ML22073A185

Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant

ML22073A185
Person / Time
Site:	Grand Gulf
Issue date:	03/08/2022
From:	Entergy Operations
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22074A009	List: ML22067A049 ML22073A167 ML22073A169 ML22073A171 ML22073A172 ML22073A173 ML22073A175 ML22073A176 ML22073A177 ML22073A178 ML22073A182 ML22073A183 ML22073A184 ML22073A185 ML22073A186
References
GNRO2022/00010
Download: ML22073A185 (201)

Text

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE OF CONTENTS CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

............................................... 1.1-1 1.1.1 Type of License Required ........................ 1.1-1 1.1.2 Identification of Applicant ..................... 1.1-2 1.1.3 Number of Plant Units ........................... 1.1-3 1.1.4 Description of Location ......................... 1.1-3 1.1.5 Type of Nuclear Steam Supply System ............. 1.1-3 1.1.6 Type of Containment ............................. 1.1-3 1.1.7 Core Thermal Power Levels ....................... 1.1-3 1.1.8 Scheduled Completion and Operation Dates ........ 1.1-4 1.1.9 Organization of Contents ........................ 1.1-4 1.1.9.1 Subdivisions .................................... 1.1-4 1.1.9.2 Standard Format ................................. 1.1-4 1.1.9.3 References ...................................... 1.1-5 1.1.9.4 Tables and Figures .............................. 1.1-5 1.1.9.5 Numbering of Pages .............................. 1.1-5 1.1.9.6 Revising the Updated FSAR ....................... 1.1-5 1.2 GENERAL PLANT DESCRIPTION .................................. 1.2-1 1.2.1 Principal Design Criteria ....................... 1.2-1 1.2.1.1 General Design Criteria ......................... 1.2-1 1.2.1.2 System Criteria ................................. 1.2-5 1.2.2 Plant Description .............................. 1.2-10 1.2.2.1 Site Characteristics ........................... 1.2-10 1.2.2.2 General Arrangement of Structures and Equipment ...................................... 1.2-13 1.2.2.3 Nuclear System ................................. 1.2-17 1.2.2.4 Nuclear Safety Systems and Engineered Safety Features ....................................... 1.2-21 1.2.2.5 Power Conversion System ........................ 1.2-29 1.2.2.6 Electrical Systems and Instrumentation Control . 1.2-32 1.2.2.7 Fuel Handling and Storage Systems .............. 1.2-35 1-i Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE OF CONTENTS 1.2.2.8 Cooling Water and Auxiliary Systems ............ 1.2-36 1.2.2.9 Radioactive Waste Management ................... 1.2-42 1.2.2.10 Radiation Monitoring and Control ............... 1.2-43 1.2.2.11 Particularly Difficult Engineering Problems .... 1.2-44 1.2.2.12 Extrapolation of Technology .................... 1.2-44 1.3 COMPARISON TABLES .......................................... 1.3-1 1.3.1 Comparisons with Similar Facility Designs ....... 1.3-1 1.3.1.1 Nuclear Steam Supply System Design Characteristics ................................. 1.3-1 1.3.1.2 Power Conversion System Design Characteristics . 1.3-1 1.3.1.3 Engineered Safety Features Design Characteristics ................................. 1.3-1 1.3.1.4 Containment Design Characteristics .............. 1.3-1 1.3.1.5 Radioactive Waste Management Systems Design Characteristics ................................. 1.3-1 1.3.1.6 Structural Design Characteristics ............... 1.3-1 1.3.1.7 Instrumentation and Electrical Systems Design Characteristics ................................. 1.3-2 1.3.2 Comparison of Final and Preliminary Information ..................................... 1.3-2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS ................... 1.4-1 1.4.1 GGNS Project .................................... 1.4-1 1.4.2 Architect Engineer .............................. 1.4-1 1.4.3 Nuclear Steam Supply System ..................... 1.4-2 1.4.4 Turbine Generator Vendor ........................ 1.4-3 1.4.5 Consultants ..................................... 1.4-3 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION ............. 1.5-1 1.5.1 Current Development Programs .................... 1.5-1 1.5.1.1 Instrumentation for Vibration ................... 1.5-1 1.5.1.2 Core Spray Distribution ......................... 1.5-1 1.5.1.3 Core Spray and Core Flooding Heat Transfer Effectiveness ................................... 1.5-2 1.5.1.4 Verification of Pressure Suppression Design ..... 1.5-2 1-ii Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE OF CONTENTS 1.5.1.5 Critical Heat Flux Testing ...................... 1.5-3 1.5.1.6 Structural Testing .............................. 1.5-4 1.6 MATERIAL INCORPORATED BY REFERENCE ......................... 1.6-1 1.7 ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS ........... 1.7-1 1.8 SYMBOLS USED IN ENGINEERING DRAWINGS ....................... 1.8-1 1.9 ABBREVIATIONS .............................................. 1.9-1 1.10 DRAWING NUMBER-FSAR FIGURE NUMBER CROSS-REFERENCE ......... 1.10-1 1-iii Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

LIST OF TABLES Table 1.3-1 Comparison of Nuclear Steam Supply System Design Characteristics Table 1.3-2 Comparison of Power Conversion System Design Characteristics Table 1.3-3 Comparison of Engineered Safety Features Design Characteristics Table 1.3-4 Comparison of Containment Design Characteristics Table 1.3-5 Radioactive Waste Management Systems Design Characteristics Table 1.3-6 Comparison of Structural Design Characteristics Table 1.3-7 Comparison of Electrical Systems Table 1.3-8 Significant Design Changes from PSAR to FSAR Table 1.4-1 Commercial Nuclear Reactors Completed, Under Construction or in Design by General Electric Table 1.6-1 Referenced Reports Table 1.7-1 Nonproprietary Electrical and Instrumentation/

Control Drawings Incorporated by Reference Table 1.9-1 Acronyms Used in FSAR Table 1.10-1 Cross-Reference List of Drawing Numbers and FSAR Figure Numbers 1-iv Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

LIST OF FIGURES Figure 1.1-1 Heat Balance at Rated Power Figure 1.2-1 Orientation of Principal Plant Structures Figure 1.2-2 General Arrangement Plan at El. 93'-0" and 100'-9" Figure 1.2-3 General Arrangement Plan at El. 113'-0",

111'-0", 119'-0", 120'-10" and 114'-6" Figure 1.2-4 General Arrangement Plan at El. 133'-0",

148'-0", 139'-0", 135'-4" and 147'-7" Figure 1.2-5 General Arrangement Plan at El. 166'-0",

161'-10" and 170'-0" Figure 1.2-6 General Arrangement Plan at El. 184'-6",

185'-0" and 189'-0" Figure 1.2-7 General Arrangement Plan at El. 208'-10" Figure 1.2-8 General Arrangement Sections "A-A" and "B-B" Figure 1.2-9A Turbine Building General Arrangement Sections "A-A" & "B-B" Figure 1.2-9B Turbine Building General Arrangement Sections "C-C", "D-D" & "E-E" Figure 1.2-9C Identification Key for Turbine Building Equipment Figure 1.2-10 Radwaste Building Plan at El. 93'-0" Figure 1.2-11 Deleted Figure 1.2-12 Deleted Figure 1.2-13 Deleted Figure 1.2-14 Radwaste Building Sections "C-C" & "D-D" Figure 1.2-15 Natural Draft Cooling Tower Figure 1.2-16 Auxiliary Cooling Tower Figure 1.8-1 P&I Legend Figure 1.8-2 P&ID Legend Figure 1.8-3 P&I Legend (General Electric) 1-v LBDCR 2019-008

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

CHAPTER

1.0 INTRODUCTION

AND GENERAL DESCRIPTION OF PLANT

1.1 INTRODUCTION

This updated Final Safety Analysis Report (FSAR) complies with the Standard Format and Content of Safety Analysis Reports (Revision 2) issued by the Nuclear Regulatory Commission (NRC) in September 1975 and 10CFR50.71(e). The original FSAR, as amended, is considered to be the licensing basis for the plant. The updated FSAR will be the reference document for purposes of communications with the NRC such as reporting of deviations from conditions as stated in the FSAR and for evaluations requiring the FSAR, such as 10CFR50.59. Reference to the FSAR by this document, by plant directives, and by pertinent Entergy Operations manuals will be understood to reference the updated FSAR. This approach is consistent with the guidance provided in Generic Letter 81-06 entitled Periodic Updating of Final Safety Analysis Reports (FSARs) dated February 26, 1981.

A discussion of the format of the updated FSAR is presented in subsection 1.1.9.

1.1.1 Type of License Required

[HISTORICAL INFORMATION] [The original FSAR was submitted in support of the application of Mississippi Power & Light Company**

for a license to operate a two-unit nuclear power facility at a core thermal power level of 3833 MWt, each, the power level equivalent to 100 percent of the design steam flow. This application was submitted under Section 103 (b) of the Atomic Energy Act of 1954, as amended, and the regulation of the Nuclear Regulatory Commission set forth in Part 50 of Title 10 to the Code of Federal Regulations (10CFR50).

In December of 1979 construction of Grand Gulf Unit 2 (NRC Docket Number 50-417) was deferred in order to concentrate resources on the completion of Unit 1. After Unit 1 had received its Commercial Operating License, Entergy Operations, Inc. formally requested the NRC to revoke the Construction Permit and officially cancel the second unit at the Grand Gulf Nuclear Station. The Construction Permit for Grand Gulf Unit 2 was formally revoked by the NRC in August 1991.]

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.1.2 Identification of Applicant

[HISTORICAL INFORMATION] [The Grand Gulf Nuclear Station is owned by System Energy Resources, Inc. (SERI*) and South Mississippi Electric Power Association (SMEPA). SMEPA maintains a 10%

interest in GGNS and, of its original 90% ownership share, SERI maintains 77.23% ownership interest. The remaining 12.77%

interest is owned by equity investors: Textron Financial Corporation and Resources Capital Management Corporation, and is leased back to SERI. SERI and SMEPA pay costs associated with their respective ownership or leased interests. Entergy Operations, Inc. (Entergy Operations) operates GGNS. SERI and Entergy Operations are wholly owned subsidiaries of Entergy Corporation, a registered public utility holding company.

Mississippi Power & Light Company (MP&L**) originally assumed responsibility for design, construction, and operation of the facility and acted as an agent for SMEPA. On December 20, 1986, SERI assumed responsibility for the control and performance of licensed activities from MP&L. On June 6, 1990 Entergy Operations assumed responsibility for the control and performance of licensed activities from SERI. As a part of the final transfer, Entergy Operations assumed responsibility for commitments originally made by MP&L and SERI. In those cases in the FSAR where Entergy Operations has either present or future responsibility, reference is made to Grand Gulf Nuclear Station, GGNS, or Entergy Operations, with no mention of MP&L or SERI. In 1996, MP&L changed its name to Entergy Mississippi, Inc., however, to address certain historical information where a reference to Entergy Operations could cause confusion, MP&L or SERI is used to represent situations where either MP&L or SERI originally had responsibility or made commitments but where Entergy Operations is now responsible. In the cases where Entergy Mississippi, Inc. has responsibility (such as offsite power),

references are made to Entergy Mississippi, Inc.]

• System Energy Resources, Inc. was originally named Middle South Energy, Inc. The name was changed to System Energy Resources, Inc.

in 1986.

• • Mississippi Power & Light Company (MP&L) changed its name to Entergy Mississippi, Inc. as approved by Amendment 127 to the facility operating license. Historical references to MP&L are contained in the FSAR.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.1.3 Number of Plant Units

[HISTORICAL INFORMATION] [This application was submitted for both Units 1 and 2 of the Grand Gulf Nuclear Station which were docketed in November 1972 on NRC Docket Numbers 50-416 and 50-417, respectively.

In December of 1979 construction of Grand Gulf Unit 2 (NRC Docket Number 50-417) was deferred in order to concentrate resources on the completion of Unit 1. After Unit 1 had received its Commercial Operating License, Entergy Operations, Inc. formally requested the NRC to revoke the Construction Permit and officially cancel the second unit at the Grand Gulf Nuclear Station. The Construction Permit for Grand Gulf Unit 2 was formally revoked by the NRC in August 1991.]

1.1.4 Description of Location

[HISTORICAL INFORMATION] [The facility is located in Claiborne County, Mississippi, on the east side of the Mississippi River approximately 25 miles south of Vicksburg and 37 miles north-northeast of Natchez, Mississippi.]

1.1.5 Type of Nuclear Steam Supply System

[HISTORICAL INFORMATION] [Grand Gulf has a BWR-6 boiling water reactor (251-inch vessel with 800 fuel assemblies) designed and supplied by General Electric Company.]

1.1.6 Type of Containment

[HISTORICAL INFORMATION] [The Grand Gulf containment is the Mark III BWR containment incorporating the drywell/pressure suppression concept. The containment is a steel-lined reinforced concrete structure designed by Bechtel Power Corporation.]

1.1.7 Core Thermal Power Levels The information presented in this updated FSAR pertains to the Grand Gulf reactor with a rated power level of 4408 Mwt. This power level represents a 15% increase from the original license of 3833 Mwt. The station utilizes a single-cycle forced circulation boiling water reactor (BWR) provided by General Electric-Hitachi (GEH). The heat balance for rated power is shown in Figure 1.1-1.

The station is designed to operate at a gross electrical power output of approximately 1523.5 MWe.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.1.8 Scheduled Completion and Operation Dates

[HISTORICAL INFORMATION] [The fuel loading for Unit 1 was completed in August 1982. Commercial operation for Unit 1 was declared on July 1, 1985. Construction of Unit 2 was deferred in December 1979 in order to concentrate resources on completion of Unit 1. After completion of Unit 1, Entergy Operations, Inc.

formally requested the NRC to revoke the Unit 2 Construction Permit (NRC Docket Number 50-417). The Unit 2 Construction Permit was revoked in August 1991.]

1.1.9 Organization of Contents 1.1.9.1 Subdivisions The updated FSAR is organized into 18 chapters, each of which consists of a number of sections that are numerically identified by two numerals separated by a decimal (e.g., 3.4 is the fourth section of Chapter 3). Further subdivisions are referred to as subsections.

1.1.9.2 Standard Format The updated FSAR has been written to comply with the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (Revision 2) as issued by the Nuclear Regulatory Commission in October 1975. The updated FSAR uses the same chapter, section, and subsection headings as those used in the standard format except in cases where this format is not applicable to plant design. Where appropriate, the updated FSAR is subdivided beyond the extent of the standard format to isolate all information specifically requested in that document. Where information has been presented that is not specifically requested by the standard format and this information is identified numerically (chapter, section, or subsection), this information is presented under the appropriate general headings as a subdivision containing information specifically requested by the standard format. (For example, subsection 1.1.9 is not requested in the standard format. Since it apparently belonged in Section 1.1, it was placed after the eight subsections containing information requested by the standard format).

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.1.9.3 References References to another location in the updated FSAR are made by chapter or section number. References to another document are indicated by the notation (Ref. 1). The reference section is located at the end of the applicable text and before any tables in the section.

1.1.9.4 Tables and Figures Tabulations of data are designated as tables. They are identified by the section number, followed by a number according to its order of mention in the section (e.g., Table 3.3-5 is the fifth table of Section 3.3). Tables are located at the end of the applicable section. Drawings, sketches, curves, graphs, and engineering diagrams are all identified as figures and are numbered according to the order of mention in the section (Figure 3.4-2 is the second figure of Section 3.4). Figures are located at the end of the applicable section.

1.1.9.5 Numbering of Pages Pages are numbered sequentially within each section. For example, 1.1-2 is the second page of Section 1.1. When it becomes necessary during revision of this updated FSAR to insert a page(s) between two existing pages within a section, letters will be used (for example, to insert two pages between 3.2-4 and 3.2-5, the following page sequence would appear: 3.2-4, 3.2-4a, 3.2-4b, 3.2-5).

1.1.9.6 Revising the Updated FSAR When it becomes necessary to submit additional information or to revise information presently contained in the updated FSAR, the following procedures will be followed:

a. When a change is made to the updated FSAR text, those pages affected will be marked with the page change identification (date of revision or change number or both) and a change indicator (e.g. vertical line) drawn in the margin adjacent to the portion actually changed. Further revising of previously revised sections will delete the original labeled vertical change bar if the entire portion is revised.

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b. Figures will be revised by indicating the page change identification (date of revision or change number or both) on the Figure.

c. Revisions containing updated information shall be submitted on a replacement-page basis and shall be accompanied by a list which identifies the current pages of the FSAR following page replacement.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2 GENERAL PLANT DESCRIPTION 1.2.1 Principal Design Criteria The principal design criteria are presented in two ways. First, they are classified as either a power generation function or a safety function. Second, they are grouped according to system.

Although the distinctions between power generation or safety functions are not always clear cut and are sometimes overlapping, the functional classification facilitates safety analyses, while the grouping by system facilitates the understanding of both the system function and design.

1.2.1.1 General Design Criteria 1.2.1.1.1 Power Generation Design Criteria

a. The station is designed to produce steam for direct use in a turbine-generator unit.

b. Heat removal systems are provided with sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions and abnormal operational transients.

c. Backup heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage.

d. The fuel cladding, in conjunction with other plant systems, is designed to retain integrity throughout the range of normal operational conditions and abnormal operational transients.

e. The fuel cladding accommodates, without loss of integrity, the pressures generated by fission gases released from fuel material throughout the design life of the fuel.

f. Control equipment is provided to allow the reactor to respond automatically to minor load changes, major load changes, and abnormal operational transients.

g. Reactor power level is manually controllable.

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h. Control of the reactor is possible from a single location.

i. Reactor controls, including alarms, are arranged to allow the operator to rapidly assess the condition of the reactor system and locate system malfunctions.

j. Interlocks or other automatic equipment are provided as backup to procedural controls to avoid conditions requiring the functioning of nuclear safety systems or engineering safety features.

1.2.1.1.2 Safety Design Criteria

a. The station is designed, fabricated, constructed, and operated in such a way that the normal release of radioactive materials to the environment is significantly less than the requirements of 10 CFR 20.

b. The station is designed, fabricated, erected, and operated in such a way that the release of radioactive materials to the environment resulting from abnormal transients and accidents is less than the requirements of 10 CFR 100, 10 CFR 50.67 and 10 CFR 50 GDC 19.

c. The reactor core is designed so its nuclear characteristics do not contribute to a divergent power transient.

d. The reactor is designed so there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the reactor with other appropriate plant systems.

e. Gaseous, liquid, and solid waste disposal facilities are designed so the discharge of radioactive effluents and offsite shipment of radioactive materials can be made in accordance with applicable regulations.

f. The design provides means by which plant operators are alerted when limits on the release of radioactive material are approached.

g. Sufficient indications are provided to allow determination that the reactor is operating within the envelope of conditions considered by plant safety analysis.

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h. Radiation shielding is provided and access control patterns are established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any mode of normal plant operations.

i. Those portions of the nuclear system that form part of the reactor coolant pressure boundary are designed to retain integrity as a radioactive material containment barrier following abnormal operational transients and accidents.

j. Nuclear safety systems function to assure that no damage to the reactor coolant pressure boundary results from internal pressures caused by abnormal operational transients and accidents.

k. Where positive, precise action is immediately required in response to abnormal operational transients and accidents, such action is automatic and requires no decision or manipulation of controls by plant operations personnel.

l. Essential safety actions are provided by equipment of sufficient redundance and independence that no single failure of active components can prevent the required actions. For systems or components to which IEEE 2791971, Criteria for Protection Systems for Nuclear Power Generating Stations, applies, single failures of both active and passive electrical components are considered in recognition of the higher anticipated failure rates of passive electrical components relative to passive mechanical components.

m. Provisions are made for control of active components of nuclear safety systems and engineered safety features from the control room.

n. Nuclear safety systems and engineered safety features are designed to permit demonstration of their functional performance requirements. The ability and the extent that systems can be tested during operation is discussed further in each individual system subsection.

o. The design of nuclear safety systems and engineered safety features includes allowances for natural environmental disturbances such as earthquakes, floods, and storms at the station site.

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p. Standby electrical power sources have sufficient capacity to power all nuclear safety systems and engineered safety features requiring electrical power.

q. Standby electrical power sources are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.

r. A containment is provided that completely encloses the reactor system, drywell, and suppression pool. The containment employs the pressure suppression concept.

s. It is possible to test primary containment integrity and leak tightness at periodic intervals.

t. A secondary containment is provided that completely encloses the primary containment. This secondary containment provides a method for controlling the release of radioactive materials from the primary containment.

u. The primary containment and secondary containment, in conjunction with other engineered safety features, limit radiological effects of accidents resulting in the release of radioactive material to the containment volumes to less than the requirements of 10 CFR 100.

v. Provisions are made for removing energy from the primary containment as necessary, to maintain the integrity of the containment system following accidents that release energy to the containment.

w. Piping that penetrates the primary containment and could serve as a path for the uncontrolled release of radioactive material to the environs is isolated whenever such uncontrolled radioactive material release is threatened. Such isolation is effected in time to limit radiological effects to less than the requirements of 10 CFR 100.

x. Emergency core cooling systems are provided to limit fuel cladding temperature to less than that which could cause fragmentation in the event of a loss-of-coolant accident.

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y. The emergency core cooling systems provide for continuity of core cooling over the complete range of postulated break sizes in the reactor coolant pressure boundary.

z. Operation of the emergency core cooling systems is initiated automatically when required, regardless of the availability of offsite power supplies and the normal generating system of the station.

aa. The control room is shielded against radiation so that continued occupancy under accident conditions is possible.

bb. In the event that the control room becomes inaccessible, it is possible to bring the reactor from power range operation to cold shutdown conditions by utilizing the local controls and equipment that are available outside the control room.

cc. Backup reactor shutdown capability is provided independent of normal reactivity control provisions. This backup system has the capability to shut down the reactor from any normal operating condition and subsequently to maintain the shutdown condition.

dd. Fuel handling and storage facilities are designed to prevent inadvertent criticality and to maintain shielding and cooling of spent fuel.

1.2.1.2 System Criteria The principal design criteria for particular systems are listed in the following subsections.

1.2.1.2.1 Nuclear System Criteria

a. The fuel cladding is designed to retain integrity as a radioactive material barrier throughout the design power range. The fuel cladding is designed to accommodate, without loss of integrity, the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel.

b. The fuel cladding, in conjunction with other plant systems, is designed to retain integrity throughout any abnormal operational transient.

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c. Those portions of the nuclear system that form part of the reactor coolant pressure boundary are designed to retain integrity as a radioactive material barrier during normal operation and following abnormal operational transients and accidents.

d. Heat removal systems are provided in sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational transients. The capacity of such systems is adequate to prevent fuel cladding damage.

e. Heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage. The reactor is capable of being shut down automatically in sufficient time to permit decay heat removal systems to become effective following loss of operation of normal heat removal systems.

f. The reactor core and reactivity control system are designed so that control rod action is capable of bringing the core subcritical and maintaining it so, even with the rod of highest reactivity worth fully withdrawn and unavailable for insertion.

g. The reactor core is designed so that its nuclear characteristics do not contribute to a divergent power transient.

h. The nuclear system is designed so there is no tendency for divergent oscillation of any operating characteristic, considering the interaction of the nuclear system with other appropriate plant systems.

1.2.1.2.2 Power Conversion Systems Criteria Components of the power conversion systems are designed to perform the following basic objectives.

a. Produce electrical power from the steam coming from the reactor, condense the steam into water, and return the water to the reactor as heated feedwater, with a major portion of its gases and particulate impurities removed.

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b. Assure that any fission products or radioactivity associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures.

1.2.1.2.3 Electrical Power Systems Criteria Sufficient normal and standby auxiliary sources of electrical power are provided to attain prompt shutdown and continued maintenance of the station in a safe condition under all credible circumstances. The power sources are adequate to accomplish all required essential safety actions under postulated design-bases accident conditions.

1.2.1.2.4 Radwaste System Criteria

a. The gaseous and liquid radwaste systems are designed to minimize the release of radioactive effluents from the station to the environs. Such releases as may be necessary during normal operations are limited to values that meet the requirements of applicable regulations including 10 CFR 20 and 10 CFR 50.

b. The solid radwaste disposal systems are designed so that inplant processing and offsite shipments are in accordance with all applicable regulations, including 10 CFR 20, 10 CFR 71, and 49 CFR 171 through 179, as appropriate.

c. The system's design provides means by which station operations personnel are alerted whenever specified limits on the release of radioactive material may be approached.

1.2.1.2.5 Auxiliary Systems Criteria

a. Fuel handling and storage facilities are designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel. Provisions are made for maintaining the cleanliness of spent fuel cooling and shielding water.

b. Auxiliary systems which are required for safe shutdown or to mitigate the consequences of an accident are designed to function during normal and/or accident conditions.

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c. Auxiliary systems that are not required to effect safe shutdown of the reactor or maintain it in a safe condition are designed such that a failure of these systems shall not prevent the essential auxiliary systems from performing their design functions.

1.2.1.2.6 Nuclear Safety Systems and Engineered Safety Features Criteria Principal design criteria for nuclear safety systems and engineered safety features are as follows:

a. These criteria correspond to criteria j through q, x through z, bb and cc in subsection 1.2.1.1.2.

b. Standby electrical power sources have sufficient capacity to power engineered safety features requiring electrical power.

c. Standby electrical power sources are provided to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available.

d. In the event that the control room is inaccessible, it is possible to bring the reactor from power range operation to a cold shutdown condition by manipulating controls and equipment that are available outside the control room.

e. Backup reactor shutdown capability is provided independent of normal reactivity control provisions. This backup system has the capability to shut down the reactor from any operating condition and subsequently to maintain the shutdown condition.

1.2.1.2.7 Process Control Systems Criteria The principal design criteria for the process control systems are discussed in this subsection.

1.2.1.2.7.1 Nuclear System Process Control Criteria

a. Control equipment is provided to allow the reactor to respond automatically to load changes within design limits.

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b. It is possible to control the reactor power level manually.

c. Control of the nuclear system is possible from a central location.

d. Nuclear systems process controls and alarms are arranged to allow the operator to rapidly assess the condition of the nuclear system and to locate process system malfunctions.

e. Interlocks or other automatic equipment are provided as a backup to procedural controls to avoid conditions requiring the actuation of engineered safety features.

1.2.1.2.7.2 Power Conversion Systems Process Control Criteria

a. Control equipment is provided to control the reactor pressure throughout its operating range.

b. The turbine is able to respond automatically to minor changes in load.

c. Control equipment in the feedwater system maintains the water level in the reactor vessel at the optimum level required by steam separators.

d. Control of the power conversion equipment is possible from a central location.

e. Interlocks or other automatic equipment are provided in addition to procedural controls to avoid conditions requiring the actuation of engineered safety features.

1.2.1.2.7.3 Electrical Power System Process Control Criteria

a. The Class IE power systems are designed as a three Division system. The ESF systems of any two of the three divisions provide for the minimum safety functions necessary to shutdown the unit and maintain it in a safe shutdown condition.

b. Protective relaying is used to detect and isolate faulted equipment from the system with a minimum of disturbance in the event of equipment failure.

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c. Voltage relays or bistables are used on the emergency equipment buses to isolate these buses from the normal electrical system in the event of loss of offsite power and to initiate starting of the standby emergency power system diesel generators.

d. The standby emergency power diesel generators are started by control relays. The generators are also loaded by a control system to meet the existing emergency condition.

e. Electrically operated breakers are controllable from the control room.

f. Instruments for monitoring the operation of essential generators, transformers, and circuits are provided in the control room.

1.2.2 Plant Description 1.2.2.1 Site Characteristics 1.2.2.1.1 Location

[HISTORICAL INFORMATION] [Grand Gulf Nuclear Station is located in Claiborne County in southwestern Mississippi. The plant site is on the east side of the Mississippi River about 25 miles south of Vicksburg and 37 miles north-northeast of Natchez.] The Grand Gulf Military Park borders a portion of the north side of the plant site property, and the community of Grand Gulf is about 1-1/

2 miles to the north. The town of Port Gibson is about 6 miles southeast of the plant site.

The site and its environs consist primarily of woodlands and farms. The total area of the plant site is approximately 2100 acres. Within this area are two lakes, Gin Lake and Hamilton Lake.

These lakes were once the channel of the Mississippi River and average about 8 to 10 feet in depth.

The western half of the plant site consists of materials deposited by the Mississippi River and extends eastward from the river about 0.8 mile. This area is generally 55 to 75 feet above mean sea level (msl).

The eastern half of the plant site is rough and irregular with steep slopes and deep-cut stream valleys and drainage courses.

Elevations in this portion of the plant site range from about 80 1.2-10 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) feet above msl to more than 200 feet above msl at the inland of the site. Elevations of about 400 feet above mean sea level occur on the hilltops east and northeast of the site.

The orientation of the principal plant structures on the site is shown in Figure 1.2-1.

1.2.2.1.2 Meteorology

[HISTORICAL INFORMATION] [The climate at the site is generally subtropical and humid in character, but is subject to important polar influence from time to time. Maximum rainfall in a 68-year period of record at Vicksburg, Mississippi, was 9.97 inches in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the maximum average monthly rainfall was about 16.5 inches. Prevailing winds are from the south-southeast. Maximum wind speeds at Municipal Airport, Jackson, Mississippi, in a 50-year period of record were 68 mph and occurred in March 1952.

During 92 years of record, 65 hurricanes, or post-hurricane path centerlines, passed within 100 miles of the site. There have been two damaging tornadoes in a 50-year period of record (1916-1966) within a 25-mile radius. This is typical of tornado frequency in the site region. An onsite meteorological measurement program was initiated in 1972 to provide data to assess limits to be set later on radioactive gas releases. Safety-related structures are design-ed for a maximum tornado load of 360 mph and wind load of 90 mph.]

1.2.2.1.3 Hydrology

[HISTORICAL INFORMATION] [The site for the Grand Gulf Nuclear Station is located on the east side of the Mississippi River in the vicinity of river mile 406 about 25 miles south of Vicksburg and 6 miles northwest of Port Gibson. It is bounded on the west by the Mississippi River and on the east by loessial bluffs (forming part of the hilly region which extends from Vicksburg to Baton Rouge). The Mississippi River floodplain adjacent to the site is relatively low and flat with elevations of 55 to 75 ft msl.

The plant site is located in the loessial uplands with a plant grade elevation of 132.5 ft msl. This elevation is well above the probable maximum flood (PMF) elevation in the Mississippi River.

The design project flood (DPF) and 100-year flood elevation of the Mississippi River in the plant vicinity are at elevations of 96.2 and 93.1 ft msl, respectively.

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The plant makeup and service water is supplied by a series of radial collector wells located in the floodplain parallel to the Mississippi River. These collector wells have been constructed by sinking cylindrical concrete caissons into the alluvial aquifer, sealing the bottom with a concrete plug, and projecting perforated pipes horizontally into the aquifer.

The principal ground water-bearing zones in the site vicinity are the Mississippi River alluvium, the terrace deposits, and the Catahoula formation.

The Mississippi River alluvium is the principal aquifer at the site and is the source of plant service water supply. The ground water is unconfined and the water level is generally controlled by the Mississippi River stage. The terrace deposits contain local permeable zones that yield several hundred gallons of water per minute. The regional water table occurs within the terrace deposits and adjacent Mississippi River alluvium; however, several perched water zones also occur within the terrace deposits. The Catahoula formation underlies the alluvium and terrace deposits and comprises a source of ground water for domestic wells. The ground water in the Catahoula formation occurs during semi-confined conditions.]

1.2.2.1.4 Geology

[HISTORICAL INFORMATION] [Surface material at the site is Pleistocene loess. This material erodes easily forming very steep slopes along stream channels. One such slope, along the Mississippi River floodplain, divides the site so that it lies in two subprovinces of the Central Gulf Coastal Plain physiographic province. The subprovinces are the Loess or Bluff Hills to the east and the Mississippi alluvial plain to the west.

The site is underlain by approximately 18,000 ft of Cretaceous through Cenozoic sands, gravels, clays, marls, claystones, sandstones, and limestones. These sediments were deposited on middle Jurassic evaporites, the parent material for salt domes found in the area. Regional dip is southward and becomes progressively steeper toward the Gulf Coast. As a result of the steepened dip, most formations tend to be wedge shaped, thickening coastward.

Several domal or structural uplift areas are found within the Gulf Coast Basin. The nearest of these, located about 50 miles east-northeast of the site, is the Jackson Dome. Formation of this 1.2-12 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) structure began in the early Cretaceous period and ended in the middle Tertiary period. A salt dome has been formed as near as 8 miles from the site. The dome was formed from the late Cretaceous period through the Oligocene epoch. No nearer salt domes are known.

Petroleum exploration drilling near the site has generally been unsuccessful. Within a 6-mile radius of the site, 13 wildcat oil wells have been drilled; all were dry. The nearest of these was 3-1/2 miles from the site. At least 50 wells have been drilled in Claiborne County and only two have discovered hydrocarbons.]

1.2.2.1.5 Seismology and Design Response Spectra

[HISTORICAL INFORMATION] [The site area is not seismically active; however, distant earthquakes may have been felt there.

The New Madrid, Missouri, earthquakes of 1811-1812, which occurred 325 miles north of the site, had maximum intensities of MM XI-XII. These events are conservatively estimated to have had a maximum intensity of MM VI at the site.

The largest event known to have occurred in the Gulf Coast Basin, not associated with a structure, is the strong intensity MM VI Donaldsonville, Louisiana, earthquake of October 19, 1930. If this earthquake occurred at the site, a peak acceleration of 0.07-0.10 g would result, according to the intensity-acceleration curves of Neumann (1954). A safe shutdown peak horizontal acceleration of 0.15 g and vertical acceleration of 0.10 g were selected for plant design giving additional conservatism. Design spectra for the safe shutdown earthquake with horizontal acceleration of 0.15 g and for a variety of damping values have been used for analysis of plant structures and equipment.]

1.2.2.1.6 Unusual Site Characteristics There are no unusual site characteristics.

1.2.2.2 General Arrangement of Structures and Equipment

[HISTORICAL INFORMATION] [The principal buildings and structures include the containment structure, the turbine building, the auxiliary building, the control building, the diesel generator building, the standby service water cooling towers and basins, the enclosure building, the radwaste building, the auxiliary cooling tower, and the natural draft cooling tower.] A structure which houses the administration offices, clean machine shop, and 1.2-13 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) guardhouse is provided. A building is also provided to store the site fire truck, foam chemicals, and miscellaneous fire fighting apparatus.

Bulk storage facilities are provided for hydrogen and oxygen in support of the hydrogen water chemistry system on the plant north end of the Unit 2 cooling tower basin. The bulk liquid hydrogen facility includes a 20,000 gallon cryogenic tank, cryogenic pumps, atmospheric vaporizers and gas storage tubes to supply high pressure gas to the hydrogen water chemistry, generator cooling and primary water tank blanket systems.

The bulk liquid oxygen facility includes a 9,000 gallon cryogenic tank and atmospheric vaporizers to supply low pressure gas to the hydrogen water chemistry system.

A Large Component Storage Building (LCSB) is located in the Northwest laydown area. This building houses components that were replaced during the GGNS EPU. The components include the steam dryer, both moisture separator reheaters, 9 feedwater heaters, both reactor feedpump turbines and their inner casings and the high pressure turbine rotor.

These buildings and structures are founded upon suitable material for their intended function. Structures essential to the safe operation and shutdown of the plant are designed to withstand more extreme loading conditions than normally considered in conventional nonnuclear design practice. The buildings and internal structures so designated are designed to provide protection as required from tornadoes, earthquakes, and the failure of equipment producing flooding, missiles, and pipe whip.

Additional discussion of design considerations may be found in Chapter 3.

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Location and orientation of the buildings on the site are shown in Figure 1.2-1. The general arrangement of the buildings and equipment locations is shown in Figures 1.2-2 through 1.2-16.

a. The containment structure, shown in Figures 1.2-2 through 1.2-8, is a seismic Category I structure which encloses the reactor coolant system, the drywell, suppression pool, upper pool, and some of the engineered safety feature systems and supporting systems. The functional design basis of the containment, including its penetrations and isolation valves, is to contain with adequate design margin the energy released from a design basis loss-of-coolant accident and to provide a leaktight barrier against the uncontrolled release of radioactivity to the environment, even assuming a partial loss of engineered safety features.

b. The turbine building, shown in Figures 1.2-2 through 1.2-8, houses all equipment associated with the main turbine generator. Other auxiliary equipment is also located in this building. There are safety-related instruments in the turbine building, but the building will not collapse onto or otherwise adversely affect the systems of which those instruments are a part in the event of a postulated accident.

c. The auxiliary building, shown in Figures 1.2-2 through 1.2-8, is a seismic Category I structure that contains safety systems, fuel storage and shipping equipment and necessary auxiliary support systems. Redundant safety trains in the auxiliary building and all other areas of the plant are separated and protected so that a loss of function of one train will not prevent the other train from performing its safety function.

d. The control building, shown in Figures 1.2-2 through 1.2-8, is a seismic Category I, multistoried, concrete and steel structure, in which many of the control and electrical systems, including required support systems directly related to safety or necessary for plant operations, are located.

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e. The diesel generator building, shown in Figure 1.2-4 is seismic Category I and is constructed of reinforced concrete. The building contains the three diesel generators, three fuel oil day tanks, six starting air receivers-compressors, air intake vents and filters, mufflers, and controls. Each diesel generator and its associated equipment is in an individual room within the diesel generator building. The building interior and exterior walls that separate the diesel generators and associated equipment constitute a fire barrier wall having a 3-hour fire resistance rating.

f. The enclosure building, shown in Figure 1.2-8, is a limited leakage seismic Category I structure that encloses the upper portion of the containment above the auxiliary building roof level. The enclosure building provides a boundary for the standby gas treatment system, which maintains a negative pressure in the volume between the containment and enclosure building to ensure that leakage of radioactive materials from the containment is filtered prior to release to the environment in the unlikely event of a loss-of-coolant accident.

g. The radwaste building, shown in Figures 1.2-10 through 1.2-14, has six major areas; the collection tankage area, a processing area, a pipeway area, a personnel area, a solidification area, and a storage area. The radwaste systems process liquid, solid, and gaseous radioactive wastes generated by the plant.

h. The natural draft cooling tower is a concrete, natural draft, hyperbolic structure and is shown in Figure 1.2-15.

The tower is designed to operate alone or in conjunction with the auxiliary cooling tower to dissipate all excess heat removed from the main condensers and accomplishes this function by the use of a spray network, a film type heat transfer surface, a tower basin, and circulating water pumps, piping, and valves.

i. The Ultimate Heat Sink (see Figure 1.2-1) is comprised of two separate, seismic Category I, mechanical draft cooling tower/pumphouse/basin structures. Each tower consists of four cells; each cell with a separate stack. Only four cells are required to support Unit 1 operation. The towers are constructed of a reinforced concrete frame with air 1.2-16 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) intake louvers in the sides. Cells A and B of SSW cooling tower A and B contain ceramic fill blocks within the frame. Cells C and D of SSW cooling tower A and B contain stainless steel fill within the frame. Each tower is located over a separate concrete cooling water basin. Each pumphouse is located over the southwest corner of the basins, contains vertical wet pit pumps, and is provided with separate tornado missile protection walls on all sides, and on the roof.

j. The auxiliary cooling tower is a multi-cell mechanical draft fiberglass reinforced plastic structure with a concrete basin/foundation and is shown in Figure 1.2-16.

The auxiliary cooling tower is designed to operate in conjunction with the natural draft cooling tower to dissipate excess heat removed from the main condensers. It accomplishes this function by the use of a spray network, a film type heat transfer surface, electric motor driven fans, a tower basin, a discharge flume connected to the natural draft cooling tower basin, piping, valves, and associated electric equipment contained in the auxiliary cooling tower power and control building.

1.2.2.3 Nuclear System The nuclear system includes a direct cycle, forced circulation, General Electric boiling water reactor that produces steam for direct use in the steam turbine. A heat balance showing the major parameters of the nuclear system for the rated power conditions is shown in Figure 1.1-1.

Extended Power Uprate On September 8, 2010, Entergy requested approval of an amendment to the Grand Gulf Nuclear Station, Unit 1 (GGNS) Operating License and Technical Specifications to increase the maximum reactor core power operating limit authorized in the Operating License from 3898 megawatts thermal (MWt) to 4408 MWt. The license amendment request included NEDC-33477P, Safety Analysis Report for Grand Gulf Nuclear Station Constant Pressure Power Uprate, Revision 0(PUSAR), which the NRC also reviewed in conjunction with the amendment request. On July 18, 2012, the NRC approved the license amendment request.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.3.1 Reactor Core and Control Rods Fuel for the reactor core consists of slightly enriched uranium dioxide pellets sealed in Zircaloy-2 tubes. These tubes (or fuel rods) are assembled into individual fuel assemblies. Gross control of the core is achieved by movable, bottom-entry control rods. The control rods are cruciform in shape and are dispersed throughout the lattice of fuel assemblies. The control rods are positioned by individual control rod drives.

Each fuel assembly has several fuel rods with gadolinia (Gd2O3) mixed in solid solution with the UO2. The Gd2O3 is a burnable poison which diminishes the reactivity of the fresh fuel. It is depleted as the fuel reaches the end of its first cycle.

A conservative limit of plastic strain is the design criterion used for fuel rod cladding failure. The peak linear heat generation for steadystate operation is well below the fuel damage limit even late in life. Experience has shown that the control rods are not susceptible to distortion and have an average life expectancy many times the residence time of a fuel loading.

1.2.2.3.2 Reactor Vessel and Internals The reactor vessel contains the core and supporting structures; the steam separators and dryers; the jet pumps; the control rod guide tubes; the distribution lines for the feedwater, core sprays, and standby liquid control; the in-core instrumentation; and other components. The main connections to the vessel include steam lines, coolant recirculation lines, feedwater lines, control rod drive and in-core nuclear instrument housings, core spray lines, residual heat removal lines, standby liquid control line, core differential pressure line, jet pump pressure sensing lines, and water level instrumentation.

The reactor vessel is designed and fabricated in accordance with applicable codes for a pressure of 1250 psig. The nominal operating pressure in the steam space above the separators is 1040 psia. The vessel is fabricated of low alloy steel and is clad internally with stainless steel (except for the top head, nozzles, and nozzle weld zones which are unclad).

The reactor core is cooled by demineralized water that enters the lower portion of the core and boils as it flows upward around the fuel rods. The steam leaving the core is dried by steam separators 1.2-18 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) and dryers located in the upper portion of the reactor vessel. The steam is then directed to the turbine through the main steam lines. Each steam line is provided with two isolation valves in series; one on each side of the containment barrier.

1.2.2.3.3 Reactor Recirculation System The reactor recirculation system pumps reactor coolant through the core. This is accomplished by two recirculation loops external to the reactor vessel but inside the containment. Each external loop contains motor-operated maintenance valves and one hydraulically operated flow control valve. The variable position hydraulic flow control valve operates in conjunction with a low frequency motor-generator set to control reactor power level through the effects of coolant flow rate on moderator void content.

The internal portion of the loop consists of the jet pumps which contain no moving parts. The jet pumps provide a continuous internal circulation path for the major portion of the core coolant flow. The jet pumps are located in the annular region between the core shroud and the vessel inner wall. Any recirculation line break would still allow core flooding to approximately two-thirds of the core height - the level of the inlet of the jet pumps.

1.2.2.3.4 Residual Heat Removal System The residual heat removal (RHR) system is a system of pumps, heat exchangers, and piping that fulfills the following functions:

a. Removes decay and sensible heat during and after plant shutdown.

b. Injects water into the reactor vessel, following a loss-of-coolant accident, rapidly enough to reflood the core and maintain fuel cladding below fragmentation temperature independent of other core cooling systems. This is discussed in subsection 1.2.2.4.8, Emergency Core Cooling Systems.

c. Removes heat from the containment, following a loss-of-coolant accident, to limit the increase in containment pressure. This is accomplished by cooling and 1.2-19 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) recirculating the suppression pool water (containment cooling) and by spraying the containment air space (containment spray) with suppression pool water.

1.2.2.3.5 Reactor Water Cleanup System The reactor water cleanup system recirculates a portion of reactor coolant through a filter-demineralizer to remove particulate and dissolved impurities from the reactor coolant. It also removes excess coolant from the reactor system under controlled conditions.

1.2.2.3.6 Nuclear Leak Detection System The nuclear leak detection system consists of temperature, pressure, flow, and fission-product sensors with associated instrumentation and alarms. This system detects and annunciates leakage in the following systems:

a. Main steam lines

b. Reactor water cleanup (RWCU) system

c. Residual heat removal (RHR) system

d. Reactor core isolation cooling (RCIC) system

e. Fuel pool cooling and cleanup (FPCC) system

f. High pressure core spray (HPCS) system

g. Low pressure core spray (LPCS) system

h. Instrument lines Small leaks generally are detected by temperature and pressure changes, fillup rate of drain sumps, and fission product concentration inside the containment. Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.4 Nuclear Safety Systems and Engineered Safety Features 1.2.2.4.1 Reactor Protection System The reactor protection system (RPS) initiates a rapid, automatic shutdown (scram) of the reactor. It acts in time to prevent fuel cladding damage and any nuclear system process barrier damage following abnormal operational transients. The reactor protection system overrides all operator actions and process controls and is based on a fail-safe design philosophy that allows appropriate protective action even if a single failure occurs.

1.2.2.4.2 Neutron Monitoring System Although not all portions of the neutron monitoring system qualify as a nuclear safety system, those that provide high neutron flux signals to the reactor protection system do. The intermediate range monitors (IRMs) and average power range monitors (APRMs), which monitor neutron flux via in-core detectors, signal the reactor protection system to initiate a scram in time to prevent excessive fuel cladding damage as a result of overpower transients. The source range monitors (SRMs) prevent rod motion in the startup mode when certain conditions discussed in subsection 7.6.1.6.2d are not satisfied.

1.2.2.4.3 Control Rod Drive System When a scram is initiated by the reactor protection system, the control rod drive system inserts the negative reactivity necessary to shut down the reactor. Each control rod is controlled individually by a hydraulic control unit. When a scram signal is received, high pressure water stored in an accumulator in the hydraulic control unit or reactor pressure forces its control rod into the core.

1.2.2.4.4 Control Rod Drive Housing Supports Control rod drive housing supports are located underneath the reactor vessel near the control rod housings. The supports limit the travel of a control rod in the event that a control rod housing is ruptured. The supports prevent a nuclear excursion as a result of a housing failure and thus protect the fuel barrier.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.4.5 Control Rod Velocity Limiter A control rod velocity limiter is attached to each control rod to limit the velocity at which a control rod can fall out of the core should it become detached from its control rod drive. This action limits the rate of reactivity insertion resulting from a rod drop accident. The limiters contain no moving parts.

1.2.2.4.6 Nuclear System Pressure Relief System A pressure relief system consisting of safety/relief valves mounted on the main steam lines is provided to prevent excessive pressure inside the nuclear system following either abnormal operational transients or accidents.

1.2.2.4.7 Reactor Core Isolation Cooling System The reactor core isolation cooling system (RCIC) provides makeup water to the reactor vessel when the vessel is isolated. The RCIC system uses a steam-driven turbine-pump unit and operates automatically in time and with sufficient coolant flow to maintain adequate water level in the reactor vessel.

1.2.2.4.8 Emergency Core Cooling Systems (ESF System)

Four emergency core cooling systems are provided to maintain fuel cladding below fragmentation temperature in the event of a breach in the reactor coolant pressure boundary that results in a loss of reactor coolant. The systems are:

High pressure core spray (HPCS) system Automatic depressurization (ADS)

Low pressure core spray (LPCS)

Low pressure coolant injection (LPCI), an operating mode of the residual heat removal system

a. High Pressure Core Spray - The HPCS system provides and maintains an adequate coolant inventory inside the reactor vessel to maintain fuel cladding temperatures below fragmentation temperature in the event of breaks in the reactor coolant pressure boundary. The system is initiated by either high pressure in the drywell or low water level in the vessel. It operates independently of all other 1.2-22 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) systems over the entire range of pressure differences from greater than normal operating pressure to zero. The HPCS cooling decreases vessel pressure to enable the low pressure cooling systems to function. The HPCS system pump motor is powered by a diesel generator if auxiliary power is not available, and the system may also be used as a backup for the RCIC system.

b. Automatic Depressurization - The automatic depressurization system rapidly reduces reactor vessel pressure in a loss-of-coolant (LOCA) accident situation in which the HPCS system fails to maintain the reactor vessel water level. The depressurization provided by the system enables the low pressure emergency core cooling systems to deliver cooling water to the reactor vessel. The ADS uses some of the relief valves that are part of the nuclear system pressure relief system. The automatic relief valves are arranged to open on conditions indicating both that a break in the reactor coolant pressure boundary has occurred and that the HPCS system is not delivering sufficient cooling water to the reactor vessel to maintain the water level above a preselected value. The ADS will not be activated unless either the LPCS or LPCI pumps are operating. This is to ensure that adequate coolant will be available to maintain reactor water level after the depressurization.

c. Low Pressure Core Spray - The LPCS system consists of one independent pump and the valves and piping to deliver cooling water to a spray sparger over the core. The system is actuated by conditions indicating that a breach exists in the reactor coolant pressure boundary but water is delivered to the core only after reactor vessel pressure is reduced. This system provides the capability to cool the fuel by spraying water into each fuel channel. The LPCS loop functioning in conjunction with the ADS or HPCS can maintain the fuel cladding below the prescribed temperature limit following a loss-of-coolant accident.

d. Low Pressure Coolant Injection - Low pressure coolant injection is an operating mode of the residual heat removal (RHR) system, but is discussed here because the LPCI mode acts as an engineered safety feature in conjunction with the other emergency core cooling systems.

LPCI uses the pump loops of the RHR to inject cooling 1.2-23 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) water into the pressure vessel. LPCI is actuated by conditions indicating a breach in the reactor coolant pressure boundary, but water is delivered to the core only after reactor vessel pressure is reduced. LPCI operation provides the capability of core reflooding, following a loss-of-coolant accident, in time to maintain the fuel cladding below the prescribed temperature limit.

1.2.2.4.9 Containment Systems 1.2.2.4.9.1 Containment Functional Design The containment design for this plant has been given the name Mark III. This containment design incorporates the drywell/pressure suppression feature of previous BWR containment designs into a dry-containment type of structure.

In fulfilling its design basis as a fission product barrier in case of an accident, the Mark III containment is a low-leakage structure even at the elevated pressures that could follow a main steam line rupture or a recirculation line break.

The main features of the design include the following:

a. A drywell surrounding the reactor pressure vessel (RPV) and a large part of the reactor coolant pressure boundary

b. A suppression pool that serves as a heat sink during normal operational transients and accident conditions

c. A containment upper pool for shielding, refueling operations, and makeup to the suppression pool

d. A steel-lined reinforced concrete containment structure The containment functional design is described in more detail in subsection 6.2.1.

1.2.2.4.9.2 RHR/Suppression Pool Cooling The suppression pool cooling subsystem of RHR is placed in operation to limit the temperature of the water in the suppression pool following a design basis loss-of-coolant accident, to control the pool temperature during normal operation of the safety-relief valves and the RCIC system, and to reduce the pool temperature following an isolation transient. In the suppression 1.2-24 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) pool cooling mode of operation, the RHR main system pumps take suction from the suppression pool and pump the water through the RHR heat exchangers where cooling takes place by transferring heat to the service water. The fluid is then discharged back to the suppression pool or the reactor pressure vessel.

1.2.2.4.9.3 RHR/Containment Spray (ESF System)

A containment spray system is provided to function, by automatic initiation, to condense steam which may bypass the suppression pool to prevent over-pressurization of the containment following a LOCA. The containment spray system consists of two redundant subsystems, each with its own full-capacity spray header. Each subsystem is supplied from a separate redundant RHR subsystem.

The containment spray system also serves as an iodine removal system to reduce doses to the environment following a LOCA.

1.2.2.4.9.4 Combustible Gas Control (ESF System)

In the unlikely event of a loss-of-coolant accident, hydrogen and oxygen will be generated in the drywell and containment. The combustible gas control system will ensure that hydrogen concentrations are kept below the limits specified in NRC Regulatory Guide 1.7, Revision 1. For postulated degraded core accidents, the combustible gas control system will preclude the potential for local detonations and ensure the integrity of the containment. The systems used will include a drywell purge system, hydrogen control systems, and a backup containment purge system.

The drywell purge compressor also performs the function diluting the drywell source term with the containment and suppression pool environment by pressurizing the drywell and discharging the drywell source term through the drywell suppression pool vents.

With the implementation of the alternative source term (Amendment 145), this dilution of drywell source term is no longer credited in the Equipment Qualification analysis which is presented in FSAR Section 3.11.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.4.10 Containment and Reactor Vessel Isolation Control System (ESF System)

The containment and reactor vessel isolation control system automatically initiates closure of isolation valves to close off all process lines that are potential leakage paths for radioactive material to the environs. This action is taken upon indication of a breach in the reactor coolant pressure boundary.

1.2.2.4.10.1 Main Steam Line Isolation Valves Although all pipelines that both penetrate the containment and offer a potential release path for radioactive material are provided with redundant isolation capabilities, the main steam lines, because of their large size and large mass flow rates, are given special isolation consideration. Automatic isolation valves are provided in each main steam line. Each is powered by both air pressure and spring force. These valves fulfill the following objectives:

a. Prevent excessive damage to the fuel barrier by limiting the loss of reactor coolant from the reactor vessel resulting from either a major leak from the steam piping outside the containment or a malfunction of the pressure control system resulting in excessive steam flow from the reactor vessel.

b. Limit the release of radioactive materials by isolating the reactor coolant pressure boundary in case of a gross release of radioactive materials from the fuel to the reactor cooling water and steam.

c. Limit the release of radioactive materials by closing the containment barrier in case of a major leak from the nuclear system inside the containment.

1.2.2.4.10.2 Main Steam Line Flow Restrictors A venturi-type flow restrictor is installed in each steam line.

These devices limit the loss of coolant from the reactor vessel before the main steam line isolation valves are closed in case of a main steam line break outside the containment.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.4.11 Process Radiation Monitoring System 1.2.2.4.11.1 Main Steam Line Radiation Monitoring Subsystem The main steam line radiation monitoring subsystem consists of four gamma radiation monitors located externally to the main steam lines just outside the containment. The monitors are designed to detect a gross release of fission products from the fuel. On detection of high radiation, the trip signals generated by the monitors are used to initiate closure of the Rx water sample line drywell isolation valves and trip the mechanical vacuum pump and valves.

1.2.2.4.11.2 Ventilation Exhaust Radiation Monitoring System The process ventilation radiation monitoring systems consist of a number of radiation monitors arranged to monitor the activity level of the air exhaust from the containment and drywell, auxiliary building fuel handling and pool sweep areas, and air intake into the control room.

1.2.2.4.12 Standby Gas Treatment System (ESF System)

The standby gas treatment system has been designed to minimize exfiltration of contaminated air from the enclosure building, the auxiliary building, and the containment following an accident or abnormal condition that could result in abnormally high airborne radioactivity in these areas.

All necessary equipment and surrounding structures have been designed to seismic Category I specifications.

All components of the standby gas treatment system will be operable during a loss of offsite power supply.

1.2.2.4.13 Auxiliary Building Isolation Control System The auxiliary building isolation control system automatically initiates closure of isolation valves on selected lines that penetrate the auxiliary building to preserve the integrity of the standby gas treatment boundary. This action is taken upon indication of a breach in the reactor coolant pressure boundary.

1.2-27 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.4.14 Safety-Related Electrical Power Systems Standby ac power is supplied by three diesel generators. Each engineered safety features (ESF) division is supplied by a separate diesel generator. There are no provisions for transferring ESF division buses between standby ac power supplies or supplying more than one ESF division from one diesel generator.

The one-to-one relationship between diesel generator and ESF division ensures that a failure of one diesel generator can affect only one ESF division.

Three independent Class IE 125-volt dc systems exist, one per ESF division of the Class IE electric power system.

1.2.2.4.15 Standby Liquid Control System Although not intended to provide prompt reactor shutdown, as the control rods are, the standby liquid control system provides a redundant, independent, and alternate way to bring the nuclear fission reaction to subcriticality and to maintain subcriticality as the reactor cools. The system makes possible an orderly and safe shutdown in the event that not enough control rods can be inserted into the reactor core to accomplish shutdown in the normal manner. The system is sized to counteract the positive reactivity effect from rated power to the cold shutdown condition.

1.2.2.4.16 Safe Shutdown from Outside the Control Room In the event that the control room becomes inaccessible, the reactor can be brought from power range operation to cold shutdown conditions by the use of the local controls and equipment that are available outside the control room.

1.2.2.4.17 Main Steam Line Isolation Valve Leakage Control System (ESF System)

The main steam line isolation valve leakage control system (MSIVLCS) is designed to minimize the fission products which could bypass the standby gas treatment system after a LOCA. This is accomplished by directing the leakage through the closed main steam line isolation valves to a space serviced by the Standby Gas Treatment System (SGTS).

1.2-28 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.4.18 Feedwater Leakage Control System (ESF System)

The feedwater leakage control system is designed to minimize the fission products which could bypass the SGTS after a LOCA. This is accomplished by filling the feedwater lines between the containment isolation valves with suppression pool water and maintaining a water seal at a pressure slightly higher than the containment pressure.

1.2.2.4.19 Suppression Pool Make-up System (ESF System)

The suppression pool make-up system provides water from the upper containment pool to the suppression pool by gravity flow following a LOCA. The quantity of water provided is sufficient to maintain required drywell upper-most vent coverage for all postulated accidents.

1.2.2.4.20 Control Room HVAC (ESF System)

The control room HVAC system provides an environment in the control room suitable for the operation of equipment necessary for the safe shutdown of the plant and will function in the event of a LOCA. The system shall protect the plant operators from the results of any accident which could impair their safety and therefore compromise the safety of the plant.

1.2.2.5 Power Conversion System 1.2.2.5.1 Turbine Generator The turbine generator is an 1800-rpm, tandem-compound, six-flow, 46-inch last-stage buckets, reheat unit with electrohydraulic control (EHC) for normal operation. The EHC system is equipped with two independent levels of speed sensing. The approximate rating of the turbine generator is 1,352,907 kw.

The generator is a direct-driven, three-phase, 60-Hz, 22,000 volt, 1800-rpm, hydrogen cooled with water cooled stator and rotor windings, synchronous generator rated at 1600 MVA at 0.9 power factor, 75 psig hydrogen pressure, and 0.58 short-circuit ratio.

1.2-29 LBDCR 2018-076

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.5.2 Main Steam System The main steam system delivers steam from the nuclear boiler system via four 28-inch OD steam lines to the turbine generator, second-stage reheaters, steam jet air ejectors, offgas preheater, and to the reactor feed pump turbines, seal steam generators, and main condenser hotwell at startup and low loads.

1.2.2.5.3 Main Condenser Steam from the low-pressure turbine is exhausted directly downward into the condenser shells through exhaust openings in the bottom of the turbine casings and is condensed. The condenser is a three-section, multipressure condenser, each section serving one double-flow, low-pressure turbine section. The condenser also serves as a heat sink for the turbine bypass system, feedwater heater and drain tank high-level dumps, relief valve discharges during transient conditions and reactor feed pump turbine exhausts.

1.2.2.5.4 Main Condenser Evacuation System The main condenser evacuation system removes the noncondensable gases from the main condenser and exhausts them to the gaseous radwaste system. Two twin-element, two-stage, steam jet air ejectors (100-percent-capacity each), complete with intercondenser, are provided for air removal during normal operation. A mechanical vacuum pump is used during startup.

1.2.2.5.5 Turbine Gland Sealing System The turbine gland sealing system provides clean, nonradioactive steam to the seals of the turbine valve stem glands and the turbine shaft glands. The sealing steam is supplied by a separate seal steam generator using condensate from the condensate storage tank during normal plant operation. The unit auxiliary boiler provides an auxiliary steam supply for startup and when the seal steam generator is not available. The seal steam condenser collects and condenses the air and steam mixture and discharges the air leakage to the turbine building vent, using a motor-driven exhauster. Contaminated gland seal heating steam is condensed in Feedwater Heater No. 4.

1.2-30 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.5.6 Steam Bypass System and Pressure Control System A turbine bypass system is provided which passes steam directly to the main condenser under the control of the pressure controller.

Steam is bypassed to the condenser whenever the reactor steaming rate exceeds the load permitted to pass to the turbine generator.

The capacity of the turbine bypass system is 30.4% of the reactor rated steam flow. The pressure control system provides main turbine control valve and bypass valve position demands so as to maintain a nearly constant reactor pressure during normal plant operation.

1.2.2.5.7 Circulating Water System The circulating water system provides the main condenser with a continuous supply of cooling water to remove the heat rejected from the cycle. The circulating water system is a closed system utilizing a natural draft cooling tower and a mechanical draft auxiliary cooling tower. Two vertical motor-driven pumps circulate the cooling water from the cooling tower basin through the main condenser and then back to the cooling towers. Makeup water, to compensate for drift, blowdown, and evaporation losses, is supplied from the plant service water system.

1.2.2.5.8 Condensate and Feedwater Systems Three condensate pumps take the deaerated condensate from the hotwell of the intermediate-pressure shell of the main condenser and deliver it, in turn, through the condensate full flow filters and the condensate demineralizers. Filtered and demineralizer effluent then passes to the three condensate booster pumps, and the condensate booster pumps then discharge through four stages of low-pressure feedwater heaters to the two turbine-driven reactor feed pumps. Drains from moisture-separator and reheaters, and the fifth- and sixth-stage feedwater heaters, are pumped forward by two heater drain pumps, and the drains from first-,

second-, third-, and fourth-stage lower-pressure heaters are cascaded back to the main condenser. The reactor feed pumps discharge the total feedwater flow through the fifth- and the sixth-stage high-pressure feedwater heaters to the reactor.

Contaminated gland sealing steam from the reactor feed pump turbines is condensed in the main condenser.

1.2-31 Revision 2018-076

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.5.9 Condensate Cleanup System The 133-percent-capacity condensate cleanup system consists of eight units of multiple deep-bed-type demineralizers (with two units as spares) that operate in parallel. The system also includes three precoat filters (precoat filter B abandoned in place)used for startup and normal operation.

The condensate cleanup system maintains the required purity of feedwater flowing to the reactor.

1.2.2.5.10 Hydrogen Water Chemistry System A hydrogen water chemistry (HWC) system is provided to further reduce the susceptibility of reactor recirculation piping and reactor vessel internal materials to intergranular stress corrosion cracking. This is accomplished by injecting hydrogen into the condensate booster pump suction header to suppress the formation of radiolytic oxygen in the reactor coolant. Oxygen is injected into the offgas system to maintain a stoichiometric balance of hydrogen and oxygen entering the offgas recombiners.

1.2.2.6 Electrical Systems and Instrumentation Control 1.2.2.6.1 Electrical Power Systems The station generator power is fed to a main step-up transformer bank through the isolated phase bus system that was modified for EPU. The main step-up transformer bank transforms the power generated at 20.9 kV (originally 22 kV) to 500 kV. Then it is fed to the switchyard where the distribution of power to the utility grid via the transmission lines and to the station for station ac power requirements takes place.

The switchyard is fed by three 500 kV transmission lines on separate right-of-ways. The station offsite power is fed by two 500 kV circuits from the switchyard and one independent 115 kV offsite circuit. Each 500 kV circuit feeds a service transformer which provides engineered safety features (ESF) and balance-of-plant transformers with 34.5 kV power for further voltage transformations. The independent 115 kV offsite circuit feeds a third ESF transformer. The ESF transformers provide only the ESF buses with 4.16 kV ac power, and the balance-of-plant transformers supply the 4.16 kV ac power, and the balance-of-plant transformers supply the 4.16, 6.9, and 13.8 kV ac power 1.2-32 LBDCR 2018-049

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) requirements for the balance-of-plant load groups. The ac power is also distributed to the ESF and balance-of-plant loads at 480 volts from the associated load centers and motor control centers.

Changes in electrical load demand associated with EPU are two nonsafety-related changes, the addition of Radial Wells and the Auxiliary Cooling Tower (ACT) expansion.

Three independent Class IE 125-volt dc systems exist, one per ESF division of the Class IE electric power system. For the balance-of-plant electric system, three 125-volt dc systems are provided; two of these are connected in series to provide a 250-volt dc system for large dc loads.

Standby ac power is supplied by three diesel generators. Each ESF division is supplied by a separate diesel generator. There are no provisions for transferring ESF division buses between standby ac power supplies or supplying more than one ESF division from one diesel generator. The one-to-one relationship between diesel generator and ESF division ensures that a failure of one diesel generator can affect only one ESF division.

1.2.2.6.2 Nuclear System Process Control and Instrumentation 1.2.2.6.2.1 Rod Control and Information System The rod control and information system provides the means by which control rods are positioned from the control room for power control. The system operates valves in each hydraulic control unit to change control rod position. One control rod or a group of rods can be manipulated at a time. The system includes the logic that restricts control rod movement (rod block) under certain conditions as a backup to procedural controls.

1.2.2.6.2.2 Recirculation Flow Control System During normal power operation a variable position discharge valve is used to control flow. Adjusting this valve changes the coolant flow rate through the core and thereby changes the core power level. For startup and shutdown flow changes at lower power, the pump speed is changed by adjusting the frequency of the electrical power supply.

1.2-33 Revision 2016-000

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.6.2.3 Neutron Monitoring System The neutron monitoring system is a system of in-core neutron detectors and out-of-core electronic monitoring equipment. The system provides indication of neutron flux, which can be correlated to thermal power level for the entire range of flux conditions that can exist in the core. The source range monitors (SRMs) and the intermediate range monitors (IRMs) provide flux level indications during reactor startup and low power operation.

The local power range monitors (LPRMs) and average power range monitors (APRMs) allow assessment of local and overall flux conditions during power range operation. The traversing in-core probe system (TIP) provides a means to calibrate the individual LPRM sensors. The Neutron Monitoring System provides inputs to the Rod Control and Information System to initiate rod block trips if preset flux limits are exceeded, and inputs to the Reactor Protection System to initiate a scram if other limits are exceeded.

1.2.2.6.2.4 Refueling Interlocks A system of interlocks that restricts movement of refueling equipment and control rods when the reactor is in the refueling and startup modes is provided to prevent an inadvertent criticality during refueling operations. The interlocks back up procedural controls that have the same objective. The interlocks affect the refueling platform, refueling platform main hoist, and control rods.

1.2.2.6.2.5 Reactor Vessel Instrumentation In addition to instrumentation for the nuclear safety systems and engineered safety features, instrumentation is provided to monitor and transmit information that can be used to assess conditions existing inside the reactor vessel and the physical condition of the vessel itself. This instrumentation monitors reactor vessel pressure, water level, coolant temperature, reactor core differential pressure, coolant flow rates, and reactor vessel head inner seal ring leakage.

1.2.2.6.2.6 Core Performance Monitoring System An on-line core performance monitoring system is provided to monitor and log process variables and to make certain analytical computations.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.6.3 Power Conversion Systems Process Control and Instrumentation 1.2.2.6.3.1 Pressure Regulator and Turbine-Generator Control The pressure controller maintains control of the turbine control valves and turbine bypass valves to allow proper generator and reactor response to system load demand changes while maintaining the nuclear system pressure essentially constant.

In normal operating conditions, the turbine-generator operates in the turbine-follow-reactor mode. In this mode, it controls throttle pressure by modulation of the turbine control valves and bypass valves, but can respond to increases in grid frequency by closing turbine control valves and opening the bypass valves. The system does not respond to decreases in frequency.

The turbine-generator speed-load controls act to maintain the turbine speed (generator frequency) constant and respond to load changes by adjusting the reactor recirculation flow control system and pressure control set point.

1.2.2.6.3.2 Feedwater Control System The feedwater control system automatically controls the flow of feedwater into the reactor pressure vessel to maintain the water within the vessel at predetermined levels. A conventional three element control system is used to accomplish this function.

1.2.2.7 Fuel Handling and Storage Systems 1.2.2.7.1 New and Spent Fuel Storage New and spent fuel storage racks are designed to prevent inadvertent criticality and load buckling. Sufficient coolant and shielding are maintained to prevent overheating and excessive personnel exposure, respectively. The design of the fuel pool provides for corrosion resistance, adherence to seismic Category I requirements, and prevention of keff from exceeding 0.95 under flooded conditions. This subject is further discussed in Section 9.1.

1.2-35 Revision 2018-076

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.7.2 Fuel Handling System The fuel handling equipment includes a 125-ton cask and a 150-ton cask crane, new fuel bridge crane, fuel handling platform, fuel inspection stand, fuel preparation machine, fuel assembly transfer mechanism, containment refueling platform, containment polar crane, and other related tools for reactor servicing. All equipment conforms with applicable codes and standards. The principal function of the cask crane is to handle the spent fuel cask. The new fuel bridge crane normally transfers new fuel while in the fuel handling area until it is placed in the fuel preparation machine. The fuel handling platform transfers the fuel assemblies between the transfer pool, storage pools, and cask. Fuel assemblies are transferred through the transfer tube between the containment and the auxiliary building. The fuel assemblies inside the containment are handled by the refueling platform.

The disassembly and reassembly of the reactor head, internals, and drywell head during refueling is done using the containment polar crane.

All tools and servicing equipment necessary to meet the reactor general servicing requirements are designed for efficiency and safe serviceability.

1.2.2.8 Cooling Water and Auxiliary Systems 1.2.2.8.1 Standby Service Water System The standby service water (SSW) system is designed to cool reactor auxiliaries essential to a safe reactor shutdown, to minimize the leakage of radioactive contamination from these auxiliaries to the environment, to provide a means of flooding the drywell and containment, and to provide a backup source of makeup water to the spent fuel pool. The system consists of two independent trains, each capable of cooling the engineered safety features following a LOCA and rejecting this heat to the atmosphere through one of the two redundant standby service water cooling towers. The system is designed to meet seismic Category I requirements. In the unlikely event that radioactive contamination occurs in either train, the radiation monitors of the system will alarm and permit the operator to isolate the portion of the system that is contaminated.

1.2-36 Revision 2018-076

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.8.2 Component Cooling Water System The component cooling water (CCW) system is a closed-loop system that provides parallel flow cooling to auxiliary equipment in the containment, drywell, and auxiliary buildings. The closed loop provides a barrier between contaminated systems and the service water discharged to the environment. Heat is removed from the closed loop by the plant service water system. The system has no safety-related function or required for a safe shutdown of the reactor and it is not designed to seismic Category I requirements.

However, piping and valves associated with fuel pool heat exchangers and piping and valves forming a part of containment boundary are safety-related and designed to seismic Category I requirements. Radiation monitors are provided to detect contaminated leakage into the closed system.

1.2.2.8.3 Turbine Building Cooling Water System This system is designed to cool the auxiliary plant equipment associated with the power conversion systems over the full range of normal plant operation.

1.2.2.8.4 Ultimate Heat Sink The ultimate heat sink, consisting of the standby service water (SSW) system cooling towers and makeup basins, provides heat rejection and makeup water required for the dissipation of heat to permit the safe shutdown and cooldown of the plant and to maintain it in a safe shutdown condition. The SSW cooling towers are seismic Category I.

1.2.2.8.5 Condensate Storage and Transfer System The condensate storage and transfer system maintains the required capacity and flow of the condensate for the RCIC and HPCS systems and maintains the required level in the condenser hotwell. The system also: Stores and transfers upper containment pool water during refueling, and cask storage pool water during fuel shipping cask loading; receives and stores the process effluent from the liquid radwaste system and provides makeup to other plant systems where required; provides storage space for the suppression pool water during plant shutdown, and provides condensate to the control rod drive (CRD) hydraulic system.

The system consists of a condensate storage tank, two condensate transfer pumps, and the necessary controls and instrumentation.

1.2-37 Revision 2018-076

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.8.6 Makeup Water Treatment System The makeup water treatment system furnishes suitable water as makeup for the plant. The permanent plant equipment consists of two trains, each containing a mixed bed cation exchanger, a mixed bed anion exchanger, and a charcoal filter. Connections are available for a mobile vendor supplied water treatment system.

1.2.2.8.7 Domestic Water System and Sanitary Waste Water System The domestic water system provides the necessary supply of domestic water for the plant. Construction water is used as the domestic water system supply.

The sanitary waste water system is designed to maintain the sewer waste water quality in accordance with the applicable quality criteria limits.

1.2.2.8.8 Chilled Water Systems Chilled water is produced by mechanical chilling units and supplied to area cooling units through closed recirculating piping systems. Chemical water treatment is provided for scale and corrosion control.

1.2.2.8.9 Compressed Air Systems The service, instrument and plant air systems provide a continuous supply of compressed air of suitable quality and pressure for instrument control and general plant use. The plant air compressors discharge into their respective air receivers.

The air is then distributed throughout the plant. Instrument air is additionally filtered and dried by plant air dryers prior to distribution throughout the plant.

1.2.2.8.10 Process Sampling Systems The process sampling system is furnished to provide process information that is required to monitor plant and equipment performance and changes to operating parameters. Representative liquid and gas samples are taken automatically and/or manually during normal plant operation for laboratory or on-line analyses.

1.2-38 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.8.11 Plant Floor and Equipment Drainage The floor and equipment drainage system is designed to collect liquid waste throughout the plant and discharge the radioactive and potentially radioactive waste to the radwaste system for processing. Separate drainage facilities are provided for nonradioactive waste.

The drainage system is also used to detect abnormal leakage in the emergency safety features rooms, the drywell, and containment.

1.2.2.8.12 Heating, Ventilating, and Air Conditioning Systems The plant heating, ventilating, and air conditioning systems are designed to provide an environment with controlled temperature and humidity to ensure the comfort and safety of personnel and the integrity of plant equipment.

Plant heating, ventilating, and air conditioning systems serving engineered safety features equipment are designed with sufficient redundancy to ensure operation during emergency conditions.

1.2.2.8.13 Fire Protection System The fire protection system is designed to provide an adequate supply of water or chemicals to points throughout the plant area where fire protection may be required. Diversified fire-alarm and fire-suppression types are selected to suit the particular areas or hazards being protected. The water for the system is taken from two 300,000-gallon tanks that are replenished automatically from the plant service water system. In addition to the tanks, the system consists of one electric-driven pump, two diesel engine-driven pumps, one jockey pump, and the associated piping, valves, and hydrants.

Chemical fire-fighting systems (e.g., CO2 and Halon 1301) are also provided as additions to or in lieu of the water fire-fighting systems.

The necessary instrumentation and controls are provided for the proper operation of the fire-fighting systems and for fire detection and annunciation.

1.2-39 Revision 2019-044

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.8.14 Communications Systems Diverse systems have been provided for intra-plant and plant-to-offsite communication. A detailed description of the systems is provided in section 9.5.2.

1.2.2.8.15 Lighting Systems The design of the lighting facilities is based on standards of the Illuminating Engineering Society. Special attention is given to areas where proper lighting is imperative during normal and emergency operations. The system design precludes the use of mercury vapor fixtures in the containment and the fuel handling area except where specifically evaluated and approved. The normal lighting systems are fed from the normal buses. Essential lighting fixtures are supplied by engineered safety features buses and are backed up by diesel-generator units. Emergency lighting fixtures are backed up by inverters off the station batteries and self-contained batteries. Normal operation and regular simulated offsite power-loss tests verify system integrity.

1.2.2.8.16 Diesel Generator Fuel-Oil System The purpose of this system is to supply and store the fuel oil required to operate the diesel-generator units during post-LOCA maximum load demands. The principal design criteria associated with this system consist of the following:

a. Seven-day fuel oil capacity to meet the conditions above is provided for each diesel

b. Seismic Category I design

c. Missile protection 1.2.2.8.17 Auxiliary Steam System An auxiliary steam system is provided to furnish a separate and independent steam supply. Process steam is generated in packaged, high voltage, electrode boilers and distributed through the plant by an auxiliary steam header. Auxiliary steam is required for 1.2-40 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) condensate deaeration/heating, pump testing, and main turbine shaft seal steam during startup. Auxiliary steam is also used for plant heating and other miscellaneous plant processes.

1.2.2.8.18 Plant Service Water System The plant service water (PSW) system is designed to cool plant auxiliaries that are not potential sources of radioactive contamination during normal operation, that are not required for safe reactor shutdown, and that can be efficiently cooled by raw well water. During refueling outages, PSW may provide cooling water for the alternate decay heat removal subsystem of RHR. The PSW system also provides makeup to the circulating water system and the water treatment system. The system draws water from the radial well system, pumps the coolant through the heat exchangers, and discharges to the circulating water system.

1.2.2.8.19 Containment Ventilation The containment ventilation system consists of a normally operating containment ventilation system, a containment purge system, and a drywell purge system.

The containment ventilation system has been designed to provide a reliable source of fresh air, and to filter the containment air by recirculation through filter trains.

The containment purge system has been designed to purge the containment completely, when required, at a minimum rate of one air change per 5-hour period.

The drywell purge system has been designed either to purge the drywell at a minimum rate of one air change per hour or to serve as a drywell cleanup system for the removal of airborne contamination at a minimum recirculation rate of one air change per hour.

1.2.2.8.20 Fuel Pool Cooling and Cleanup System The fuel pool cooling and cleanup system maintains acceptable temperature, clarity, and radioactivity levels of the water in the upper containment, fuel storage, and cask pools. The system includes two heat exchangers, each with the capacity for removing 1.2-41 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 15.0 x 106 Btu/hr from the pool with 140 F pool water and 95 F cooling water and having the capacity to pass the system flow or greater to maintain the desired purity level.

Detailed system operation is provided in Section 9.1.3.

1.2.2.9 Radioactive Waste Management 1.2.2.9.1 Gaseous Radwaste System The purpose of the gaseous radwaste system is to process and control the release of gaseous radioactive wastes to the site environs so that the total radiation exposure to persons outside the controlled area does not exceed the maximum limits of the applicable 10 CFR 20 regulations even with some defective fuel rods.

The offgases from the main condenser are the major source of gaseous radioactive waste. The treatment of these gases includes volume reduction through a catalytic hydrogen-oxygen recombiner, water vapor removal through a condenser, decay of short-lived radioisotopes through a holdup line, further condensation and cooling filtration, adsorption of isotopes on activated charcoal beds, further filtration through high efficiency filters, and final releases.

Continuous radiation monitors are provided which indicate radioactive release from the reactor and from the charcoal absorbers. The radiation monitors are used to isolate the offgas system on high radioactivity in order to prevent gas of unacceptably high activity from release.

1.2.2.9.2 Liquid Waste System The liquid waste system, consisting of equipment drain, floor drain, and chemical waste subsystems, is designed to collect and process waste generated throughout the plant. Processing of the waste is sufficient to allow recycle of the wastewater. Ties exist between all the subsystems to provide backup processing in the event of failure.

Continuous radiation monitors in the discharge line provide indications and records of radioactivity release and automatically discontinue flow in the event of high activity levels.

1.2-42 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.9.3 Solid Waste System The solid waste system is designed to handle and dispose of solid waste produced by the plant. The waste, depending on activity and type, will be packaged for offsite shipment in accordance with all applicable regulations.

1.2.2.10 Radiation Monitoring and Control 1.2.2.10.1 Process Radiation Monitoring System Process radiation monitoring systems are provided to monitor and control radioactivity in process and effluent streams and to activate appropriate alarms and controls.

A process radiation monitoring system is provided for indicating and recording radiation levels associated with plant process streams and effluent paths leading to the environment. All effluents from the plant which are potentially radioactive are monitored.

Process radiation monitoring is also discussed in subsections 7.6.1.2 and 12.3.4.

1.2.2.10.2 Area Radiation Monitoring System The area radiation monitoring system functions to alert plant personnel of increasing or abnormally high radiation levels which could possibly result in inadvertent overexposure. The system consists of detectors located throughout the plant, along with local alarms, and has readout, alarming, and recording provisions in the control room.

1.2.2.10.3 Offsite Radiological Monitoring System The important pathways to man are monitored by radiological measurements, including surveys, passive dosimeters, and samples collected for laboratory analyses. These include airborne, aquatic, and terrestrial pathways. The radiological monitoring program is implemented at least one year prior to reactor criticality. The program is designed to document background levels of direct radiation and concentrations of radionuclides that exist in aquatic and terrestrial ecosystems before and after plant operation and document the concentrations of radionuclides that could be attributable to operation of the Grand Gulf Nuclear Station.

1.2-43 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2.2.11 Particularly Difficult Engineering Problems In general, particularly difficult engineering problems can be defined as those requiring development work or vendor testing to finalize the design. Such areas are discussed in Section 1.5.

1.2.2.12 Extrapolation of Technology There are no significant extrapolations of technology incorporated in the Grand Gulf Nuclear Station.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2-45 LBDCR 2016-005

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2-49 LBDCR 2016-195

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2-57 LBDCR 2016-041

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

FIGURE 1.2-11: Deleted (See Figure 12.3-6) 1.2-58 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

FIGURE 1.2-12: Deleted (See Figure 12.3-7) 1.2-59 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

FIGURE 1.2-13: Deleted (See Figure 12.3-8) 1.2-60 LBDCR 2019-008

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.2-61 Revision 2016-00

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.3 COMPARISON TABLES 1.3.1 Comparisons with Similar Facility Designs

[HISTORICAL INFORMATION] [This subsection highlights the principal design features of the plant and compares its major features with other boiling water reactor facilities. The design of this facility is based on proven technology obtained during the development, design, construction; and operation of boiling water reactors of similar types. The data, performance, characteristics, and other information presented here represent a current, firm design. The comparisons presented here were considered valid at the time the operating license was issued.]

1.3.1.1 Nuclear Steam Supply System Design Characteristics

[HISTORICAL INFORMATION] [Table 1.3-1 summarizes the design and operating characteristics for the nuclear steam supply systems.

Parameters are related to rated power output for a single plant unless otherwise noted.]

1.3.1.2 Power Conversion System Design Characteristics

[HISTORICAL INFORMATION] [Table 1.3-2 compares the power conversion system design characteristics.]

1.3.1.3 Engineered Safety Features Design Characteristics

[HISTORICAL INFORMATION] [Table 1.3-3 compares the engineered safety features design characteristics.]

1.3.1.4 Containment Design Characteristics

[HISTORICAL INFORMATION] [Table 1.3-4 compares the containment design characteristics.]

1.3.1.5 Radioactive Waste Management Systems Design Characteristics

[HISTORICAL INFORMATION] [Table 1.3-5 compares the radioactive waste management design characteristics.]

1.3.1.6 Structural Design Characteristics

[HISTORICAL INFORMATION] [Table 1.3-6 compares the structural design characteristics.]

1.3-1 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.3.1.7 Instrumentation and Electrical Systems Design Characteristics

[HISTORICAL INFORMATION] [Table 1.3-7 compares the electrical systems design characteristics. Table 7.1-2 compares the instrumentation and control systems design characteristics.]

1.3.2 Comparison of Final and Preliminary Information All of the significant changes that have been made in the facility design since submission of the PSAR are listed in Table 1.3-8.

Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR.

1.3-2 Revision 2016-00

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION THERMAL AND HYDRAULIC DESIGN (See Section 4.4)

Rated power, MWt 3833 2436 2436 3579 Design power, MWt 3993 2550 2550 3758 (ECCS design basis)

Steam flow rate, lb/hr 16.419 x 106 10.03 X 106 10.477 X 106 15.396 X 106 Core coolant flow rate, lb/hr 112.5 x 106 78.5 X 106 78.5 X 106 105.0 X 106 1.3-3 Feedwater flow rate, lb/hr 16.379 x 106 10.445 X 106 10.477 x 106 15.358 x 106 System pressure, nominal in 1045 1020 1020 1040 steam dome, psia Average power density, kW/liter 54.1 51.2 50.51 56.0 Maximum thermal output, kW/ft 13.4 13.4 13.4 13.4 Average thermal output, kW/ft 5.92 7.11 5.45 6.04 Maximum heat flux, Btu/hr-ft2 362,000 428,300 354,000 354,300 Average heat flux, Btu/hr-ft2 159,700 159,700 143,900 159,600 Maximum UO2 temperature, F 3430 4380 3325 3337 Revision 2016-00 Average volumetric fuel 1100 1100 1100 1100 temperature, F Average cladding surface 558 558 558 558 temperature, F THERMAL AND HYDRAULIC DESIGN

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 Updated Final Safety Analysis Report (UFSAR) 251-800 218-560 218-560 238-732 GRAND GULF NUCLEAR GENERATING STATION Minimum critical power ratio (MCPR) 1.23 1.9* 1.21 1.24 Coolant enthalpy at core inlet, 527.9 526.2 527.4 527.9 Btu/lb Core maximum exit voids within 76 79 75 76 assemblies Core average exit quality,% steam 14.7 12.9 13.6 14.9 Feedwater temperature, F 417 387.4 420 420 1.3-4 Design Power Peaking Factor (See Section 4.4)

Maximum relative assembly power 1.40 1.40 1.40 1.40 Local peaking factor 1.13 1.24 1.24 1.13 Axial peaking factor 1.40 1.5 1.4 1.40 Total peaking factor 2.26 2.6 2.43 2.22 NUCLEAR DESIGN (First Core)

(See Section 4.3)

Revision 2016-00 Water/UO2 volume ratio (cold) 2.70 2.53 2.41 2.70 Reactivity with strongest control <0.99 <0.99 <0.99 <0.99 rod out, keff

• For Hatch, minimum critical heat flux ratio (MCHFR) was used.

Moderator void coefficient Hot, no voids, k/k - % void -1.0 x 10-3 -1.0 X 10-3 -1.0 X 10-3 -0.3 X 10-5

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION At rated output, k/k - % void -1.6 x 10-3 -1.6 x 10-3 -1.6 x 10-3 -1.0 x 10-5 Fuel temperature doppler coefficient At 68 F, k/k - F fuel -1.3 x 10-5 -1.3 X 10-5 -1.3 X 10-5 -1.6 X 10-5 Hot, no voids, k/k - F fuel -1.2 x 10-5 -1.2 X 10-5 -1.2 X 10-5 -1.3 X 10-5 At rated output, k/k - F fuel -1.3 x 10-5 -1.3 X 10-5 -1.3 X 10-5 -1.2 X 10-5 Initial average U-235 enrichment 1.70 2.23 1.90 1.90 wt. %

1.3-5 Fuel average discharge exposure, 15,000 19,000 15,053 13,000*

MWd/short ton CORE MECHANICAL DESIGN (Initial GGNS Core)

Fuel Assembly (See Section 4.2)

Number of fuel assemblies 800 560 560 732 Fuel rod array 8 x 8 7 X 7 8 X 8 8 X 8 Revision 2016-00

• Average - first core CORE MECHANICAL DESIGN (Continued)

Overall dimensions, in. 176 176 176 176 Weight of UO2 per assembly lb 458 490.4 465.15 472 (pellet type) (chamfered) (undished) (Chamfered) 483.4 (dished)

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Weight of fuel assembly, lb 699 681 698 (including (undished) channel) 675 (dished)

Fuel Rods (See Section 4.2) 62 49 63 63 Number per fuel assembly Outside diameter, in. 0.483 0.563 0.493 0.493 1.3-6 Cladding thickness, in. 0.032 0.032 0.034 0.034 Gap, pellet to cladding, in. 0.0045 0.006 0.0045 0.009 Length of gas plenum, in. 10 16 14 12 Cladding material* Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2

• Free-standing loaded tubes CORE MECHANICAL DESIGN (Continued)

Fuel Pellets Revision 2016-00 (See Section 4.2)

Material UO2 UO2 UO2 UO2 Density, % of theoretical 95 95 95 94 Diameter, in. 0.410 0.487 0.416 0.416 Length, in. 0.410 0.5 0.420 0.420

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 Updated Final Safety Analysis Report (UFSAR) 251-800 218-560 218-560 238-732 GRAND GULF NUCLEAR GENERATING STATION Fuel Channel (See Section 4.2)

Overall dimension, length, in. 166.9 166.9 166.9 166.9 Thickness, in. 0.120 0.080 0.100 0.120 Cross section dimensions, in. 5.46 x 5.46 5.44 X 5.44 5.48 X 5.48 5.52 X 5.52 Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core Assembly 1.3-7 (See Section 4.2)

Fuel weight as UO2, lb. 366,400 272,850 260,538 345,500 Core diameter (equivalent), in. 191.5 160.2 160.2 183.2 Core height (active fuel), in. 150 144 146 148 CORE MECHANICAL DESIGN (Continued)

Reactor Control System (See Chapters 4 and 7)

Method of variation of Movable Movable Movable Movable Revision 2016-00 reactor power control control control control rods and rods and rods and rods and variable variable variable variable forced forced forced forced coolant coolant coolant coolant flow flow flow flow

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Number of movable control rods 193 137 137 177 Shape of movable control rods Cruciform Cruciform Cruciform Cruciform Pitch of movable control rods 12.0 12.0 12.0 12.0 Control material in movable rods B4C B4C B4C B4C granules granules granules granules compacted compacted compacted compacted in SS tubes in SS tubes in SS tubes in SS tubes Type of control rod drives Bottom Bottom Bottom Bottom 1.3-8 entry entry entry entry locking locking locking locking piston piston piston piston CORE MECHANICAL DESIGN (Continued)

Type of temporary reactivity control Burnable Burnable Burnable Burnable for initial core poison; poison; poison; poison; gadolinia- gadolinia- gadolinia- gadolinia-urania fuel urania fuel urania fuel urania fuel Revision 2016-00 rods rods rods rods Incore Neutron Instrumentation (See Chapters 4 and 7)

Number of incore neutron detectors 176 124 124 164 (fixed)

Number of incore detector assemblies 44 31 31 41

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 Updated Final Safety Analysis Report (UFSAR) 251-800 218-560 218-560 238-732 GRAND GULF NUCLEAR GENERATING STATION Number of detectors per assembly 4 4 4 4 Number of flux mapping neutron 5 4 4 5 detectors Range (and number) of detectors Source range monitor Source to Source to Source to Source to 0.001% 0.001% 0.001% 0.001%

power (6) power (4) power (4) power Intermediate range monitor 0.001% to 0.001% to 0.001% to 0.001% to 1.3-9 10% power 10% power 10% power 10% power (8) (8) (8)

CORE MECHANICAL DESIGN (Continued)

Local power range monitor 5% to 125% 5% to 125% 5% to 125% 5% to 125%

power (176) power (124) power (124) power Average power range monitor 2.5% to 2.5% to 2.5% to 2.5% to 125% power 125% power 125% power 125% power*

(8)* (6)* (6)*

Number and type of incore neutron 7 Sb-Be 5 Sb-Be 5 Sb-Be 7 Sb-Se Revision 2016-00 sources REACTOR VESSEL DESIGN

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION 251-800 218-560 218-560 238-732 (See Section 5.3)

Material Low-alloy/ Carbon Carbon Carbon Steel steel/ steel/ steel/

stainless stainless stainless stainless clad clad clad clad Design pressure, psig 1250 1265 1250 1250 Design temperature, F 575 575 575 575 Inside diameter, ft-in. 20-11 18-2 18-2 19-10 1.3-10 Inside height, ft-in. 73-0 69-4 69-4 70-10

• Channels of monitors from LPRM detectors REACTOR VESSEL DESIGN (Continued)

Minimum base metal thickness 6.14 5.53 5.375 5.70 Revision 2016-00 (cylindrical section), in.

Minimum cladding thickness, in 1/8 1/8 1/8 1/8 Reactor Coolant Recirculation Design (See Chapter 5)

Number of recirculation loops 2 2 2 2 Design pressure:

Inlet leg, psig 1250 1148 1250 1250

TABLE 1.3-1: [HISTORICAL INFORMATION] COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 Updated Final Safety Analysis Report (UFSAR) 251-800 218-560 218-560 238-732 GRAND GULF NUCLEAR GENERATING STATION Outlet leg, psig 1625*; 1274 1675*; 1675*;

1525** 1575** 1575**

Design temperature, °F 575 562 575 575 Pipe diameter, in. 24 28 20 22/24 Pipe material, ANSI 304/316 304/316 304/316 304 Recirculation pump flow rate, gpm 44,900 42,200 33,880 35,400 1.3-11 Number of jet pumps in reactor 24 20 20 20

• Pump and discharge piping to and including discharge block valve

• • Discharge piping from discharge block valve to vessel MAIN STEAMLINES (See Section 5.4)

Number of steamlines 4 4 4 4 Revision 2016-00 Design pressure, psig 1250 1146 1250 1250 Design temperature, F 575 563 575 575 Pipe diameter, in. 28 24 24 26 Pipe material Carbon Carbon Carbon Carbon steel steel steel steel

TABLE 1.3-2: [HISTORICAL INFORMATION] COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS GG Bailly Limerick Zimmer Turbine Generator (See Section 10.2)

Net generator output (MW) 1331.5 626 1,092 835.9 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Turbine cycle heat rate (Btu/KW-hr) 9815 9602 10,287 9959 Type/LSB length (line) TC6F-46 TC4F/28 TC6F/38 TC4F/40 Cylinders (No.) 1-HP, 3-LP 1-HP, 2-LP 1-HP, 3-LP 1-HP, 2-LP Steam Conditions at throttle valve Flow (lb/hr) 15.655 x 106 8.29 x 106 14.14 X 106 10.477 x 106 Pressure (psia) 997.86 965 965 965 Temperature (F) 544 510 540 540 Moisture Content (%) 0.66 0.40 0.40 0.40 1.3-12 Turbine cycle arrangement (See Section 10.4)

Steam reheat stages (No.) 2 2 None 2 Feedwater heating stages (No.) 6 6 6 6 Strings of feedwater heaters (No.) 2-HP, 3-LP 2 3 2 Heaters in condenser necks (No.) 4 1 2 1 Heater drain system Pumped forward Pumped forward Cascade Pumped forward Condensate pumps (No.) 3 3 3 3 Condensate booster pumps (No.) 3 3 None 3 Heater drain pumps (No.) 2 2 None 2 Reactor feed pumps (No.) 2 2 3 2 Main steam line Revision 2016-00 Steam lines (No.) 4 4 4 4 Design pressure (psig) 1250 1250 1250 1250 Design temperature (F) 575 575 575 575 Pipe diameter (in.) 28 20 26 24 Pipe material Carbon steel Carbon steel Carbon steel Carbon steel Main steam bypass capacity (%) 35 25 25 25

TABLE 1.3-2: [HISTORICAL INFORMATION] COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS (CONTINUED)

GG Bailly Limerick Zimmer Final feedwater temperature (F) 417 420 420 420 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Condenser (See Section 10.4)

Type Multiple Single Multiple Single pressure pressure pressure pressure Condenser shells (No.) 3 2 3 2 Design pressure (in. Hg abs) 3.62/2.91/2.37 3.2 2.81/3.56/4.67 3.5 Total condenser duty (Btu/hr) 8.506 x 109 4.25 x 109 7.8 x 109 7.053 x 109 Circulating water system (Section 10.4)

Type Closed/ND & Closed/ND Closed/ND Closed/ND 1.3-13 Mech Draft cooling tower cooling tower cooling tower cooling tower Flow (gpm) 572,000 376,000 113,000 (each) 450,000 Circulating water pumps (No.) 2 2 4 3 (1/2 capacity) (1/2 capacity)

Revision 2016-00

TABLE 1.3-3: [HISTORICAL INFORMATION] COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 EMERGENCY CORE COOLING SYSTEMS 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION (Systems sized on design power)

Low Pressure Core Spray Systems (See Section 6.3)

Number of loops 1 2 1 1 Flow rate, gpm 7115 at 4625 at 4725 at 6000 at 128 psid 120 psid 119 psid 122 psid High Pressure Core Spray System 1.3-14 (See Section 6.3)

Number of loops 1 1a 1 1 Flow rate, gpm 1650 at 4250 1330 at 1465 at 1147 psid 1110 psid 1130 psid 7115 at 4725 at 6000 at 200 psid 200 psid 200 psid Revision 2016-00 Automatic Depressurization System (See Section 6.3)

Number of relief valves 8 7 7 8 Low Pressure Coolant Injectionb (See Section 6.3)

TABLE 1.3-3: [HISTORICAL INFORMATION] COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 EMERGENCY CORE COOLING SYSTEMS 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Number of loops 3 2 3 3 Number of pumps 3 4 3 3 Flow rate, gpm/pump 7450 at 7700 at 5050 at 7100 at 24 psid 20 psid 20 psid 20 psid AUXILIARY SYSTEMS 1.3-15 Residual Heat Removal System (See Section 5.4)

Reactor Shutdown cooling Mode:

Number of loops 2 2 2 2 Number of pumps 2 4 2 2 Revision 2016-00 Flow rate, gpm/pumpc 7450 7700 5050 7100 Duty, Btu/hr/heat exchangerd 50 x 106 32 X 106 30.8 X 106 45.0 X 106 Number of heat exchangers 2 2 2 2 Primary containment cooling mode:

TABLE 1.3-3: [HISTORICAL INFORMATION] COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 EMERGENCY CORE COOLING SYSTEMS 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Flow rate, gpm 7450e 30,800 5050e 7100e Standby Service Water System (See Section 9.2)

Flow rate, gpm/heat exchanger 25,300 8000 5000 7000 Number of pumps 3 4 4 5 (2 @ 12,000 1.3-16 gpm)

(1 @ 1,300 gpm)

Reactor Core Isolation Cooling System (See Section 5.4)

Flow rate, gpm 800 at 400 at 400 at 700 at 1120 psid 1120 psid 1120 psid 1120 psid Fuel Pool Cooling and Cleanup System Revision 2016-00 (See Section 9.1)

Capacity, Btu/hr 15.0 x 106 5.7 X 106 6.6 X 106 11.8 X 106 Notes a

High-pressure cooling injection system utilized

TABLE 1.3-3: [HISTORICAL INFORMATION] COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS (CONTINUED)

GRAND GULF HATCH 1 ZIMMER GESSAR BWR 6 BWR 4 BWR 5 BWR 6 EMERGENCY CORE COOLING SYSTEMS 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION b

A mode of the RHR system c

Capacity during reactor flooding mode with more than one pump running d

Heat exchanger duty at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following reactor shutdown e

Flow per heat exchanger 1.3-17 Revision 2016-00

TABLE 1.3-4: [HISTORICAL INFORMATION] COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (See Chapter 3)

Grand Gulf Zimmer Bailly Limerick Type Mark III. Mark II. Over-and- Mark II. Over-and- Mark II. Over-and-Updated Final Safety Analysis Report (UFSAR)

Reinforced under primary under primary under primary GRAND GULF NUCLEAR GENERATING STATION concrete containment, containment, containment, containment, but enclosing drywell enclosing drywell enclosing drywell with pressure and suppression and suppression and suppression suppression. pool. Enclosed by pool. Enclosed by pool. Enclosed by Containment reactor building. reactor building. reactor building.

encloses drywell and suppression pool.

Leak rate (%/day) 0.35 0.5 0.5 0.5 1.3-18 Containment Construction Reinforced Not applicable Not applicable Not applicable concrete cylindrical structure (not prestressed) with hemispherical head; steel lined.

Internal design 185 Not applicable Not applicable Not applicable temperature (F)

Design pressure (psig) 15 Not applicable Not applicable Not applicable Free (air) volume 1.40 x 106 Not applicable Not applicable Not applicable (cu ft) (excluding drywell)

Revision 2016-00 Drywell Construction Reinforced Prestressed Prestressed Prestressed concrete. concrete. Drywell concrete. Drywell concrete. Drywell Basically is frustum of a is frustum of a is frustum of a cylindrical; Flat cone; steel lined. cone; steel lined. cone; steel lined.

concrete roof with a steel refueling head

TABLE 1.3-4: [HISTORICAL INFORMATION] COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS (See Chapter 3) (Continued)

Grand Gulf Zimmer Bailly Limerick Internal design 330 340 340 340 temperature (F)

Updated Final Safety Analysis Report (UFSAR)

Design pressure (psig) 30 +45, -2 +45, -2 +55, -5 GRAND GULF NUCLEAR GENERATING STATION Free (air) volume, 270,000 287,000 263,800 390,450 total (cu ft)

Suppression Pool Construction Reinforced Prestressed Prestressed Prestressed concrete, steel concrete. Pool is concrete. Pool is concrete. Pool is lined. Basically cylindrical; steel cylindrical; steel cylindrical; steel cylindrical. lined. lined. lined.

Internal design 185 340 340 340 temperature (F)

Design pressure (psig) 15 +45, -2 +45, -2 +55, -5 1.3-19 Water volume (cu ft) 136,000 106,000 73,500 122,400 Break area/total vent 0.008 0.008 0.012 0.019 area Revision 2016-00

TABLE 1.3-5: [HISTORICAL INFORMATION] RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS GRAND GULF HATCH 1 ZIMMER BWR 6 BWR 4 BWR 5 GASEOUS RADWASTE 251-800 218-560 218-560 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION (See Section 11.3)

Design Bases, noble gases, 100,000 100,000 100,000 Ci/sec at 30 min at 30 min at 30 min Process treatment Chilled Recombiner Chilled charcoal Ambient Charcoal Charcoal Number of beds 8 12 5 Design condenser 40 40 12.5 In-leakage, cfm 1.3-20 Release point-height 31.5 394 172 above ground, ft (Radwaste Bldg)

LIQUID RADWASTE (See Section 11.2)

Treatment of:

1. Floor drains Filtered, F, D, and R F, E, and R demineralized, evaporated, and returned to Revision 2016-00 condensate storage

2. Equipment drains Filtered, F, D, and R F, D, and R demineralized, evaporated, and returned to condensate storage

• See legend, Sheet 2

TABLE 1.3-5: [HISTORICAL INFORMATION] RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS (Continued)

GRAND GULF HATCH 1 ZIMMER BWR 6 BWR 4 BWR 5 LIQUID RADWASTE 251-800 218-560 218-560 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION (Cont.)

3. Chemical drains Neutralized, F, discharged E, E, D, evaporated, and solid to concentrates to returned to radwaste solid radwaste equipment drain distillate R collector tank

4. Laundry drains NONE Diluted and sent Reverse osmosis (Laundry will be to circulating discharge processed water discharge 1.3-21 offsite by an authorized contractor.)

5. Expected annual avg. release, 110,000 20,000 10,900 Ci (excluding tritium)

• Legend:

D = demineralized F = filtered E = evaporator/concentrator Revision 2016-00 R = recycled, i.e., returned to condensate storage

TABLE 1.3-6: [HISTORICAL INFORMATION] COMPARISON OF STRUCTURAL DESIGN CHARACTERISTICS GRAND GULF HATCH 1ZIMMERGESSAR BWR 6 BWR 4 BWR 5 BWR 6 Seismic Design 251-800 218-560 218-560 238-732 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION (See Sections 3.2 and 3.7)

Operating Basis Earthquake

- horizontal g 0.075 0.08 0.10 0.15

- vertical g 0.05 0.05 0.07 0.15 Safe shutdown earthquake 1.3-22

- horizontal g 0.15 0.15 0.20 0.3

- vertical g 0.10 0.10 0.14 0.3 Wind Design (See Section 3.3)

Maximum sustained - mph 90 105 90 130 Revision 2016-00 Tornados (See Section 3.3)

Translational - mph 60 60 60 70 Tangential - mph 300 300 300 290

TABLE 1.3-7: [HISTORICAL INFORMATION] COMPARISON OF ELECTRICAL SYSTEMS (See Chapter 8)

System Grand Gulf Bailly Zimmer Limerick (1 unit) (1 unit) (2 unit) (2 unit)

Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Number of offsite circuits 3 9 4 4 Number of auxiliary 3 service transformers 3-1 unit auxiliary 2-1 unit auxiliary 4-2 unit auxiliary power sources (1 exclusively for transformers transformers 2 startup esf) 1 reserve auxiliary 2 startup transformers transformer transformers 1 emergency reserve auxiliary transformer Number of preferred power 3 3 (except 2 for HPCS) 2 2 1.3-23 circuits for esf buses Number of esf buses per 3 3 3 4 unit Number of standby a-c 3 (1/esf bus) 3 (1/esf bus) 3 (1/esf bus) 4 (1/2) esf buses) power supplies Number of 125 V d-c 3 (1/esf bus) 3 (1/esf bus) 3 (1/esf bus) 4 (1/2 esf buses) systems supplying buses Revision 2016-00 Sharing of standby power Diesels, batteries supplies and d-c buses shared between None None interconnections between interconnected corresponding buses safety buses of both units Note: esf = engineered safety features

TABLE 1.3-8: SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR ITEM CHANGE REASON FOR CHANGE FSAR SECTION IN WHICH CHANGE IS DISCUSSED Nuclear fuel The number of water rods Improved fuel 4.2.2.3.2 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION in each fuel bundle has performance been changed from 1 to

2. Five different U-235 enrichments are now used in the fuel assemblies instead of previous four types.

Control rod Changed to 11 wire probe Improved reliability 4.2 drive position and solid state and increased frequency indication of checking actual rod 1.3-24 position Feedwater The thermal sleeve was To eliminate failure, 5.3 sparger changed to provide leakage, and provide improved slip fit design for possilbe inservice of sparger to nozzle. inspection.

Standby liquid Interlocks on the SLC To prevent inadvertent 7.4.1.2, control (SLC) system were revised. boron injection during 9.3.5 system system testing.

Revision 2016-00 RCIC system Each component of the Improved testability 5.4 RCIC system has been made capable of functional testing during normal plant operation.

TABLE 1.3-8: SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR (CONTINUED)

ITEM CHANGE REASON FOR CHANGE FSAR SECTION IN WHICH CHANGE IS DISCUSSED Automatic The interlocks on the To meet IEEE-279 7.3.1.1 depressurization automatic requirements Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION system (ADS) depressurization system were revised.

Leak detection system The leak detection To meet IEEE-279 and 7.6.1.4 system was revised to Reg. Guide 1.45 upgrade the capability requirements and incorporate the requirements of IEEE-279 Added additional monitors to increase adequacy of detection 1.3-25 Control rod drive Increased system Provides increased 3.9.4.1, fast scram pressure from 1750 to reactivity 4.6 2000 psi, enlarged control, especially at insert/withdraw draw end of fuel cycle.

lines, and increased Provides increased accumulator volume to thermal margin, and provide faster scram reduces amount of time operation of steam relief Revision 2016-00 Reactor Recirc. Pumps tripped on signals Reduces transient 4.6.4, pump trip from turbine control or core flow and 5.4.1, stop valves upon reactivity. Works with 7.6.1.8 generator load rejection fast scram to provide or turbine trip increased thermal margin

TABLE 1.3-8: SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR (CONTINUED)

ITEM CHANGE REASON FOR CHANGE FSAR SECTION IN WHICH CHANGE IS DISCUSSED Fuel storage racks Added 48 more fuel Increases capacity to 9.1.1, storage castings for use handle more onsite fuel 9.1.2 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION in spent fuel, new fuel storage and containment pool storage areas Fuel Pool Cooling Upgraded calculations to Provides for increased 9.1.3 verify pool cooling fuel capacity system able to handle increased fuel storage Fuel Pool Cooling Upgraded system (except To meet Reg. Guide 1.13 9.1.3 1.3-26 for filter/demineralizer which can be isolated) to meet Seismic 1 classification High Pressure Changed motor control Design improvement 7.3 Core Spray System center capacity to handle increased electrical loads Reactor Protection Changes for control Provides improved 7.2.2.2 Revision 2016-00 System system instrument test testability reliability ability. Changed from switches to transmitters and added calibration units

TABLE 1.3-8: SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR (CONTINUED)

ITEM CHANGE REASON FOR CHANGE FSAR SECTION IN WHICH CHANGE IS DISCUSSED Gauged Control Changed logic and Improves operating time 3.9.4.1, Rod Withdrawal control rod drive for control maneuvering 4.6 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION hydraulic system to move and startup groups of control rods.

Added stabilizing hydraulic valves Reactor In-Core Changed replacement from Improves time for 7.6.1.5 Monitors top to bottom of core replacement during monitor entry outages Reactor Recirc. Added vibration sensors Improves reliability Ch 5 Pump to record and alarm when high shaft vibration 1.3-27 encountered on pump or motor Reactor Recirc. Added Motor-Generator Provides improved 7.7.1.3 Pump Motor Sets to provide control operation Controls for reduced flow during startup and shutdown Reactor Recirc. Removed pump bypass Design Improvement 5.4.3 System lines for reduction of region potentially sensitive to stainless Revision 2016-00 steel stress corrosion problems Feedwater Added system to plant To eliminate through- 6.7.2 Leakage Control line bypass leakage System

TABLE 1.3-8: SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR (CONTINUED)

ITEM CHANGE REASON FOR CHANGE FSAR SECTION IN WHICH CHANGE IS DISCUSSED Switchyard Changed configuration Safety 8.2.2 Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION of 500 Kv switchyard to Evaluation Report (SER) provide two 500 Kv by NRC offsite sources Guard Pipe Design criteria To comply with BTPs MEB 3.6 Assemblies 3-1 and APCSB 3-1 Pipe Break Design criteria NRC requirement/design 3.6 1.3-28 Criteria improvement ISI ISI criteria ASME, Code,Section XI 5.2.4, requirements 6.6 Suppression Pool Added system To reduce doses inside 9.3.6 Clean-up System containment Tornado Missile Spectrum Changed spectrum To comply with SRP 3.5.1.4 3.5.1.4 Draft Rev 1 Containment Leakage Raised from 0.1%/day to Improved meteorological 6.2.6 Revision 2016-00 0.35%/day data (see Section 2.3),

also, taking credit for Iodine removal via containment sprays (see Section 6.5)

Radial Wells - Added radial wells, To improve water 9.2.10 Intake Structure deleted intake quality of cooling structure water systems

TABLE 1.3-8: SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR (CONTINUED)

ITEM CHANGE REASON FOR CHANGE FSAR SECTION IN WHICH CHANGE IS DISCUSSED Service Water Rerouted discharge line Discharge line Ch. 2 Discharge Line into the barge slip originally routed to Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION the middle of the Mississippi River; changed to reduce hazards to shipping Auxiliary Building Removed double Double valve isolation 6.2.3 Isolation Valve isolation valves on not necessary for SGTS Arrangement smaller piping operation based on single failure analysis 1.3-29 Control Room Increased allowable Improved dose 6.4 Inleakage inleakage from 60 scfm calculations to 263 scfm Iodine Removal Via Iodine removal credit Improved dose 6.5 Containment Spray accounted for in dose calculations calculations Turbine Building Deleted Improved dose 9.4, Revision 2016-00 Ventilation Charcoal calculations show 11.3 Filters releases are within Appendix I requirements without these filters

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1 GGNS Project

[HISTORICAL INFORMATION] [The Grand Gulf Nuclear Station (GGNS) is owned or leased by System Energy Resources, Inc. (SERI) and South Mississippi Electric Power Association (SMEPA). GGNS is operated by Entergy Operations, Inc. (Entergy Operations). SERI and Entergy Operations are wholly owned subsidiaries of Entergy Corporation, formerly Middle South Utilities, Inc. SERI provides financing for construction and maintains title ownership of the facility. Entergy Operations assumes responsibility for design, construction, and operation of the facility.

MP&L, (Now Entergy Mississippi, Inc.), Middle South Energy, Inc.

(now System Energy Resources, Inc.) and SMEPA were co-applicants in the licensing proceedings for GGNS Unit 1; Entergy Operations, SERI, Entergy Mississippi, Inc., and SMEPA are co-licensees.

Entergy Operations, SERI, MP&L and SMEPA were co-applicants in the licensing proceedings for GGNS Unit 2 prior to cancellation of the Unit 2 construction permit.

During construction of GGNS Unit 1, MP&L did not maintain engineering and construction staffs but used reputable engineering and construction firms for these purposes. For the work covered by this FSAR, Bechtel Corporation was retained to provide engineering, procurement, quality assurance, and construction management services. The engineering firms and consultants used during construction of GGNS Unit 1 are given in the following subsections to Section 1.4. The current GGNS engineering staff is provided in Section 13.1.]

1.4.2 Architect Engineer

[HISTORICAL INFORMATION] [Bechtel Corporation has been continuously engaged in construction or engineering since 1898.

For the last 35 years, Bechtel has been active in the fields of pipelines, petroleum, power generation and distribution, harbor development, mining and metallurgy, and chemical and industrial processing. The Bechtel organization has grown progressively to become one of the world's largest engineer-constructors for industrial facilities. Since the close of World War II, Bechtel Corporation has been responsible for the design of over 200 thermal power-generating units; this represents more than 115,000,000 kilowatts of new generating capacity, of which more than 65,000,000 kilowatts are nuclear. Bechtel Corporation is 1.4-1 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) qualified to provide and does provide required services for station design, construction management, equipment procurement, construction, and startup.]

1.4.3 Nuclear Steam Supply System

[HISTORICAL INFORMATION] [The General Electric Company was awarded the contracts to design, fabricate, and deliver the single-cycle, boiling water nuclear steam supply system, to fabricate the first core of nuclear fuel, and to provide technical direction for the installation and startup of this equipment.

General Electric has engaged in the development, design, construction, and operation of boiling water reactors (BWR) since 1955. Thus, General Electric has substantial experience, knowledge, and capability to design, manufacture, and furnish technical assistance for the installation and startup of the reactors. See Table 1.4-1 for a list of nuclear facilities utilizing GE designed reactors.]

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.4.4 Turbine Generator Vendor

[HISTORICAL INFORMATION] [Allis-Chalmers Power Systems, Inc. (A-CPSI) has supplied the turbine generators and provided technical assistance for installation and startup of this equipment. The A-CPSI is a jointly owned company of Allis-Chalmers Corporation (A-CC) and Kraftwerk Union AG (KWU) of West Germany and employs the extensive experience and capabilities of both of its parent firms in the steam turbine generator and nuclear power field. The technology and design of A-CPSI turbine generators is provided by KWU under a license agreement. In the past 25 years, KWU and its parent firms, Siemens and AEG of Germany (at the beginning of 1977, Siemens purchased the AEG share of KWU and thus became the sole owner), have designed and built nearly 600 steam turbine generator units for fossil-fueled and nuclear power plants. At the present time, KWU and A-CPSI have, either in service or on order, a total of 40 nuclear turbine generators rated 350 Mw or larger for BWR and PWR applications. The design of the turbine generator for Grand Gulf has been based directly on similar KWU units in service or being manufactured at the present time. KWU and A-CC also have related experience in the design and construction of BWR and PWR reactors and complete turnkey nuclear power plants.

The Grand Gulf turbine generator was manufactured partly by A-CC and partly by KWU, and certain components (such as heat exchangers, pumps, motors, and prefabricated piping) have been procured directly by A-CPSI. Manufacturing by A-CC has been conducted primarily at its facilities in West Allis, Wisconsin, where many turbine generators and other equipment have been built. KWU has manufactured its portion at its Muelheim/Ruhr facilities, where turbines and generators are built; and in Erlangen, where electrical control equipment is designed and built.

Technical assistance for installation and startup has been provided by the A-CPSI Product Service Department, which is staffed with personnel trained and experienced in this work.]

1.4.5 Consultants

[HISTORICAL INFORMATION] [Woodward-Clyde Consultants has been retained to assist in evaluating the potential impact that the construction and operation of the nuclear facility has on the 1.4-3 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) environment. They provide environmental consulting services in the areas of meteorology, demography, hydrology, biology, and radiological surveys.

Memphis State University provides, through its Center for Nuclear Studies, instructions which include a training program in Basic Reactor Fundamentals for employees of Entergy Operations as part of their qualifications for licensing as nuclear power plant operating and/or maintenance personnel.

Southern Nuclear Engineering, Inc. was retained to provide engineering consultant services in the areas of:

a. Plant site review

b. Writing of licensing documents

c. Review of licensing documents and Environmental Reports

d. Review of plant component and system designs

e. Design of special equipment and systems

f. Performing or checking calculations required in the design and/or licensing of the nuclear plant(s)

g. Presentation of expert witness testimony or technical information at licensing meetings or hearings and at Public Hearings

h. Design and/or operation of meteorological and other environmental stations Engineering Data Systems, Inc. (EDS) has been retained to provide consulting engineering services as required by GGNS. These services are typically required in the following areas:

a. Review and assistance in the development of a QA program

b. Review of equipment specifications for QA/QC requirements

c. Review of equipment specifications for seismic requirements

d. Technical assistance in review of nuclear plant systems 1.4-4 Revision 2016-00

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EDS shall provide such other consulting engineering services as may be required by GGNS.

Eberline Instrument Corporation (EIC) provides consultation services for radiation exposure control related programs as requested by authorized personnel of Entergy Operations. Such consultation includes but is not necessarily limited to the following activities:

a. Assist with the development of an operating philosophy

b. Assist with the preparation of technical specifications dealing with in-plant exposure control or radioactivity released to the environment. Assist with the preparation of amendments to these technical specifications that will permit maximum flexibility consistent with the Entergy Operations operating philosophy and NRC requirements.

c. Review facility and equipment design to identify potential exposure control problems and suggest modifications that would help limit radiation exposures to as low as reasonably achievable (ALARA).

d. Assist in development of radiation protection training programs.

e. Provide backup radiological control personnel for non-routine activities.

f. Help specify instrumentation for radiation exposure control and effluent documentation.

EIC provides other such consulting engineering services as required by GGNS.

General Electric Co. (GEH), I&SE Division, provides consultation and nondestructive testing services in connection with the inservice inspection of the GGNS. This includes:

a. Performance of the inservice inspection.

b. Engineering consulting to assure that the inspection can be performed.

c. Data analysis and appropriate recommendations.

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

d. Preparation of a criterion used in the design of Class I and Class II systems.

e. Review and comment on drawings and specifications of pipes, welds, hangers, access provisions, insulation, shielding, etc.

f. Preparation of input to documents required for licensing.

g. Assistance to Entergy Operations, Inc. with the Nuclear Regulatory Commission in the area of inservice inspection.

h. Prepare or assist in preparation of bid specifications or requirements for inservice inspection.

i. Keep Entergy Operations updated on Code changes and new codes affecting GGNS and inservice inspection requirements.

Nuclear Services Corporation (NSC) provides engineering services as required by GGNS in the following typical areas:

a. Assist with the development of an operating philosophy

b. Assist with the preparation of technical specifications

c. Assist with preparation of procedures for routine operation, maintenance, inspection, and refueling activities and non-routine or emergency plans Betz, Calgon, or other contract chemical service companies as required provide consultation services for water chemistry and other chemical analysis as GGNS deems necessary.]

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TABLE 1.4-1: [HISTORICAL INFORMATION] COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION OR IN DESIGN BY GENERAL ELECTRIC Rating Year of Year of Station Utility (MWe) Order Startup Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Dresden 1 Commonwealth Edison 200 1955 1960 Humboldt Bay Pacific G&E 69 1958 1963 Kahl Germany 15 1958 1961 Garigliano Italy 150 1959 1964 Big Rock Point Consumers Power 70 1959 1963 JPDR Japan 11 1960 1963 KRB Germany 237 1962 1967 Tarapur 1 India 190 1962 1969 Tarapur 2 India 190 1962 1969 1.4-7 GKN Holland 52 1963 1968 Oyster Creek JCP&L 640 1963 1969 Nine Mile Point 1 Niagara Mohawk 625 1963 1970 Dresden 2 Commonwealth Edison 809 1965 1970 Pilgrim Boston Edison 644 1965 1972 Millstone 1 NUSCO 642 1965 1971 Tsuruga Japan 340 1965 1970 Nuclenor Spain 440 1965 1971 Fukushima 1 Japan 439 1966 1971 BKW KKM Switzerland 306 1966 1972 Revision 2016-00 Dresden 3 Commonwealth Edison 809 1966 1971 Monticello Northern States 545 1966 1971 Quad Cities 1 Commonwealth Edison 800 1966 1972 Browns Ferry 1 TVA 1,098 1966 1974 Browns Ferry 2 TVA 1,098 1966 1974 Quad Cities 2 Commonwealth Edison 800 1966 1972

TABLE 1.4-1: [HISTORICAL INFORMATION] COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION OR IN DESIGN BY GENERAL ELECTRIC (CONTINUED)

Rating Year of Year of Station Utility (MWe) Order Startup Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Vermont Yankee Vermont Yankee 514 1966 1972 Peach Bottom 2 Philadelphia Electric 1,065 1966 1974 Peach Bottom 3 Philadelphia Electric 1,065 1966 1974 Fitzpatrick PASNY 821 1966 1975 Bailly NIPSCO 660 1967 1977 Shoreham LILCO 819 1967 1978 Cooper Nebraska PPD 778 1967 1974 Browns Ferry 3 TVA 1,098 1967 1975 Limerick 1 Philadelphia Electric 1,098 1967 1981 Hatch 1 Georgia 786 1967 1975 1.4-8 Fukushima 2 Japan 762 1967 1974 Brunswick 1 Carolina P&L 821 1968 1976 Brunswick 2 Carolina P&L 821 1968 1975 Arnold Iowa ELP 569 1968 1974 Fermi 2 Detroit Edison 1,123 1968 1979 Limerick 2 Philadelphia Electric 1,065 1969 1982 Hope Creek 1 PSE&G 1,067 1969 1981 Hope Creek 2 PSE&G 1,067 1969 1983 Zimmer 1 CCDPP 810 1969 1978 Revision 2016-00 Chinshan 1 Taiwan 610 1969 1977 Caorso 1 Italy 827 1969 1975 Hatch 2 Georgia Power Company 795 1970 1978 La Salle 1 Commonwealth Edison 1,078 1970 1978 La Salle 2 Commonwealth Edison 1,078 1970 1979 Susquehanna 1 Pennsylvania P&L 1,052 1970 1980

TABLE 1.4-1: [HISTORICAL INFORMATION] COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION OR IN DESIGN BY GENERAL ELECTRIC (CONTINUED)

Rating Year of Year of Station Utility (MWe) Order Startup Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Susquehanna 2 Pennsylvania P&L 1,052 1970 1982 Chinshan 2 Taiwan 610 1970 1978 WPPSS 2 WPPSS 1,103 1971 1977 Nine Mile Point 2 Niagara Mohawk 1,080 1971 1979 Grand Gulf 1 Entergy Operations, Inc. 1,290 1971 1980 Kaiseraugst Switzerland 915 1971 1978 Fukushima 6 Japan 1,135 1971 1976 Tokai 2 Japan 1,135 1971 1976 Riverbend 1 Gulf States 934 1972 1980 Riverbend 2 Gulf States 934 1972 1981 1.4-9 Perry 1 Cleveland Electric 1,205 1972 1979 Perry 2 Cleveland Electric 1,205 1972 1980 Douglas Point 1 PEPCO 1,178 1972 1985 Douglas Point 2 PEPCO 1,178 1972 1987 Hartsville 1 TVA 1,228 1972 1980 Hartsville 2 TVA 1,228 1972 1981 Hartsville 3 TVA 1,228 1972 1981 Hartsville 4 TVA 1,228 1972 1982 Laguna Verde 1 Mexico 660 1972 1977 Leibstadt Switzerland 940 1972 1978 Kuosheng 1 Taiwan 992 1972 1978 Revision 2016-00 Kuosheng 2 Taiwan 992 1972 1979 Clinton 1 Illinois Power 955 1973 1981 Clinton 2 Illinois Power 955 1973 1984 Montague 1 NUSCO 1,220 1973 1982

TABLE 1.4-1: [HISTORICAL INFORMATION] COMMERCIAL NUCLEAR REACTORS COMPLETED, UNDER CONSTRUCTION OR IN DESIGN BY GENERAL ELECTRIC (CONTINUED)

Rating Year of Year of Station Utility (MWe) Order Startup Updated Final Safety Analysis Report (UFSAR) GRAND GULF NUCLEAR GENERATING STATION Allens Creek 1 Houston L&P 1,150 1973 1980 Allens Creek 2 Houston L&P 1,150 1973 1982 Skagit 1 Puget SD 1,290 1973 1981 Skagit 2 Puget SD 1,290 1973 1983 Blackfox 1 Oklahoma 950 1973 1983 Blackfox 2 Oklahoma 950 1973 1985 Laguna Verde 2 Mexico 660 1973 1978 Enel 6 Italy 982 1974 1980 1.4-10 Enel 8 Italy 982 1974 1980 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION 1.5.1 Current Development Programs 1.5.1.1 Instrumentation for Vibration

[HISTORICAL INFORMATION] [Vibration testing for reactor internals has been performed on virtually all GE-BWR plants. At the time of issue of NRC Regulatory Guide 1.20, test programs for compliance were instituted. The first BWR 6 plant of each size will be considered a prototype design and will be instrumented and subjected to both cold and hot, two-phase flow testing to demonstrate that flow-induced vibrations similar to those expected during operation will not cause damage. Subsequent plants which have internals similar to those of the prototypes will be tested in compliance to the requirements of Regulatory Guide 1.20 to confirm the adequacy of the design with respect to vibration.]

1.5.1.2 Core Spray Distribution

[HISTORICAL INFORMATION] [Due to slight changes in core dimensions and core spray sparger geometry, the core spray flow distribution header has been tested to assure that each fuel assembly in the reactor core would receive adequate cooling water in the event of a LOCA. These tests are regarded as confirmatory only since the basic spray header design has been successfully tested over a wide range of similar geometrical conditions.

The tests demonstrate that each fuel assembly receives adequate cooling water flow for any spray system flow rate between the rated flow and the runout flow condition.

GEH has completed development of a core spray methodology, consisting of single nozzle tests in steam, computer calculations, and multiple nozzle tests in air, to calculate minimum bundle flow. Application of the methodology for Grand Gulf shows a minimum calculated bundle flow of 3.1 gpm. This compares to a minimum required bundle flow of approximately 1 gpm as described in the questions and answers to NEDO-10846, April 1973. Additional descriptions of the tests and computer codes may be found in NEDO-20566, Amendment 3, April 1977; NRC letter, Review of General Electric Topical Report, NEDO-20566, Amendment 3, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K-Effect of Steam Environment of BWR Core Spray Distribution, June 13, 1978; and 1.5-1 Revision 2016-00

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NRC letter from P. S. Check to R. L. Tedesco, Evaluation of NEDO-24712, Core Spray Distribution Methodology Confirmation Tests, September 8, 1980.]

1.5.1.3 Core Spray and Core Flooding Heat Transfer Effectiveness

[HISTORICAL INFORMATION] [Due to the incorporation of an 8 x 8 fuel rod array with unheated water rods, tests have been conducted to demonstrate the effectiveness of ECCS in the new geometry.

These tests are regarded as confirmatory only, since the geometry change is very slight and the water rods provide an additional heat sink in the inside of the bundle which improves heat transfer effectiveness.

There are two distinct programs involving the core spray. Testing of the core spray distribution has been accomplished, and the Licensing Topical Report NEDO-10846, BWR Core Spray Distribution, April, 1973, has been submitted. The other program concerns the testing of core spray and core flooding heat transfer effectiveness. The results of testing with stainless steel cladding were reported in the Licensing Topical Report NEDO-10801, Modeling the BWR/6 Loss-of-Coolant Accident: Core Spray and Bottom Flooding Heat Transfer Effectiveness, March, 1973.

The results of testing using Zircaloy cladding were reported in the Licensing Topical Report, NEDO-20231, Emergency Core Cooling Tests of an Internally Pressurized, Zircaloy Clad, 8x8 Simulated BWR Fuel Bundle, December, 1973.]

1.5.1.4 Verification of Pressure Suppression Design

[HISTORICAL INFORMATION] [The General Electric Company has conducted a large scale test program to verify the performance characteristics of the Mark III containment. Large scale testing was started in November 1973 following completion of a two-year small scale test program.

The large scale test program utilizes a facility which represents a segment of a Mark III containment. The original character of the programs was to be a confirmatory exercise to verify the short term analytical model. The scope of the total program included testing beyond design basis conditions to investigate the margins available in pressure suppression systems. As a result of this testing, GE proposed a new analytical model to evaluate the Mark 1.5-2 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

III design. This model is entitled The General Electric Mark III Pressure Suppression Containment System Analytical Model, and is described in NEDO-20533.

During early tests it was observed that containment structures could be subject to significant suppression pool hydrodynamic loads during blowdown. This resulted in several additional test series whose objective was to generate design basis loads to be incorporated in the design of the affected containment structures.

Eleven large scale test series have been completed to date. The primary objective of three series of these tests was to verify short-term analytical models for horizontal vents (and centerline submergences). The objectives of two others were to obtain scoping data regarding pool dynamic response and impact loads on structures located above the suppression pool. Other tests were designed to measure froth impingement loads on the Hydraulic Control Unit floor and to determine pool swell motion characteristics, to measure pool impact loads on representative containment structures, and to determine pool motion characteristics for large air mass fraction vent flows and to compare these scale results to the previous full scale air tests.

Additional tests will be conducted to indicate comparability of liquid blowdown to steam blowdowns and to investigate pool stratification and vent chugging effects.

Tests will be performed with the suppression pool at an initial elevated temperature to determine steam condensation characteristics under such conditions. A multi-vent series will be run to consider possible vent interactions. In plant testing of the safety relief valves to verify that the design basis safety relief valve discharge loads inside the suppression pool are adequately conservative will be performed on Grand Gulf Unit 1 prior to full power operation.]

1.5.1.5 Critical Heat Flux Testing

[HISTORICAL INFORMATION] [A program for Critical Heat Flux testing was established and was to be similar to that described in the report APED-5286, Design Basis for Critical Heat Flux Condition in Boiling Water Reactors, September 1966. Since that time, however, a new analysis has been performed and the GETAB program initiated. The results of that analysis and related testing is described in the approved Licensing Topical Report, 1.5-3 Revision 2016-00

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NEDO-10958-A, General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, January 1977.

These results and correlation are applicable to GEH Initial core fuel for cycle 1.

A similar program has been established by Exxon Nuclear Company (ENC), Inc. as part of the process of developing the capability to license fuel for nuclear reactor reloads. The result of this program is the XN-3 (Revision 1) critical heat flux correlation which is used to predict the onset of transition boiling. This effort is described in the approved licensing topical report XN-NF-512(P)(A) Revision 1 and XN-NF-512(P)(A) Revision 1 Supplement 1, XN-3 Critical Power Correlation, Exxon Nuclear Co., October 1982.]

1.5.1.6 Structural Testing

[HISTORICAL INFORMATION] [Although tests are being conducted to determine the effects of vibration on fuel assembly spacers and to determine the forces to which the assemblies are subjected during shipment, there is no special program at present concentrating on structural testing, and no topical report is anticipated.]

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GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.6 MATERIAL INCORPORATED BY REFERENCE Table 1.6-1 is a list of GE and Bechtel topical reports and any other reports of documents which are incorporated in whole or in part by reference in this FSAR and has been filed with the NRC.

Additional documents which are referenced in this FSAR are listed at the end of the sections in which they have been referenced.

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TABLE 1.6-1: REFERENCED REPORTS Referenced Report in FSAR Number Title Section A.

APED-4827 Maximum Two-Phase Vessel Blowdown 6.2 from Pipes (April 1965)

APED-4986 Consequences of Operating Zircaloy-2 4.2 Clad Fuel Rods Above the Critical Heat Flux (October 1965)

APED-5286 Design Basis for Critical Heat Flux 1.5 Condition in BWRs (September 1966)

APED-5458 Effectiveness of Core Standby Cooling 5.4 Systems for General Electric Boiling Water Reactors (March 1968)

APED-5460 Design and Performance of General 3.9 Electric BWR Jet Pumps (July 1968)

APED-5555 Impact Testing on Collet Assembly for 4.6 Control Rod Drive Mechanism 7RDB144A (November 1967)

APED-5640 Xenon Considerations in Design of 4.1, 4.3 Boiling Water Reactors (June 1968)

APED-5652 Stability and Dynamic Performance of 4.1 the General Electric Boiling Water Reactor APED-5706 In-Core Neutron Monitoring System for 7.6.1.5, General Electric Boiling Water 7.7.1.7, Reactors (November 1968, Revised 7.6.2.5 April 1969) 1.6-2 Revision 2016-00

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TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section APED-5736 Guidelines for Determining Safe Test 6.3 Intervals and Repair Times for Engineered Safeguards (April 1969)

APED-5750 Design and Performance of General 5.4 Electric Boiling Water Reactor Main Steam Line Isolation Valves (March 1969)

APED-5756 Analytical Methods for Evaluating the 15.4 Radiological Aspects of the General Electric Boiling Water Reactor (March 1969)

GEAP-10546 Theory Report for Creep-Plast 4.1 Computer Program (January 1972)

GEAP-13112 Thermal Response and Cladding 4.2 Performance of an Internally Pressurized, Zircaloy-Clad, Simulated BWR Bundle Cooled by Spray Under Loss-of-Coolant Conditions (April 1971)

NEDC-33477P Safety Analysis Report for Grand Gulf Referenced Nuclear Station Constant Pressure in Chapters Power Uprate (March 2012 as corrected 1 through June 2016) 12 & 15 NEDE-10313 PDA-Pipe Dynamic Analysis Program for 3.6 Pipe Rupture Movement (Proprietary Filing)

NEDE-11146 Design Basis for New Gas System (July 11.3 1971) (Company Proprietary)

NEDE-20386 Fuel Channel Deflections 4.2 1.6-3 LBDCR 2016-178

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TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section NEDE-21156 Supplemental Information for Plant 4.4 Modification to Eliminate Significant In-Core Vibration (January 1976)

NEDE-21175-P Fuel Assembly Evaluation of Combined 3.9 BWR/6 Safe Shutdown Earthquake (SSE) and Loss-of-Coolant Accident (LOCA)

Loadings (November 1976)

NEDE-21354-P PWR Fuel Channel Mechanical Design 3.9 and Deflection (September 1976)

NEDE-24196 Basis for BWR 6 8x8 Fuel Thermal 4.4, 4.3 (Proprietary) Analysis Application, General Electric Information Report NEDE-23014 HEX 01 User's Manual (July 1976) 15.2 NEDM-10735 Densification Considerations in BWR 4.2 Fuel Design and Performance (December 1972)

NEDO-10173 Current State of Knowledge, High 4.2, 11.1 Performance BWR Zircaloy-Clad U02 Fuel (May 1970)

NEDO-10174 Consequences of a Postulated Fuel 4.2 Blockage Incident in a Boiling Water Reactor (May 1970)

NEDO-10299 Core Flow Distribution in a Modern 4.4 Boiling Water Reactor as Measured in Monticello (January 1971)

NEDO-10320 The General Electric Pressure 6.2 Suppression Containment Analytical Model (April 1971) Supplement (May 1971) 1.6-4 Revision 2016-00

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TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section NEDO-10329 Loss-of-Coolant Accident and 4.3 Emergency Core Cooling Models for General Electric Boiling Water Reactors (April 1971) Supplement 1 (April 1971) Addenda (May 1971)

NEDO-10349 Analysis of Anticipated 15.8 Transients Without Scram (March 1971)

NEDO-10466-A Power Generation Control Complex 7.1.2.2 (February 1979) and Addendum l (December 1979)

NEDO-10505 Experience with BWR Fuel Through 4.2, 11.1 September 1971 (May 1972)

NEDO-10527 Rod Drop Accident Analysis for 4.3, 15.4 Large Boiling Water Reactors (March 1972) Supplement l (July 1972) Supplement 2 (January 1973 NEDO-10585 Behavior of Iodine in Reactor 15.6 Water During Plant Shutdown and Startup (August 1972)

NEDO-10602 Testing of Improved Jet Pumps for 3.9 the BWR/6 Nuclear System (June 1972)

NEDO-10734 A General Justification for 11.3 Classification of Effluent Treatment System Equipment as Group D (February 1973)

NEDO-10739 Methods for Calculating Safe Test 6.3 Intervals and Allowable Repair Times for Engineered Safeguard Systems (January 1973) 1.6-5 Revision 2016-00

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TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section NEDO-10751 Experimental and Operational 11.3 Confirmation of Offgas System Design Parameters (January 1973)

(Company Proprietary)

NEDO-10801 Modeling the BWR/6 Loss-of- 1.5 Coolant Accident: Core Spray and Bottom Flooding Heat Transfer Effectiveness (March 1973)

NEDO-10802 Analytical Methods of Plant 4.4, 5.2, Transient Evaluations for General 15.1 Electric Boiling Water Reactor (February 1973)

NEDO-10846 BWR Core Spray Distribution 1.5 (April 1973)

NEDO-10899 Chloride Control in BWR Coolants 5.2 (June 1973)

NEDO-10905 High Pressure Core Spray Power 8.3.1.2 Supply Unit NEDO-10958 General Electric BWR Thermal 4.3, 4.4, Analysis Basis (GETAB): Data, 15.0 Correlation, and Design Application (November 1973)

NEDO-10958-A General Electric BWR Thermal 1.5, 15.4, Analysis Basis (GETAB): Data, 16.1 Correlation, and Design Application (January 1977)

NEDO-10959 General Electric BWR Thermal 15.0 Analysis Basis (GETAB): Data, Correlation, and Design Application (November 1973) 1.6-6 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section NEDO-10977 Drywell Integrity Study: 6.3 Investigation of Potential Cracking in BWR/6 Mark III Containment NEDO-20231 Emergency Core Cooling Tests of an 1.5 Internally Pressurized, Zircaloy-Clad, 8x8 Simulated BWR Fuel Bundle (December 1973)

NEDO-20340 Process Computer Performance 4.3 Evaluation Accuracy (June 1974)

NEDO-20360 General Electric Boiling Water 4.2, 15.4 Reactor Generic Reload Application for 8x8 Fuel (May 1975)

NEDO-20360-IP General Electric Boiling Water 4.2 Reactor Generic Reload Application for 8x8 Fuel (March 1976)

NEDO-20533 The General Electric Mark III 1.5 Pressure Suppression Containment System Analytical Model (June 1974)

NEDO-20566 General Electric Company Model for 3.9, 4.3, Loss-of-Coolant Accident Analysis 6.3, 1.5 in Accordance with 10 CFR 50, Appendix K (January 1976)

NEDO-20605 Creep Collapse Analysis of BWR 4.2 and NEDO-20606 Fuel Using Safe Collapse Model (August 1974)

NEDO-20626 Studies of BWR Designs for 15.8 Mitigation of Anticipated Transients without Scrams (October 1974) 1.6-7 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section NEDO-20626-1 Studies of BWR Designs for 15.8 Mitigation of Anticipated Transients without Scrams (June 1975)

NEDO-20626-2 Studies of BWR Designs for 15.8 Mitigation of Anticipated Transients without Scrams (July 1975)

NEDO-20913 Lattice Physics Methods (June 1975) 4.3 NEDO-20922 Experience with BWR Fuel Through 4.2, 11.1 September 1974 (June 1975)

NEDO-20939 Lattice Physics Methods 4.3 Verification (August 1975)

NEDO-20943 Urania-Gadolinia Nuclear Fuel 4.2 Physical and Material Properties (January 1977)

NEDO-20944 BWR/4 and BWR/5 Fuel Design 4.1, 4.3 (October 1976)

NEDO-20946 BWR Simulator Methods Verification 4.3 (May 1976)

NEDO-20948-P Fuel Design (June 1976) 4.2 BWR/6 NEDO-20953 Three-Dimensional Boiling Water 15.4 Reactor Core Simulator (May 1976) 1.6-8 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section NEDO-20964 Generation of Void and Doppler 4.3 Reactivity Feedback for Application to BWR Plant Transient Analysis (August 1975)

NEDO-21142 Realistic Accident Analysis for 15.4, 15.6, General Electric Boiling Water 15.7 Reactor - The RELAC Code and User's Guide (September 1978)

NEDO-21143 Conservative Radiological 15.4, 15.6, Accident Evaluation - The 15.7 CO/NAC01 Code (March 1976)

NEDO-21159 Airborne Release from BWRs for 11.1 Environment Impact Evaluations (March 1976)

NEDO-21174 BWR Fuel Channel Deflections 4.2 NEDO-21231 Banked Position Withdrawal 4.3 Sequence (September 1976)

NEDO-21291 Group Notch Mode of the RSCS for 15.4 Cooper (June 1976)

NEDO-21708 Radiation Effects in BWR Pressure 5.3 Vessel Steels NEDO-21985 Functional Capability Criteria 3.9 for Essential Mark II Piping (September 1978)

NEDO-24083 Recirculation Pump Shaft Seal 5.5 Leakage Analysis (November 1978)

NEDO-24142 Fast Scram Control Rod Drive 4.6 1.6-9 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section NEDO-24154 Qualification of the One- 5A3.1.1 Dimensional Core Transient Model for BWR, NEDO-24154, October 1978 NEDO-24708A Additional Information for NRC 18.1.29 Staff Generic Report on BWRs, Volumes 1 and 2 (December 1980)

NEDO-24712 Core Spray Design Methodology 1.5 Confirmation Tests (August 1979)

Qualification Program (October 1978)

NEDO-26453 3D BWR Core Simulator (May 1976) 4.3 Oyster Creek Station, FSAR 1.5 Amendment 10 "Summary Memorandum on Excursion 4.3, 15.0 Analysis Uncertainties," Dresden Nuclear Power Station, Unit 3, Plant Design Analysis Report Amendment 3 Hatch Nuclear Plant, Unit 1, PSAR 15.5; Amendment 10, Appendix L; and 7.6.1.5 Amendment 7.

Millstone Nuclear Power Station, 6.3 PSAR Amendment 14 Pilgrim Nuclear Power Station, 6.3 PSAR Amendment 14 Quad Cities Station, Units 1 and 4.3 2, PSAR Amendment 9 1.6-10 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section 22A4365 Interim Containment Loads Report 3.8 (ICLR), Mark III Containment, Appendix Revision 2 B. Other Referenced Reports AE-RTL-788 Void Measurements in the Region 4.4 of Subcooled and Low Quality Boiling (April 1966)

ANL-5621 Boiling Density in Vertical 4.4 Rectangular Multichannel Sections with Natural Circulation (November 1956)

ANL-6385 Power-to-Void Transfer Functions 4.4 (July 1961)

AGN-TM-407 AGN-GAM, and IBM 7090 Code to 4.3 Calculate Spectra and Multigroup Constants (April 1965)

ANL-7460 Reactor Development Program 4.3 Progress Report, p. 121-122 (June 1968)

ANL-7527 Reactor Development Program 4.3 Progress Report, p. 132 (December 1968)

BNL-5826 THERMOS-A Thermalization 4.3 Transport Code for Reactor Design (June 1961)

BNWL-340 "Computer Code Abstracts, 4.3 Computer Code-HRG," Reactor Physics Dept., Technical Activities Quarterly Report, July, August, September, 1966 (October 15, 1966) 1.6-11 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section BHR/DER 70-1 Radiological Surveillance 11.1 Studies at a Boiling Water Nuclear Power Reactor (March 1970)

BHR/DER 70-1 Radiological Surveillance 11.1 Studies at a Boiling Water Nuclear Power Reactor (March 1970)

BMI-1163 Vapor Formation and Behavior 4.4 in Boiling Heat Transfer (February 1957)

CF 59-6-47 Removal of Fission Product Gases 11.3 (ORNL) from Reactor Off-Gas Streams by Adsorption (June 11, 1959)

HCOG-GGNS-004 Hydrogen Control Final 6.2.5.3.1 Analysis Report (10 CFR 50.44), Rev. 0 (October 28, 1993)

IDO-ITR-105 The Response of Waterlogged 4.2 U02 Fuel Rods to Power Bursts (April 1969)

IN-ITR-111 The Effects of Cladding Material 4.2 and Heat Treatment on the Response of Waterlogged U02 Fuel Rods to Power Bursts (January 1970)

STl-372-38 Kinetic Studies of 4.4 Heterogeneous Water Reactors (April 1966)

TID-4500 Relap 3 - A Computer Program for 3.6 Reactor Blowdown Analysis IN-1321 (June 1970)

WACP-6065 Melting Point of Irradiated 4.2 Uranium Dioxide (February 1965) 1.6-12 LBDCR 2019-011

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section WAPD-BT-19 A Method of Predicting Steady- 4.4 State Boiling Vapor Fractions in Reactor Coolant Channels (June 1960)

WAPD-TM-283 Effects of High Burnup on 4.2 Zircaloy-Clad, Bulk U02 Plate Fuel Element Samples (September 1962)

WAPD-TM-416 WIGLE - A Program for the Solution 4.3 of the Two-Group Space-Time Diffusion Equations in Slab Geometry (1964)

WAPD-TM-629 Irradiation Behavior of Zircaloy- 4.2 Clad Fuel Rods Containing Dished End U02 Pellets (July 1967)

C. Bechtel Corporation Reports BN-TOP-1 "Testing Criteria for Integrated 3.8.3, Leak Rate Testing of Primary 3.8.1 Containment Structures for Nuclear Power Plants," Revision 1, November 1972 BN-TOP-2 "Design for Pipe Rupture Effects," 3.6, Revision 2, May 1974 3.8 BN-TOP-4 "Subcompartment Pressure Analysis" 6.2.1.2 Revision 0, July 1976, and Appendix 3E Revision 1 BC-TOP-1 "Containment Building Liner Plate 3.8.1, Design Report," December 1972, 3.8.3 Rev. 1 1.6-13 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.6-1: REFERENCED REPORTS (CONTINUED)

Referenced Report in FSAR Number Title Section BC-TOP-3A "Tornado and Extreme Wind Design 3.3, 3.8 Criteria for Nuclear Power Plants,"

Revision 3, August 1974 BC-TOP-4 "Seismic Analysis of Structures and 3.7, 3.8 Equipment for Nuclear Power Plants," Revision 1, September 1972, including Addendum 1 dated April, 1973.

BC-TOP-5A "Prestressed Concrete Nuclear 3.8.1 Reactor Containment Structures,"

Rev. 3, February 1975 BC-TOP-9A "Design of Structures for Missile 3.5.3 Impact," Rev. 2, September 1974 1.6-14 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.7 ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS Table 1.7-1 contains a list of non-proprietary electrical, instrumentation, and control drawings which are incorporated in the FSAR by reference. This table lists those drawings which are considered to be necessary to evaluate the safety-related features in Chapter 7 and 8. These tables will be updated in future amendments as necessary.

1.7-1 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J0200 Sh 1 Logic Symbols J0204 Sh 8 P64 Fire Protection Sys Aux Bldg Isln Valves Unit 1 J0240 Sh 0 Z51 Control Room HVAC Sys J-0251 Sh 96 G17 Floor Drain Filter Outlet Valve J-0251 Sh 96 G17 Equipment Drain Filter Outlet Valve J-0251 Sh 98 G17 Equipment Drain Floor Drain Filter Bypass Valves J0300 Sh 0 A21 Loop Diagram Legend Index J0340 Sh 0 Z51 Control Room HVAC Sys Index J0400 Control Room Panel Location J0401 Upper Cable Spreading Room Panel Location J0402 Lower Cable Spreading Room Panel Location J0419 Control Room Vent VB SH13 P855 J1202 Sh 0 P21 Makeup Water Treatment Sys 1.7-2 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J1203 Sh 0 P66 Domestic Water Sys J1216 Sh 7 P11 Cond & Refueling Water Transfer & Stg Sys Isln Valves J1216 Sh 11 P11 Cond & Refueling Water Transfer Sys Aux Bldg Isln Valves J1216 Sh 12 P11 Cond & Refueling Water Transfer & Stg Sys Isln Valves J1221 Sh 0 P41 Standby Service Water Sys Index J1222 Sh 6 P44 Plant Service Water Sys Aux Bldg Isln Valves J1222 Sh 9 P44 Standby Service Water to Plant Service Water Crosstie Valves J1222 Sh 16 P44 Plant Service Water Instr

& Svc Air Cprsrs Cut-out Valve J1223 Sh 11 SP43 Turbine Bldg Cool Wtr Service Air Compsr Iso Valves J1224 Sh 0 P42 Component Cooling Water System J1225 Sh 4 P71 Plant Chilled Water Isln Valves 1.7-3 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J1225 Sh 5 P71 Plant Chilled Water Sys Aux Bldg Isln Valves J1226 Sh 3 P52 Service Air Isln Control Air-Operated Valves J1226 Sh 4 P52 Service Air Sys Aux Bldg Isln Valves J1226 Sh 5 P52 Service Air Isln Control Motor-Operated Valve J1228 Sh 1 C11 CRD Pump Suction Aux Bldg Isln Valve J1231 Sh 1 M41 Containment Cooling Sys Containment Isln Valves J1231 Sh 2 M41 Containment Cooling Sys Drywell Isln Valves J1231 Sh 14 M41 Containment Cooling Sys Aux Bldg Isln Valves J1231 Sh 16 M41 Containment Cooling Sys Containment Isln Values J1233 Sh 6 T41 Aux Bldg Vent Sys Isln Valves J1234 Sh 1 T42 Fuel Handling Area Vent Sys Isln Valves J1235 Sh 0 T51 Emergency Pump Room Vent Sys Index 1.7-4 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J1236 Sh 0 T48 Standby Gas Treatment Sys Index J1237 Sh 0 E61 Combustible Gas Control Rooms Ventilation System Index J1241 Sh 0 X77 Diesel Generator Rooms Ventilation System Index J1250 Sh 4 E31 Leak Detection Trip Unit Fault Monitor Alarms J1254 Sh 0 P75 Standby Diesel Generator System Index J1255 Sh 6 P72 Drywell Chilled Water Sys Isln Valves J1256 Sh 1 P45 Floor & Equipment Drain Sys Isln Valves J1256 Sh 2 P45 Floor & Equipment drain Sys Aux Bldg Isln Valves J1256 Sh 30 P45 Drywell Chemical Waste Isln Valves J1256 Sh 44 P45 Containment Isolation Valves J1258 Sh 0 Y47 Standby Service Water Pump House Vent System Index 1.7-5 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J1259 Sh 0 Z77 Safeguard Switchgear and Btry Rooms Vent System Index J1260 Sh 0 M71 Containment Drywell & Aux Bldg Instm and Control J1261 Sh 0 P81 HPCS Diesel Generator System Index J1262 Sh 6 P53 Instrument Air Isln Control Air Operated Valve J1262 Sh 8 P53 Instrument Air Sys Aux Bldg Isln Valves J1262 Sh 11 P53 Instrument Air Isln Control Motor Operated Valve J1267 Sh 0 T46 Engineered Safety Features Elec Switchgear Rooms Cooling System Index J1271 Sh 0 E12 Residual Heat Removal System J1272 Sh 2 G46 FPCC Filter-Demin Sys Backwash Aux Bldg Isln Valve J1277 Sh 1 G36 RWCU Backwash Rcvg Tank Containment Isln Valves J1277 Sh 2 G36 RWCU Backwash Rcvg Tank Aux Bldg Isln Valves 1.7-6 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J1279 Sh 0 E30 Suppression Pool Makeup System Index J1281 Sh 0 B21 Nuclear Boiler System Index J1284 Sh 1 G33 Reactor Water Cleanup Aux Bldg Isln Valves J1293 Sh 3 E38 Block Flow/Ctmt Isln Valves A & B J1297 Sh 2 P60 Suppression Pool Cleanup Sys Containment & Aux Bldg Isln Valves J1298 Sh 0 E38 Feedwater Leakage Control System J1321 Sh 0 P41 Standby Service Water System Index J1322 Sh 0 P44 Plant Service Water System J1324 Sh 0 P42 Component Cooling Water System Index J1328 Sh 0 C11 CRD Hydraulic System Index J1336 Sh 0 T48 Standby Gas Treatment System Index J1337 Sh 0 E61 Combustible Gas Control System Index 1.7-7 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J1341 Sh 0 X77 Diesel Generator Bldg Vent System Index J1350 Sh 0 E31 Leak Detection Sys Index J1354 Sh 0 P75 Standby Diesel Generator System Index J1358 Sh 0 Y47 Standby Service Water Pump House Vent System Index J1359 Sh 0 Z77 Safeguard Switchgear and Battery Rooms Vent Sys Index J1360 Sh 0 M71 Containment Drywell and Aux Bldg Instrumentation and Control Index J1361 Sh 0 P81 HPCS Diesel Generator System Index J1367 Sh 0 T46 Engineered Safety Features Electrical Switchgear Rooms Cooling System Index J1368 Sh 0 C71 Reactor Protection System Index J1369 Sh 0 C61 Remote Shutdown System Index J1374 Sh 0 D21 Area Radiation Monitoring System J1375 Sh 0 D23 Drywell Monitoring System 1.7-8 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J1379 Sh 0 E30 Suppression Pool Level J1398 Sh 0 E38 Feedwater Leakage Control System J0400 Control Room Panel Location J0401 Upper Cable Spreading Room Panel Location J0402 Lower Cable Spreading Room Panel Location J1414 Diesel Generator BB 1H13-P864 J1416 Div IV Engineered Safety Features Logic VB 1H13-P878 J1417 Div I Engineered Safety Features Logic VB 1H13-P871 J1418 Div II Engineered Safety Features Logic VB 1H13-P872 J1431 Div III Engineered Safety Features Logic VB 1H13-P877 J1487A thru D Remote Shutdown Control Panel 1H22-P150 J1488A & B Remote Shutdown Control Panel 1H22-P151 1.7-9 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 1, Instrumentation/Control (Bechtel)

Drawing No.

(FSAR Figure No.) Title J0501 Instrument Location - Control Bldg El 93-0, 113-0, 133-0, and 148-0 J0502 Instrument Location - Control Bldg El 166-0, 175-0, and 189-0 J1502 Instrument Location Turbine Bldg El 113-0 J1503 Instrument Location Turbine Bldg El 133-0 J1504 Instrument Location Turbine Bldg El 166-0 J1505 Instrument Location Aux and Cntmt Bldg El 93-0 and 100-9 J1506 Instrument Location Aux and Cntmt Bldg El 119-0 and 114-6 J1507 Instrument Location Aux and Cntmt Bldg El 135-4, 139-0 and 147-7 J1508 Instrument Location Aux and Cntmt Bldg El 161-10 and 166-0 J1509 Instrument Location Aux and Cntmt Bldg El 184-6 and 185-0 1.7-10 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E0001 Main One Line Diagram (8.1-1)

E0002 Phasing Diagram E0004 Phasing Diagram E0010 Synchronizing Diagram Engineered Safety Features Buses 15AA, 16AB, 17AC, 25AA, 26AB, 27AC E0013 One Line Meter and Relay Diag Aux Elec Dist Sys and Boiler Bus 19UD E0014 One Line Meter and Relay Diag Aux Elec Dist Sys and Boiler Bus 29UE E0021 Ground Detection Sch for 3 Wire Ungrounded DC Sys E0022 Ground Detection Sch for 2 Wire Ungrounded DC Sys E0028 Three Line Meter and Relay Diagram Engineered Safety Features Transformers E0030-000 General Symbols, Notes and Shs A-71 Details E0032 One Line Meter and Relay Diagram (8.3-7a) 120/240V AC Uninterruptible Power Supplies 1.7-11 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E0111-000 4.16kV Switchgear Typical Shs 0-2 Circuit Breaker Internal Details E0116-000 480V Switchgear Typical Circuit Shs 0-2 Breaker Internal Details E0131-000 Control Room HVAC Sys Shs 0-30 E0231 Sh 0 Fire Protection System Index E0231 Sh A Fire Protection System Relay Tabulation E0231 Sh 14 Fire Protection System E0231 Sh 19 Fire Protection System Aux Bldg Isln Valve F282A E0231 Sh 22 Fire Protection System BOP Computer Points E0232-000 Domestic Water System Index and Shs 0-9 Relay Tabulation E0627 Lighting and Comm Plan Control Bldg El 148-0 E0628 Lighting and Comm Plan Control Bldg El 166-0 E0630 Lighting and Comm Plan Control Bldg El 177 1.7-12 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E0637 Lighting and Comm Plan Control Bldg El 111-0 E0638 Lighting and Comm Plan Control Bldg El 133-0 E0648 Control and Aux Bldgs Lighting System E-0649A Public Address System, Sound (9.5-009B) Power Telephone & Warning Light Diagram El 93'-0" & El 103'-0" E-0649B Public Address System, Sound (9.5-009C) Power Telephone & Warning Light Diagram El 133'-0", 113'-0",

118'-0" & Partial Plan 111'-0" E-0649C Public Address System, Sound (9.5-009D) Power Telephone & Warning Light Diagram El 133'-0", 136'-0",

139'-0" & Partial Plan El 148'-

0" E-0649D Public Address System, Sound (9.5-009E) Power Telephone & Warning Light Diagram El 166'-0", 161'-0" &

Partial Plan El 166'-0" & 177'-

0" E-0649E Public Address System, Sound (9.5-009F) Power Telephone & Warning Light Diagram El 185'-0", 189'-0" &

Partial Plan El 208'-10" 1.7-13 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E-0649F Public Address System, Sound (9.5-009G) Power Telephone & Warning Light Diagram Site Plan E-0649G Public Address System (9.5-009H) Administration Building E-0649H Public Address System, M&E (9.5-009I) Building E-0660 Site Raceway Plan E-0663 Enlarged Site Raceway Plan E-0672 Enlarged Site Raceway Plan E-0674 Enlarged Site Raceway Plan E-0688 Raceway Plan Control Bldg El 111-0 Area 25A E-0689 Raceway Plan Control Bldg El 133-0 Area 25A E-0690 Raceway Plan Control Bldg El 148-0 Area 25A E-0691 Raceway Plan Control Bldg El 166-0 Area 25A E-0692 Raceway Plan Control Bldg El 189-0 Area 25A E-0693 Raceway Plan Control Bldg El 177-0 Area 25A 1.7-14 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E-0694 Raceway Plan Control Bldg Area 25A Ceiling El 93-0 E-0695 Raceway Sections and Details Control Bldg Area 25A E-0700 Raceway Plan Control Bldg El 93 Ceiling Area 25B E-0701 Raceway Plan Control Bldg El 111-0 Area 25B E-0701A Raceway Plan Control Bldg El 111-0 Area 25B E-0702 Raceway Plan Control Bldg El 133-0 Area 25B (Note: Abandon in place per EC81639)

E-0703 Raceway Plan Control Bldg El 148-0 Area 25B E-0703A Raceway Plan Control Bldg El 148-0 Area 25B E-0704 Raceway Plan Control Bldg El 166-0 Area 25B E-0705 Raceway Plan Control Bldg El 189-0 Area 25B E-0705A Raceway Plan Control Bldg El 1890 Area 25B 1.7-15 Revision 2019-077 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E-0706 Raceway Plan at Ceiling Control Bldg El 177-0 Area 25B E-0716 Raceway Plan Control Bldg Sections and Details E-0724 Raceway Sections and Details Control Bldg Area 25B E-0725-000 Raceway Notes, Symbols and Shs A-51 Details (except Shs 15, 16, 17, 36, 41, 42, 43, 45)

E-0950 Raceway Plan Control Bldg El 93-0, 111-0, 133-0, 148-0 Fire

& Smoke Detection System Units 1 & 2 E-0951 Raceway Plan Control Bldg El 166-0, 177-0, 189-0 Fire &

Smoke Detection System Units 1 & 2 E-0961 Raceway Plan Radwaste Bldg El 93-0 Fire & Smoke Detection System Units 1 & 2 E-0962 Raceway Plan Radwaste Bldg El 118-0 Fire & Smoke Detection System Units 1 & 2 E-0963 Raceway Plan Radwaste Bldg El 136-0 Fire & Smoke Detection System Units 1 & 2 1.7-16 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E-0964 Raceway Plan Misc. Bldgs Fire &

Smoke Detection System Units 1 & 2 E-0965 Raceway Plan Water Treatment Bldg El 133-0 and Stdby Wtr Pump HS Basin A & B Fire &

Smoke Detection System Units 1 & 2 E-1004 One Line Meter and Relay Diag 6.9kV BOP Buses 11 HD and 12 HE, Unit 1 E-1008 One Line Meter and Relay Diag (8.3-1) 4.16kV ESF System E-1009 One Line Meter and Relay Diag 4.16kV ESF System E-1017 One Line Meter and Relay Diag 480V Buses 15BA1, 15BA2, 15BA3, 15BA4 E1018 One Line Meter and Relay Diag 480V Buses 16BB1, 16BB2, 16BB3, 16BB4 E1019 One Line Meter and Relay Diag 480V Buses 15BA5, 16BB5 E1020 One Line Meter and Relay Diag 480V Buses 15BA6 and 16BB6 E1022 One Line Meter and Relay (8.3-10B) Diagram 125V Bus 1.7-17 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1023 One Line Meter and Relay Diag (8.3-10) 125V DC Buses 11DA, 11DB, and 11DC E1024 One Line Meter and Relay Diag (8.3-7) 120/240V AC Uninterruptible Power Supplies E1026 One Line Meter and Relay Diag (8.3-7b) 120 V AC ESF Uninterruptible Power Supplies E1032-000 208 - 120V AC Engineered Safety Shs 0-15 Features Power Panels E1034-000 120V AC Power Supplies to Cont Shs 0-9 and Instr Panels E1036-000 125V DC Power Supplies to Cont Shs 0-4 and Instr Panels E1027 One Line Meter and Relay Diag (8.3-10a) 125 V DC Buses 11DK and 11DL E1039 Logic Diagram Load Shedding and (8.3-9) Sequencing Panel E1042 Diesel Logic Diagram Engineered (8.3-8) Safety Features Div I E1043 Diesel Logic Diagram Engineered Safety Features Div II 1.7-18 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1053 Three Line Meter and Relay Diagram ESF Div I E1054 Three Line Meter and Relay Diagram ESF Div II E1057 Shs 1,2 480V ESF MCC 15B41 Aux Bldg E1058 Shs 1,2 480V ESF MCC 16B21 Aux Bldg E1059 480V ESF MCC 17B11 Cont Bldg E1081 Shs 1,2 480V ESF MCC 15B11 Aux Bldg E1082 Shs 1-2 480V ESF MCC 15B31 Aux Bldg E1083 Shs 1-2 480V ESF MCC 15B21 Aux Bldg E1084 Sh 1 480V ESF MCC 15B61 Cntrl Bldg E1085 480V ESF MCC 15B51 Standby Service Water Pump House E1086 Shs 1-3 480V ESF MCC 16B11 Aux Bldg E1087 Shs 1-2 480V ESF MCC 16B31 Aux Bldg E1088 Shs 1-2 480V ESF MCC 26B41 Aux Bldg E1089 480V ESF MCC 16B61 Containment Bldg 1.7-19 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1090 480V ESF MCC 16B51 Standby Service Water Pump House E1091 480V ESF MCC 17B01 Cont Bldg E1098 Shs 1-2 480V ESF MCC 16B42 Aux Bldg E1099 Shs 1-2 480V ESF MCC 15B42 Aux Bldg E1100 Shs 1-2 Motor Control Cabinet Tabulation Index E1109-000 4.16kV Engineered Safety Shs 0-27 Features System (except Shs 4, 9, 8, 10, 11, 13, 14, 15, 16, 19)

E1110-000 Standby Diesel Generator System Shs 0-23 Division I (except Shs 2, 5, 6, 7)

E1111-000 Standby Diesel Generator System Shs 0-23 Division II (except Shs 2, 5, 6, 7)

E1112-000 HPCS Diesel Generator Fuel Oil Shs 0-5 Transfer (except Sh 2)

E1115-000 480V Load Center ESF Division I Shs 0-13 1.7-20 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1116-000 480V Load Center ESF Division Shs 0-13 II E1117-000 125V DC ESF Distribution System Shs 0-4 E1118-000 125V Battery Chargers Shs 0-1 E1120-000 Load Shedding and Sequencing Shs 0-7 Tables E1155-000 Feedwater Leakage Control Shs 0-4 System E1159-000 Nuclear Boiler System Shs 0-3 (except Sh 1)

E1167-000 Control Rod Drive System Unit 1 Shs 0-3 E1174 Schematic Diagram C71 RPS MG (8.3-14) Set Control System E1180-000 Residual Heat Removal System Shs 0-5 E1186-000 Combustible Gas Control System Shs 0-45 (except Shs 10, 21,26,28,35,36,39) 1.7-21 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1203-000 Reactor Water Cleanup System Index and Relay Tabulation E1205-000 Filter/Demineralizer Sys Index and Aux Rly Tab E1205 Sh 1 Filter/Demineralizer Sys Backwash Rcvg Tk Containment Isln Valve F101 E1205 Sh 2 Filter/Demineralizer Sys Backwash Rcvg Tank Aux E1205 Sh 6 Filter/Demineralizer Sys 120V AC Fuse Panel Power Supplies E1205 Sh 8 Filter/Demineralizer Sys Backwash Rcvg Tank Containment Bldg Isln Valve F106 E1208-000 Fuel Pool Cooling and Cleanup Index E1208 Sh A Fuel Pool Cooling and Cleanup Filter/Demin Sys Relay Tabulation E1208 Sh 1 Fuel Pool Cooling and Cleanup Filter/Demin Sys Backwash Aux Bldg Isln Valve F253 1.7-22 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1208 Sh 8 Fuel Pool Cooling and Cleanup Filter/Demin Sys 120V AC Power Supply E1213 Sh 0 Containment Cooling Sys Index E1213 Sh 1 Containment Cooling Sys Relay Tabulation E1213 Sh 2 Containment Cooling Sys Containment Isolation Valve F011 E1213 Sh 3 Containment Cooling Sys Containment Isolation Valve F012 E1213 Sh 4 Containment Cooling Sys Drywell Isolation Valve F015 E1213 Sh 5 Containment Cooling Sys Drywell Isolation Valve F013 E1213 Sh 27 Containment Cooling Sys 125V DC and 120V AC Fuse Panel Power Supply E1213 Sh 28 Containment Cooling Sys 120V AC Fuse Panel Power Supply E1213 Sh 29 Containment Cooling Sys Control Room Ann E1213 Sh 30 Containment Cooling Sys Control Room Ann 1.7-23 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1213 Sh 32 Containment Cooling Sys Computer Points E1215 Sh 10 P72 Drywell Chilled Water System Isolation Control F121-A E1215 Sh 9 P72 Drywell Chilled Water System Isolation MOV F123-B E1219-000 Containment Drywell and Aux Shs 0-24 Bldg Instrumentation and (except Shs 15, 16) Control E1220-000 Suppression Pool Makeup Sys and Shs 0-13 Aux Relay Tabulation (except Shs 5, 6)

E1221 Sh 0 Condensate and Refueling Water Storage and Transfer Index Unit 1

E1221 Sh 10 Condensate and Refueling Water Storage and Transfer Aux Bldg Isln Valve F062 E1221 Sh 11 Condensate and Refueling Water Storage and Transfer Aux Bldg Isln Valve F064 E1221 Sh 13 Condensate and Refueling Water Storage and Transfer 120V AC &

125V DC Fuse Panel Power Supplies 1.7-24 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1221 Sh 17 Condensate and Refueling Water Storage and Transfer E1221 Sh 18 Condensate and Refueling Water Storage and Transfer E1222 Sh 0 Makeup Water Treatment Sys Index & Relay Tabulation E1222 Sh 1 Makeup Water Treatment Sys Isln Motor Operated Valve F018-B E1222 Sh 2 Makeup Water Treatment Sys Isln Motor Operated Valve F017-A E1222 Sh 3 Makeup Water Treatment Sys Aux Bldg Isln Valve F024 E1222 Sh 4 Makeup Water Treatment Sys 120V AC Power Supplies E1222 Sh 5 Makeup Water Treatment Sys Computer Points E1225-000 Standby Service Water Sys Shs 0-56 (except Shs 12, 38, 47)

E1226 Sh 0 Component Cooling Water System Index E1228 Sh 0 Plant Service Water Sys Index 1.7-25 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1228 Sh A Plant Service Water Sys Relay Tabulation E1228 Sh 7 Plant Service Water Sys SSW Crosstie Motor Operated Valve F067 E1228 Sh 8 Plant Service Water Sys SSW to Plant Service Water Crosstie Motor-Operated Valve F054 E1228 Sh 9 Standby Service Water System Aux Bldg Outboard Valve F068 E1228 Sh 17 Plant Service Water Sys Power Distribution E1228 Sh 21 Plant Service Water Sys Aux Bldg Isln Valve F121 E1228 Sh 22 Plant Service Water Sys Computer Points E1228 Sh 23 Plant Service Water Sys Computer Points E1229 Sh 0 Instrument Air System Index E1229 Sh A Instrument Air System Relay Tabulation E1229 Sh 6 Instrument Air System 120V AC Fuse Panel Power Supplies 1.7-26 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1229 Sh 9 Instrument Air System Aux Bldg Isln Valve F026A E1229 Sh 10 Instrument Air System Isln Control Motor Operated Valve F003A E1229 Sh 11 Instrument Air System Isln Control Motor Operated Valve F005B E1229 Sh 13 Instrument Air System Computer Points E1229 Sh 14 Instrument Air System Computer Points E1234 Sh 0 Plant Chilled Water System Index E1234 Sh A Plant Chilled Water Sys Relay Tabulation E1234 Sh 3 Plant Chilled Water Sys Isln Valve F148 E1234 Sh 4 Plant Chilled Water Sys Isln Valve F149 E1234 Sh 5 Plant Chilled Water Sys Aux Bldg Isln Valve F306 1.7-27 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1234 Sh 13 Plant Chilled Water Sys 120V AC

& 125V DC Fuse Panel Power Supplies E1234 Sh 14 Plant Chilled Water Sys 120V AC Fuse Panel Power Supplies E1234 Sh 16 Plant Chilled Water Sys Computer Points E1239 Sh 0 Service Air System Index E1239 Sh A Service Air System Relay Tabulation E1239 Sh 3 Service Air System Isln Valve F105 E1239 Sh 4 Service Air System Isln Valve F221A E1239 Sh 5 Service Air System Isln Valve F195B E1239 Sh 7 Service Air System 120V AC Fuse Panel Power Supplies E1239 Sh 9 Service Air System Computer Points E1240 Sh 0 Suppression Pool Cleanup System 1.7-28 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1253 Sh 0 Aux Bldg Ventilation System Index E1253 Sh 6 Aux Bldg Ventilation Sys Aux Bldg Vent Sys Isln Valve F007 E1253 Sh 11 Aux Bldg Ventilation Sys ESF 120V AC Fuse Panel Power Supplies E1254 Sh 0 Fuel Handling Area Vent System Index E1254 Sh A Fuel Handling Area Vent Sys Relay Tabulation E1254 Sh 1 Fuel Handling Area Vent Sys Isln Valve F004 E1254 Sh 2 Fuel Handling Area Vent Sys Isln Valve F011 E1254 Sh 3 Fuel Handling Area Vent Sys Isln Valve F019 E1254 Sh 26 Fuel Handling Area Vent Sys 120V AC & 125V DC Fuse Panel Power Supplies E1257-000 Standby Gas Treatment Sys Shs 0-25 (except Sh 9) 1.7-29 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1258-000 Emergency Pump Room Vent Sys Shs 0-5 Relay Tabulation E1265-000 Diesel Generator Room Vent Sys Shs 0-10 except Sh 4)

E1266-000 Standby Service Water Pump Shs 0-12 House Vent Sys (except Sh 5)

E1267 Sh 12 Safeguard Switchgear and Battery Rooms Vent Sys and Relays E1269-000 Engineered Safety Features Shs 0-2 Electrical Switchgear Room Cooling Sys E1271 Sh 0 Floor and Equipment Drains Sys E1271 Sh A Floor and Equipment Drains Sys Relay Tabulation E1271 Sh 13 Floor and Equipment Drains Sys Fl & Eqpt Dr Isln Valve F004 E1271 Sh 14 Floor and Equipment Drains Sys Fl & Eqpt Dr Isln Valve F099 E1271 Sh 15 Floor and Equipment Drains Aux Bldg Isln Valve F158 1.7-30 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1271 Sh 16 Floor and Equipment Drains Sys Drywell Chem Waste Isln Motor Operated Valve F096A E1271 Sh 17 Floor and Equipment Drains Sys 120V AC Power Supplies E1271 Sh 18 Floor and Equipment Drains Sys Computer Points E1283 Control Room PGCC Isolators, Digital and Analog E1284 Control Room PGCC Isolators, Digital and Analog E1285 Control Room PGCC Isolators, Digital and Analog E1286-000 Local Isolators Shs 0-6 E-1358-1F Appendix R Alternate Shutdown Engraving -

1H22-P295 E-1358-1G Appendix R Alternate Shutdown Engraving -

1H22-P296 E-1358-1J Appendix R Alternate Shutdown Engraving -

1H22-P298 1.7-31 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E-1358-1K Appendix R Alternate Shutdown Engraving -

1H22-P299 E1625 Lighting and Communication Plan Aux & Containment Bldg El 119-0 and 120-10 E1626 Lighting and Communication Plan Aux and Containment Bldg El 139-0 & 145-4 E1627 Lighting and Communication Plan Aux and Containment Bldg El 161-10, 116-0 and 170-0 E1672 Raceway Plan Aux Bldg El 93-0 Area 7 E1673 Raceway Plan Aux Bldg El 93-0 Area 8 E1675 Raceway Plan Aux Bldg El 93-0 Area 10 E1676 Raceway Plan Aux Bldg El 119-0 Area 7 E1677 Raceway Plan Aux Bldg El 119-0 Area 8 E1678 Raceway Plan Aux Bldg El 119-0 Area 9 1.7-32 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1679 Raceway Plan Aux Bldg El 119-0 Area 10 E1680 Raceway Plan Aux Bldg El 139-0 Area 7 E1681 Raceway Plan Aux Bldg El 139-0 Area 8 E1682 Raceway Plan Aux Bldg El 139-0 Area 9 E1683 Raceway Plan Aux Bldg El 139-0 Area 10 E1684 Raceway Plan Aux Bldg El 166-0 and 170-0 Area 7 E1685 Raceway Plan Aux Bldg El 166-0 and 170-0 Area 8 E1686 Raceway Plan Aux Bldg El 166-0 and 167-6 Area 9 E1687 Raceway Plan Aux Bldg El 166-0 and 167-6 Area 10 E1688 Raceway Plan Aux Bldg El 185-0 Area 9 E1689 Raceway Plan Aux Bldg El 185-0 Area 10 1.7-33 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1690 Raceway Plan Aux Bldg El 208-10 Area 9 E1691 Raceway Plan Aux Bldg El 208-10 Area 10 E1692 Raceway Plan Aux Bldg El 245-0 and 228-0 Area 9 E1693 Aux Bldg Vertical Cable Tray Chase E1694 Raceway Aux Bldg Misc Sect and Details E1695 Raceway Plan Aux Bldg Misc Sect and Details E1700 Raceway Plan Containment Bldg El 93-0 and 100-9 Area 11 E1701 Raceway Plan Containment Bldg El 114-6 and 120-10 Area 11 E1702A Raceway Plan Containment Bldg El 135-4 Azimuth 0 to 90 Area 11 E1702B Raceway Plan Containment Bldg El 135-4 AZ 90 to 180 Area 11 E1702C Raceway Plan Containment Bldg El 135-4 AZ 180 to 270 Area 11 1.7-34 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1702D Raceway Plan Containment Bldg El 135-4 AZ 270 to 0 Area 11 E1702F Raceway Plan Hydrogen Igniter System E1703 Raceway Plan Containment Bldg El 161-10 and 170-0 Area 11 E1704 Raceway Plan Containment Bldg El 184-6 Area 11 E1705 Raceway Plan Containment Bldg El 208-10 Area 11 E1706 Raceway Containment Bldg Misc Sect and Details E1707 Raceway Containment Bldg Misc Sect and Details E1708 Raceway Plan Containment Bldg Developed View of Drywell Wall Inside Drywell AZ 90 to 270 E1709 Raceway Plan Containment Bldg RPIS Channel Under RPV El 114-6 Area 11 E1710 Raceway Plan Containment Bldg RPS Channel Under RPV El 113-6 Area 11 1.7-35 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E1711 Raceway Details Containment Bldg Cable and Channel Under RPV Area 11 E-1712 Raceway Containment Bldg Developed View of Drywell Wall Inside Drywell AZ 270 to 90 E-1713 Raceway Containment Bldg Misc Sects and Details E-1714 Raceway Plan Diesel Generator Bldg Area 12 El 133-0 E-1715 Raceway Plan Diesel Generator Bldg Area 12 El 158-0 E-1716 Raceway Plans Cooling Towers (SSW) No. 1 & 2 E-1719 Raceway Plan Diesel Generator Bldg Misc Sects & Details E-1805 Raceway Plan Turbine Bldg El 93-0 Fire & Smoke Detection System Unit 1 E-1806 Raceway Plan Turbine Bldg El 113-0 Fire & Smoke Detection System Unit 1 1.7-36 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E-1807 Raceway Plan Turbine Bldg El 133-0 Fire & Smoke Detection System Unit 1 E-1808 Raceway Plan Turbine Bldg El 166-0 Fire & Smoke Detection System Unit 1 E-1809 Raceway Plan Aux Bldg & Cntmt El 93-0, 100-9 Fire & Smoke Detection System Unit 1 E-1800 Raceway Plan Aux Bldg & Cntmt El 119-0, 120-10, 114-6 Fire &

Smoke Detection System Unit 1 E-1801 Raceway Plan Aux Bldg & Cntmt El 139-0, 135-4, 147-7 Fire &

Smoke Detection System Unit 1 E-1802 Raceway Plan Aux Bldg & Cntmt El 161-10, 166-0 Fire & Smoke Detection System Unit 1 E-1803 Raceway Plan Aux Bldg & Cntmt El 184-6, 185-0 Fire & Smoke Detection System Unit 1 E-1804 Raceway Plan Aux Bldg & Cntmt El 208-10 Fire & Smoke Detection System Unit 1 1.7-37 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 2, Electrical (Bechtel)

Drawing No.

(FSAR Figure No.) Title E-7177 SH 0 Plant Air System Index & Relay Tabulation E-7177 SH 5 Plant Air System Heater Control and Power Distribution Section 3, Electrical and Instrumentation/Control (General Electric) 828E234BA Standby Liquid Control System Shs 1-4 (SLC) 828E525BA Feedwater Control System Shs 1-5 828E231BA Control Rod Drive - Hydraulic Shs 1-4 System 865E344BA Reactor Water Cleanup System Shs 1-4 (RWCS) 828E447 Jet Pump Instrumentation System Shs 1-3 828E446 Reactor Recirculation System Shs 1-35 828E549BA Nuclear Boiler Process Shs 1-5 Instrumentation System 828E534BA Residual Heat Removal System Shs 1-20 (RHR) 1.7-38 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 3, Electrical and Instrumentation/Control (General Electric)

Drawing No.

(FSAR Figure No.) Title 828E536BA High-Pressure Core Spray System Shs 1-7 (HPCS) 828E537BA HPCS - Power Supply Shs 1-15 828E535BA Low-Pressure Core Spray System Shs 1-7 (LPCS) 828E539BA Reactor Core Isolation Cooling Shs 1-13 System (RCIC) 828E444BA Automatic Depressurization Shs 1-12 System (ADS) 828E445BA Nuclear Steam Supply Shutoff Shs 1-15 System E-1187 Leak Detection System Shs 1-16 E-1177 Process Radiation Monitoring Shs 1-36 System E-1176 Neutron Monitoring System Shs 1-6 (NMS) - Startup Range Detector Drive Control E-1171 NMS - Startup Range Shs 1-22 1.7-39 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 3, Electrical and Instrumentation/Control (General Electric)

Drawing No.

(FSAR Figure No.) Title E-1170 NMS - Traversing Incore Probe Shs 1-5 E-1172 NMS - Power Range Shs 1-60 (except Shs 21, 22, 53, 55)

E-1173 Reactor Protection System (RPS)

Shs 1-28 (7.2-1) 828E532BA RPS Interconnection Shs 1-3 Scheme E-1174 RPS MG Set Control Remote (8.3-11) Shutdown System E-1151 Offgas System Shs 1-30 E-1210 Main Steamline Isolation Shs 1-3 Valve Leakage Control Shs 6-21 System (MSIV-LCS)

E-1165 Rod Control and Information Shs 1-22 System E-1206 RWCU - F/D Control Shs 1-24 E-1207 Fuel Pool Cooling and Cleanup Shs 1-16 System 1.7-40 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 3, Electrical and Instrumentation/Control (General Electric)

Drawing No.

(FSAR Figure No.) Title E-1209 Fuel Pool Filter Demineralizer Shs 1-24 System 762E401 Nuclear Boiler System FCD Shs 1-4 105D4920 Reactor Recirculation System Shs 1-5 FCD 762E429 Control Rod Drive Hydraulic Shs 1-7 System FCD 762E434 Standby Liquid Control System FCD 762E459 Neutron Monitoring System FCD Shs 1-7 944E453 Residual Heat Removal System Shs 1-5 FCD 762E294BA Low-Pressure Core Spray System Shs 1-2 FCD 851E892BA High-Pressure Core Spray System Shs 1-3 FCD 105D5046 HPCS Power Supply FCD Shs 1-5 105D5116 Leak Detection System FCD Shs 1-3 1.7-41 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 3, Electrical and Instrumentation/Control (General Electric)

Drawing No.

(FSAR Figure No.) Title 865E343 MSIV Leakage Control System FCD Shs 1, 4 762E297BA Reactor Core Isolation Cooling Shs 1-5 System FCD 762E298BA, Rev. 5 Div 3 HPCS Power System Shs 1-2 (8.3 - 12a, b) 762E298BA Div 3 HPCS (8.3 - 13) ESF - DC System 762E407 Reactor Core Cleanup System FCD 762E414BA Fuel Pool Cooling & Cleanup Shs 1-2 System FCD 807E523BA Offgas System FCD Shs 1-5 944E990 Reactor Protection System IED Shs 1-3 (7.2-1a,b,c) 762E293WJ Leak Detection System IED Shs 1-4 (7.6-2a,b,c,d) 3636-120-001 HPCS Jacket Cooling Water (9.5-15) System 1.7-42 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.7-1: NONPROPRIETARY ELECTRICAL AND INSTRUMENTATION/

CONTROL DRAWINGS (CONTINUED)

Section 3, Electrical and Instrumentation/Control (General Electric)

Drawing No.

(FSAR Figure No.) Title 3636-130-001 HPCS Air Start System (9.5-16) 3636-119-1 HPCS Lubrication System (9.5-18) 3636-119-2 (9.5-18) 1.7-43 Revision 2020-080

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.8 SYMBOLS USED IN ENGINEERING DRAWINGS The symbols applicable to piping and instrumentation diagrams (P&IDs) used throughout this report are shown in Figures 1.8-1 through 1.8-3.

1.8-1 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

C:\dgn\m0030a.dgn 11/21/2017 10:29:49 AM 1.8-2 LBDCR 2016-076

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.8-3 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.8-4 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.9 ABBREVIATIONS Table 1.9-1 is a list of the abbreviations used in this Updated Final Safety Analysis Report.

1.9-1 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.9-1: ABBREVIATIONS AASHO American Association of State Highway Officials A-CC Allis-Chalmers Corporation A-CPSI Allis-Chalmers Power Systems Inc.

ACI American Concrete Institute ACRS Advisory Committee for Reactor Safeguards ADHRS Alternate Decay Heat Removal Subsystem ADS Automatic Depressurization System AE Architect Engineer AISC American Institute of Steel Constructors ALARA as low as reasonably achievable ANI American Nuclear Insurers ANSI American National Standards Institute APRM Average Power Range Monitor ARI Alternate Rod Insertion ARM Area Radiation Monitor ARSS Alternate Reactor Scram System ASCE American Society of Civil Engineers ASCS Agricultural Stabilization Conservation Service ASDC Alternate Shutdown Cooling ASTM American Society for Testing Materials ATWS Anticipated Transients Without Scram AWS American Welding Society BEA Bureau of Economic Analysis BOF Bottom of Active Fuel BOP Balance of Plant BTP Branch Technical Position BWR Boiling Water Reactor CAMS Continuous Air Monitors CAV Cumulative Average Velocity CCW Component Cooling Water CFFF Condensate Full Flow Filtration System CFR Code of Federal Regulations CGCS Combustible Gas Control System CHF Critical Heat Flux CM Center of Mass CMAA Crane Manufacturing Association of America CP Construction Permit CPR Critical Power Ratio 1.9-2 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.9-1: ABBREVIATIONS (Continued)

CR Center of Rigidity CRACIS Control Room Atmospheric Control and Isolation System CRD Control Rod Drive CRDA Control Rod Drop Accident CRPI Control Rod Position Indication CRVICS Containment and Reactor Vessel Isolation Control System CRWE Control Rod Withdrawal Error CRWST Condensate and Refueling Water Storage and Transfer CST Condensate Storage Tank CTO Checkout and Turnover Organization DBA Design-Basis Accident DCS Distributed Control System DELS Diesel Engine Lubrication System DG Diesel Engine-Generator DGCAIES Diesel Generator Combustion Air Intake and Exhaust System DGCWS Diesel Generator Cooling Water System DGSS Diesel Generator Starting System DOP Dioctyl Pathalate DPA Displacements Per Atom DPF Design Project Flood ECA Engineering Change Authorization ECCS Emergency Core Cooling System ECN Engineering Change Notice ECP Electrochemical Corrosion Potential EDS Engineering Data Systems EFCV Excess Flow Check Valve EHC Electrohydraulic Control EIC Eberline Instrument Corporation EOC End of Cycle EPU Extended Power Uprate ER Environmental Report ERTS Earth Resources Technology Satellite ESF Engineered Safety Feature FA Full Arc (Mode of TCV Operation)

FANP Framatome-ANP FAP Fatigue Analysis Program FCD Functional Control Diagram 1.9-3 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.9-1: ABBREVIATIONS (Continued)

FCV Flow Control Valve FEDS Floor and Equipment Drainage System FDDR Field Deviation Disposition Request FHA Fuel Handling Accident FLECHT Full-Length Emergency Cooling Heat Transfer FM&IS Flow Monitoring and Isokinetic Sampling FMD Forced Helium Dehydration FMEA Failure Modes and Effects Analysis FPCC Fuel Pool Cooling and Cleanup FWCF Feedwater Controller Failure FWLCS Feedwater Leakage Control System FSAR Final Safety Analysis Report GE General Electric Company GEH GE-Hitachi Nuclear Energy Americas LLC GESSAR General Electric Standard Safety Analysis Report GETAB General Electric Thermal Analysis Basis GGNS Grand Gulf Nuclear Station GNP Gross National Product HCU Hydraulic Control Unit HEPA High-Efficiency Particulate Air/Absolute (referring to filters)

HMI Human Machine Interface HPCS High Pressure Core Spray HPU Hydraulic Power Unit HTGR High-Temperature Gas-Cooled Reactor H&V Heating and Ventilating HVAC Heating, Ventilating, and Air-conditioning HWC Hydrogen Water Chemistry HX Heat Exchanger IAC Interim Acceptance Criteria (NRC)

IBA Intermediate Break Accident IDS Instrument Data Sheet IEEE Institute of Electrical and Electronic Engineers IGSCC lntergranular Stress Corrosion Cracking IRM Intermediate Range Monitor ISI Inservice Inspection IST Inservice Testing KWU Kraftwerk Union LCD Local Climatological Data LCO Limiting Condition of Operation LCR Logarithm of Count Rate 1.9-4 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.9-1: ABBREVIATIONS (Continued)

LCS Leakage Control System LDS Leak-Detection System LFMGS Low Frequency Motor-Generator Set LHGR Linear Heat Generation Rate LOCA Loss-of-Coolant Accident LOFWH Loss of Feedwater Heating LOOP Loss-of-Offsite Power LPCI Low-Pressure Coolant Injection LPCS Low-Pressure Core Spray LPMS Loose Parts Monitoring System LPRM Local Power Range Monitor LPZ Low Population Zone LRNB Generator Load Reject W/0 Bypass LSSS Limiting Safety System Setting MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCC Motor Control Center MCPR Minimum Critical Power Ratio MELLLA+ Maximum Extended Load Line Limit Analysis Plus MELLLA Maximum Extended Load Line Limit Analysis MEOD Maximum Extended Operating Domain MG Motor-Generator Set MLD Mean Low Water Datum MP&L Mississippi Power & light MPM MP Machinery and Testing, LLC MPC Maximum Permissable Concentration MSL Mean Sea Level MSL Main Steam Line MSIV Main Steam Isolation Valve MSIV-LCS Main Steam Isolation Valve Leakage Control System NAPSIC North American Power Systems Interconnection Committee NB Nuclear Boiler NBR Nuclear Boiler Rated (power)

NCC Network Control Center NCIG Nuclear Construction Issues Group NCMA National Concrete Masonry Association NDT Nil-Ductility Transition/Nondestructive Testing NED Nuclear Energy Division NFPA National Fire Protection Association NMS Neutron-Monitoring System NPDES National Pollutant Discharge Elimination System 1.9-5 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.9-1: ABBREVIATIONS (Continued)

NPSH Net Positive Suction Head NPSHA Net Positive Suction Head Available NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NSSSS Nuclear Steam Supply Shutoff System NSOA Nuclear Safety Operational Analysis ODCM Offsite Dose Calculation Manual OSRC On-Site Safety Review Committee OBE Operating Basis Earthquake OFS Orificed Fuel Support OL Operating License OPRM Oscillation Power Range Monitor ORE Occupational Radiation Exposures OSHA Occupational Safety & Health Administration PA Public Address PAM Post Accident Monitoring PASS Post Accident Sampling System PBS Power Systems Branch PCIOMR Preconditioning Interim Operational Management Recommendation PCS Process Computer System PCT Peak Cladding Temperature P&ID Piping and Instrumentation Diagram PMF Probable Maximum Flood PMF Probable Maximum Flood PPA Peak to Peak Pressure Amplitude PQL Product Quality Checklist PRA Peak Recording Accelerometers PRFDS Pressure Regulator Failure- Down Scale PRM Power Range Monitor PRM Process Radiation Monitoring PSAR Preliminary Safety Analysis Report PSRC Plant Safety Review Committee PSS Process Sampling System PSTF Pressure Suppression Test Facility PSW Plant Service Water PSWRW Plant Service Water Radial Well PTLR Pressure-Temperature Limits Report PUSAR Power Uprate Safety Analysis Report PVS Plant Vent Stack PWR Pressurized Water Reactor 1.9-6 LBDCR 2018-095

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.9-1: ABBREVIATIONS (Continued)

QA/QC Quality Assurance/Quality Control QGC Quality Group Classification RBM Rod Block Monitor RCA Radiological Controlled Area RCIC Reactor Core Isolation Cooling RCIS Rod Control and Information System RCPB Reactor Coolant Pressure Boundary RFP Reactor Feed Pump RFWT Reduced Feedwater Temperature R.G. Regulatory Guide RHR Residual Heat Removal RMS Radiation Monitoring System RO Reactor Operator RPCS Rod Pattern Control System RPIS Rod Position Information System RPTS Recirculation Pump Trip System RPIS Rod Position Information System RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RRS Required Response Spectra RSO Reactor System Outline RWCU Reactor Water Cleanup RWE Rod Withdrawal Error RWL Rod Withdrawal Limiter RWM Rod Worth Minimizer RWST Refueling Water Storage Tank SACF Single Active Component Failure SAF Single Active Failure SAR Safety Analysis Report SBA Small Break Accident SBO Station Blackout SCC Stress Corrosion Cracking SDIV Scram Discharge Instrument Volume SDV Scram Discharge Volume SER Safety Evaluation Report SERI System Energy Resources, Inc SFP Spent Fuel Pool SGTS Standby Gas Treatment System SJAE Steam Jet Air Ejector SLCS Standby Liquid Control System 1.9-7 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.9-1: ABBREVIATIONS (Continued)

SLO Single Loop Operation SMA Strong Motion Accelerometers SMEPA South Mississippi Electric Power Association SOE Single Operator Error SPC Siemens Power Corporation SPCU Suppression Pool Cleanup SPMU Suppression Pool Makeup SQCF Seismic Qualification File SQRT Seismic Qualification Review Team SRDI Safety-Related Display Information SRFI Slow Recirculation Flow Increase SRLR Supplemental Reload Licensing Report SRM Source Range Monitor SRO Senior Reactor Operator SRP Standard Review Procedure SRSS Square Root of the Sum of the Squares SRV Safety/Relief Valve SRVDLs Safety Relief Valve Discharge Line SS Safe Shutdown SSE Safe Shutdown Earthquake SSW Standby Service Water TAF Top of Active Fuel TBCWS Turbine Building Cooling Water System TCV Turbine Control Valve TG Turbine-Generator TIP Traversing Incore Probe TRM Technical Requirements Manual TRS Test Response Spectra TSVC Turbine Stop Valve Closure Load TTNB Turbine Trip W/O Bypass UBC Uniform Building Code UHS Ultimate Heat Sink UPS Uninterruptible Power Supply URC Ultrasonic Cleaning USGS U.S. Geological Survey UT Ultrasonic Testing Vac Volts-Alternating Current VBWR Vallecitos Boiling Water Reactor VCT Vertical Cask Transport VRF Velocity Range Factor 1.9-8 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR) 1.10 DRAWING NUMBER-FSAR FIGURE NUMBER CROSS-REFERENCE Table 1.10-1 is a list of Bechtel drawing numbers cross-referenced with their corresponding FSAR figure numbers. This information has been provided to supplement FSAR figures containing P&IDs or SFDS which have been flagged to indicate that specific process lines are continued on other drawings.

1.10-1 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS Drawing No. System Identification FSAR Figure No.

C-0012 Orientation of Principal Structures 1.2-1 M-0001 General Arrangement Plan at El. 1.2-2 93'-0" and 100'-9" M-0002 General Arrangement Plan at El. 1.2-3 113'-0", 111'-0", 119'-0", 112'-0",

and 114'-6" M-0003 General Arrangement Plan at El. 1.2-4 133'-0", 148'-0", 139'-0", 135'-4",

and 147'-7" M-0004 General Arrangement Plan at El. 1.2-5 166'-0", 161'-10", and 170'-0" M-0005 General Arrangement Plan at El. 1.2-6 184'-6", 184'-0", and 189'-0" M-0006 General Arrangement Plan at El. 1.2-7 208'-10" M-0007 General Arrangement Sections "AA" 1.2-8 and "BB" M-1001 Turbine Building Plan at EL. 93'-0" 1.2-9c M-1006 Turbine Building General 1.2-9A Arrangement Sections "A-A" and "B-B" M-1007 Turbine Building General 1.2-9B Arrangement Sections "C-C," "D-D,"

and "E-E" M-0017 Radwaste Building Plan at El. 93'- 1.2-10 0"

M-0018 Radwaste Building Plan at El. 118'- 12.3-6 0"

M-0019 Radwaste Building Plan at El. 136'- 12.3-7 0"

M-0020 Radwaste Building Plan Sections "A- 12.3-8 A" and "B-B" M-0021 Radwaste Building Sections "C-C" 1.2-14 and "D-D" M-015.0- Natural Draft Cooling Tower 1.2-15 N1W20W001N-1.1-001 M-7005 Auxiliary Cooling Tower 1.2-16 1.10-2 LBDCR 2019-008

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-0030 A P&ID Legend 1.8-1 M-0030 B P&ID Legend 1.8-2 M-0033 A Makeup Water Treatment System 9.2-11 M-0033 B Makeup Water Treatment System 9.2-12 M-0034 A Domestic Water System 9.2-13 M-0034 B Domestic Water System 9.2-14 M-0035 A Fire Protection System 9.5-1 M-0035 B Fire Protection System 9.5-2 M-0035 D Fire Protection System 9.5-4 M-0035 E Fire Protection System 9.5-5 M-0035 F Fire Protection System 9.5-6 M-0035 G Fire Protection System 9.5-7 M-0035 H Fire Protection System 9.5-8 M-0035 J Fire Protection System 9.5-8a M-0035 K Fire Protection System 9.5-8b M-0035 L Fire Protection System 9.5-8c M-0035 R Fire Protection System 9.5-8e M-0036 B Auxiliary Steam System 9.5-20 (Sh. 1)

M-0036 C Auxiliary Steam System 9.5-20a M-0036 D Auxiliary Steam System 9.5-20 (Sh. 2)

M-0039 K Liquid Radwaste System 11.2-1 M-0039 L Liquid Radwaste System 11.2-2 M-0039 M Liquid Radwaste System 11.2-3 M-0039 N Liquid Radwaste System 11.2-4 M-0039 P Liquid Radwaste System 11.2-5 M-0039 Q Liquid Radwaste System 11.2-6 M-0039 R Liquid Radwaste System 11.2-7 M-0039 S Liquid Radwaste System 11.2-8 M-0039 T Liquid Radwaste System 11.2-9 M-0039 U Liquid Radwaste System 11.2-10 M-0039 V Liquid Radwaste System 11.2-11 M-0039 W Liquid Radwaste System 11.2-12 M-0039 X Liquid Radwaste System 11.2-12a M-0039 Y Liquid Radwaste System 11.2-12b M-0040 A Solid Radwaste System 11.4-1 M-0040 B Solid Radwaste System 11.4-1a M-0040 C Solid Radwaste System 11.4-1b 1.10-3 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-0040 D Solid Radwaste System 11.4-1c M-0041 Floor and Equipment Drain System 9.3-12 M-0045 A Embedded and Suspended Drains 9.3-13 Radwaste Bldg.

M-0045 B Embedded and Suspended Drains 9.3-14 Radwaste Bldg.

M-0045 C Embedded and Suspended Drains Pipe 9.3-14a Tunnel & Waste Treatment Building Units 1 & 2 M-0046 Sewage Treatment Plant 9.2-15 M-0047 A Radwaste Building Ventilation 9.4-4 System M-0047 B Radwaste Building Ventilation 9.4-5 System M-0047 C Radwaste Building Ventilation 9.4-005B System M-0049 Control Room HVAC System 9.4-1 M-0050 A Control Building HVAC System 9.4-16a M-0050 B Control Building HVAC System 9.4-16b M-0050 C Hot Machine Shop/Decontamination 9.4-16c Facility HVAC M-0051 A Miscellaneous Building Ventilation 9.4-14 System M-0051 B Miscellaneous Building Ventilation 9.4-15 System M-0051 C Miscellaneous Building Ventilation 9.4-15a System M-0052 A Plant Service Water Radial Well 9.2-27 (Sh. 1)

System M-0052 B Plant Service Water Radial Well 9.2-27 (Sh. 2)

System M-0052 C Plant Service Water Radial Well 9.2-27 (Sh. 3)

System M-0053 Process Sampling System, Part 4 9.3-8a M-1044 A Hydrogen and Carbon Dioxide System 10.2-001-1 M-1044 B Hydrogen and Carbon Dioxide System 10.2-001-2 M-1051 A Main and Reheat Steam System 10.3-1 (Sh. 1) 1.10-4 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1051 B Main and Reheat Steam System 10.3-2 M-1051 C Main and Reheat Steam System 10.3-3 M-1051 D Main and Reheat Steam System 10.3-1 (Sh. 2)

M-1052 Extraction Steam System 10.3-4 M-1053 A Condensate System 10.4-10 (Sh. 1)

M-1053 B Condensate System 10.4-11 M-1053 C Condensate System 10.4-12 M-1053 E Condensate System 10.4-10 (Sh. 2)

M-1054 Feedwater System 10.4-13 M-1054 B RFPT EHC System 10.4-13B M-1055 A Heater Vents and Drains 10.3-5 (Sh. 1)

M-1055 B Heater Vents and Drains 10.3-6 M-1055 C Heater Vents and Drains 10.3-7 M-1055 D Heater Vents and Drains 10.3-5 (Sh. 2)

M-1055 E Heater Vents and Drains 10.3-5 (Sh. 3)

M-1056 A Moisture Separator-Reheater Vents 10.3-8 (Sh. 1) and Drains M-1056 B Moisture Separator-Reheater Vents 10.3-8 (Sh. 2) and Drains M-1057 A Main and RFP Turbine Steam Seal and 10.4-3 Drain System M-1057 B Main and RFP Turbine Steam Seal and 10.4-4 Drain System M-1059 A Circulating Water System 10.4-5 M-1059 B Circulating Water System 10.4-6 M-1059 C Circulating Water System 10.4-7 M-1059 D Circulating Water System 10.4-007a M-1059 E Circulating Water System 10.4-005-02 M-1060 A Condenser Air Removal System 10.4-1 M-1060 B Condenser Air Removal System 10.4-2 M-1061 A Standby Service Water System 9.2-1 M-1061 B Standby Service Water System 9.2-2 M-1061 C Standby Service Water System 9.2-3 M-1061 D Standby Service Water System 9.2-4 M-1062 A Turbine Bldg. Cooling Water System 9.2-24 M-1062 B Turbine Bldg. Cooling Water System 9.2-25 M-1062 C Turbine Bldg. Cooling Water System 9.2-26 1.10-5 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1062 D Turbine Bldg. Cooling Water System 9.2-26a M-1063 A Component Cooling Water System 9.2-9 M-1063 B Component Cooling Water System 9.2-10 M-1064 C Condensate Cleanup System 10.4-9a M-1064 D Condensate Cleanup System 10.4-9b M-1064 E Condensate Cleanup System 10.4-9c M-1064 F Condensate Cleanup System 10.4-9d M-1064 G Condensate Cleanup System 10.4-9e M-1064 H Condensate Cleanup System 10.4-9f M-1064 J Condensate Cleanup System 10.4-9g M-1065 Condensate and Refueling Water 9.2-16 Storage and Transfer System M-1067 A Instrument Air System 9.3-1 M-1067 B Instrument Air System 9.3-2 M-1067 C Instrument Air System 9.3-2b M-1067 D Instrument Air System Auxiliary 9.3-2c Building M-1067 E Instrument Air System 9.3-2d M-1067 F Instrument Air System 9.3-2e M-1067 G Instrument Air System 9.3-2f M-1067 H Instrument Air System Containment 9.3-002j M-1067 M Instrument Air 9.3-001b M-1068 A Service Air System 9.3-3 (Sh. 1)

M-1068 B Service Air System 9.3-4 M-1068 C Service Air System 9.3-3 (Sh. 3)

M-1068 D Service Air System 9.3-3 (Sh. 2)

M-1069 A Process Sampling System 9.3-5 M-1069 B Process Sampling System 9.3-6 M-1069 C Process Sampling System 9.3-7 M-1069 D Process Sampling System 9.3-7a M-1070 A Standby Diesel Generator System 9.5-11 M-1070 B Standby Diesel Generator System 9.5-12 M-1070 C Standby Diesel Generator System 9.5-11a M-1070 D Standby Diesel Generator System 9.5-12a M-1072 A Plant Service Water System 9.2-22 M-1072 B Plant Service Water System 9.2-23 (Sh. 1)

M-1072 D Plant Service Water System 9.2-23b 1.10-6 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1072 E Plant Service Water System 9.2-23c M-1072 F Plant Service Water System 9.2-23 (Sh. 2)

M-1072 H Plant Service Water System 9.2-23a M-1077 A Nuclear Boiler System 5.2-6 (Sh. 1)

M-1077 B Nuclear Boiler System 5.2-7 M-1077 C Nuclear Boiler System 5.2-8 (Sh. 1&2)

M-1077 D Nuclear Boiler System 5.2-6 (Sh. 2)

M-1078 A Reactor Recirculation System 5.4-2 (Sh. 1)

M-1078 B Reactor Recirculation System 5.4-3 M-1078 E Reactor Recirculation System 5.4-2 (Sh. 2)

M-1079 Reactor Water Cleanup System 5.4-21 M-1080 A Filter/Demineralizer System (RWCU) 5.4-25 M-1080 B Filter/Demineralizer System (RWCU) 5.4-26 M-1081 A Control Rod Drive Hydraulic System 4.6-7 M-1081 B Control Rod Drive Hydraulic System 4.6-8 (Sh. 1)

M-1081 C Control Rod Drive Hydraulic System 4.6-8 (Sh. 2)

M-1082 Standby Liquid Control System 9.3-24 M-1083 A Reactor Core Isolation Cooling 5.4-10 System M-1083 B Reactor Core Isolation Cooling 5.4-11 System M-1085 A Residual Heat Removal System 5.4-16 (Sh. 1)

M-1085 B Residual Heat Removal System 5.4-17 M-1085 C Residual Heat Removal System 5.4-16 (Sh. 2)

M-1085 D Residual Heat Removal System 5.4-17a M-1086 High Pressure Core Spray System 6.3-1 M-1087 Low Pressure Core Spray System 6.3-4 M-1088 C Fuel Pool Cooling and Cleanup 9.1-26 (Sh. 1)

System M-1088 D Fuel Pool Cooling and Cleanup 9.1-27 System M-1088 E Fuel Pool Cooling and Cleanup 9.1-26 (Sh. 2)

System M-1089 Filter Demineralizer System (FPCC) 9.1-28 M-1090 A Leak Detection System 7.6-16 M-1090 B Leak Detection System 7.6-17 M-1091 Combustible Gas Control System 6.2-81 1.10-7 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1092 A Offgas System - Low Temperature 11.3-5 M-1092 B Offgas System - Low Temperature 11.3-6 M-1092 C Offgas System - Low Temperature 11.3-7 M-1092 D Offgas System - Low Temperature 11.3-8 M-1092 E Offgas System - Low Temperature 11.3-10 M-1093 A HPCS Diesel Generator System 9.5-13 M-1093 B HPCS Diesel Generator System 9.5-13a M-1093 C HPCS Diesel Generator System 9.5-13b M-1094 A Floor and Equipment Drains System 9.3-9 (Sh. 1)

M-1094 B Floor and Equipment Drains System 9.3-10 M-1094 C Floor and Equipment Drains System 9.3-11 M-1094 E Floor and Equipment Drains System 9.3-9 (Sh. 2)

M-1095 Offgas Vault Refrigeration System 11.3-9 M-1096 Suppression Pool Makeup System 6.2-82 M-1097 MSIV Leakage Control System 6.7-1 M-1098 A Embedded and Suspended Floor 9.3-15 Drains, Aux. Bldg.

M-1098 B Embedded and Suspended Floor 9.3-16 Drains, Aux. Bldg.

M-1098 C Embedded and Suspended Floor 9.3-17 Drains, Turbine Bldg.

M-1098 D Embedded and Suspended Floor 9.3-18 Drains, Turbine Bldg.

M-1098 E Embedded and Suspended Floor 9.3-19 Drains, Turbine and Ctmt Bldg M-1098 F Embedded and Suspended Floor 9.3-20 Drains, Turbine and Ctmt Bldg M-1098 G Embedded and Suspended Floor 9.3-21 Drains, Ctmt Bldg and Drywell M-1098 H Embedded and Suspended Floor 9.3-22 Drains, Ctmt Bldg and Drywell M-1099 Suppression Pool Cleanup System 9.3-23 M-1100 A Containment Cooling System 9.4-11 M-1100 B Containment Cooling System 9.4-12 M-1101 Drywell Cooling System 9.4-13 M-1102 A Standby Gas Treatment System 6.5-2 M-1102 B Standby Gas Treatment System 6.5-3 1.10-8 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1103 A Auxiliary Bldg. Ventilation System 9.4-10 M-1104 A Fuel Handling Area Ventilation 9.4-2 System M-1104 B Fuel Handling Area Ventilation 9.4-3 System M-1105 A Turbine Building Ventilation 9.4-6 System, Unit 1 M-1105 B Turbine Bldg. Ventilation System 9.4-7 M-1105 C Turbine Bldg. Ventilation System, 9.4-7a Unit 1 M-1106 A Diesel Gen Rm, ESF, Electrical 9.4-9a SWGR, SSW, and Circ. Water Pumphouse M-1106 B Diesel Gen Rm, ESF, Electrical 9.4-9b SWGR, SSW, and Circ. Water Pumphouse M-1107 A Process Radiation Monitoring System 11.5-2 M-1107 B Process Radiation Monitoring System 11.5-4 M-1107 C Process Radiation Monitoring System 11.5-5 M-1107 D Process Radiation Monitoring System 11.5-3 M-1107 E Process Radiation Monitoring System 11.5-6 M-1107 F Process Radiation Monitoring System 11.5-7 M-1107 G Process Radiation Monitoring System 11.5-1 M-1107 H Process Radiation Monitoring System 11.5-8 M-1108 A Safeguard Switchgear and Battery 9.4-8a Room Ventilation System M-1108 B Safeguard Switchgear and Battery 9.4-8b Room Ventilation System M-1109 A Plant Chilled Water System 9.2-17 M-1109 B Plant Chilled Water System 9.2-18 M-1109 C Plant Chilled Water System 9.2-19 M-1109 D Plant Chilled Water System 9.2-20 M-1109 E Plant Chilled Water System 9.2-21 M-1109 F Plant Chilled Water System 9.2-21a M-1110 A Containment and Drywell Instrument 7.5-5 and Control System 1.10-9 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1110 B Ctmt and Drywell Inst and Control 7.5-6 System M-1111 A Containment Leak Rate Test System 6.2-76 M-1111 B Containment Leak Rate Test System 6.2-77 M-1111 C Containment Leak Rate Test System 6.2-78 M-1111 D Containment Leak Rate Test System 6.2-79 M-1111 E Containment Leak Rate Test System 6.2-79a M-1112 Feedwater Leakage Control System 6.7-5 M-1115 B Turbine Cycle Heat Balance 10.1-2 M-1118 Seismic Instrumentation System, 3.7-82 Unit 1 & Common M-1119 A Drywell Chilled Water System 9.2-48 M-1119 B Drywell Chilled Water System 9.2-49 M-1119 C Drywell Chilled Water System 9.2-50 M-1126 Plant Air System 9.3-31 M-1126 B Plant Air System 9.3-32 M-1126 C Plant Air System 9.3-33 M-1500 Internally Generated Missiles 3.5-6 Ctmt. El. 93'-0" & 100'-9" Area 11

- Unit 1 M-1501 Internally Generated Missiles 3.5-13 Ctmt. El. 114'-6" & 120'-10" Area 11 - Unit 1 M-1502 Internally Generated Missiles 3.5-7 Ctmt. El. 135'-4", 140'-0" & 147'-

7" Area 11 - Unit 1 M-1503 Internally Generated Missiles 3.5-15 Ctmt. El. 161'-10" & 170'-0" Area 11 - Unit 1 M-1504 Internally Generated Missiles 3.5-16 Ctmt. - Unit 1 Section "A-A" M-1505 Internally Generated Missiles 3.5-17 Ctmt. - Unit 1 Section "B-B" M-1506 Internally Generated Missiles 3.5-5 Ctmt. - Unit 1 Misc. Sections &

Details 1.10-10 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1507 Internally Generated Missiles 3.5-19 Ctmt. - Unit 1 Misc. Sections &

Details M-1508 Internally Generated Missiles 3.5-10 Aux. Bldg. El. 93'-0" Area 10 Unit 1

M-1509 Internally Generated Missiles 3.5-12 Aux. Bldg. El. 119'-0" Area 8 Unit 1

M-1510 Internally Generated Missiles 3.5-8 Aux. Bldg. El. 139'-0" Area 7 Unit 1

M-1511 Internally Generated Missiles 3.5-11 Aux. Bldg. El. 139'-0" Area 8 Unit 1

M-1512 Internally Generated Missiles 3.5-14 Aux. Bldg. Sections Unit 1 M-1513 Internally Generated Missiles 3.5-18 Aux. Bldg. Sections Unit 1 M-1514 Internally Generated Missiles 3.5-20 Aux. Bldg. Sections Unit 1 M-1515 Internally Generated Missiles 3.5-9 Aux. Bldg. Partial Plans Unit 1 M-1516 Internally Generated Missiles 3.5-21 Control Bldg. El. 148'-0" Area 25A Unit 1 M1550 A High Energy Pipe Break Main Steam 3.6A-1A "A" & "B" Inside Ctmt. Unit 1 M1550 B High Energy Pipe Break Main Steam 3.6A-1B "C" & "D" Inside Ctmt. Unit 1 M1551 High Energy Pipe Break Main Steam 3.6A-2 System Outside Ctmt. Unit 1 M-1552 High Energy Pipe Break Stm. Supply 3.6A-3 to RCIC & RHR from Main Stm. "A" Inside Ctmt. Unit 1 M-1553 High Energy Pipe Break Stm. Supply 3.6A-4 to RCIC & RHR Outside Ctmt. Unit 1 1.10-11 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1554 A High Energy Pipe Break Feedwater 3.6A-5A System "A" Inside Ctmt. Unit 1 M-1554 B High Energy Pipe Break Feedwater 3.6A-5B System "B" Inside Ctmt. Unit 1 M-1555 High Energy Pipe Break FW System 3.6A-6 Including RHR & RWCU Inj. Piping Outside Ctmt. Unit 1 M-1556 A High Energy Pipe Break Reactor 3.6A-7(Sh. 1)

Water to DCB-3 M-1556B High Energy Pipe Break RWCU System 3.6A-7(Sh. 2)

Inside Ctmt. Unit 1 M-1556C High Energy Pipe Break Reactor 3.6A-7A Water from Reactor to DBA-11 M-1557 High Energy Pipe Break RWCU System 3.6A-8 Outside Ctmt. Unit 1 M-1558 High Energy Pipe Break RHR Suction 3.6A-9 Off of Recirc. Loop "B" Inside Ctmt. Unit 1 M-1559 High Energy Pipe Break HPCS Piping 3.6A-10 Inside Ctmt. Unit 1 M-1560 High Energy Pipe Break LPCS Piping 3.6A-11 Inside Ctmt. Unit 1 M-1561 High Energy Pipe Break RHR-LPCI 3.6A-12 Piping & RPV Head Spray Inside Ctmt. Unit 1 M-1562A High Energy Pipe Break Main Steam 3.6A-13A Drains Inside Ctmt. Unit 1 M-1562B High Energy Pipe Break Main Steam 3.6A-13B Drains Outside Ctmt. Unit 1 M-1563 High Energy Pipe Break Reactor 3.6A-13C Steam Unit 1 M-1564 High Energy Pipe Break DRW Vents & 3.6A-13D Drains Unit 1 M-1565A High Energy Pipe Break Main Steam 3.6A-13E (Sh.

Line Drain Outside Ctmt. from 1)

Isolation Valves Unit 1 1.10-12 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

M-1565B High Energy Pipe Break Main Steam 3.6A-13E (Sh.

Line Drain Outside Ctmt. from 2)

Isolation Valves Unit 1 M-1565C High Energy Pipe Break Main Steam 3.6A-13E (Sh.

Line Drains Outside Ctmt. from 3)

Isolation Valves Unit 1 M-1565D High Energy Pipe Break Main Steam 3.6A-13E (Sh.

Line Drain Outside Ctmt. from 4)

Isolation Valves Unit 1 M-1565E High Energy Pipe Break Main Steam 3.6A-13E (Sh.

Line Drain Outside Ctmt. from 5)

Isolation Valves Unit 1 M-1566A High Energy Pipe Break Sodium 3.6A-13F (Sh.

Pentaborate to RPV Unit 1 1)

M-1566B High Energy Pipe Break Sodium 3.6A-13F (Sh.

Pentaborate to RPV Unit 1 2)

M-1556D High Energy Pipe Break PWCU System 3.6A-7 (Sh. 3)

Inside Containment M-1556E High Energy Pipe Break PWCU System 3.6A-7 (Sh. 4)

Inside Containment M-1556F High Energy Pipe Break PWCU System 3.6A-7 (Sh. 5)

Inside Containment M-1556G High Energy Pipe Break PWCU System 3.6A-7 (Sh. 6)

Inside Containment M-1568 High Energy Pipe Break Condensate 3.6A-13G to H.P. Condensers Unit 1 M-1569A High Energy Pipe Break Steam to 3.6A-13H (Sh.

MSIV Leakage Control System Unit 1 1)

M-1569B High Energy Pipe Break Steam to 3.6A-13H (Sh.

MSIV Leakage Control System Unit 1 2)

M-1570 High Energy Pipe Break Feedwater 3.6A-13I Leakage Control System Unit 1 M-1571 High Energy Pipe Break Reactor 3.6B-4a Recirculation Loops A & B Unit 1 SFD-0039 A Liquid Radwaste System 11.2-13 SFD-0039 B Liquid Radwaste System 11.2-14 SFD-0039 C Liquid Radwaste System 11.2-15 SFD-0039 D Liquid Radwaste System 11.2-16 1.10-13 Revision 2016-00

GRAND GULF NUCLEAR GENERATING STATION Updated Final Safety Analysis Report (UFSAR)

TABLE 1.10-1: CROSS-REFERENCE LIST OF DRAWING NUMBERS AND FSAR FIGURE NUMBERS (Continued)

Drawing No. System Identification FSAR Figure No.

SFD-0039 E Liquid Radwaste System 11.2-17 SFD-0039 F Liquid Radwaste System 11.2-18 SFD-0040 Solid Radwaste System 11.4-2 SFD-0049 Control Room HVAC System 6.5-1 SFD-1077 Nuclear Boiler System, Unit 1 5.2-9 SFD-1079 Reactor Water Cleanup System 5.4-22 SFD-1081 Control Rod Drive Hydraulic System 4.6-10 SFD-1082 Standby Liquid Control System 9.3-25 SFD-1083 A Reactor Core Isolation Cooling 5.4-12 System, Unit 1 SFD-1083 B Reactor Core Isolation Cooling 5.4-13 System, Unit 1 SFD-1085 Residual Heat Removal System 5.4-18 762E445 High Pressure Core Spray System 6.3-2 SFD-1087 Low Pressure Core Spray System 6.3-5 SFD-1088 Fuel Pool Cooling and Cleanup 9.1-30 System SFD-1089 Filter Demineralizer System (FPCC) 9.1-29 SFD-1092 A Offgas System - Low Temperature 11.3-5 SFD-1092 C Offgas System - Low Temperature 11.3-7 SFD-1094 Sh 1 Floor and Equipment Drains System 9.3-28 SFD-1094 Sh 2 Floor and Equipment Drains System 9.3-29 SFD-1097 MSIV Leakage Control System 6.7-2 SFD-1102 Standby Gas Treatment System 6.5-4 SFD-1111 Containment Leak Rate Test System 6.2-80 1.10-14 Revision 2016-00