ML18106A634
| ML18106A634 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/31/2018 |
| From: | Peter Hearn NRC/NRO/DNRL/LB4 |
| To: | Whitley B Southern Nuclear Engineering |
| HEARN P/415-1189 | |
| Shared Package | |
| ML18106A626 | List: |
| References | |
| LAR 17-042, EPID L-2017-LLA-0409 | |
| Download: ML18106A634 (39) | |
Text
ATTACHMENT TO LICENSE AMENDMENT NO. 125 TO FACILITY COMBINED LICENSE NO. NPF-91 DOCKET NO.52-025 Replace the following pages of the Facility Combined License No. NPF-91 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Combined License No. NPF-91 REMOVE INSERT 2
2 3
3 4
4 5
5 6
6 7
7 8
8 9
9 10 10 11 11 12 12 16 16 Appendix C to Facility Combined License No. NPF-91 REMOVE INSERT C-51 C-51 C-52 C-52 C-52a C-52a C-55 C-55 C-56 C-56 C-57 C-57 C-93 C-93 C-101 C-101 C-172 C-172
Appendix C to Facility Combined License No. NPF-91 (continued)
REMOVE INSERT C-197 C-197 C-199 C-199 C-215 C-215 C-227 C-227 C-311 C-311 C-319 C-319 C-370 C-370 C-381 C-381 C-382 C-382 C-403 C-403 C-404 C-404 C-405 C-405 C-406 C-406 C-443 C-443 C-445 C-445 C-447a C-447a
2 Amendment No. 125 D.
SNC2 is technically qualified to engage in the activities authorized by this license in accordance with the Commission regulations set forth in 10 CFR Chapter I.
SNC and the VEGP owners together are financially qualified to engage in the activities authorized by this COL in accordance with the Commission regulations set forth in 10 CFR Chapter I; E.
SNC and the VEGP owners have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements;"
F.
The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; G.
After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering reasonable available alternatives, the issuance of this license subject to the conditions for protection of the environment set forth herein is in accordance with Subpart A of 10 CFR Part 51 and all applicable requirements have been satisfied; and H.
The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by this license will be in accordance with the applicable regulations in 10 CFR Parts 30, 40, and 70.
- 2.
On the basis of the foregoing findings regarding this facility, COL No. NPF-91 is hereby issued to SNC, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and the City of Dalton, Georgia (the licensees) to read as follows:
A.
This license applies to the VEGP Unit 3, a light-water nuclear reactor and associated equipment (the facility), owned by the VEGP Owners. The facility would be located adjacent to existing VEGP Units 1 and 2 on a 3,169-acre coastal plain bluff on the southwest side of the Savannah River in eastern Burke County, GA, approximately 15 miles east-northeast of Waynesboro, GA, and 26 miles southeast of Augusta, GA, and is described in the licensees' updated final safety analysis report (UFSAR), as supplemented and amended.
B.
Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:
(1)
SNC pursuant to Sections 103 and 185b. of the Act and 10 CFR Part 52, to construct, possess, use, and operate the facility at the designated location in accordance with the procedures and limitations set forth in this license; (2)
The VEGP owners pursuant to the Act and 10 CFR Part 52, to possess but not operate the facility at the designated location in Burke County, GA, in accordance with the procedures and limitations set forth in this license; 2 SNC is authorized by the VEGP owners to exercise responsibility and control over the physical construction, operation, and maintenance of the facility.
3 Amendment No. 125 (3)
(a)
SNC pursuant to the Act and 10 CFR Part 70, to receive and possess at any time, special nuclear material as reactor fuel, in accordance with the limitations for storage and in amounts necessary for reactor operation, described in the UFSAR, as supplemented and amended; (b)
SNC pursuant to the Act and 10 CFR Part 70, to use special nuclear material as reactor fuel, after a Commission finding under 10 CFR 52.103(g) has been made, in accordance with the limitations for storage and in amounts necessary for reactor operation, described in the UFSAR, as supplemented and amended; (4)
(a)
SNC pursuant to the Act and 10 CFR Parts 30 and 70, to receive, possess, and use, at any time before a Commission finding under 10 CFR 52.103(g), such byproduct and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts, as necessary; (b)
SNC pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, after a Commission finding under 10 CFR 52.103(g), any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as necessary; (5)
(a)
SNC pursuant to the Act and 10 CFR Parts 30 and 70, to receive, possess, and use, before a Commission finding under 10 CFR 52.103(g),
in amounts not exceeding those specified in 10 CFR 30.72, any byproduct or special nuclear material that is (1) in unsealed form; (2) on foils or plated surfaces, or (3) sealed in glass, for sample analysis or instrument calibration or other activity associated with radioactive apparatus or components; (b)
SNC pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, after a Commission finding under 10 CFR 52.103(g), in amounts as necessary, any byproduct, source, or special nuclear material without restriction as to chemical or physical form, for sample analysis or instrument calibration or other activity associated with radioactive apparatus or components but not uranium hexafluoride; and (6)
SNC pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
The license is subject to, and the licensees shall comply with, all applicable provisions of the Act and the rules, regulations, and orders of the Commission, including the conditions set forth in 10 CFR Chapter I, now or hereafter in effect.
4 Amendment No. 125 D.
The license is subject to, and SNC shall comply with, the conditions specified and incorporated below:
(1)
Changes during Construction (a)
SNC may request use of a preliminary amendment request (PAR) process, for license amendments, at any time before a Commission finding under 10 CFR 52.103(g). To use the PAR process, SNC shall submit a written request to the Office of New Reactors (NRO) in accordance with COL-ISG-025, Changes during Construction under Part 52.
(b)
Before NROs issuance of a written PAR notification, SNC shall submit the license amendment request (LAR). Thereafter, NRO will issue a written PAR notification, setting forth whether SNC may proceed in accordance with the PAR, LAR, and COL-ISG-025. If SNC elects to proceed and the LAR is subsequently denied, SNC shall return the facility to its current licensing basis.
(2)
Pre-operational Testing (a)
SNC shall perform the design-specific pre-operational tests identified below:
- 1.
In-Containment Refueling Water Storage Tank (IRWST)
Heatup Test (first plant test as identified in UFSAR Section 14.2.9.1.3 Item (h));
- 2.
Pressurizer Surge Line Stratification Evaluation (first plant test as identified in UFSAR Section 14.2.9.1.7 Item (d));
- 3.
Reactor Vessel Internals Vibration Testing (first plant test as identified in UFSAR Section 14.2.9.1.9);
- 4.
Core Makeup Tank Heated Recirculation Tests (first three plants test as identified in UFSAR Section 14.2.9.1.3 Items (k) and (w)); and
- 5.
Automatic Depressurization System Blowdown Test (first three plants test as identified in UFSAR Section 14.2.9.1.3 Item (s)).
(b)
SNC shall review and evaluate the results of the tests identified in Section 2.D.(2)(a) of this license and confirm that these test results are within the range of acceptable values predicted or
5 Amendment No. 125 otherwise confirm that the tested systems perform their specified functions in accordance with UFSAR Section 14.2.9, (c)
SNC shall notify the Director of NRO, or the Directors designee, in writing, upon successful completion of the design-specific pre-operational tests identified in Section 2.D.(2)(a) of this license; and (d)
SNC shall notify the Director of NRO, or the Directors designee, in writing, upon the successful completion of all the ITAAC included in Appendix C to this license.
(3)
Nuclear Fuel Loading and Pre-critical Testing (a)
Until the submission of the notification required by Section 2.D.(2)(c) of this license, SNC shall not load fuel into the reactor vessel; (b)
Upon submission of the notification required by Section 2.D.(2)(c) of this license and upon a Commission finding in accordance with 10 CFR 52.103(g) that all the acceptance criteria in the ITAAC in Appendix C to this license are met, SNC is authorized to perform pre-critical tests in accordance with the conditions specified herein; (c)
SNC shall perform the pre-critical tests identified in UFSAR Section 14.2.10.1; (d)
SNC shall review and evaluate the results of the tests identified in Section 2.D.(3)(c) of this license and confirm that these test results are within the range of acceptable values predicted or otherwise confirm that the tested systems perform their specified functions in accordance with UFSAR Section 14.2.10; and (e)
SNC shall notify the Director of NRO, or the Directors designee, in writing, upon successful completion of the pre-critical tests identified in Section 2.D.(3)(c) of this license.
(4)
Initial Criticality and Low-Power Testing (a)
Upon submission of the notification required by Section 2.D.(3)(e) of this license, SNC is authorized to operate the facility at reactor steady-state core power levels not to exceed 5-percent thermal power in accordance with the conditions specified herein; (b)
SNC shall perform the initial criticality and low-power tests identified in UFSAR Sections 14.2.10.2 and 14.2.10.3, respectively, the Natural Circulation (first plant test) identified in UFSAR Section 14.2.10.3.6, and the
6 Amendment No. 125 Passive Residual Heat Removal Heat Exchanger (first plant test) identified in UFSAR Section 14.2.10.3.7; (c)
SNC shall review and evaluate the results of the tests identified in Section 2.D.(4)(b) of this license and confirm that these test results are within the range of acceptable values predicted or otherwise confirm that the tested systems perform their specified functions in accordance with UFSAR Sections 14.2.10.2 and 14.2.10.3; and (d)
SNC shall notify the Director of NRO, or the Directors designee, in writing, upon successful completion of initial criticality and low-power tests identified in Section 2.D.(4)(b) of this license, including the design-specific tests identified therein.
(5)
Power Ascension Testing (a) Upon submission of the notification required by Section 2.D.(4)(d) of this license, SNC is authorized to operate the facility at reactor steady-state core power levels not to exceed 100-percent thermal power in accordance with the conditions specified herein, but only for the purpose of performing power ascension testing; (b)
SNC shall perform the power ascension tests identified in UFSAR Section 14.2.10.4, the Rod Cluster Control Assembly Out of Bank Measurements (first plant test) identified in UFSAR Section 14.2.10.4.6, and the Load Follow Demonstration (first plant test) identified in UFSAR Section 14.2.10.4.22; (c)
SNC shall review and evaluate the results of the tests identified in Section 2.D.(5)(b) of this license and confirm that these test results are within the range of acceptable values predicted or otherwise confirm that the tested systems perform their specified functions in accordance with UFSAR Section 14.2.10.4; and (d)
SNC shall notify the Director of NRO, or the Directors designee, in writing, upon successful completion of power ascension tests identified in Section 2.D.(5)(b) of this license, including the design-specific tests identified therein.
(6)
Maximum Power Level Upon submission of the notification required by Section 2.D.(5)(d) of this license, SNC is authorized to operate the facility at steady state reactor core power levels not to exceed 3400 MW thermal (100-percent thermal power), as described in the UFSAR, in accordance with the conditions specified herein.
7 Amendment No. 125 (7)
Reporting Requirements (a)
Within 30 days of a change to the initial test program described in UFSAR Section 14, Initial Test Program, made in accordance with 10 CFR 50.59 or in accordance with 10 CFR Part 52, Appendix D, Section VIII, Processes for Changes and Departures, SNC shall report the change to the Director of NRO, or the Directors designee, in accordance with 10 CFR 50.59(d).
(b)
SNC shall report any violation of a requirement in Section 2.D.(3),
Section 2.D.(4), Section 2.D.(5), and Section 2.D.(6) of this license within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initial notification shall be made to the NRC Operations Center in accordance with 10 CFR 50.72, with written follow up in accordance with 10 CFR 50.73.
(8)
Incorporation The Technical Specifications, Environmental Protection Plan, and ITAAC in Appendices A, B, and C, respectively of this license, as revised through Amendment No. 125, are hereby incorporated into this license.
(9)
Technical Specifications The technical specifications in Appendix A to this license become effective upon a Commission finding that the acceptance criteria in this license (ITAAC) are met in accordance with 10 CFR 52.103(g).
(10)
Operational Program Implementation SNC shall implement the programs or portions of programs identified below, on or before the date SNC achieves the following milestones:
(a)
Environmental Qualification Program implemented before initial fuel load; (b)
Reactor Vessel Material Surveillance Program implemented before initial criticality; (c)
Preservice Testing Program implemented before initial fuel load; (d)
Containment Leakage Rate Testing Program implemented before initial fuel load; (e)
- 1.
The fire protection measures in accordance with Regulatory Guide (RG) 1.189 for designated storage building areas (including adjacent fire areas that could affect the storage area) implemented before initial receipt
8 Amendment No. 125 of byproduct or special nuclear materials that are not fuel (excluding exempt quantities as described in 10 CFR 30.18);
- 2.
The fire protection measures in accordance with RG 1.189 for areas containing new fuel (including adjacent areas where a fire could affect the new fuel) implemented before receipt of fuel onsite;
- 3.
All fire protection program features implemented before initial fuel load; (f)
Standard Radiological Effluent Controls implemented before initial fuel load; (g)
Offsite Dose Calculation Manual implemented before initial fuel load; (h)
Radiological Environmental Monitoring Program implemented before initial fuel load; (i)
Process Control Program implemented before initial fuel load; (j)
Radiation Protection Program (RPP) (including the ALARA principle) or applicable portions as identified in UFSAR Section 12.5 thereof:
- 1.
RPP features applicable to receipt of by-product, source, or special nuclear materials (excluding exempt quantities as described in 10 CFR 30.18) implemented before initial receipt of such materials;
- 2.
RPP features (including the ALARA principle) applicable to new fuel implemented before receipt of initial fuel on site;
- 3.
All other RPP features (including the ALARA principle) except for those applicable to control radioactive waste shipment implemented before initial fuel load;
- 4.
RPP features (including the ALARA principle) applicable to radioactive waste shipment implemented before first shipment of radioactive waste; (k)
Reactor Operator Training Program implemented 18 months before the scheduled date of initial fuel load; (l)
Motor-Operated Valve Testing Program implemented before initial fuel load;
9 Amendment No. 125 (m)
Initial Test Program (ITP)
- 1.
Component Test Program implemented before the first component test;
- 2.
Preoperational Test Program implemented before the first preoperational test; and
- 3.
Startup Test Program implemented before initial fuel load; (n)
Special Nuclear Material Control and Accounting Program implemented before initial receipt of special nuclear material; and (o)
Special Nuclear Material Physical Protection Program implemented before initial receipt of special nuclear material on site.
(11)
Operational Program Implementation Schedule No later than 12 months after issuance of the COL, SNC shall submit to the Director of NRO, or the Directors designee, a schedule for implementation of the operational programs listed in UFSAR Table 13.4-201, including the associated estimated date for initial loading of fuel. The schedule shall be updated every 6 months until 12 months before scheduled fuel loading, and every month thereafter until all the operational programs listed in UFSAR Table 13.4-201 have been fully implemented.
(12)
Site-and Unit-specific Conditions (a)
SNC shall either remove and replace, or shall improve, the soils directly above the blue bluff marl for soils under or adjacent to Seismic Category I structures, to eliminate any liquefaction potential.
(b)
Before commencing installation of individual piping segments and connected components in their final locations, SNC shall complete the as-designed pipe rupture hazards analysis for compartments (rooms) containing those segments in accordance with the criteria outlined in the UFSAR Sections 3.6.1.3.2 and 3.6.2.5, and shall inform the Director of NRO, or the Directors designee, in writing, upon the completion of this analysis and the availability of the as-designed pipe rupture hazards analysis reports.
(c)
Before commencing installation of individual piping segments, identified in UFSAR Section 3.9.8.7, and connected components in their final locations in the facility, SNC shall complete the analysis of the as-designed individual piping segments and shall inform the Director of NRO, or the Directors
10 Amendment No. 125 designee, in writing, upon the completion of these analyses and the availability of the design reports for the selected piping packages.
(d)
No later than 180 days before initial fuel load, SNC shall submit to the Director of NRO, or the Directors designee, in writing, a fully developed set of plant-specific emergency action levels (EALs) for VEGP Unit 3 in accordance with the criteria defined in Amendment No. 77. The EALs shall have been discussed and agreed upon with State and local officials.
No later than 180 days before initial fuel load, SNC shall submit to the Director of NRO, or the Directors designee, in writing, an assessment of emergency response staffing performed in accordance with NEI 10-05, Assessment of On-Shift Emergency Response Organization Staffing and Capabilities, Revision 0.
(e)
SNC shall not revise or modify the provisions of Sections 5.3, 5.4, 5.6, 5.9, and 5.10 of the Special Nuclear Material (SNM) Physical Protection Program until the requirements of 10 CFR 73.55 are implemented.
(f)
No later than 12 months after issuance of the COL, SNC shall submit to the Director of NRO, or the Directors designee, a schedule for implementation of the following license conditions.
The schedule shall be updated every 6 months until 12 months before scheduled fuel loading, and every month thereafter until each license condition has been fully implemented. The schedule shall identify the completion of or implementation of the following:
- 1.
The construction and inspection procedures for steel concrete composite (SC) construction activities for seismic Category I nuclear island modules (including shield building SC modules) described in UFSAR Section 3.8.4.8;
- 2.
The spent fuel rack Metamic Coupon monitoring program (before initial fuel load);
- 3.
Implementation of the flow accelerated corrosion (FAC) program including construction phase activities (before initial fuel load);
- 4.
A turbine maintenance and inspection program, which must be consistent with the maintenance and inspection program plan activities and inspection intervals identified in UFSAR Section 10.2.3.6 (before initial fuel load);
- 5.
The availability of documented instrumentation uncertainties to calculate a power calorimetric uncertainty (before initial fuel load);
11 Amendment No. 125
- 6.
The availability of administrative controls to implement maintenance and contingency activities related to the power calorimetric uncertainty instrumentation (before initial fuel load);
- 7.
The site-specific severe accident management guidelines (before startup testing);
- 8.
The operational and programmatic elements of the mitigative strategies for responding to circumstances associated with loss of large areas of the plant due to explosions or fire developed in accordance with 10 CFR 50.54(hh)(2) (before initial fuel load); and
- 9.
The ITP procedures identified in U FSAR Section 14.2.3:
- a. administrative manual (before the first component test)
- b. preoperational testing (before scheduled performance)
- c. startup testing (before initial fuel load)
(g)
Before initial fuel load, SNC shall:
- 1.
Update the seismic interaction analysis in UFSAR Section 3.7.5.3 to reflect as-built information, which must be based on as-procured data, as well as the as-constructed condition;
- 2.
Reconcile the seismic analyses described in Section 3.7.2 of the UFSAR, to account for detailed design changes, including, but not limited to, those due to as-procured or as-built changes in component mass, center of gravity, and support configuration based on as-procured equipment information;
- 3.
Calculate the instrumentation uncertainties of the actual plant operating instrumentation to confirm that either the design limit departure from nucleate boiling ratio (DNBR) values remain valid or that the safety analysis minimum DNBR bounds the new design limit DNBR values plus DNBR penalties;
- 4.
Update the pressure temperature (P-T) limits using the pressure temperature limits report (PTLR) methodologies approved in the UFSAR, using the plant-specific material properties or confirm that the reactor vessel material properties meet the specifications of and use the Westinghouse generic PTLR curves;
- 5.
Verify that plant-specific belt line material properties are consistent with the properties given in UFSAR Section 5.3.3.1 and Tables 5.3-1 and 5.3-3. The verification must include a pressurized thermal shock
12 Amendment No. 125 (PTS) evaluation based on as-procured reactor vessel material data and the projected neutron fluence for the plant design objective. Submit this PTS evaluation report to the Director of NRO, or the Directors designee, in writing, at least 18 months before initial fuel load;
- 6.
Review differences between the as-built plant and the design used as the basis for the AP1000 seismic margin analysis. SNC shall perform a verification walkdown to identify differences between the as-built plant and the design. SNC shall evaluate any differences and must modify the seismic margin analysis as necessary to account for the plant-specific design and any design changes or departures from the certified design. SNC shall compare the as-built structures, systems, and components (SSC) high confidence, low probability of failures (HCLPFs) with those assumed in the AP1000 seismic margin evaluation, before initial fuel load. SNC shall evaluate deviations from the HCLPF values or assumptions in the seismic margin evaluation due to the as-built configuration and final analysis to determine if vulnerabilities have been introduced;
- 7.
Review differences between the as-built plant and the design used as the basis for the AP1000 probabilistic risk assessment (PRA) and UFSAR Table 19.59-18. SNC shall evaluate the plant-specific PRA-based insight differences and shall modify the plant-specific PRA model as necessary to account for the plant-specific design and any design changes or departure from the PRA certified in Rev. 19 of the AP1000 DCD;
- 8.
Review differences between the as-built plant and the design used as the basis for the AP1000 internal fire and internal flood analysis. SNC shall evaluate the plant-specific internal fire and internal flood analyses and shall modify the analyses as necessary to account for the plant-specific design and any design changes or departures from the design certified in Rev. 19 of the AP1000 DCD; and
- 9.
Perform a thermal lag assessment of the as-built equipment listed in Tables 6b and 6c in Attachment A of APP-GW-GLR-069, Equipment Survivability Assessment, to provide additional assurance that this equipment can perform its severe accident functions during environmental conditions resulting from hydrogen burns associated with severe accidents. SNC shall perform this assessment for equipment used for severe accident mitigation that has not been tested at severe accident conditions. SNC shall assess the ability of the as-built equipment to perform
16 Amendment No. 125 certified design. This exemption is specific to the organization and numbering scheme in the FSAR and is related to departure number VEGP DEP 1-1.
(2)
The following exemptions from regulations were granted in the rulemaking for the design certification rule that is referenced in the application. In accordance with 10 CFR Part 52, Appendix D, Section V, Applicable Regulations, Subsection B, and pursuant to 10 CFR 52.63(a)(5), the licensees are exempt from portions of the following regulations:
(a)
Paragraph (f)(2)(iv) of 10 CFR 50.34Plant Safety Parameter Display Console; (b)
Paragraph (c)(1) of 10 CFR 50.62Auxiliary (or emergency) feedwater system; and (c)
Appendix A to 10 CFR Part 50, GDC 17Second offsite power supply circuit.
(3)
For the reasons set forth below, the following specific exemptions, which are outside the scope of the design certification rule referenced in the application, are granted:
(a)
The licensees are exempt from the requirements of 10 CFR 70.22(b), 10 CFR 70.32(c), 10 CFR 74.31, 10 CFR 74.41, and 10 CFR 74.51 because the licensees meet the requirements of 10 CFR 70.17 and 74.7 as follows: The exemption is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Additionally, special circumstances are present in that the application of the regulations in this particular circumstance is not necessary to achieve the underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)) as described in the COL Application and the staff SER dated August 5, 2011.
(b)
The licensees are exempt from the requirements of 10 CFR 52.93(a)(1) as it relates to the exemption granted in Section 2.F.(1)(a) of this license because the exemption meets the requirements of 10 CFR 52.7, because the exemption is authorized by law, will not present an undue risk to the public health or safety, and is consistent with the common defense and security. Additionally, special circumstances are present in that the application of the regulation in this particular circumstance is not necessary to achieve the underlying purpose of the rule (10 CFR 50.12(a)(2)(ii)) as described in the staff SER dated August 5, 2011.
C-51 Amendment No. 125 Table 2.1.2-1 Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS/
DAS Active Function Loss of Motive Power Position Pressurizer Safety Valve RCS-PL-V005B Yes Yes No
-/-
No
-/-
Transfer Open/
Transfer Closed First-stage ADS Motor-operated Valve (MOV)
RCS-PL-V001A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is First-stage ADS MOV RCS-PL-V001B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Second-stage ADS MOV RCS-PL-V002A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Second-stage ADS MOV RCS-PL-V002B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Third-stage ADS MOV RCS-PL-V003A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Third-stage ADS MOV RCS-PL-V003B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Fourth-stage ADS Squib Valve RCS-PL-V004A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Fourth-stage ADS Squib Valve RCS-PL-V004B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Fourth-stage ADS Squib Valve RCS-PL-V004C Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Fourth-stage ADS Squib Valve RCS-PL-V004D Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is ADS Discharge Header A Vacuum Relief Valve RCS-PL-V010A Yes Yes No No/Yes No No/No Transfer Open ADS Discharge Header B Vacuum Relief Valve RCS-PL-V010B Yes Yes No No/Yes No No/No Transfer Open First-stage ADS Isolation MOV RCS-PL-V011A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is First-stage ADS Isolation MOV RCS-PL-V011B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Second-stage ADS Isolation MOV RCS-PL-V012A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is
C-52 Amendment No. 125 Table 2.1.2-1 Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS/
DAS Active Function Loss of Motive Power Position Second-stage ADS Isolation MOV RCS-PL-V012B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Third-stage ADS Isolation MOV RCS-PL-V013A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Third-stage ADS Isolation MOV RCS-PL-V013B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/Yes Transfer Open As Is Fourth-stage ADS MOV RCS-PL-V014A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No None As Is Fourth-stage ADS MOV RCS-PL-V014B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No None As Is Fourth-stage ADS MOV RCS-PL-V014C Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No None As Is Fourth-stage ADS MOV RCS-PL-V014D Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No None As Is Reactor Vessel Head Vent Valve RCS-PL-V150A Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No Transfer Closed/
Transfer Open Closed Reactor Vessel Head Vent Valve RCS-PL-V150B Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No Transfer Closed/
Transfer Open Closed Reactor Vessel Head Vent Valve RCS-PL-V150C Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No Transfer Closed/
Transfer Open Closed Reactor Vessel Head Vent Valve RCS-PL-V150D Yes Yes Yes Yes/Yes Yes (Valve Position)
Yes/No Transfer Closed/
Transfer Open Closed RCS Hot Leg 1 Flow Sensor RCS-101A Yes Yes/Yes Yes
-/-
RCS Hot Leg 1 Flow Sensor RCS-101B Yes Yes/Yes Yes
-/-
RCS Hot Leg 1 Flow Sensor RCS-101C Yes Yes/Yes Yes
-/-
RCS Hot Leg 1 Flow Sensor RCS-101D Yes Yes/Yes Yes
-/-
C-52a Amendment No. 125 Table 2.1.2-1 Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS/
DAS Active Function Loss of Motive Power Position RCS Hot Leg 2 Flow Sensor RCS-102A Yes Yes/Yes Yes
-/-
RCS Hot Leg 2 Flow Sensor RCS-102B Yes Yes/Yes Yes
-/-
RCS Hot Leg 2 Flow Sensor RCS-102C Yes Yes/Yes Yes
-/-
RCS Hot Leg 2 Flow Sensor RCS-102D Yes Yes/Yes Yes
-/-
C-55 Amendment No. 125 Table 2.1.2-1 Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS/
DAS Active Function Loss of Motive Power Position RCS Hot Leg 1 Level Sensor RCS-160A Yes Yes/Yes Yes
-/-
RCS Hot Leg 2 Level Sensor RCS-160B Yes Yes/Yes Yes
-/-
Passive Residual Heat Removal (PRHR) Return Line Temperature Sensor RCS-161 Yes Yes/Yes Yes
-/-
Pressurizer Pressure Sensor RCS-191A Yes Yes/Yes Yes
-/-
Pressurizer Pressure Sensor RCS-191B Yes Yes/Yes Yes
-/-
Pressurizer Pressure Sensor RCS-191C Yes Yes/Yes Yes
-/-
Pressurizer Pressure Sensor RCS-191D Yes Yes/Yes Yes
-/-
Pressurizer Level Reference Leg Temperature Sensor RCS-193A Yes Yes/Yes Yes
-/-
Pressurizer Level Reference Leg Temperature Sensor RCS-193B Yes Yes/Yes Yes
-/-
Pressurizer Level Reference Leg Temperature Sensor RCS-193C Yes Yes/Yes Yes
-/-
Pressurizer Level Reference Leg Temperature Sensor RCS-193D Yes Yes/Yes Yes
-/-
Pressurizer Level Sensor RCS-195A Yes Yes/Yes Yes
-/-
Pressurizer Level Sensor RCS-195B Yes Yes/Yes Yes
-/-
Pressurizer Level Sensor RCS-195C Yes Yes/Yes Yes
-/-
Pressurizer Level Sensor RCS-195D Yes Yes/Yes Yes
-/-
RCP 1A Bearing Water Temperature Sensor RCS-211A Yes Yes/Yes Yes
-/-
C-56 Amendment No. 125 Table 2.1.2-1 Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS/
DAS Active Function Loss of Motive Power Position RCP 1A Bearing Water Temperature Sensor RCS-211B Yes Yes/Yes Yes
-/-
RCP 1A Bearing Water Temperature Sensor RCS-211C Yes Yes/Yes Yes
-/-
RCP 1A Bearing Water Temperature Sensor RCS-211D Yes Yes/Yes Yes
-/-
RCP 1B Bearing Water Temperature Sensor RCS-212A Yes Yes/Yes Yes
-/-
RCP 1B Bearing Water Temperature Sensor RCS-212B Yes Yes/Yes Yes
-/-
RCP 1B Bearing Water Temperature Sensor RCS-212C Yes Yes/Yes Yes
-/-
RCP 1B Bearing Water Temperature Sensor RCS-212D Yes Yes/Yes Yes
-/-
RCP 2A Bearing Water Temperature Sensor RCS-213A Yes Yes/Yes Yes
-/-
RCP 2A Bearing Water Temperature Sensor RCS-213B Yes Yes/Yes Yes
-/-
RCP 2A Bearing Water Temperature Sensor RCS-213C Yes Yes/Yes Yes
-/-
RCP 2A Bearing Water Temperature Sensor RCS-213D Yes Yes/Yes Yes
-/-
RCP 2B Bearing Water Temperature Sensor RCS-214A Yes Yes/Yes Yes
-/-
C-57 Amendment No. 125 Table 2.1.2-1 Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS/
DAS Active Function Loss of Motive Power Position RCP 2B Bearing Water Temperature Sensor RCS-214B Yes Yes/Yes Yes
-/-
RCP 2B Bearing Water Temperature Sensor RCS-214C Yes Yes/Yes Yes
-/-
RCP 2B Bearing Water Temperature Sensor RCS-214D Yes Yes/Yes Yes
-/-
RCP 1A Pump Speed Sensor RCS-281 Yes Yes/Yes Yes
-/-
RCP 1B Pump Speed Sensor RCS-282 Yes Yes/Yes Yes
-/-
RCP 2A Pump Speed Sensor RCS-283 Yes Yes/Yes Yes
-/-
RCP 2B Pump Speed Sensor RCS-284 Yes Yes/Yes Yes
-/-
Note: Dash (-) indicates not applicable.
Table 2.1.2-2 Line Name Line Number ASME Code Section III Leak Before Break Functional Capability Required Hot Legs RCS-L001A RCS-L001B Yes Yes Yes Cold Legs RCS-L002A RCS-L002B RCS-L002C RCS-L002D Yes Yes Yes Pressurizer Surge Line RCS-L003 Yes Yes Yes ADS Inlet Headers RCS-L004A/B RCS-L006A/B RCS-L030A/B RCS-L020A/B Yes Yes Yes Safety Valve Inlet Piping RCS-L005A RCS-L005B Yes Yes Yes
C-93 Amendment No. 125 Table 2.2.1-1 (cont.)
Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS/DAS Active Function Loss of Motive Power Position Upper Personnel Hatch CNS-MY-Y03 Yes Yes
-/-
-/-
Lower Personnel Hatch CNS-MY-Y04 Yes Yes
-/-
-/-
Containment Vessel CNS-MV-01 Yes Yes
-/-
-/-
Electrical Penetration P03 DAS-EY-P03Z Yes Yes No/Yes
-/-
Electrical Penetration P01 ECS-EY-P01X Yes Yes No/Yes
-/-
Electrical Penetration P02 ECS-EY-P02X Yes Yes No/Yes
-/-
Electrical Penetration P06 ECS-EY-P06Y Yes Yes No/Yes
-/-
Electrical Penetration P07 ECS-EY-P07X Yes Yes No/Yes
-/-
Electrical Penetration P09 ECS-EY-P09W Yes Yes No/Yes
-/-
Electrical Penetration P10 ECS-EY-P10W Yes Yes No/Yes
-/-
Electrical Penetration P11 IDSA-EY-P11Z Yes Yes Yes/Yes
-/-
Electrical Penetration P12 IDSA-EY-P12Y Yes Yes Yes/Yes
-/-
Electrical Penetration P13 IDSA-EY-P13Y Yes Yes Yes/Yes
-/-
Electrical Penetration P14 IDSD-EY-P14Z Yes Yes Yes/Yes
-/-
Electrical Penetration P15 IDSD-EY-P15Y Yes Yes Yes/Yes
-/-
Electrical Penetration P16 IDSD-EY-P16Y Yes Yes Yes/Yes
-/-
Electrical Penetration P17 ECS-EY-P17X Yes Yes No/Yes
-/-
Electrical Penetration P18 ECS-EY-P18X Yes Yes No/Yes
-/-
Electrical Penetration P19 ECS-EY-P19Z Yes Yes No/Yes
-/-
Electrical Penetration P20 ECS-EY-P20Z Yes Yes No/Yes
-/-
Electrical Penetration P21 EDS-EY-P21Z Yes Yes No/Yes
-/-
Electrical Penetration P22 ECS-EY-P22X Yes Yes No/Yes
-/-
C-101 Amendment No. 125
C-172 Amendment No. 125 Table 2.2.5-1 (cont.)
Equipment Name Tag No.
ASME Code Section III Seismic Cat. I Remotely Operated Valve Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Control PMS Active Function Loss of Motive Power Position Eductor Bypass Isolation Valve VES-PL-V046 Yes Yes No
-/-
No Transfer Open Pressure Regulating Valve A VES-PL-V002A Yes Yes No
-/-
No Throttle Flow Pressure Regulating Valve B VES-PL-V002B Yes Yes No
-/-
No Throttle Flow MCR Air Delivery Isolation Valve A VES-PL-V005A Yes Yes Yes Yes/No Yes Yes Transfer Open Open MCR Air Delivery Isolation Valve B VES-PL-V005B Yes Yes Yes Yes/No Yes Yes Transfer Open Open Temporary Instrument Isolation Valve A VES-PL-V018 Yes Yes No
-/-
No No Transfer Open Temporary Instrument Isolation Valve B VES-PL-V019 Yes Yes No
-/-
No No Transfer Open MCR Pressure Relief Isolation Valve A VES-PL-V022A Yes Yes Yes Yes/No Yes Yes Transfer Open Open MCR Pressure Relief Isolation Valve B VES-PL-V022B Yes Yes Yes Yes/No Yes Yes Transfer Open Open
C-197 Amendment No. 125 Table 2.3.2-4 Inspections, Tests, Analyses, and Acceptance Criteria No.
ITAAC No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 311 2.3.02.11a.iii Not used per Amendment No. 113 312 2.3.02.11a.iv Not used per Amendment No. 113 313 2.3.02.11b Not used per Amendment No. 113 314 2.3.02.12a Not used per Amendment No. 113 315 2.3.02.12b Not used per Amendment No. 113 316 2.3.02.13 Not used per Amendment No. 113 317 2.3.02.14
- 14. The nonsafety-related piping located inside containment and designated as reactor coolant pressure boundary, as identified in Table 2.3.2-2, has been designed to withstand a seismic design basis event and maintain structural integrity.
Inspection will be conducted of the as-built piping as documented in the CVS Seismic Analysis Report.
The CVS Seismic Analysis Reports exist for the non-safety related piping located inside containment and designated as reactor coolant pressure boundary as identified in Table 2.3.2-2.
Figure 2.3.2-1 Chemical and Volume Control System u
cvs RCS--~~~~-.!)Ic:I---U:::::IJ---o(] [J -
-, PURIFlCATlON
- a.
cVS-"Mi:-D2 LOOP IRC: ORC rL, MIXED BED
~CVS MV 01B CYS-MY-01AL~I~~~~ -
~~Ltl' I
CA TlON BED 0 CVS-MV-02 I DEMINERAUZER I
~--~
SUCTION -- -* -
RCS AUXJUARY SPRAY VD85
~-----l
_lREACTOR COOLANT CYS-MV-OJB tr Fll TERS
~CYS-MV-L-----l ER--ixJ--L-fl-M---WSS SPENT RESIN STORAGE TANK CYS-PL V217 C'IS MAKEUP PUMP CYS-MP-01A
-- ZINC CVS ADDITlON MAKEUP H.
CYS-PL ADDITlON V219 PUMP CYS-MP-01B BORIC ACID STORAGE TANK
..J CYS-MT-01 I
L"r'.
_"r'J
~--ows CVS-PL CVS-PL V1J6A V1J6B C-199 Amendment No. 125
C-215 Amendment No. 125
- 12. a) The motor-operated and check valves identified in Table 2.3.6-1 perform an active safety-related function to change position as indicated in the table.
b) After loss of motive power, the remotely operated valves identified in Table 2.3.6-1 assume the indicated loss of motive power position.
- 13. Controls exist in the MCR to cause the pumps identified in Table 2.3.6-3 to perform the control function.
- 14. Displays of the RNS parameters identified in Table 2.3.6-3 can be retrieved in the MCR.
C-227 Amendment No. 125 Table 2.3.6-4 Inspections, Tests, Analyses, and Acceptance Criteria No.
ITAAC No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 382 2.3.06.11a
- 10. Safety-related displays identified in Table 2.3.6-1 can be retrieved in the MCR.
Inspection will be performed for retrievability of the safety-related displays in the MCR.
Safety-related displays identified in Table 2.3.6-1 can be retrieved in the MCR.
11.a) Controls exist in the MCR to cause those remotely operated valves identified in Table 2.3.6-1 to perform active functions.
Stroke testing will be performed on the remotely operated valves identified in Table 2.3.6-1 using the controls in the MCR.
Controls in the MCR operate to cause those remotely operated valves identified in Table 2.3.6-1 to perform active functions.
11.b) The valves identified in Table 2.3.6-1 as having PMS control perform active safety functions after receiving a signal from the PMS.
Testing will be performed using real or simulated signals into the PMS.
The valves identified in Table 2.3.6-1 as having PMS control perform the active function identified in the table after receiving a signal from the PMS.
12.b) After loss of motive power, the remotely operated valves identified in Table 2.3.6-1 assume the indicated loss of motive power position.
Testing of the remotely operated valves will be performed under the conditions of loss of motive power.
Upon loss of motive power, each remotely operated valve identified in Table 2.3.6-1 assumes the indicated loss of motive power position.
- 13. Controls exist in the MCR to cause the pumps identified in Table 2.3.6-3 to perform the control function.
Testing will be performed to actuate the pumps identified in Table 2.3.6-3 using controls in the MCR.
Controls in the MCR cause pumps identified in Table 2.3.6-3 to perform the control function.
- 14. Displays of the RNS parameters identified in Table 2.3.6-3 can be retrieved in the MCR.
Inspection will be performed for retrievability in the MCR of the displays identified in Table 2.3.6-3.
Displays of the RNS parameters identified in Table 2.3.6-3 are retrieved in the MCR.
383 2.3.06.11b Not used per Amendment No. 113 384 2.3.06.12a.i 12.a) The motor-operated and check valves identified in Table 2.3.6-1 perform an active safety-related function to change position as indicated in the table.
i) Tests or type tests of motor-operated valves will be performed that demonstrate the capability of the valve to operate under its design conditions.
i) A test report exists and concludes that each motor-operated valve changes position as indicated in Table 2.3.6-1 under design conditions.
ii) Inspection will be performed for the existence of a report verifying that the as-built motor-operated valves are bounded by the tested conditions.
ii) A report exists and concludes that the as-built motor-operated valves are bounded by the tested conditions.
385 2.3.06.12a.ii Not used per Amendment No. 85
C-311 Amendment No. 125 Table 2.5.4-1 (cont.)
Minimum Inventory of Controls, Displays, and Alerts at the RSW Description Control Display Alert(1)
PRHR Outlet Temperature Yes Yes Passive Containment Cooling System (PCS) Storage Tank Water Level Yes No PCS Cooling Flow Yes No IRWST to Normal Residual Heat Removal System (RNS)
Suction Valve Status Yes Yes Remotely Operated Containment Isolation Valve Status Yes No Containment Area High-range Radiation Level Yes Yes Containment Pressure (Extended Range)
Yes No Core Makeup Tank (CMT) Level Yes No Manual Reactor Trip (also initiates turbine trip)
Yes Manual Safeguards Actuation Yes Manual CMT Actuation Yes Manual Automatic Depressurization System (ADS) Stages 1, 2, and 3 Actuation Yes Manual ADS Stage 4 Actuation Yes Manual PRHR Actuation Yes Manual Containment Cooling Actuation Yes Manual IRWST Injection Actuation Yes Manual Containment Recirculation Actuation Yes Manual Containment Isolation Yes Manual Main Steam Line Isolation Yes Manual Feedwater Isolation Yes Manual Containment Hydrogen Igniter (Nonsafety-related)(2)
Yes Manual Containment Vacuum Relief Yes Note: Dash (-) indicates not applicable.
- 1. These parameters are used to generate visual alerts that identify challenges to the critical safety functions. For the RSW, the visual alerts are embedded in the nonsafety-related displays as visual signals.
- 2. Containment hydrogen igniter control is provided as a soft control.
C-319 Amendment No. 125 f) The ECS provides a reverse-power trip of the generator circuit breaker which is blocked for at least 15 seconds following a turbine trip.
- 5. Controls exist in the main control room (MCR) to cause the circuit breakers identified in Table 2.6.1-3 to perform the listed functions.
- 6. Displays of the parameters identified in Table 2.6.1-3 can be retrieved in the MCR.
Table 2.6.1-1 Equipment Name Tag No.
Seismic Category I Class 1E/
Qual. for Harsh Envir.
Safety-Related Display Reactor Coolant Pump (RCP) Circuit Breaker ECS-ES-31 Yes Yes/No (Trip open only)
Yes RCP Circuit Breaker ECS-ES-32 Yes Yes/No (Trip open only)
Yes RCP Circuit Breaker ECS-ES-41 Yes Yes/No (Trip open only)
Yes RCP Circuit Breaker ECS-ES-42 Yes Yes/No (Trip open only)
Yes RCP Circuit Breaker ECS-ES-51 Yes Yes/No (Trip open only)
Yes RCP Circuit Breaker ECS-ES-52 Yes Yes/No (Trip open only)
Yes RCP Circuit Breaker ECS-ES-61 Yes Yes/No (Trip open only)
Yes RCP Circuit Breaker ECS-ES-62 Yes Yes/No (Trip open only)
Yes
C-370 Amendment No. 125 Table 2.7.1-2 Line Name Line Number ASME Code Section III Leak Before Break Functional Capability Required Main Control Room Supply VBS-L311 Yes No No Main Control Room Exhaust VBS-L312 Yes No No Main Control Room Toilet Exhaust VBS-L313 Yes No No Main Control Room Sanitary Vent Line SDS-PL-L016 Yes No No Main Control Room Sanitary Drain Line SDS-PL-L179 Yes No No Main Control Room Sanitary Drain Line SDS-PL-L182 Yes No No Main Control Room Water Line PWS-PL-L319 Yes No No Main Control Room Water Line PWS-PL-L320 Yes No No Main Control Room Waste Water Line WWS-PL-L808 Yes No No Main Control Room Waste Water Line WWS-PL-L851 Yes No No Table 2.7.1-3 Equipment Tag No.
Display Control Function Supplemental Air Filtration Unit Fan A VBS-MA-03A Yes (Run Status)
Start Supplemental Air Filtration Unit Fan B VBS-MA-03B Yes (Run Status)
Start MCR/CSA Supply Air Handling Units (AHU) A Fans VBS-MA-01A VBS-MA-02A Yes (Run Status)
Start MCR/CSA Supply AHU B Fans VBS-MA-01B VBS-MA-02B Yes (Run Status)
Start Division "A" and "C" Class 1E Electrical Room AHU A Fans VBS-MA-05A VBS-MA-06A Yes (Run Status)
Start Division "A" and "C" Class 1E Electrical Room AHU C Fans VBS-MA-05C VBS-MA-06C Yes (Run Status)
Start Division "B" and "D" Class 1E Electrical Room AHU B Fans VBS-MA-05B VBS-MA-06B Yes (Run Status)
Start
CHILLER PUMP VWS-MP-02 AIR-COOLED CHILLER VWS-MS-02 0 ru >
MCR/CSA SUPPLY I
- I AIR HANDLING UNIT VBS-MY-C01A CLASS 1E ELECTRicAL EQUIPMENT ROOM AIR HANDLING UNIT VBS-MY-C02A CLASS 1E ELECTRICAL EQUIPMENT ROOM AIR HANDLING UNIT VBS-MY-C02D CVS PUMP ROOM I
- I UNIT COOLER VAS-MY-C07A
- '}
I cvs PUMP ROOM UNIT COOLER FAN A VAS-MA-07A I
- I RNS PUMP ROOM UNIT COOLER VAS-MY-C12B RNS PUMP ROOM UNIT COOLER VAS-MY-C06A
~
RNS PUMP ROOM UNIT COOLER FAN A VAS-MA-OBA Figure 2.7.2-1 (Sheet 1 of 2)
Central Chilled Water System C-381 Amendment No. 125
C-382 Amendment No. 125 Figure 2.7.2-1 (Sheet 2 of 2)
Central Chilled Water System
C-403 Amendment No. 125
- 6. The RSR provides a suitable workspace environment, separate from the MCR, for use by the RSW operators.
- 7. The HSI resources available at the RSW include the alarm system displays, the plant information system, and the controls.
- 8. The RSW and the available HSI permit execution of tasks by licensed operators to establish and maintain safe shutdown.
- 9. The capability to access displays and controls is provided (controls as assigned by the MCR operators) for local control and monitoring from selected locations throughout the plant.
Table 3.2-1 Inspections, Tests, Analyses, and Acceptance Criteria No.
ITAAC No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 739 3.2.00.01a
- 1. The HFE verification and validation program is performed in accordance with the HFE verification and validation implementation plan and includes the following activities:
a) HSI Task support verification a) An evaluation of the implementation of the HSI task support verification will be performed.
a) A report exists and concludes that: Task support verification was conducted in conformance with the implementation plan and includes verification that the information and controls provided by the HSI match the display and control requirements generated by the function-based task analyses and the operational sequence analyses.
740 3.2.00.01b
- 1. The HFE verification and validation program is performed in accordance with the HFE verification and validation implementation plan and includes the following activities:
b) HFE design verification b) An evaluation of the implementation of the HFE design verification will be performed.
b) A report exists and concludes that: HFE design verification was conducted in conformance with the implementation plan and includes verification that the HSI design is consistent with the AP1000 specific design guidelines developed for each HSI resource.
C-404 Amendment No. 125 Table 3.2-1 Inspections, Tests, Analyses, and Acceptance Criteria No.
ITAAC No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 741 3.2.00.01c.i
- 1. The HFE verification and validation program is performed in accordance with the HFE verification and validation implementation plan and includes the following activities:
c) Integrated system validation c) (i) An evaluation of the implementation of the integrated system validation will be performed.
c) (i) A report exists and concludes that: The test scenarios listed in the implementation plan for integrated system validation were executed in conformance with the plan and noted human deficiencies were addressed.
742 3.2.00.01c.ii
- 1. The HFE verification and validation program is performed in accordance with the HFE verification and validation implementation plan and includes the following activities:
c) Integrated system validation c) (ii) Tests and analyses of the following plant evolutions and transients, using a facility that physically represents the MCR configuration and dynamically represents the MCR HSI and the operating characteristics and responses of the AP1000 design, will be performed:
Normal plant heatup and startup to 100% power Normal plant shutdown and cooldown to cold shutdown Transients: reactor trip and turbine trip Accidents:
Small-break LOCA Large-break LOCA Steam line break Feedwater line break Steam generator tube rupture c) (ii) A report exists and concludes that: The test and analysis results demonstrate that the MCR operators can perform the following:
Heat up and start up the plant to 100% power Shut down and cool down the plant to cold shutdown Bring the plant to safe shutdown following the specified transients Bring the plant to a safe, stable state following the specified accidents 743 3.2.00.01d
- 1. The HFE verification and validation program is performed in accordance with the HFE verification and validation implementation plan and includes the following activities:
d) Issue resolution verification d) An evaluation of the implementation of the HFE design issue resolution verification will be performed.
d) A report exists and concludes that: HFE design issue resolution verification was conducted in conformance with the implementation plan and includes verification that human factors issues documented in the design issues tracking system have been addressed in the final design.
C-405 Amendment No. 125 Table 3.2-1 Inspections, Tests, Analyses, and Acceptance Criteria No.
ITAAC No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 744 3.2.00.01e
- 1. The HFE verification and validation program is performed in accordance with the HFE verification and validation implementation plan and includes the following activities:
e) Plant HFE/HSI (as designed at the time of plant startup) verification e) An evaluation of the implementation of the plant HFE/HSI (as designed at the time of plant startup) verification will be performed.
e) A report exists and concludes that: The plant HFE/HSI, as designed at the time of plant startup, is consistent with the HFE/HSI verified in 1.a) through 1.d).
745 3.2.00.02
- 2. The MCR includes reactor operator workstations, supervisor workstation(s),
safety-related displays, and safety-related controls.
An inspection of the MCR workstations and control panels will be performed.
The MCR includes reactor operator workstations, supervisor workstation(s),
safety-related displays, and safety-related controls.
746 3.2.00.03.i Not used per Amendment No. 85 747 3.2.00.03.ii Not used per Amendment No. 85 748 3.2.00.03.iii Not used per Amendment No. 85 749 3.2.00.03.iv Not used per Amendment No. 85 750 3.2.00.03.v Not used per Amendment No. 85 751 3.2.00.04
- 4. The HSI resources available to the MCR operators include the alarm system, plant information system (nonsafety-related displays), wall panel information system, nonsafety-related controls (soft and dedicated), and computerized procedure system.
An inspection of the HSI resources available in the MCR for the MCR operators will be performed.
The HSI (at the time of plant startup) includes an alarm system, plant information system (nonsafety-related displays), wall panel information system, nonsafety-related controls (soft and dedicated), and computerized procedure system.
752 3.2.00.05
- 5. The RSW includes reactor operator workstation(s) from which licensed operators perform remote shutdown operations.
An inspection of the RSW will be performed.
The RSW includes reactor operator workstation(s).
C-406 Amendment No. 125 Table 3.2-1 Inspections, Tests, Analyses, and Acceptance Criteria No.
ITAAC No.
Design Commitment Inspections, Tests, Analyses Acceptance Criteria 753 3.2.00.06.i Not used per Amendment No. 85 754 3.2.00.06.ii Not used per Amendment No. 85 755 3.2.00.06.iii Not used per Amendment No. 85 756 3.2.00.07
- 7. The HSI resources available at the RSW include the alarm system displays, the plant information system, and the controls.
An inspection of the HSI resources available at the RSW will be performed.
The as-built HSI at the RSW includes the alarm system displays, the plant information system, and the controls.
757 3.2.00.08
- 8. The RSW and the available HSI permit execution of tasks by licensed operators to establish and maintain safe shutdown.
Test and analysis, using a workstation that physically represents the RSW and dynamically represents the RSW HSI and the operating characteristics and responses of the AP1000, will be performed.
A report exists and concludes that the test and analysis results demonstrate that licensed operators can achieve and maintain safe shutdown conditions from the RSW.
758 3.2.00.09
- 9. The capability to access displays and controls is provided (controls as assigned by the MCR operators) for local control and monitoring from selected locations throughout the plant.
An inspection of the local control and monitoring capability is provided.
The capability for local control and monitoring from selected locations throughout the plant exists.
C-443 Amendment No. 125
- 1. The seismic Category I equipment identified in Table 3.5-1 can withstand seismic design basis loads without loss of safety function.
- 2. The Class 1E equipment identified in Table 3.5-1 as being qualified for a harsh environment can withstand the environmental conditions that would exist before, during, and following a design basis accident without loss of safety function for the time required to perform the safety function.
- 3. Separation is provided between system Class 1E divisions, and between Class 1E divisions and non-Class 1E cable.
- 4. Safety-related displays identified in Table 3.5-1 can be retrieved in the main control room (MCR).
- 5. The process radiation monitors listed in Table 3.5-2 are provided.
- 6. The effluent radiation monitors listed in Table 3.5-3 are provided.
- 7. The airborne radiation monitors listed in Table 3.5-4 are provided.
- 8. The area radiation monitors listed in Table 3.5-5 are provided.
Table 3.5-1 Equipment Name Tag No.
Seismic Cat. I Class 1E Qual. for Harsh Envir.
Safety-Related Display Containment High Range Monitor PXS-RE160 Yes Yes Yes Yes Containment High Range Monitor PXS-RE161 Yes Yes Yes Yes Containment High Range Monitor PXS-RE162 Yes Yes Yes Yes Containment High Range Monitor PXS-RE163 Yes Yes Yes Yes MCR Radiation Monitoring Package A(1)
VBS-JS01A Yes Yes No Yes MCR Radiation Monitoring Package B(1)
VBS-JS01B Yes Yes No Yes Containment Atmosphere Monitor (Gaseous)
PSS-RE026 Yes No No No Containment Atmosphere Monitor (particulate, for RCS pressure boundary leakage detection)
PSS-RE027 Yes No No No Notes: (1) Each MCR Radiation Monitoring Package includes particulate, iodine and gaseous radiation monitors.
C-445 Amendment No. 125 Table 3.5-4 Airborne Radiation Monitors Equipment List Equipment No.
Fuel Handling Area Exhaust Radiation Monitor VAS-RE001 Auxiliary Building Exhaust Radiation Monitor VAS-RE002 Auxiliary Building Exhaust Radiation Monitor VAS-RE003 Annex Building Exhaust Radiation Monitor VAS-RE008 Health Physics and Hot Machine Shop Exhaust Radiation Monitor VHS-RE001 Radwaste Building Exhaust Radiation Monitor VRS-RE023 Table 3.5-5 Area Radiation Monitors Primary Sampling Room RMS-RE008 Containment Area - Upper Personnel Hatch Operating Deck RMS-RE009 Main Control Room RMS-RE010 Chemistry Laboratory RMS-RE011 Fuel Handling Area 1 RMS-RE012 Rail Car Bay/Filter Storage Area (Auxiliary Building Loading Bay)
RMS-RE013 Liquid and Gaseous Radwaste Area RMS-RY014 Control Support Area RMS-RE016 Radwaste Building Mobile Systems Facility RMS-RE017 Hot Machine Shop RMS-RE018 Annex Staging and Storage Area RMS-RE019 Fuel Handling Area 2 RMS-RE020 Containment Area - Lower Personnel Hatch Maintenance Level RMS-RE021
C-447a Amendment No. 125 Table 3.5-7 Component Name Tag No.
Component Location Containment High Range Radiation Monitor PXS-RE160 Containment Containment High Range Radiation Monitor PXS-RE161 Containment Containment High Range Radiation Monitor PXS-RE162 Containment Containment High Range Radiation Monitor PXS-RE163 Containment MCR Radiation Monitoring Package A VBS-JS01A Auxiliary Building MCR Radiation Monitoring Package B VBS-JS01B Auxiliary Building Containment Atmosphere Radiation Monitor (Gaseous)
PSS-RE026 Auxiliary Building Containment Atmosphere Radiation Monitor (particulate, for RCS pressure boundary leakage detection)
PSS-RE027 Auxiliary Building Steam Generator Blowdown Radiation Monitor BDS-RE010 Turbine Building Steam Generator Blowdown Radiation Monitor BDS-RE011 Turbine Building Component Cooling Water Radiation Monitor CCS-RE001 Turbine Building Main Steam Line Radiation Monitor SGS-RY026 Auxiliary Building Main Steam Line Radiation Monitor SGS-RY027 Auxiliary Building Service Water Blowdown Radiation Monitor SWS-RE008 Turbine Building