCS specific activity concentrations for each isotope is based on the assumed fuel defect level (1%), which is multiplied by the ratio of its dose conversion factor to the I-131 dose conversion factor for iodines and Xe-133 dose conversion factor for noble gases to obtain its I-131 or Xe-133 equivalence. The dose conversion factors used are the thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11 for iodines and the effective dose conversion factors from Table III.1 of EPA Federal Guidance Report No. 12, as identified in the revised Definitions for DOSE EQUIVALENT I-131 and revised DOSE EQUIVALENT XE-133 that are included in the Technical Specification markups contained in Enclosure IV, Section 9. The total of the isotopic dose equivalencies is the total I-131 or Xe-133 equivalence corresponding to the design fuel defect level of 1%. Each of the isotopic concentration values at the design defect level is then divided by the total I-131 or Xe-133 dose equivalence to obtain the isotopic concentrations at the applicable Technical Specification 3.4.16, RCS Specific Activity, limits of
<1.0 µCi/gm for DE I-131 and <500 µCi/gm for DE Xe-133.
Note that I-130, Kr-85, and Xe-131m are not listed in the Technical Specification Definitions for DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133. These nuclides do not contribute to the total of the isotopic dose equivalencies; however, their fuel defect level activities are divided by the total I-131 of Xe-133 dose equivalencies and scaled to the Technical Specification limits for DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133.
Westinghouse Non-Proprietary Class 3 Page 109 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment ARCB-RAI-38 SRP 16.0 states, in part, that In TS change requests for facilities with TS based on previous STS [Standard Technical Specifications], licensees should comply with comparable provisions in these STS NUREGs to the extent possible or justify deviations from the STS.
Several of the proposed changes to the TSs and TS bases (an example is provided below) do not align with the STS. Please provide a justification for deviations from the STS.
For example, proposed page B 3.6.3-2 removes detail from the bases of TS 3.6.3, Containment Isolation Valves, which is inconsistent with the STS, NUREG 1431, and Revision 4. The detail provides the limiting total response time for the isolation of containment. Please provide the value used in the DBA analyses assumed for containment isolation.
Response
As noted in the above question, SRP Section 16.0, Technical Specifications, Rev. 3, Subsection II., Acceptance Criteria, states, in part:
In TS change requests for facilities with TS based on previous [emphasis added] STS, licensees should comply with comparable provisions in these STS NUREGs to the extent possible or justify deviations from the STS.
This subsection further states:
Acceptable justifications for deviation would include retention of existing TS requirements, non-adoption of Standard Technical Specifications (STS) requirements not represented in existing TS (e.g., an LCO in STS but not in existing TS), editorial preference, facility design, and a technically justified alternative presentation equivalent to the STS intent. In some cases, comparison to the previous STS may help evaluate the proposed changes by clarifying the TS intent. The previous [emphasis added] STS NUREGs are as follows:
NUREG-0103, STS, Babcock and Wilcox Plants NUREG-0452, STS, Westinghouse Plants NUREG-0212, STS, Combustion Engineering Plants NUREG-0123, STS, General Electric Plants
Westinghouse Non-Proprietary Class 3 Page 110 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment From the above, the quoted subsection specifies the TS change requests based on previous STS (i.e., NUREG-0452, not NUREG-1431) should comply with comparable provisions in these STS NUREGs (i.e., NUREG-1431) to the extent possible or justify deviations from the STS. The current WCGS TSs are based on NUREG-1431, Revision 1. WCNOC submitted a license amendment request on May 15, 1997 (Reference 6) that provided a conversion of the WCGS TSs that were based on NUREG-0452 to TSs that were based on NUREG-1431, Revision 1.
The improved TSs were issued with the issuance of Amendment No. 123 on March 31, 1999 (Reference 7). The process for converting to the improved TSs requires justifications for deviations from the STS.
Subsection III of SRP 16.0, states, in part:
The TS bases are not parts of the Operating License (OL) or Current Operating License (COL); however, the reference TS bases have a wealth of information on safety limit or LCO purposes and action and surveillance requirements. Particularly important is the bases description of how the system or parameter addressed by the LCO satisfies 10 CFR 50.36(c)(2)(ii). Regardless of any technical differences with the reference TS bases, the TS review should verify whether the plant-specific TS bases are consistent with the FSAR plant-specific accident analysis and system description. In addition, the TS bases should describe the basis for each TS requirement accurately.
Nuclear Reactor Regulation (NRR) Office Instruction No. LIC-100, Revision 1, Control of Licensing Bases for Operating Reactors, Section 3.2, Technical Specification Bases Section, states (page 3.10), in part:
If the licensee and staff decide to process TS Bases pages, even though NRC review and approval is not required [emphasis added], the staff may issue the revised pages via a letter to the licensee (or as part of an amendment).
Section 2 of Enclosure VI of Reference 1 provided a description of the changes to the TSs and an associated justification for the change. The changes were further supported by the technical analysis in Section 4 of Enclosure VI of Reference 1.
Regarding the specific changes to TS Bases page B 3.6.3-2, these changes were made for consistency with the AST analyses and to correct a legacy issue associated with the current analyses of record for WCGS. The LOCA and rod ejection dose consequence analyses, which are the only two analyses that model isolation of containment, assumed the containment was isolated at the beginning of event, with the exception of the mini-purge releases for the first 10 seconds of the LOCA event. This is consistent with the current analyses of record for WCGS.
The basis for deleting the wording from the Bases for TS 3.6.3 is that it refers to the containment purge and exhaust valves and not the mini-purge valves.
Westinghouse Non-Proprietary Class 3 Page 111 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment Reference 4 RAIs SRXB-RAI-1:
In the Summary of Analysis Codes Utilized in Postulated Accident Analyses, there are two Non-Loss-of-Coolant Accident (LOCA) Safety Analysis Codes listed for the Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, RETRAN, and LOFTRAN. Please provide the information that is obtained from each code and used as the basis for the updated analysis.
Response
The departure from nucleate boiling (DNB) analysis of the rod withdrawal at power (RWAP) event is performed using the RETRAN code, consistent with the approved methodology in WCAP-14882-P-A, RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses. RETRAN is also the non-LOCA code used to perform DNB analyses for other Updated Safety Analysis Report (USAR) Chapter 15 events.
The RWAP DNB cases performed consider 10%, 60%, and 100% power cases, at various reactivity insertion rates for minimum and maximum reactivity feedback conditions. Since the Revised Thermal Design Procedure (RTDP) DNB methods were used, initial condition uncertainties are accounted for in the DNB ratio (DNBR) safety analysis limit. As such, the initial conditions are set to their nominal values. A relatively small number of cases are performed in order to determine the most limiting set of conditions (e.g., a sensitivity study to determine conservative treatment of initial condition uncertainties is not required).
For RWAP cases analyzed to address RCS overpressurization concerns, the limiting set of conditions is established through varying a number of parameters (power, conservative treatment of initial condition uncertainties, reactivity insertion rates, etc.) resulting in a large number of cases to be considered. As discussed in WCAP-14882-P-A, the RETRAN and LOFTRAN codes are essentially equivalent. The LOFTRAN code is more suited for running a large number of cases than RETRAN; as such, it is the preferred transient analysis code for addressing this RCS pressure acceptance criterion. The LOFTRAN code (WCAP-7907-P-A, LOFTRAN Code Description) has historically been the approved transient analysis code for non-LOCA analyses performed by Westinghouse and has previously been used to perform RWAP RCS overpressurization analyses for plants that otherwise utilize the RETRAN code for the DNB analyses. As a recent precedent, this approach was utilized for the Turkey Point Extended Power Uprate (Licensing Report Table 2.8.5.0-7: ADAMS Ascension Number ML103560177, and Safety Evaluation Report: ADAMS Ascension Number ML11293A365).
Enclosure I, Section 2.5.2.1.3 of this LAR has been updated to include the above discussion.
Westinghouse Non-Proprietary Class 3 Page 112 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment
References:
- 1. WCNOC Letter ET 13-0023, License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, August 13, 2013. ADAMS Accession No. ML13247A075.
Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), April 3, 2014. ADAMS Accession No. ML14083A400.
- 3. WCNOC Letter ET 14-0017, Withdrawal of License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, June 18, 2014. ADAMS Accession No. ML14175A119.
Request for Additional Information RE: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), April 30, 2014. ADAMS Accession No. ML14111A100.
- 5. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Wolf Creek Generating Station - Issuance of Amendment Re: Elimination of Requirements for Post Accident Sampling Systems (TAC NO. MB0678), March 2, 2001. ADAMS Accession No. ML010670233.
- 6. WCNOC letter ET 97-0050, Technical Specification Conversion Application, May 15, 1997. ADAMS Accession No. 9705210318 (Public Legacy Library).
- 7. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Conversion to Improved Technical Specification for Wolf Creek Generating Station - Amendment No. 123 to Facility Operating License No. NPF-42 (TAC NO. M98738), March 31, 1999. ADAMS Accession No. ML022050061.
- 8. Letter from A. J. Mendiola, USNRC, to Technical Specifications Task Force, Potential Issues with Plant-Specific Adoption of Travelers TSTF-51, Revision 2, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations, TSTF-286, Revision 2, Operations Involving Positive Reactivity Additions, and TSTF-471, Revision 1, Eliminate Use of Term Core Alterations in Actions and Notes, November 7, 2013. ADAMS Accession No. ML13246A358.
- 9. Not used.
- 10. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Wolf Creek Generating Station - Issuance of Amendment Re: Equipment Hatch Open During Refueling (TAC NO.
MB2599), July 30, 2002. ADAMS Accession No. ML022120428.
Westinghouse Non-Proprietary Class 3 Page 113 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment
- Amendment No. 95 to Facility Operating License No. NPF-42 (TAC NO. M94113),
February 28, 1996. ADAMS Accession No. ML022040718.
- 12. Letter from J. N. Donohew, USNRC, to O. L. Maynard, WCNOC, Wolf Creek Generating Station - Issuance of Amendment Re: Use of Administrative Controls for Open Containment Penetrations During Refueling (TAC NO. MA9293), September 12, 2000.
ADAMS Accession No. ML003750021.
- 13. NUREG-0881, Supplement 5, Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No. 1, March 1985.
- 14. SNUPPS letter SLNRC 86-1, Steam Generator Tube Rupture Analysis - SNUPPS, January 8, 1986.
- 15. Letter from P. W. OConnor, USNRC, to G. L. Koester, Kansas Gas & Electric Company, Request for Additional Information Related to the SNUPPS Steam Generator Tube Rupture Analysis, November 12, 1986.
- 16. WCNOC letter WM 87-0145, Response to RAI Regarding SGTR Analysis, May 15, 1987.
- 17. Letter from D. V. Pickett, USNRC, to B. D. Withers, WCNOC, Safety Evaluation Report for the Wolf Creek Generating Station Steam Generator Tube Rupture Analysis (TAC NO.
57363), May 7, 1991.
- 18. WCAP-10698-P-A, SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, August 1987.
- 19. WCAP-11002 (Proprietary) and WCAP-11003 (Non-Proprietary), Evaluation of Steam Generator Overfill Due to a SGTR Accident, February 1986.
- 20. WCNOC letter ET 14-0003, Response to Request for Additional Information Regarding License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, January 28, 2014. ADAMS Accession No. ML14035A224.
- 21. WCNOC letter WO 14-0031, Response to Request for Additional Information Regarding License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, March 20, 2014. ADAMS Accession No. ML14091A245.
- 22. WCNOC letter WO 14-0032, Supplemental Information for Response to Request for Additional Information Regarding License Amendment Request for the Transition to Westinghouse Core Design and Safety Analysis, March 26, 2014. ADAMS Accession No. ML14091A261.
Westinghouse Non-Proprietary Class 3 Page 114 of 114 LTR-OPB-PMO-SP-MEP-16-017 Rev. 2 - NP Attachment
- 23. Letter from R. F. Kuntz, USNRC, to C. M. Crane, Exelon Generation Company, Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 - Issuance of Amendment Re: Alternative Source Term (TAC NOS. MC6221, MC6222, MC6223, and MC6224), September 8, 2006. ADAMS Accession No. ML062340420.
- 24. Letter from C. F. Lyon, USNRC, to M. W. Sunseri, WCNOC, Wolf Creek Generating Station - Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), December 13, 2013. ADAMS Accession No. ML13345B335.
Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC NO. MF2574), March 5, 2014. ADAMS Accession No. ML14058A088.
- 26. Email from Fred Lyon, USNRC, to Steve G Wideman, WCNOC, FW: Wolf Creek Generating Station - Request for Additional Information Re: Transition to Westinghouse Core Design and Safety Analysis (TAC No. MF2574), December 23, 2013. ADAMS Accession No. ML13357A250.
- 27. Letter from C. F. Lyon, USNRC, to M. W. Sunseri, WCNOC, WOLF CREEK GENERATING STATION- REQUEST FOR ADDITIONAL INFORMATION RE:
TRANSITION TO WESTINGHOUSE CORE DESIGN AND SAFETY ANALYSIS (TAC NO.
MF2574), January 28, 2014. ADAMS Accession No. ML14027162.