RS-18-103, Response to Request for Additional Information Concerning License Amendment Request to Revise Alternate Source Term Dose Calculation

From kanterella
Jump to navigation Jump to search

Response to Request for Additional Information Concerning License Amendment Request to Revise Alternate Source Term Dose Calculation
ML18242A156
Person / Time
Site: Clinton Constellation icon.png
Issue date: 08/29/2018
From: Simpson P
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2018-LLA-0004, RS-18-103
Download: ML18242A156 (9)
v • d • e


Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-18-103 10 CFR 50.90 August 29, 2018 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket Nos. 50-461

Subject:

Response to Request for Additional Information Concerning License Amendment Request to Revise Alternate Source Term Dose Calculation (EPID: L-2018-LLA-0004)

References:

1. Letter from Patrick R. Simpson (Exelon Generation Company, LLC (EGC)) to U.S. NRC, "License Amendment Request to Incorporate Revised Alternative Source Term Dose Calculation," dated January 9, 2018
2. Email from J. Wiebe (NRC) to K. M. Nicely (EGC), "Preliminary Request for Additional Information Regarding Clinton Revised Alternate Source Term (EPID: L-2018-LLA-0004)," dated July 12, 2018 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Facility Operating License (FOL) No. NPF-62 for Clinton Power Station (CPS), Unit 1. The proposed change requested: 1) a TS change for the acceptance criteria for feedwater penetration leakage,
2) a revised element of analysis methodology pertaining to mixing of activity in the Secondary Containment, and 3) a more than minimal increase in Control Room post-loss of coolant accident (LOCA) dose consequences.

In Reference 2, the NRC requested additional information required to complete its review of Reference 1.

August 29, 2018 U.S. Nuclear Regulatory Commission Page2 EGC is providing the requested information, as described in Reference 2, in the Attachment to this letter.

There are no regulatory commitments contained within this letter.

If you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 29th day of August 2018.

Respectfully, Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC

Attachment:

Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term cc: NRC Regional Administrator, Region Ill NRC Senior Resident Inspector- Clinton Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST TO INCORPORATE REVISED ALTERNATIVE SOURCE TERM DOSE CALCULATION CLINTON POWER STATION, UNIT 1 DOCKET NO. 50-461 The current loss of coolant accident (LOCA) dose calculation for Clinton Power Station, Unit 1 (CPS) is based on alternative source term (AST) methodology in accordance with 10 CFR 50.67, "Accident source term," that was approved by NRC letter dated September 19, 2005 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML052570461)

(Amendment 167 to CPS license NPF-62). The current CPS Updated Safety Analysis Report (USAR) reflects the changes approved by Amendment 167. By letter dated January 9, 2018 (ADAMS Accession No. ML18009B037), Exelon Generation Company, LLC (EGC) proposed changes to the analysis of record in the USAR. The changes proposed include a revision to the analysis methodology pertaining to mixing of activity in the secondary containment and a lower acceptance criteria for primary containment to secondary containment leakage through the feedwater penetration. The NRC staff requires additional information to complete its review.

RAI SCPB-1 to Exelons January 9, 2018, letter contains a markup of the revised USAR and the Technical Specification (TS) bases pages. The markup in USAR page 15.6-10 contains a new item (6) added to Section 15.6.5.5.1.2 "Fission Product Transport to the Environment." Item (6) states "Credit is taken for mixing primary containment leakage in 50% of the secondary containment volume in accordance with RG [Regulatory Guide] 1.183, Appendix A, Section 4.4."

RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants," Appendix A, Section 4.4, states in part:

Credit for dilution in the secondary containment may be allowed when adequate means to cause mixing can be demonstrated. Otherwise, the leakage from the primary containment should be assumed to be transported directly to exhaust systems without mixing. Credit for mixing, if found to be appropriate, should generally be limited to 50%."

Provide information to demonstrate the means available to cause mixing of primary containment leakage with the secondary containment volume and a description of any supporting review/analysis performed at CPS to justify the 50% mixing assumption within secondary containment.

EGC Response to RAI SCPB-1 The current Clinton Power Station (CPS) loss of coolant accident (LOCA) analysis credits mixing in 50 percent (%) of the Secondary Containment volume consistent with NRC Standard Review Plan (SRP) Section 6.5.3, "Fission Product Control Systems and Structures," and Regulatory Guide 1.183, Appendix A, "Assumptions for Evaluating the Radiological Consequences of a [Light Page 1 of 7

ATTACHMENT Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term Water Reactor] Loss-Of-Coolant Accident," Section 4.4. The leakage from the Primary Containment to the Secondary Containment is 0.598%/day and the containment volume (excluding the Drywell) is 1,512,341 cubic feet (ft3). The flow rate from the Primary Containment is therefore approximately six (6) cubic feet per minute (cfm) into a Secondary Containment volume of approximately 1,700,000 ft3. The containment leakage, if any, would most likely be associated with piping penetrations, which are located in the lower part of the containment (e.g.,

Elevation (El.) 707'-6" mean sea level), or door seals. The small amount of containment leakage would have to diffuse through the secondary containment prior to being exhausted by the Standby Gas Treatment System (SGTS) to the environment. Significant mixing would occur as the leakage travels through the Secondary Containment prior to entering the SGTS. If power is available, the Secondary Containment heating, ventilation, and air conditioning (HVAC) system would provide additional mixing by supplying air from clean areas to potentially contaminated areas in accordance with good as-low-as-reasonably-achievable (ALARA) radiation protection practices. The Secondary Containment volume is 1,700,000 ft3 (i.e., rounded down from 1,704,995 ft3) so the mixing volume used in the analysis is 8.50E+05 ft3.

The CPS design includes the SGTS which is a safety-related HVAC system. Although this system is not a mixing system per se, the system does take suction directly or indirectly from every compartment in the Auxiliary Building which provides mixing. The following rooms located at El. 707'-6" are serviced by the SGTS (i.e., total flow of approximately 1000 cfm for these areas):

  • RHR "A" Heat Exchanger (Hx) Room,
  • RHR "B" Pump Room,
  • RHR "B" Hx Room,
  • RHR "C" Pump Room,
  • Accessible Area El. 715-0" The SGTS also services the following areas (i.e., total flow approximately 3000 cfm for these areas):
  • Radwaste Pipe Tunnel
  • Combustible Gas Control Boundary (various elevations)
  • Auxiliary Building Aisles Although no specific transport analysis was performed due to the unknown location of the potential leakage, there is sufficient justification for the assumption that the leakage is mixed in 50% of the Secondary Containment volume.

Page 2 of 7

ATTACHMENT Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term RAI SCPB-2 By letter dated June 21, 1996 (ADAMS Accession No. ML020990619), the NRC staff approved CPS adoption of Option B "Performance Based Testing Requirements" of Appendix J to 10 CFR 50, "Primary Reactor Containment Leakage Rate Testing for Water Cooled Power Reactors." The exemptions to the requirements 10 CFR Part 50, Appendix J, are stated in Section 2.D of the CPS facility Operating License.

CPS TS 5.5.13, "Primary Containment Leakage Rate Testing," states in part that "A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012 [ADAMS Accession No. ML12221A202], and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008 [ADAMS Accession No. ML100620847], as modified by the following exception: (1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A tests."

Appendix J to 10 CFR 50, Option B, Section III.B, "Performance-Based Requirements for Type B and C Tests," requires pneumatic tests to measure containment isolation valve leakage rates.

NEI 94-01, Revision 3-A, Section 8.0 "Testing Methodologies for Type A, Type B and Type C Tests" states that, "Type A, Type B and Type C tests should be performed using the technical methods and techniques specified in ANSI/ANS-56.8-2002, "Containment System Leakage Testing Requirements," or other alternative testing methods that have been approved by the NRC." Section 2, "Definitions" of ANSI/ANS-56.8-2002 defines Type C test as "A pneumatic test to measure leakage rates from containment isolation valves, which are potential gaseous leakage pathways from containment during a design-basis LOCA."

Section 2, "Detailed Description," in Attachment 1 to EGCs January 9, 2018, letter states that changes to the current LOCA dose calculation includes a reduction to the feedwater isolation valve liquid leakage from 2.0 gallons per minute (gpm) to 1.5 gpm, and a reduction in the feedwater isolation valve air leakage from 10.98 cubic feet per minute (cfm) to 8.64 cfm. Based on the USAR markup (page 15.6-10, Item 4) provided in Enclosure 2 to the LAR, the containment atmosphere leakage through the feedwater penetration applies in the first 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the liquid leakage through the feedwater penetration applies for the duration of the accident after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The letter contains an associated change to Surveillance Requirement (SR) 3.6.1.3.11. Currently, the SR requires verification that the combined leakage rate from both primary containment feedwater penetrations is 2 gpm. EGCs January 9, 2018, letter proposes to reduce the combined leakage to 1.5 gpm. The frequency of the surveillance is "in accordance with the primary containment leakage rate testing program."

Appendix J testing program only provides for pneumatic testing program, therefore, it is not clear what method or interval requirements the licensee would apply to the liquid leakage testing performed at CPS for this SR.

Describe the Appendix J leakage tests currently performed to satisfy the requirements of SR 3.6.1.3.11 and TS 5.5.13. If the tests, including the test medium, are different from the Page 3 of 7

ATTACHMENT Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term approved methods in Appendix J to 10 CFR 50 and ANSI/ANS-56.8-2002, describe where the exemptions/exceptions are explicitly stated in the CPS license documents including any prior staff approval obtained for such exemptions/exceptions.

Explain if the reduced liquid and containment atmosphere leakage values proposed by EGCs January 9, 2018, letter are already supported by recent historical results of Appendix J testing or if new testing would be necessary.

EGC Response to RAI SCPB-2 CPS implemented a feedwater leakage control system (FWLCS) in 2000 to provide an enhanced means of isolating the FW penetrations post-LOCA. With the operation of the FWLCS, the periodic leakage testing requirement for the primary containment FW penetration isolation valves is being conducted with a water leakage test in lieu of an air leakage test. The NRC has previously reviewed the CPS FWLCS modification and testing of valves with water as documented in a safety evaluation dated April 25, 2000, (i.e., Accession No. ML003710668).

The CPS 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," Program covers pneumatic tests and also contains provisions for water testing the FW isolation valves. The FW Containment Penetrations 1MC-09 and 1MC-10 are tested using a low pressure (i.e., not 1000 psig) water test in accordance with CPS Procedures 9861.05D013, "Local Leak Rate Test Data Sheet for 1MC009," and 9861.05D014, "Local Leak Rate Test Data Sheet for 1MC010." This testing is in accordance with ANS-56.8-2002, Section 3.4, "Qualified Seal System Testing Requirements." The qualified seal system is provided by the FWLCS as previously reviewed by the NRC as discussed above.

As shown in Table 1 below, the proposed reduced leakage criterion is supported by recent historical results of Appendix J testing.

Table 1: Appendix J Test History for the CPS FW Isolation Valves Combined Leakage Acceptance Criterion*

Refuel Outage (gallons per minute (gpm)) (gpm)

C1R17 (2017) 0.15 2.0 C1R15 (2015) 0.88 2.0 C1R14 (2013) 0.32 2.0

  • Revised acceptance criterion is 1.5 gpm.

Page 4 of 7

ATTACHMENT Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term RAI RMET/RHM-1 In its January 9, 2018, letter, EGC states that no credit is taken for the dual, separated outside air intake design feature for evaluating the radiological consequences of design basis accidents (DBAs). Section 3.3.2.1 of RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," dated June 2003 (ADAMS Accession No. ML15268A488), states that for dual outside air intakes in different wind direction windows which cannot be isolated by design, the /Q values for the limiting (least favorable) outside air intake should be calculated for each time interval.

Table 1 in the Safety Evaluation related to Amendment 167 (ADAMS Accession No. ML052570461) documents the staffs review and acceptance of the /Q values used in the current AST dose analysis. Those /Q values, along with the revised design basis /Q values proposed for use in this LAR, are listed in Table 1:

Table 1: Filtered Intake Control Room /Q Values (sec/m3)

Current Design Basis West Most Favorable Intake Proposed Revised Time Period Intake East Intake Divided by 4 Design Basis 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 9.45E-04 9.75E-04 2.36E-04 9.45E-04 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.58E-04 7.09E-04 1.77E-04 7.58E-04 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.28E-04 2.93E-04 7.33E-05 3.28E-04 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 2.61E-04 2.13E-04 5.33E-05 2.61E-04 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.85E-04 1.79E-04 4.48E-05 1.85E-04 Table 1 shows that for the proposed revised design basis, the most favorable /Q value was selected for the 0-2 hour time period whereas the least favorable /Q values were selected for the remaining time periods.

Provide justification for why the /Q value for the limiting (least favorable) outside air intake was not used for 0-2 hour time period for the post-LOCA control room dose calculations as suggested in RG 1.194.

EGC Response to RAI RMET/RHM-1 Regulatory Guide (RG) 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," Section 3.3.2, "Dual Ventilation Outside Air Intakes," addresses control room ventilation systems that have two outside air intakes, each of which meets applicable design criteria of an engineered safeguards feature (ESF), including single-failure criterion, missile protection, seismic criteria, and operability under loss-of-offsite AC power conditions. This RAI references RG 1.194, Section 3.3.2.1 as the basis for the NRC question.

Section 3.3.2.1 addresses the situation when both dual intakes are located within the same wind direction window. For this condition, Section 3.3.2.1 would require use of an average, or a weighted average, of the two atmospheric relative concentration values (X/Q) depending on the Page 5 of 7

ATTACHMENT Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term flow rates for the intakes. Because the CPS dual intakes are not in the same wind direction window, Section 3.3.2.1 is not applicable.

Section 3.3.2.2 addresses dual outside air intakes that are not in the same wind direction but cannot be isolated by design. In accordance with this section, the Control Room X/Q would be based on half of the maximum X/Q of the two separate intakes or a weighted average dependent on the respective flows. The CPS Control Room HVAC system is designed so that the intakes are not in the same wind direction window, and allows either intake to be isolated. Consequently, Section 3.3.2.2 is not applicable to CPS.

Section 3.3.2.3 addresses the situation in which the ventilation system design allows the operator to manually select the least contaminated outside air intake as a source of outside air makeup and to close the other intake. Section 3.3.2.3 further states that the X/Q value for the limiting intake should be used for the time interval prior to intake isolation. This X/Q value may be reduced by a factor of 2 to account for dilution by the flow from the other intake. The X/Q values for the favorable intake are used for the subsequent time intervals. The X/Q values for the favorable intake may be reduced by a factor of 4 to account for the dual inlet and the expectation that the operator will make the proper intake selection. This section would be applicable to CPS except for the inability to switch to the more favorable intake following a loss of divisional power. Regulatory Guide 1.194 requires that the operator have the ability to select the remote intake with the lowest activity to take credit for dual Control Room remote intakes. Following a loss of divisional power (i.e., a single failure), the ability to select the more favorable remote intake from a Control Room radiological dose perspective would not be possible at CPS. Consequently, the previous credit taken for a dual Control Room remote intake can no longer be taken due to single failure issues with the current system design. Based on this, the reduction in the atmospheric dispersion by a factor of four to account for selection of the most favorable remote intake is no longer applied.

The post-accident scenarios to be considered relative to credit for dual intakes involve the availability of divisional power. If a loss of divisional power does not occur, the operator would be able to select the most favorable intake and the X/Q value would be reduced by a factor of four in accordance with RG 1.194, Section 3.3.2.3. If a loss of divisional power occurs, either the East or West intake will fail closed. For the revised LOCA analysis, the intake which gave the highest overall Control Room dose from 0-720 hours (i.e., the West intake) was used. The Control Room dose results using the West intake and the East intake are shown in Table 2 below. As seen, if the East intake is used the Control Room dose at 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> is 2.3% lower confirming that the use of the West intake in this analysis is conservative.

Page 6 of 7

ATTACHMENT Response to Request for Additional Information Regarding Clinton Revised Alternate Source Term Table 2: Control Room Dose Results for West and East Intakes AOR East Intake (West Intake)

Case TEDE TEDE

[REM]

[REM]

CNMT Leakage 1.09 1.03 Bypass Leakage 0.62 0.62 MSIV Leakage 1.16 1.13 FW Water Leakage 1.21 1.21 FW Air Leakage 0.08 0.08 Purge Leakage 0.06 0.05 ECCS Leakage 0.17 0.16 RCIC Leakage 0.22 0.22 Gamma Shine 0.277 0.277 Total 4.89 4.78 The use of the East intake X/Q for the initial two-hour post-accident interval and then switching to the West intake is not appropriate because this would imply no loss of divisional power in which case it would be possible to select the most favorable intake and reduction of the Control Room X/Q by a factor of four would be justified. Because a loss of divisional power would result in the loss of one or the other intake, CPS should be considered as having two separate/single remote intakes. Either the East or West intake will be operable following a loss of divisional power (i.e., a single failure). As shown above, the West intake produces the highest control room dose from 0-720 hours, making it the limiting intake. Therefore, the X/Q values for the West intake were used for all time intervals.

Page 7 of 7

Retrieved from "https://kanterella.com/w/index.php?title=RS-18-103,_Response_to_Request_for_Additional_Information_Concerning_License_Amendment_Request_to_Revise_Alternate_Source_Term_Dose_Calculation&oldid=2399530"