L-MT-24-002, Submittal of Revision 41 to the Updated Safety Analysis Report
| ML24109A129 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/17/2024 |
| From: | Brown G Northern States Power Company, Minnesota, Xcel Energy |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24109A128 | List:
|
| References | |
| L-MT-24-002 | |
| Download: ML24109A129 (18) | |
Text
ENCLOSURE 5 CONTAINS SECURITY-R ELATED INFORMATION WITHHOLD UNDER 10 CFR 2.390 WHEN SEPARATE FROM ENCLOSURE 5 THIS DOCUMENT IS DECONTROLLED fl Xcel Energy 2807 West County Road 75 Monticello, MN 55362
April 17, 2024 L-MT-24-002 10 CFR 50.71(e) 10 CFR 50.59(d)(2) 10 CFR 54.37(b)
ATTN: Document C ontrol D esk U.S. Nuclear Re gulatory Co mmission Washington, D C 2 0555- 0001
Monticello Nuclear G enerating Plant Docket N o. 50- 263 Renewed F acility O perating License No. D PR-22
Submittal of Revision 41 to the U pdated S afety A nalysis R eport
Pursuant to 10 CFR 50.71(e), Revision 41 to the Monticello Nuclear Generating Plant (MNGP)
Updated Safety Analysis Report (USAR) is provided. This revision completes an update of the information in the USAR for the period from October 29, 20 22, to December 21, 2023.
The changes in this revision reflect the incorporation of modifications, license amendments, and editorial corrections and clarifications. These changes are made in accordance with the guidance provided in Nuclear Energy Institute (NEI) Report NEI 98-03, "Guidelines for Updating Final Safety Analysis Reports", Revision 1, dated June 1999 and Regulatory Guide 1.181, "Content of the Updated Final Safety Analysis Report in accordance with 10 CFR 50.71(e)", dated September 1999.
, Report of Changes, Tests and Experim ents, states that there were three (3), 10 CFR 50.59 evaluations performed. This report is provided as required by 10 CFR 50.59(d)(2).
, "Report of Changes t o Licensee Docketed Commitments ", indicates that in accordance with the guidance provided in NEI 99- 04, "Guidelines for Managing NRC Commitment Changes", dated July 1999, for this period, there w ere no commitment retirements and no changes to commitments.
, "Summary of Information Removed from the USAR ", provides the information removed from the USAR for this revision cycle. This information is provided in accordance with Revision 1 of NEI 98- 03 and Regulatory Guide 1.181.
ENCLOSURE 5 CONTAINS SECURITY-RELATED INFORMATION WITHHOLD UNDER 10 CFR 2.390 WHEN SEPARATE FROM ENCLOSURE 5 THIS DOCUMENT IS DECONTROLLED
L-MT-24-002 Page 2 of 2 (Non-Proprietary) provides a copy of the USAR that has been redacted for public use. Enclosure 5 (Proprietary) provides a copy of the USAR that contains sensitive information to be withheld from public disclosure per 10 CFR 2.390(d)(1 ).
, "Report of Changes to the Monticello Fire Protection Program", provides a summary of changes to the program. Changes to the Fire Protection Program are provided in accordance with the guidance contained in Generic Letter 86-10, "Implementation of Fire Protection Requirements", dated April 24, 1986.
, This summary report is in lieu of adding a level of detail to the MNGP USAR that is greater in the remainder of the USAR, including the License Renewal Supplement in Appendix K (USAR-K). The contents of this report are consistent with the requirements of 10 CFR 54.37(b) and the guidance of Regulatory Issue Summary (RIS) 2007-16, Rev.1, "Implementation of the Requirements of 10 CFR 54.37(b) for Holders of Renewed Licenses".
Summary of Commitments
This letter makes no new commitments and no revisions to existing commitments.
In accordance with 10 CFR 50.71(e)(2), I certify that the information presented herein, accurately presents changes made since the previous submittal prepared pursuant to Commission requirement and identifies changes made under the provisions of 10 CFR 50;,~ot previously submitted to the Commission.
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G~egory D. t3ro~,,.,, *" // _.c* l--~~ U y--------.
Plant Manager, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota
Enclosures (7)
cc: Administrator, Region 111, USNRC Resident Inspector, Monticello, USNRC (without enclosure 4, enclosure 5)
State of Minnesota, Department of Commerce (letter with enclosure 1)
ENCLOSURE 1
MONTICELLO NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS AND EXPERIMENTS
The following includes a brief description and summary of the 10 CFR 50.59 evaluations performed between October 29, 2022 and December 21, 2023 for those changes, tests and experiments that were carried out without prior U.S. Nuclear Regulatory Commission (NRC) approval, pursuant to the requirements of 10 CFR 50.59(d)(2).
Evaluation No. SCR-20-0130 Revision 1:
Use of ATRIUM 10XM Lead Test Assemblies
ECR 601000002262 - Cycle 32 Core Reload Modification
Activity
Description:
As part of the core reload during refueling outage 31 (spring 2023), three material test specimen lead test assemblies (LTAs) will be loaded. The proposed activity evaluated in this 50.59 evaluation includes in-core operation of the LTAs for the duration of Cycle 32, activities to support in-core operation, and document updates required to support Cycle 32 operation. These activities are collectively referred to as operation of LTAs in this evaluation. A 50.59 evaluation is required for operation of LTAs because this is a test/experiment not described in detail in the USAR and is not within the bounds of tests/experiments described in detail in the USAR, although loading of LTAs is noted in the Monticello USAR section 3.4.4.2, Irradiated Fuel Rods." Revision 0 of this evaluation addressed loading four LTAs in the Cycle 31 core. Revision 1 addresses loading three LTAs in the Cycle 32 core.
ECR 601000002262 implements the Cycle 32 core reload, including loading of LTAs M31050, M31051, M31052. Aspects of operation not related to the LTAs have 50.59 applied as identified in the ECR.
Summary of Evaluation:
A 50.59 evaluation was performed for lead test assemblies (LTAs) and associated documentation changes.
The 50.59 evaluations response to each of the eight 10 CFR 50.59(c)(2) criterion is No. Therefore, the evaluation concluded that NRC approval for LTAs and associated documentation changes is not required. This is because no new failure modes are introduced, the LTAs remain non-limiting compared to other assemblies that are analyzed in accordance with standard reload methods, and only minor conservative changes are made to the previously NRC approved
Page 1 of 3 critical power correlation to show that the LTAs are non-limiting with no other analysis method changes. The changes have no impact on accident frequency, are not an initiator of any malfunction, and create no new failure modes. A conservative change to an element of a previously NRC approved methodology to show that the existing design basis in the USAR and COLR for co-resident fuel analyzed with approved methods described in TS 5.6.3 is more limiting is not considered a departure from a USAR method of evaluation.
Evaluation No. SCR-23-0012 Revision 0:
Atrium 11, New Methods, and Cycle 32 Core Reload Evaluation
Calculations04-040, 04-041,20-013, and USAR-14 updates
Activity
Description:
As part of the Cycle 32 core reload modification, ATRIUM 11 fuel is being loaded for the first time. To take advantage of the new fuel type, the average core exposure limitation of 37 GWD/MTU is revised to 41 GWD/MTU for the control rod drop accident (CRDA) and fuel handling accident (FHA). The higher exposure changes dose consequences. Only the CRDA and FHA are evaluated at this time due to open issues on the LOCA that do not allow calculation revision at this time. Consequently, the core average exposure cannot be increased beyond 37 GWD/MTU in actual plant operation until the LOCA dose analysis is revised for the higher core average exposure.
Concurrently, performance of dose consequences analysis has been brought in-house. RADTRAD is a computer code/method of evaluation noted in Chapter 14.7 of the USAR (incorporated by reference to the Alternative Source Term license amendment and various analyses). As a result of performing analysis in-house, the code version for RADTRAD has changed from the previous AAC version 3.03 to Zachry version 1.3. The new version has been procured as safety related software as documented in the initial purchase agreement (Contract Purchase Order: 4000016729). Evaluation has shown that the new version of RADTRAD produces essentially the same results for the Monticello analysis. At this time, the design basis calculations and the USAR results are only being revised to use the new RADTRAD for the CRDA and FHA, since open issues on the LOCA analysis do not allow calculation revision at this time. However, this evaluation justifies the use of the new version of RADTRAD for any of the three analyzed events (LOCA, CRDA, and FHA).
In addition to the calculation changes, the affected USAR sections are attached as Attachment 2. These changes only include the changes required for this Evaluation, not changes needed for other activities associated with the Cycle 32 core reload modification.
Summary of Evaluation:
The 50.59 evaluations response to each of the eight 10 CFR 50.59(c)(2) criterion is No. Therefore, the evaluation concluded that NRC approval for the
Page 2 of 3 calculation revisions and associated documentation changes is not required. This is because the changes have no impact on accident frequency, are not an initiator of any malfunction, and create no new failure modes. Use of a different version of the RADTRAD code is not a change to a method of evaluation described in the USAR because the same method is implemented in both code versions.
Evaluation No. SCR-23-0043 Revision 0:
Impacts of Opening MO-2565
602000030041 and 602000030042 for two ARPs
Activity
Description:
PCR 602000030041 PRO/C.6-259-A-07/MT/00 DRYWELL HIGH RADIATION PCR 602000030042 PRO/C.6-259-A-03/MT/00 DRYWELL HI-HI RADIATION
The two PCRs (602000030041 and 602000030042) will incorporate new remote manual operator action to open MO-2565 during accident conditions of a Loss-of-cooling Accident (LOCA) when damage to fuel is suspected. Opening of MO-2565 creates a supplemental drain path that will allow post-LOCA MSIV leakage to drain to the condenser. This supplemental drain path will not replace the credited drain path.
Current design and licensing basis assumes that MO-2565 is closed and post-LOCA MSIV leakage drains to the condenser through restricti on orifice RO-2567.
These new manual operator actions are being incorporated into alarm response procedures such that if steam leaks in the turbine building occur due to a degraded condition, these new manual operator actions that establish a supplemental MSIV leakage drain path could be credited as a compensatory measure to support the operability of the control room.
Summary of Evaluation:
The opening of MO-2565 as outlined in the two ARPs is primarily beneficial because it complements the design function of RO-2567 during the accident scenario and provides additional assurance that the radionuc lides associated with MSIV/SCB Leakage are directed to the condenser as intended. This is a USAR described design function in USAR-14 "Plant Safety Analysis". The activity has been thoroughly reviewed against the design requirements and 10CFR 50.59. The potentially adverse effects identified during the 50.59 process are discussed within this evaluation. This evaluation has shown that any adverse effects still satisfy the 8 criteria. The conclusion of this 50.59 Evaluation is that the activity does not require prior NRC approval.
Page 3 of 3 ENCLOSURE 2
MONTICELLO NUCLEAR GENERATING PLANT REPORT OF CHANGES TO LICENSEE DOCKETED COMMITMENTS
Commitments are identified and reported to the Commission in accordance with guidance provided in NEI Technical Report 99 -04 Revision 0, "Guidelines for Managing NRC Commitment Changes."
This enclosure provides a brief description and a summary of changes to commitments established with the NRC by the Monticello Nuclear Generating Plant (MNGP) per NEI 99-04 guidelines.
For Revision 41 :
There were no changes to commitments that requi re additional NRC notification.
Page 1 of 1 ENCLOSURE 3
MONTICELLO NUCLEAR GENERATING PLANT
SUMMARY
OF INFORMATION REMOVED FROM THE USAR
Consistent with the guidance in Nuclear Energy Institute (NEI) Report NEI 98-03, Guidelines for Updated Final Safety Analysis Reports, Revision 1 and Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance With 10 CFR 50.71(e), Information removed from the Monticello Nuclear Generating Plant Updated Safety Analysis Report (USAR) is summarized below.
USAR Change 604000000460 Affected sections:
Section USAR-10, Plant Auxiliary Systems 10.4.2.4, Inspection and Testing Deleted the specified quarterly frequency for full flow testing of the RHR Service Water System. Testing frequency is instead described more generally to preclude potential conflicts with Technical Specifications Surveillance Requirement frequencies.
USAR Change 604000000920 Affected sections:
Section USAR-05, Containment System 5.2.4.3, Containment Isolation Valves Deleted the specified refuel outage frequency for testing of excess flow check valves on instrument lines connected to the reactor coolant system penetrating containment. Testing frequency is instead referenced to Technical Specifications Surveillance Requirement 3.6.1.3.8.
Page 1 of 1 ENCLOSURE 4
MONTICELLO NUCLEAR GENERATING PLANT
USAR REVISION 41
(NON-PROPRIETARY, REDACTED FOR PUBLIC USE)
ENCLOSURE 5 CONTAINS SECURITY-RELATED INFORMATION WITHHOLD UNDER 10 CFR 2.390
ENCLOSURE 5
MONTICELLO NUCLEAR GENERATING PLANT
USAR REVISION 41
(PROPRIETARY, SENSITIVE INFORMATION, NOT FOR PUBLIC USE)
ENCLOSURE 5 CONTAINS SECURITY-RELATED INFORMATION WITHHOLD UNDER 10 CFR 2.390 ENCLOSURE 6
MONTICELLO NUCLEAR GENERATING PLANT REPORT OF CHANGES TO THE MONTICELLO FIRE PROTECTION PROGRAM
This enclosure contains a report of changes to the Monticello Fire Protection Program (FPP) in accordance with the provisions of Generic Letter (GL) 86-10.
In conformance with GL 86-10, the Updated Fire Hazards Analysis (UFHA) and the Safe Shutdown Analysis (SSDA) are incorporated directly into the Updated Safety Analysis Report (USAR) as Appendix J.5 and J.4 respectively. The following summarizes fire protection program document changes since the previous submittal.
- 1. USAR J.04, SAFE SHUTDOWN ANALYSIS
Revision 41P No changes to J.04 in revision 41P
- 2. USAR J.05, FIRE HAZARDS ANALYSIS
Revision 41P Licensing Document Change 604000000891 o Pg 547, Fire Hazards Analysis Matrix for Fire Zone 1D updated to include a comment for fire penetration FZ-1833 o Pg 638, Fire Hazards Analysis Matrix for Fire Zone 12D updated to include a comment for fire penetration FZ-1833 o Pg. 769, Attachment D Calculation/Evaluation Index updated to include FPEE 23-001, Evaluation of Fire Penetration FZ-1833 Seal Design.
- 3. FIREPROTECT, FIRE PROTECTION PROGRAM PLAN (was 4 AWI-08.01.00)
Revision 03 PCR 602000029732 o Pg 7, Program Engineering Supervisor title updated to say Program Engineering Manager o Pg 15, Changed Fire Protection Inspector education requirements to align with PI and added or equivalent for Plant Systems Introduction to make it consistent with other training bullets.
o Pgs 18-22, 25, 28, 35 revised to correct superseded procedure IDs with current procedure IDs.
Revision 04 PCR 602000031832
Page 1 of 2 o Pg 37, Updated Figure 5.3 barrier inspection criteria to remove criteria that is not supported (QIM 501000075989) and included language that is already in the procedure in a different location.
- 4. B.08.05-05, FIRE PROTECTION - SYSTEM OPERATION, TABLES A.2-1, A.2-2, A.2-3 AND A.2-4
None
In the 2021 enclosure submittial, Revision 80 was listed but it was actually Revision 81. Revision 80 was included in the 2020 enclosure submittal. The only issue was the incorrect revision number as the all the relevant information to the changes made to the procedure were correct. QIM CAP 501000079482
Page 2 of 2 Enclosure 7 Report Consistent with 10 CFR 54.37(b)
This summary report is in lieu of adding a level of detail to the Monticello Nuclear Generating Plant (MNGP) Updated Safety Analysis Report (USAR) that is greater in the remainder of the USAR, including the License Renewal Supplement in Appendix K (USAR-K). The contents of this report are consistent with the requirements of 10 CFR 54.37(b) and the guidance of Regulatory Issue Summary (RIS) 2007-16, Rev. 1, Implementation of the Requirements of 10 CFR 54.37(b) for Holders of Renewed Licenses.
Corrective actions from condition reports 501000072537 and 501000068729 identified new material and environment combinations that result in new Aging Effects Requiring Management (AERM) for various sections of piping within scope of license renewal. These were identified during the extent of condition (EOC) completed in response to the Control Rod Drive Suction Line Leak prior to refueling outage 31.
Control Rod Drive (CRD) Suction Line Leak Summary - New AERMs and Environments.
Resulting from the CRD suction line leak, an extent of condition was performed to inspect additional piping segments that traverse the seismic gap between the reactor building and turbine building. As part of this EOC, new material and environment combinations were identified for piping in scope of license renewal.
The CRD Suction Line (CRD-3-GB) was replaced during RFO31 in addition to Radwaste piping (RWN60-1 1/2-HC) with stainless steel and are thus not subject to the provisions of 10 CFR 54.37(b). These newly installed sections of pipe are at an elevation that is susceptible to the effects of groundwater making contact.
Pipe segments within the following systems were added to the System Condition Monitoring Aging Management Program (AMP) with a below grade and/or outside air environment :
Material/Environment Combination Carbon Steel/Outside Air Carbon Steel/ Stainless Below Grade Steel/Outside Air Condensate and Feedwater Radwaste Radwaste Radwaste Reactor Water Cleanup Demineralized Water Main Steam Condensate Storage Alternate Nitrogen Systems Reactor Water Cleanup Emergency Service Water Heating and Ventilation Service and Seal Water
1 of 7 The Material/Environment/Aging Effects figure within the System Condition Monitoring AMP procedure EWI-11.01.13 has been updated to reflect these changes. The following material and environment combinations were not described in the license renewal application:
Table 1: Condensate and Feedwater System List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Management Program Effect Evaluation (AERMs) (AMPs)
(AEE)
Piping and Carbon Steel Outside Air MNGP Loss of EWI-Fittings LRA Material due 11.01.13 3.4.2-2 to general MNGP and galvanic System corrosion Condition Monitoring program (LRA Section B2.1.32)
2 of 7 Table 2: Radwaste List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Management Program Effect Evaluation (AERMs) (AMPs)
(AEE)
Piping and Carbon Steel Outside Air MNGP Loss of EWI-Fittings LRA Material due to 11.01.13 Table general and MNGP 3.3.2-galvanic System 13 corrosion Condition Monitoring program (LRA Section B2.1.32)
Below MNGP Loss of EWI-Grade LRA Material due to 11.01.13 Table Crevice, MNGP 3.3.2-Galvanic, System 13 general, MIC, Condition and Pitting Monitoring corrosion program (LRA Section B2.1.32)
Stainless Steel Outside MNGP None None Air LRA Table 3.3.2-13
3 of 7 Table 3: Main Steam List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Effect Management Program Evaluation (AERMs) (AMPs)
(AEE)
Piping and Carbon Steel Outside Air MNGP LRA Loss of EWI-Fittings Table 3.4.2-Material due to 11.01.13 4 general and MNGP galvanic System corrosion Condition Monitoring program (LRA Section B2.1.32)
Table 4: Reactor Water Cleanup List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Effect Management Program Evaluation (AERMs) (AMPs)
(AEE)
Piping and Carbon Steel Outside Air MNGP LRA Loss of EWI-Fittings Table 3.3.2-Material due to 11.01.13 15 general and MNGP galvanic System corrosion Condition Monitoring program (LRA Section B2.1.32)
Below MNGP Loss of EWI-Grade LRA Table Material due to 11.01.13 3.3.2-15 Crevice, MNGP Galvanic, System General, MIC, Condition and Pitting Monitoring corrosion program (LRA Section B2.1.32)
4 of 7 Table 5: Demineralized Water System List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Effect Management Program Evaluation (AERMs) (AMPs)
(AEE)
Piping and Stainless Steel Outside Air MNGP LRA None None Fittings Table 3.3.2-5
Table 6: Alternate Nitrogen List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Effect Management Program Evaluation (AERMs) (AMPs)
(AEE)
Piping and Stainless Steel Outside Air MNGP LRA None None Fittings 3.3.2-1
Table 7: Emergency Service Water List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Effect Management Program Evaluation (AERMs) (AMPs)
(AEE)
Piping and Carbon Steel Outside Air MNGP LRA Loss of EWI-Fittings Table 3.3.2-Material due to 11.01.13 8 general and MNGP galvanic System corrosion Condition Monitoring program (LRA Section B2.1.32)
5 of 7 Table 8: Heating and Ventilation System List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Effect Management Program Evaluation (AERMs) (AMPs)
(AEE)
Piping and Carbon Steel Outside Air MNGP LRA Loss of EWI-Fittings Table 3.3.2-Material due to 11.01.13 11 general and MNGP galvanic System corrosion Condition Monitoring program (LRA Section B2.1.32)
Table 9: Condensate Storage System List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Effect Managemen Program t
(AEE)
Piping and Carbon Steel Below MNGP LRA Loss of EWI-Fittings Grade Table 3.4.2-Material due 11.01.13 1 to Crevice, MNGP Galvanic, System General, MIC, Condition and Pitting Monitoring corrosion program (LRA Section B2.1.32)
6 of 7 Table 10: Service and Seal Water System List all the components or commodity groups that apply to the following Aging Management Review Table (Attach a list to this form if the number of items is extensive):
Complete additional tables so all components or commodity groups from the scope of this form are reviewed.
Attach additional sheets if necessary.
Component Material Environment 3.x.2 Table Aging Effect Aging Type Internal External Ref. or Requiring Management Aging Management Program Effect Evaluation (AERMs) (AMPs)
(AEE)
Piping and Carbon Steel Outside Air MNGP Loss of EWI-Fittings LRA Material due to 11.01.13 Table general and MNGP 3.3.2-galvanic System 16 corrosion Condition Monitoring program (LRA Section B2.1.32)
7 of 7