ML20078S123

Rev 0 to Grand Gulf Nuclear Station Engineering Rept for Individual Plant Exam of External Events

ML20078S123
Person / Time
Site:	Grand Gulf
Issue date:	12/19/1994
From:	Locke M, Williamson E ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20078S124	List: ML20078S123
References
GGNS-94-0054, GGNS-94-0054-R00, GGNS-94-54, GGNS-94-54-R, GL-88-20, NUDOCS 9412280110
Download: ML20078S123 (53)

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ENGINEERING REPORT NO. GGNS-94 0054 PAGE 1 OF 50 REVISION 0 GRAND GULF NUCLEAR STATION ENGINEERING REPORT FOR l

INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS

SUMMARY

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TABLE OF CONTENTS 1.0 EXECUTIVE S UMMARY..........................

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1.1 Background

1.2 Plant Familiarization..............................

6 1.3 Overall Methodology...............................

7 1.4 Summary ofMajor Findings...........................

8 8

2.0 EXAMINATION DESCRIPTION......................

9 2.1 Introduction...................................

2.2 Conformance with Generic Letter and Supporting Material............

9 2.3 General Methodology 9

2.3.1 Seismic Analysis.................................. 10 2.3.2 Fire Analysis...............................

10 2.3.3 10 High Winds, Floods and Others...........

2.4 Information Assembly I1

............................. I1 3.0 SEISMIC ANALYSIS................................

12 3.0 Methodology Selection....................

3.1 EPRI Seismic Margins Method (SMM).....................

12 3.1.1 12 General Plant Information and Seismic Input...............

13 3.1.1.1 General Plant Information................13 3.1.1.2 Seismic In System Analya.....put to Structures and Equipment.......16 3.1.2

.......................17 3.1.2.1 Overall Approach..........

17 3.1.2.2 Assumptions.............

17 3.1.2.3 Principal Safety Functions................

18 3.1.2.4 System Success Criteria..................

18 3.1.2.5 Overall Success Path Logic Diagram..........

19 4

3.1.2.6 Primary and Alternate Success Path Selection,.....

21 3.1.2.6.1 Preferred Path Selection......... 21 3.1.2.6.2 Alternate Path Selection.........

22 3.1.2.7 Systems Used in Success Paths......

23 3.1.2.7.1 Front-line System Jescriptions.....

23 3.1.2.7.2 Support Systems Descriptions......

25 3.1.2.8 Equipment Identification and Selection.

26 3.1.3 Seismic Walkdown...................

f Approach...............,........

27

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3.1.3.1 27 3.1.3.2 Screening Walkdown.....

Walkdowns................

28 3.1.3.3 29 3.1.3.4 Seismic Analysis Results

................ 30 TABLE OF CONTENTS

y ENGINEERING REPORT NO. GGNS-94 0054 PAGE 4 OF 50 REVISION 0 3.2 Conclusions of Seismic Analysis........................

. 30 4.0 INTERNAL FIRES ANALYSIS.....................

t 30 5.0 HIGH WINDS, FLOODS, AND OTHERS................

30 5.1 High VTmds.........

5.2 Floods............

......................... 30 Hydrologic Conditions and Existing Flood Protectio.........................

31 5.2.1 5.2.2 Original Design Basis Evaluations..........n.........

31 5.2.3 Evaluation ofRevised Hazards Due to Flo 32 of the Mississippi River..........oding 5.2.4 Review ofFlood Hazards Due to Preci 33 over the Site Watershed....... pitation 5.2.5 Effects of the New PMP Data on Roof Loading.............

33 5.3 Transportation and Nearby Facility Accidents................

34 5.3.1 Industrial and Military Facilities..................

35 5.3.2 Transportation Facilities and Routes................

35 5.3.3 Mississippi River Accidents......

35 5.3.4 Significant Changes...........

36 5.3.5 Conclusion....

36 5.4 Others......................................

3 6

... 36 6.0 LICENSEE PARTICIPATION AND INTERNAL REVIEW TEAM

.37 6.1 IPEEE Program Organization.................

6.2 Cor.: position ofIndependent Review Team...........

37 6.3 37 Areas ofReview and Major Comments.....................

37 6.3.1 Seismic Review......................

High Winds, Floods, and Others....................

37 h

6.3.2 38 6.4 Resolution of Comments............................ 38 7.0 PLANT IMTROVEMENTS AND UNIQUE SAFETY FEATURES.....

38 7.1 Seismic Analysis.........

38 7.2 Internal Fire Analysis

......s 39 4

7.3 High Winds, Floods, and Others.

39 7.3.1 High Winds and Others.........................

39 7.3.2 Floods 39

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l ENGINEERING REPORT NO. GGNS-94-0054 PAGE 5 OF 50 REVISION O i

l TABLE OF CONTENTS 8.0

SUMMARY

AND CONCLUSIONS (INCLUDING PROPOSED RESOLUTION OF USIs AND GIs)......................... 40 8.1 Seismic Analysis................................

41 8.2 Fire Analysis...............

41 8.3 High Winds, Floods, and Others Analysis...................

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9.0 REFERENCES

.................42 10.0 Ta bles.........................................

44 10.1 Front-line Systems...............................

44 10.2 Front-line to Support System Dependency Matrix..............

45 10.3 Support System to Support System Dependency Matrix...........

46 Notes for Tables 3.2 and 3.3..........................

47 11.0 Fig u res......................,.................

4 8 11.1 GGNS Success Path Logic Diagram......................

48 11.2 GGNS Perferred Success Path.........................

49 11.3 GGNS Alternate Success Path.........................

50 LIST OF A'ITACHMENTS A'ITACHMENT 1 Safety Evaluation Applicability Review form

ENGINEERING REPORTNO. GGNS-944054 PAGE6 OF $0 REVISION 0 1.0 EXECUTIVE

SUMMARY

NRC Generic Letter 88-20, Supplement 4, (4) requested that licensees perfonn an " Individual Plant Examination of External Events (IPEEE)" to assess severe accident vulnerabilities This report is presented to document the completion of all portions of that evaluation except those portions of the evaluation related to severe accident vulnerabilities===aci=+M with plant fires. Intemal Flooding was included in the Individual Plant Examination (IPE) for intemal events, which was submitted on De=he 23,1992 to the NRC via letter no. GNRO-92/00157 (2),

1.1 Background and Objectives In the Commission policy statement on severe accidents in nuclear power plants issued on August 8,1985, the Commission concluded, based on available information, that existing plants pose no undue risk to the public health and safety and that there was no present basis for ir==Mi=*e action on any rege3 story requirements for these plants. However, the Commission recognized, based on NRC and industry experience with plant-specific probabilistic risk assessments (PRAs),'that systematic examinations are beneficial in identifying plant-specific vulnerabilities to severe accidents that could be fixed with low-cost improvements. As part of the implementation of the Severe Accident Policy, the Commision issued Generic Letter 88-20 on November 23,1988, requesting that each licensee conduct c.n Individual Plant Examination (IPE) for intemally initiated events including internal flooding.

Many PRAs indicate that, in some instances, the risk from external events could contribute significantly to core damage. In December 1987, an External Events Steering Group (EESG) was established by the NRC to make recommendations regarding the scope, methods and coordination of the Individual Plant Examination of external events (IPEEE).

In June 1991, the NRC issued Supplement 4 to Generic Letter 88-20 requesting a plant specific analysis of external events. Jointly issued with Supplement 4, NUREG 1407 (5) was issued to give procedural and submittal guidance for the IPEEE.

The objectives of the IPEEE, as outlined in NUREG-1407, are:

1.

To develop an appreciation of severe accident behavior.

2.

To understand the most likely severe accident sequences that could occur at Grand Gulf Nuclear Station (GGNS) under full power operating conditions.

3.

To gain a qualitative understanding of the overall likelihood of core damage and fission product releases.

If necessary, to reduce the overall likelihood of core damage and fission product releases by 4.

modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.

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ENGD4EERING REPORT NO. GGNS-944054 PAGE 7 OF SO REVISION 0 With the exception ofInternal Fires, Entergy Operations, Incorporated (EOI) has completed and documented an IPEEE for GGNS. This report contains a summary of the methods, results, and conclusions for the IPEEE. With the exception ofInternal Fires, which will be submitted at a later date, this report complies with the NRC request for information contained in Generic Letter 88-20, Supplement 4 and NUREG 1407.

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1.2 Plant Famliiarisation Grand Gulf Nuclear Station is located on the east bank of the Mississippi River in Claiborne County, Mississippi about six miles northwest ofPort Gibson, 25 miles southwest of Vh*al=g and 37 miles northeast of Natchez. GGNS is a General Electric BWR/6 reactor with a Mark III Containment. It has a rated and licensed core thermal output of 3833 MWt that corresponds to a net electrical output of 1254 MWe when the contribution from the reactor coolant pumps is considered. The d sign power level is 4025 MWt. The Mark III containment incorporates the -

drywell/ pressure suppression concept. This site was originally designed to support two unita, but Unit 2 has been canceled. GGNS received its full power operating license on August 31,1984 and commenced commercial operation on July 1,1985.

1 The nuclear boiler system includes a direct cycle, forced circulation boiling water reactor that produces steam for direct use in the steam turbine. The reactor vessel is a 251 inch BWR/6 vessel

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with 800 fuel assemblies. Overpressure protection is provided by 20 safety / relief valves that i

discharge into the suppression pool.

The containment is designed to contain the energy released from the design basis loss of coolant accident (1.OCA) and provide a leak tight barrier against the uncontrolled release of radioactivity to the environment. Its design incorporates a drywell surrounding the reactor pressure vessel and a large part of the reactor coolant pressure boundary; a suppression pool that servet as a heat sink during normal operational transients and accident conditions; and a steel lined reinforced concrete containment structure.

The plant design incorporates redundant Engineered Safety Features (ESF) which in conjuction with the containment ensure off site radiological consequences do not exceed regulations. Some of the important ESF systems include: Emergency Core Cooling Systems [which include High Pressure Core Spray (HPCS), Automatic Depressurization System (ADS), Low Pressure Core Spray (LPCS) and Low Pressure Core Injection (LPCI)]; Containment Spray (CS) subsystem of Residual Heat Removal (RHR); Standby Gas Treatment System (SGTS); and the Suppression Pool Makeup (SPMU) system. Other systems credited in the IPEEE for mitigating accidents include Reactor Core Isolation Cooling (RCIC), Control Rod Drive (CRD), Power Conversion System (PCS), Standby Service Water (SSW) cross-tie to LPCI, Firewater aligned for vessel injection, the Suppression Pool Cooling (SPC) and Shutdown Coolirs (SDC) modes of RHR, Containment Venting, Alternate Rod Insertion (ARI), Recirculation Pump Trip (RPT), and Standby Liquid Control System (SLCS).

l GGNS generates electrical power at 22kV that is connected to the 500kV switchyard through a main step up transformer. In the event ofloss of normal and preferred auxiliary power sources, the i

l ENGINEERING REPORT NO. GGNS-94 0054 PAGE 8 OF 50 REVISION O ESF loads and various non-safety loads (under certain conditions) can be supplied from the on-site emergency power sources. The emergency on-site power sources consist of two indap-tad and completely segregated emergency diesel generators, each of which has an adequate capacity to meet the loads required for safe shutdown of the reactor. A third diesel is used as a backup power source to the HPCS. Three divisions of emergency DC power are also available to supply the ESF loads as required.

The SSW system provides required cooling water to ESF equipment served by the system as well as to various non-safety related portions of the plant. That portion of the SSW required for safe shutdown of the plant is designed to meet Seismic Category I requirements and the single failure criterion. Non-safety related portions are isolated from the SSW during events resulting in ESF actuation.

Heating Ventilating and Cooling (HVAC) systems are present to cool ESF system equipment such as ECCS pump rooms, diesel generator rooms, SSW pump rooms and electrical switchgear.

1.3 Overall Methodology The IPEEE consist of three separate analyses:

Seismic Analysis Internal Fire Analysis (Later)

High Winds, Floods and Others Analysis For Seismic analysis, the EPRI Seismic Margins Methodology (1) was used to perform the analysis.

Although not included in this report, a Fire PRA will be performed for the Internal Fire analysis.

For High Winds, Floods and Others analysis, a review was performed to demonstrate conformance to the 1975 SRP, As required in NUREG 1407, evaluations for the effects oflocalized Probable Maximum Precipitation (PMP) consider the revised estimates published in Hydrometeorological Report (HMR) Nos. 51 and 52.

1.4 Summary of Major Findings For the Seismi: Analysis, Grand Gulf Nuclear Station is identified as a reduced scope plant by NUREG 1407. Therefore, the Safe Shutdown Earthquake (SSE) ground response spectra and corresponding in-structure response spectra were used as the Review Level Earthquake (RLE).

The conclusions of the seismic analysis is that Grand Gulf Nuclear Station is seismically rugged and that all components identified in the Safe Shutdown Path have adequately considered the seismic input.

All anchorage to these components was found to be rugged.

Only one potential vulne.sbility to a seismic event was identified, which has been corrected.

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ENGINEERING REPORT NO. OGNS-944)S4 PAGE 9 OF SO REVISION 0 For Flooding, Transportation & Nearby Facility Accidents, and Others, GGNS is in compliance with criteria in the '75 Standard Review Plan (SRP). Significant changes to site features were not noted during IPEEE inspections.

With regards to High Winds and Tornadoes, GGNS is in compliance with the 75 SRP with only few exceptions.

1 A probabilistic evaluation" for frequency of occurrence for these exceptions I

detennined a frequency of 0.77 X E-8/yr. This low frequency of occurrence is well below the screening criteria in IPEEE.

As a result of the Flood analysis, enhancements were identified which would help ensure continued compliance with the SRP and mitigate adverse effects of the new PMP criteria. Section 7.3.2 contains details of these enhancements.

2.0 EXAMINATION DESCRIPTION 2.1 Introduction With the exception ofInternal Fires, E0I has completed an IPEEE for Grand Gulf Nuclear Station.

This Section provides details on the conformance with the Generic Letter and supporting materia the general methodology and the information assembly, i

i 2.2 Conformance With Generic Letter and Supporting Material This report conforms with Generic Letter 88-20, Supplement 4 and its supporting material.

NUREG-1407 was followed closely in preparing this report. The content and format of this repo conforms with the requirements of NUREG-1407.

A major thrust of GL 88-20 is that the Utility should gain the insights into severe accident behavior.

EOI has expanded significant resources developing in-house personnel to perform the IPEEE. The majority of contractor work was performed on-site and in-house personnel worked closely with contractors.

Because of this working philosophy, the insights and knowledge gained in performance of the IPEEE have been retained by the utility.

i Technical adequacy of the IPEEE is assured by a combination of:

Use ofinformation from reliable documents.

Use of knowledgeable individuals Evaluation by multiple individuals (where appropriate).

Performance of a PEER review.

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ENGINEERING REPORT NO. GGNS 94 0054 PAGE 10 OF SO REVISION 0 For the Seismic analysis, the Level 1 PRA was used in determining what components are required for safe shutdown. Two " Seismic Capability Engineers" as defined in EPRI NP6041(1) evaluated the component for seismic adequacy.' Where questions arose on the seismic walkdowns, additional sources ofinformation, such as existing calculations, were referred to.

There is a high confidence in the technical accuracy of the IPEEE. Knowledgeable individuals -

performed the analysis using reliable sources of information. Additionally a Peer review was performed which provides additional assurance that the IPEEE is technically accurate.

2.3 General Methodology The IPEEE consists of three separate analyses:

Seismic Analysis Internal Fire Analysis (Later)

IIigh Winds, Floods and Others Analysis This Section gives a brief description of the methodology used in performing these analyses.

2.3.1 Seismic Analysis Grand Gulf Nuclear Station (GGNS) is classified a reduced scope plant as dermed NUREG-1407 based on the low seismicity. Therefore, a seismic review of the plant was performed to the plant's original design basis. This was accomplished by performing a Seismic Margins Assessment (SMA) of the Safe Shutdown Equipment List (SSEL) with plant walkdowns in accordance with the guidelines and procedures documented in Electrical Power Research Institute (EPRI) Report NP-6041-SL.m Since Grand Gulf Nuclear Station is a reduced scope plant, the original design basis Safe Shuswn Earthquake (SSE) ground response spectra and corresponding in-structure response spectra were used as the Review Level Earthquake (RLE) input for the walkdown and evaluation, as requested by NUREG-1407. No new in-structure response spectra were developed and those described in the Grand Gulf Nuclear Station Updated Final Safety Analysis Report (UFSAR) (15) were utilized.

Safe shutdown success paths were developed to identify the systems that must function to successfully shutdown and cool the reactor following the occurrence of a RLE. A safe shutdown success path is a string of systems which is used to accomplish all of the required safe shutdown functions. This success path can be depicted in a Shutdown Path Logic Diagram (SPLD). The SPLD identified 44 systems and 567 components which required evaluation for seismic adequacy.

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2.3.2 Fire Analysis LATER.

l ENGINEERING REPORT NO. GONS-944054 PAGE 11 OF 50 REVISION 0 2.3.3 High Winds, Floods and Others High Winds, Floods and Others are external events other than seismic, internal fire, or internal l

flooding events that may be initiators of accider.t sequences leading to core damage. Such phenomena are potentially important because they affect multiple components An accident involving a number of different component failures may be nearly incredible in the absence of some external influence, but may be possible or even likely by the occurrence of a tornado, for instance.

As recommended in Generic Letter 88-20, 4 9plement 4, the methodology employed for is.Jyig other external events at GGNS was a screeniag approach. The first step in the screening approach was to determine if the criteria of the 1975 Standard Review Plan (6) are met.

The progressive screening appnach as outlined in Generic Letter 88-20, Supplement 4, was used in

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the evaluation of transpor!%o vid nearby facility accidents. Guidance from NUREG-1407 and NUREG/CR-5042 (16) wr, 41sc used in the evaluation of specific hazards for the determination of potential vulnerabilities and iiaght for screening criteria.

As required in NUREG 1407, evaluations for the effects of localized Probable MWm9=

Precipitation (PMP) on site flooding and roof ponding consider the revised PMP estimates published by the National Oceanic and Atmospheric Administration in HMR No. 51. Local intense precipitation on adjacent streams, and the site yard are evaluated applying HMR No. 52, Regulatory Guide 1.59, Regulatory Guide 1.102, and ANS 2.8-1992. A backwater analysis is performed to determine plant water elevations. Food water entering the powerblock is estimated and evaluated to determine ifit can jeopardize components required to safely shutdown the plant and/or maintain the plant in a shutdown condition. Evaluations for the effects of local intense precipitation include assessments of the potential hazards resulting from roof ponding.

2.4 Information Assembly Evaluations were performed using current revisions of As-Built drawings, specifications, standards, plant procedures, and other sources. Design chant,e packages and non-conformance reports were reviewed where applicable. The As-Built conditions of plant features were verified by visual inspection and measurements in the field where required or appropriate. Evaluations for each section of the submittal (i.e., seismic, fire, flooding, high winds, etc.) are documented in detail in separate engineering reports, each prepared in accordance with the GGNS Quality Assurance program. These reports were also reviewed by an independent Peer Reviewer. Similarly, detailed calculations performed in support of these reports were prepared in accordance with the GGNS Quality Assurance program and made available for Peer Review.

4 ENGINEERING REPORT NO. GGNS-94 0054 PAGE 12 OF 50 REVISION 0 Detailed evaluations pertaining to this submittal are contained in the following reports:

Engia-ing Report GGNS-93-0001, Rev. 0 " Individual Plant Examination for External Events (External Flooding)".

Engineering Report GGNS-94-0053, Rev. O, "IPEEE Reduced Scope Seismic Margins Assessment (SMA)".

Engineering Report GGNS-93-0048, Rev. O, "High Wind and Tornado Assessment".

Engineering Report GGNS-93-0047, Rev. O, " Transportation & Nearby Facility Accidents".

t Engineering Report GGNS-93-0031, Rev. O, "An Assessment of the Risk From Lightning Initiators".

SEISMIC ANALYSIS I

i 3.0 Methodology Selection Grand Gulf Nuclear Station (GGNS) is classified a reduced scope plant as defined NUREG-1407 based on the low seismicity. Therefore, a seismic review of the plant was performed to the plant's original design basis. This was accomplished by performing a Seismic Margins Assessment (SMA) of the Safe Shutdown Equipment List (SSEL) with plant walkdowns in accordance with the guidelines and procedures documented in Electrical Power Research Institute (EPRI) Report NP-6041-SL (1).

Since Grand Gulf Nuclear Station is a reduced scope plant, the original design basis Safe Shutdown Earthquake (SSE) ground response spectra and corresponding in-structure response spectra were used as the Review Level Earthquake (RLE) input for the walkdown and evaluation, as requested by NUREG-1407. No new in-structure response spectra were developed and those described in the Grand Gulf Nuclear Station Updated Final Safety Analysis Report (UFSAR) were utilized.

Safe shutdown success paths were developed to identify the systems that must function to

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successfully shutdown and cool the reactor following the occurrence of a RLE. A safe shutdown success path is a string of systems which is used to accomplish all of the required safe shutdown functions. This success path can be depicted in a Shutdown Path Logic Diagram (SPLD). The SPLD identified 44 systerns and 567 components which required evaluation for seismic adequacy.

3.1 EPRI Seismic Margins Method (SMM)

The EPRI Seismic Margins Method (SMM)(1) was used to perform the seismic analysis. This section provides the de: ails about the analysis.

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ENGINEERING REPORT NO. GONS-94 0054 PAGE 13 OF 50 REVISION 0 3.1.1 General Plant Information and Seismic Input This section provides general plant information and the seismic

• put used in the analysis.

m 3.1.1.1 General Plant Information GGNS 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 Gulfis about 1% 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 takes, Gin Lake and Hamilton Lake. These lakes were once the channel of the Mississippi River and average about 8 to 10 feet above mean sea level (ms!).

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

The eastern half of the plant site is rough and irregular with steep slopes and deep-cut stream valleys and drainage coursed. Elevations in this portion of the plant site range about 400 feet above -

mean sea level (msl) occur on the hilltops east an' northeast ofthe site.

i 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 flood plain, divides the site so that it lies in two sub provinces of the Central Gulf Coastal Plain Physiographic province.

The sub provinces 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 feet of Cretaceous through Cenozoic sands, gravel's, clays, marls, claystones, sandstones and limestones. These sediments were deposited on middle Jurassic evaporates, 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 structure began in the early Cretaceous period and ended in the middle Tertiary pedod. A salt dome has been formed as near as 8 miles from the site. The dome was formed from the late Cr:taceous period through the Oligocene epoch. No nearer salt domes are known.

Most deep site borings encountered the Miocene age Catahoula formation. The Catahoula consists of a hard-to-very-hard, gray-to-gray-green, silty-to-sandy clay, and clayey silt and sand, with some l

I ENGINEERING REPORT NO. GGNS-94-0054 PAGE 14 OF 50 REVISION 0 l

locally indurated or cemented clay, sand and silt seams. The Catahoula Formation is the bearing stratum for the major plant structures. The maximum estimated thickness of the Catahoula

' formation at the site is 320 feet.

Uncomformably underlying the Catahoula formation is the Vicksburg Group, a sequence of four formation of Oligocene age. These formations, from youngest to oldest, are the Bucatunna, the Byram, the Glendon, and the Mint Spring. The Bucatunna is a 53 foot thick layer of stiff-to-hard green-black-to-black clay with thin, gray, fine sand seams The Byram Mari, underlying the Bucatunna, is hard-to-very-hard, green-to-gray, fme sandy, calcareous clay approximately 5 feet thick. The Byram Marl is discontinuous throughout the region. Conformably underlying the Byram Marl is the Glendon Formation.

It consists of a series of interbedded, light gray, fossiliferous limestones and hard-to-partly-indurated, grayish-green, fine sandy, calcareous clays.

Total thickness is about 46 feet. Underlying the Glendon is the Mint Spring Marl. Forty feet of the Mint Spring Marl was penetrated at the sit; however, the total thickness of the formation was not determined. The formation consists of hard, grayish green fossiliferous, glauconitic sand and clay.

Aerial photographic interpretation and geologic mapping of outcrops and excavations did not detect the presence of any faults or tectonb structures in the plant vicinity. Inspection of outcrops, exposures in excavation, and subsurface samples have revealed that there are no deformational zones within the Catahoula material, which is the foundation material for the major plant structures.

There are no reversals of dip of the Catahoula Formation in the vicinity of the site. Exposures of the Catahoula contain occasional preferred joint orientations of N75 E, N45'W and N45 E. The joints are tight and contain no altered materials. Core samples of Catahoula material at the site do not exhibit shear zones or fractures.

There is no evidence to suggest that surfacial or subsurface materials at the site have been affected by prior earthquake activity. No faults were encountered by the numerous site boring or exposed in any of the excavations.

The Gulf Coast Basin tectonic province, in which the site is located, is characterized by infrequent earthquakes oflow epicentral intensities (Modified Mercalli Intensity VI or less), with an attendant low seismic-risk level.

All documented earthquakes of epicentral intensity IV to V or greater which have occurred within 200 miles of the site have been investigated. The historical earthquake which occurred nearest to the site had a maximum intensity of III to IV and occurred on June 28, 1941, at Vicksburg, Mississippi, about 25 Miles north-northeast of the site. It is the only earthquake reported within 100 miles of the site.

An anomalous zone of seismicity, the new Madrid seismic zone, is located within the Mississippi Embayment. The New Madrid seismic zone is characterized by a high level of seismicity and potential intensity (epicentral intensities to XII). The great earthquake series of 1811-1812 occurred in this zone, near New Madrid Missouri, about 325 miles from the site. The nearest approach of this zone to the site is approximately 220 miles. Notwithstanding its relatively great distance from the site, the New Madrid seismic zone is a potential source of ground motion at the

ENGINEERING REPORT NO. GONS-944054 PAGE 15 OF 50 r

REVISION 0 Grand Gulf site. Accordingly, all know earthquakes ofintensity V or greater which have occurred within the northern Mississippi Embayment were considered in establishing the Seismic Design Basis for the site.

The Uniform Building Code designates the vicinity of the site as Zone 0 on the map intitled. " Map of the United States Showing Zones of Approximate Equal Seismic Probability." The U. S. Coast and Geodetic Survey indicates Zone 0 as an area ofno earthquake damage.

The principal buildings and eructures include the containment structure, the turbine building, the auxiliary building, the cor of building, the diesel generator building, the standby service water cooling towers and basira, the enclosure building, the radwaste building, and the natural draft cooling tower.

These buildings and structures are founded upon suitable material for their intended application.

Structures essential to the safe operation and shutdown of the plant are designed to withstand more extreme loading conditions than normally considered in conventional non nuclear design practice.

The buildings and internal structures so designated are designed to provide protection as required from tornadoes, canhquakes, and the failure of equipment producing flooding, missiles and pipe whip.

The Containment Structure 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 design basis loss-of-coolant-accident and to provide a leaktight barrier against the uncontrolled release ofradioactivity to the environment, even assuming a partial loss of engineered safety features.

The Turbine Building houses all equipment associated with the main turbine generator and other auxiliary equipment. 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 part in the event of a postulated accident.

The Auxiliary Building is Seismic Category I structure that contains safety systems, fuel storage and shipping equipment and necessary auxiliary support system. 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.

The Control building 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.

The Diesel Generator Building is a Seismic Category I structure constructed of reinforced concrete.

The building contains the three diesel generators, three fuel oil day tanks, six starting air receivers

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ENGINEERING REPORT NO. GGNS-94 o054 PAGE 16 OF S0 REVISION 0 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 Enclosure Building is a limited leakage Seismic Category I stmeture'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 releases to the environment in the unlikely event of a loss-of-coolant-accident.

The Radwaste Building contains 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.

The Natural Draft Cooling Tower is a Concrete, natural draft, hyperbolic structure. The tower is designed 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.

The Ultimate Heat Sink is a system comprised of two separate, Seismic Category I, mechanical draft cooling tower /pumphouse/ basin stmetures. 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 intake louvers in the sides and contains ceramic tile fill blocks 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 pit pumps and is provided with separate tornado missile protection walls on all sides and on the roof.

3.1.1.2 Seismic Input to Structures and Equipment The Operating Basis Earthquake (OBE) and Safe Shutdown Eanhquake (SSE) design response spectra defined the free field vibratory motions for the Grand Gulf site. The peak horizontal ground acceleration of the OBE has been established to be 0.075 g's and SSE has been established to be 0.15 g's based on the vibratory ground motion studies. The design spectra which were used for the plant seismic design were obtained by modifying Newmark's curves. In developing these spectra, variations in site conditions, foundation properties and effects of focal and epicentral distance from the site were considered.

Seismic Category I structures, systems, and components have also been designed to withstand the effects of vibratory motion of at least the operation basis earthquake in combination with other appropriate loads within allowable stress limits of applicable codes.

For the vertical direction of seismic motion, the corresponding design earthquake was taken to be two-thirds of that established for the horizontal direction at the site.

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ENGINEERING REPORT NO. GGNS-94 0054 PAGE 17 OF 50 REVISION O l

3.1.2 System Analysis

]

This section describes the process used in selecting the equipment for which the seismic adequacy was detennined.

3.1.2.1 Overall Approach The project used a Seismic Margins Assessment (SMA) for the means to investigate the seismic external event of the IPEEE. The project identified a preferred and an alternate =ce= path based on operational and systems considerations for GGNS which were originally based on the PRA model developed for the GGNS IPEm. Once the front-line systems were identified, the systems required to support the operation of the front-line systems were identified. The components that must function in order for each system to work were then identified. The Safe Shutdown List (SSEL) was developed which contains the listing of the required components. This list was developed by GGNS engineers and reviewed by both an outside consultant and an independent reviewer.

Development of the Shutdown Path Logic Diagrams (SPLD) and the SSEL is documented in Engineering Report GGNS-93-0017, Revision 0 (17).

The SPLD's were developed to identify the systems that must function to successfully shutdown and cool the reactor following the occurrence of a RLE.

3.1.2.2 Assumptions The following assumptions for the development of the SPLD apply to the Seismic IPEEE are:

1. Offsite power is assumed to be failed due to the SME (Seismic Margin Event) and unrecoverable during the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period ofinterest. Consider possible adverse effects if offsite power is not lost.

2. Path success is defined as the ability to achieve and maintain a stable hot or cold shutdown condition for at least a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period following the seismic event.

3. Assume a small break loss of coolant accident (LOCA) equivalent to 1" diameter.

4. With regard to success path and equipment selection the following apply:

Nonseismic caused unavailability of component or systems is not explicitly addressed but single train systems are to be treated with caution.

Identify primary and alternate Safe Shutdown Success paths.

Analyze redundant equipment where present in path (NUREG 1407).

Highlight non redundant equipment in path (NUREG 1407).

Use NUREG/CR-4826 screening approach for single train / multiple train systems (NUREG 1407).

To extent possible, select an alternate path involving different systems, piping runs, components, from the preferred success path (NUREG 1407).

.l ENGINEERING REPORT NO. OGNS-94-0054 PAGE 18 OF So REVISION 0

5. Verify that postulated operations are viable (NP 6041).

6. Equipment for Seismic Evaluation Includes:

Active and passive fluid mechanical components in the safe shutdown success paths Electdcal equipment in the safe shutdown success paths

7. Equipment exempt from seismic evaluation includes:

Check valves Valves with external operators that do not change state.

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8. No relay evaluation for GGNS 3.1.2.3 Principal Safety Functions B

The first step in the development of the SPLDs is to define the safety functions that must be accomplished to achieve and maintain a stable shutdown. The four safety functions are identified in EPRI NP-6041, These functions are reactivity control, reactor coolant system pressure control, reactor coolant inventory control and decay heat removal. The GGNS IPEm lso identified safety a

functions that must be accomplished to successfully mitigate the events analyzed in the IPE. The initiating events ofinterest for this evaluation are Loss of Offsite Power (LOSP) and small LOCA.

The safety functions are equivalent to the safety functions from EPRI NP-6041 except for the eady containment over pressure protection function for the small LOCA initiating event. This function is accomplished by the successful operation of the vapor suppression system. The vapor suppression system consists of the weir wall inside the drywell, the drywell to suppression pool vents and the suppression pool. These components are all passive in nature and need only maintain their I

structural integrity in order to accomplish the function. These will be reviewed as part of the containment structural review. Therefore, the four primary safety functions identified above will I

form the basis for the identification of the front-line systems for the safe shutdown paths.

l 3.1.2.4 System Success Criteria The systems that can accomplish each of the primary safety functions were defined. During the performance of the GGNS IPE, combinations of systems that are required to successfully function to successfully accomplish each safety function were identified. The systems that can successfully perform each safety function are summarized below.

Reactivity Control Reactor Protection System and Control Rod Drive System Alternate Rod Insertion (and CRD System) and Reactor Pump Trip Manual Rod Insertion (and CRD System) and Reactor Pump Trip Standby Liquid Control System and Reactor Pump Trip j

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1 ENGINEERING REPORT NO. GGNS-94 0054 PAGE 19 OF 50 REVISION 0 i

Reactor Pressure Steam Line Safety Relief Valves Control Power Conversion System (MSIVs and Condenser)*

1 Reactor Inventory Feedwater*

Control High Pressure Core Spray System #

Reactor Core Isolation Cooling System #

l Control Rod Drive System (injection mode)

Low Pressure Core Spray and Depressuization with at least 4 SRVs#

Low Pressure Core LJection and Deprammintion with at least 4 SRVs#

. Condensate and Depressuization with at least 4 SRVs*

i Standby Service Water Crosstic to LPCI and Depressuization with at least 4 SRVs '

Firewater and Depressuization with at least 4 SRVs Suppression Pool Makeup Decay Heat Power Conversion System

• Removal Suppression Pool Cooling Mode of RHR Containment Spray Cooling Mode of RHR Shutdown Cooling Mode of RHR Containment Venting

• Not available with Loss of Offsite Power

1. Suppression Pool Makeup required for these systems if LOCA The above systems also require various support systems for successful operation.

3.1.2.5 Overall Success Path Logic Diagram With the identification of the primary safety functions and the systems that can accomplish those safety functions an overall success path logic diagram (SPLD) can be developed for GGNS. A SPLD is a graphic representation that shows the combinations of systems whose successful operation will result in long term shutdown following the seismic margin earthquake. It can be envisioned as a simple electrical circuit diagram constructed in a series-parallel fashion. The Seismic margin earthquake is depicted as the node on the left, and the desired long-term safe shutdown condition is depicted as the node on the right. Between these two nodes are a number of system blocks arranged in a series-parallel manner, showing alternate paths of achieving the safe shutdown condition. The selected paths must represent paths that the control room operators will j

use based upon their training and procedures. Adequate instrumentation must also be available to l

the operators.

An overall SPLD for GGNS is provided in Figure 11.1.

This SPLD was i

constmeted using the success criteria for a Loss of Offsite Power and Small LOCA initiators as discussed above. Note that some of the systems that are capable of being utilized are not included in the GGNS overall SPLD since they are not the systems that operators would preferentially use.

It should be noted that support systems are also not included in Figure 11.1. Dependencies l

between front line systems and support systems are identified in Table 10.1.

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e ENGINEERING REPOP.T NO. GONS-94 o054

- PAGE 20 OF $0 REVISION O The Srst node on the Overall SPLD is the function of Reactivity Control. This node consists of two parallel paths. One path is made up of the control rod drive system which works in conjunction with the reactor protection system. This path represents the insertion of control rods into the core in response to an automatic scram signal generated by RPS to shutdown the reactor.

An automatic scram signal can be generated by any ore of several results of a seismic event including the loss of offsite power. The second paralle. path is made up of the standby liquid control system (SLCS) block. This block represents the injection of sodium pentaborate into the reactor coolant system such that the reactor is shutdown. The SLCS is only actuated manually and must be actuated quickly. EPRI NP-6041 recommends that SLCS not be relied upon H-of assumed stress levels on the operators during a seismic event. Even though GGNS does not completely agree with this assumption because of the strength of the emergency operating procedures (EOPs), the high degree of operator training and the culture of adherence to procedure, this assumption will be maintained for this analysis. Therefore, no credit for the SLC Sytem will be taken in the preferred and alternate path selections. The other methods of reactivity control are not included sense they rely on the CRD system also.

The second node in the SPLD represents the function of Reactor Pressure Control. This block represents the opening and closing of the SRVs to control reactor pressure. Because of the assumption ofloss of offsite power no credit is taken for the power conversion system as it will' lead to the isolation of the main steam isolation valves aed prevent use of the condenser to control pressure.

The third node of the SPLD represents the function of Reactor Inventory Control. This node consists of two primary parallel paths. One path represents inventory control with high pressure systems and the other represents inventory control with low pressure systems. The high pressurs path consists of two blocks in parallel. The top block represents the injection of water into the core at high pressure by the RCIC system. The lower block represents the injection of water into the core at high pressure by the HPCS system. Other high pressure inventory control methods (feedwater and CRD injection) are not included because of unavailability because of the basic assumptions or low capacity. The second path represents inventory control with low pressure systems and consists of two parts. The first part consists of the Automatic Depressurization System (ADS) block. This block represents depressurization of the reactor using the SRVs. Note that the EOPs direct the operators to inhibit automatic depressurization.

Therefore, depressurization is only performed manually since the operators will follow the EOPs.

Depressurization can be accomplished through the use of any of the SRVs but only the ADS valves are credited in this analysis. Depressurization is always required for use of the low pressure systems for inventory control. This is true even with the assumption that a small LOCA exists since the assumed break size is too small to depressurize the reactor in sufficient time to allow low pressure systems to inject prior to core damage. The top block represents the low pressure coolant injection (LPCI) mode of RHR. The bottom parallel path represents the low pressure core spray (LPCS) system. Other methods oflow pressure inventory control are not included on the SPLD.

Condensate would not be available because of the assumption of loss of offsite power. SSW crosstie to LPCI and firewater could be used but SSW crosstie is a lower priority system for injection in the EOPs and firewater would only be effective after inventory had been maintained by some other system for a period of several hours. The final block, which is in series with both

ENGINEERING REPORT NO. GGNS-94-0054 PAGE 21 OF 50 REVISION 0

. parallel paths for reactor inventory control represents the suppression pool makeup (SPMU) system. This system is required only if a LOCA is assumed. It is necessary because of the loss of inventory from the suppression pool to the drywell with a LOCA inside the drywell. Systems such as LPCI and LPCS which take suction on the suppression pool could loose not positive head if this inventory loss if not made up. Both RCIC and HPCS can take' suction from the condensate storage tank in addition to the suppression pool. However, the condensate storage tank is not credited for this analysis.

The fourth and final node of the SPLD represents the function of Decay Heat Removal. This node consists of two parallel success paths. The top path represents the removal of decay heat using the suppression pool cooling (SPC) mode of the RHR system. T' e lower block represents the decay n

heat removal with the shutdown cooling (SDC) mode of the RHR system. Both of these modes i

utilize the RHR heat exchangers. Note that for the SPC mode to work, decsy heat from the core must be rejected to the suppression pool through the SRVs. This also requires a continued means i

of making up inventory to the reactor vessel. Other methods of decay heat removal not included on the SPLD are the containment spray mode (CS) of RHR and containment venting. Containment venting is not included since this method requires instrument air which may not be available during a loss of offsite power. CS mode is not included because its use is lower in priority to SPC and SDC in the EOPs.

i 3.1.2.6 Primary and Alternate Success Path Selection The methodology requires the identification of two independent paths, a preferred and alternate path. The alternate path is selected to include equipment or a backup train of equipment so that the plant can be shut down in the event of an active failure or unavailability of a single item of equipment in the preferred path. In selecting the preferred and alternate shutdown paths, emphasis was placed on compatibility with plant procedures, operator training on the use of the systems, system unavailability, and minimizing the number of components required for performance of the plar.t walkdown and verification. In addition one of the success paths must be capable of mitigating a small break LOCA.

3.1.2.6.1 Preferred Path Selection The preferred path consists of the control rod drive system, SRV division 1, RCIC, and RHR A in suppression pool cooling mode. Figure 11.2 shows the preferred safe shut down path.

In this case if a Seismic Margin Earthquake (SME) occurs coincident with a loss of offsite power, the reactor protection system will provide a SCRAM signal and the CRD system will provide the motive force for driving in rods. (Since the RPS is a de-energize to SCRAM system, any loss of power to the RPS will SCRAM the reactor.) If a loss of offsite power does not occur, a manual SCRAM can be performed. The control room operators verify that the reactor is shut down following the SCRAM signal.

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j ENGINEERING REPORT NO. GONS-94-0054 PAGE 22 OF SO REVISION O Following a loss of offsite power Safety Relief Valves will be required to open in response to the pressure transient resulting from the closure of the turbine control valve and MSIVs. The MSIVs will close on loss of offsite power due to de-energization of the solenoids. If a loss of offsite power does not occur, the MSIVs can be manually closed or allowed to close on Low Steam line Pressure. Division 1 (actuation and control logic) of the SRV is the success front line system for the reactor pressure control function in this path.

The RCIC system will automatically start on a low reactor water level signal (Level 2). AcwrOg to procedure, operators will verify that the RCIC system is operating and providing inventon makeup. RCIC stops injection on high water level (Level 8) to prevent water carry over to the RCIC turbine. No manual actions are required for restarting injection since the system will 1

automatically restart injection on low water level (Level 2). However, the operators will manually control RCIC in accordance with Emergency Procedures to prevent the cycling of RCIC injection.

Note that Suppression Pool Makeup (Train A) is required for this path so that the inventory loss from the LOCA to the drywell can be made up. With out this makeup, NPSH for the RCIC and RHR pump could be lost.

j The RCIC system, due to poor turbine-driven pump reliability tends to have a relatively high unavailability. EPRI NP-6041-SL recommends that two redundant systems be required for the success path in cases such as these. HPCS is also capable ofinjecting high pressure water to the reactor vessel although with a much higher flow rate than RCIC. The HPCS system will also j

automatically start on a low reactor water level signal (Level 2) and isolate at high water level (Level 8). While RCIC is the preferred source because it is more easily controlled, HPCS is an alternate source of high pressure injection to the vessel which is capable providing adequate cooling to the reactor core..

I RCIC (or HPCS) will be used to maintain inventory while the SRVs will maintain reactor pressure, either automatically at the set relief pressure or manually in accordance with emergency procedures at a lower pressure range. RHR train A in SPC cooling mode is used to remove the decay heat from the suppression pool and thus protect the containment. This is performed manually in accordance with emergency procedures when the pool temperature increases because of the discharge from the RCIC turbine exhaust and/or the SRVs.

This path is capable of bringing to plant to hot shutdown.

3.1.2.6.2 Alternate Path Selection The alternate path consists of the control rod drive system, ADS Division 2 (actuation and control logic), LPCI C, and RHR B in shutdown cooling mode. Suppression Pool Makeup (Train B) is not required as no LOCA is assumed for this path. Figure 11.3 shows the alternate shutdown path.

A reactor SCRAM will be initiated as described for the preferred path. The EPRI Seismic Margins methodology recommends against using the Standby Liquid Control (SLCS) as a method for reactivity control because of the added stress on the operator. Initial pressure control is also similar to the preferred path except that Division 2 of SRVs is credited for this path. Since no high

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ENGINEERING REPORT NO. GGNS-94-0054 PAGE 23 OF 50 REVISION 0 pressure makeup systems are credited following the scram, the reactor inventory will be reduced as the SRVs are opened. Once reactor level reaches Level 1, ADS will actuate (as long as drywell pressure is high or the ADS bypass timer has timed out) to lower reactor pressure and LPCI C will commence injection when reactor pressure is low enough.

The operator will control the depressurization process manually in accordance with emergency procedures..

By the time water level is restored and stabilized, reactor pressure will be below the setpo*mt for use of shuts wn cooling for decay heat removal. Operator action will be required to manually open valve E12F008 as this valve is power from Division 1. This will allow the reactor to be brought to cold shutdown conditions.

3.1.2.7 Systems Used in Success Paths Table 10.1 shows a matrix of the front-line systerr.s used in the preferred and alternate success paths. Table 10.2 is a matrix showing front-line to support system dependencies. Table 10.3 is a matrix showing support system to support system dependencies.

3.1.2.7.1 Front-line Systems Descriptions The specific GGNS systems that are available to perform the safety functions identified in Section 3.1.2.4 are shown on the success path logic diagram shown in Figure 11.1. These systems are categorized as front-line systems as they perform a direct safety function. These front line systems are described below.

l Control Rod Drive (CRD) System - When a scram is initiated by the reactor protection system l

(RPS), the CRD system inserts the negative reactivity necessary to shut down the reactor. Each l

control rod is controlled individually by a hydraulic control unit. When a scram signal is received, high pressure water from an accumulator for each rod forces each control rod into the core.

l l

Standbv Liquid Control (SLC) System - The SLC system is an alternate means ofinserting negative l

reactivity into the reactor. The SLC system serves as an emergency backup system to the control rods and is used if the CRD system fails. The system injects a sufficient quantity of neutron l

absorber into the reactor to ensure that the reactor is shutdown.

Pressure Control System - A pressure control system consisting of safety relief valves mounted on the main steam lines is provided to prevent excessive pressure inside the nuclear steam system following either normal operations, abnormal operational transients, or accidents. These safety relief valves open automatically at specific pressure settings to ensure that the reactor coolant system is not over pressurized. The valves relieve steam from the reactor vessel into the suppression pool.

RCActor Core Isolation Cooling (RCIC) System - The RCIC system is designed to provide core cooling in the event of a loss of feedwater flow or loss of cooling in the condenser and to cool down and depressurize the plant to the point wh ;re the shutdown cooling mode of RHR can be

ENGINEEfUNG REPORT NO. GGNS-94-0054 PAGE24 OF S0 REVISION O l

utilized. RCIC uses steam from the reactor to pump water into the reactor vessel to maintain coolant inventory without the use of Emergency Core Cooling Systems. RCIC is capable of providing sufficient coolant for small LOCAs also.

The GGNS IPE identified a potential enhane>=*at concerning the RCIC system. It involves a procedural change to allow operators to bypass RCIC isolation on high eteam tunnel temperature due to a loss of steam tunnel cooling. As recommended in the GGNS IPE, this enhancement is still being pursued.

Hlah Pressure Core Sprav (HPCS) System - The HPCS system provides and maintains an adequate coolant inventory inside the reactor vessel to prevent fuel clad melting as a result of postulated small breaks in the reactor coolant system. The HPCS system will supply makeup water to the reactor if the RCIC system fails to operate or if a LOCA is large enough that RCIC does not have sufficient capacity to maintain inventory.

Automatic Depressurization (ADS) System - The ADS is provided to automatically depressurize the reactor coolant system so that flow from any of the low pressure ECCS can enter the reactor vessel in time to cool the core and limit fuel cladding temperature if the HPCS does not operate.

The ADS system consists of eight safety relief valves with redundant actuation capability. The system also includes accumulators and air receivers that maintain ADS capability beyond 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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

Removal of decay and sensible heat from the reactor coolant system during and after shutdown.

Removal of stored and decay heat from the reactor coolant system following a design basis loss ofcoolant accident (LOCA).

Removal of heat from containment following a LOCA in order to limit the increase in containment pressure. This is accomplished by cooling and recirculating the water in the j

suppression pool. This function can also be accomplished while spraying the containment.

Redundant trains of RHR are utilized in the preferred and alternate paths. The assumption is that these trains have similar seismic strength. This assumption will be verified during the walkdowns.

Low Pressure Coolant Injection (LPCI) - The LPCI is an operating mode of the RHR system.

LPCI uses the pump loops of the RHR system to inject cooling water at low pressure from the suppression pool into the reactor. LPCI is actuated by conditions that detect a break in the nuclear steam system, but water is delivered to the core only after the reactor vessel pressure is reduced.

LPCI operation provides the capability of core reflooding following a LOCA in time to prevent fuel clad melting. There are three independent LPCI trains.

l Suppression Pool Coolina (SPC) - The suppression pool cooling mode of RHR is placed in 1

operation to limit the temperature of the water in the suppression pool following a design basis LOCA. In this mode, the RHR pumps take suction from the suppression pool and pump the water i

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ENGINEERING REPORT NO. GGNS-94 0054 1

PAGE 25 OF 50 REVISION 0 through the RHR heat exchangers where cooling takes place by transferring heat to the Standby Service Water (SSW) system. The water is then returned to the suppression pool. There are two independent SPC trains.

h*Aawn Cantino Made (SDC) - The shutdown cooling mode is a Aielaa of the RHR system and is placed in operation during a normal shutdown and cooldown. Reactor cooldown is completed by pumping reactor coolant with the RHR pumps from one of the recirculation loops through the residual heat removal heat exchangers where heat is transferred to the SSW. The '

reactor coolant is then returned to the reactor vessel via feedwate. system inlet lines. Except for a common suction line, there are two independent SDC trains.

Sunoression Pool Makeup (SPMin System - The suppression pool makeup system is designed to provide sufficient water from the upper containment pool to the suppression pool to keep the drywell vents covered following a LOCA. This is necessary because of suppression pool inventory loss through the reactor coolant system break to the drywell.

3.1.2.7.2 Support Systems Descriptions In order for the front-line systems to perform their required safety functions, systems categorized as support systems must also remain operable during and after the SME. Support systems provide a support function to front-line systems and/or to other support systems. Plant specific support systems for GGNS are shown in the front-line and support systems dependency matrixes (Tables 10.2 and 10.3). The primary support systems are described below.

Standby AC Power Systemt - GGNS is designed to shutdown safely and maintain a safe condition on complete loss of offsite electrical power. Standby AC power is supplied by independent and redundant diesel generators to provide power to systems and components required for safe reactor shutdown. Two identical, independent diesel generators provide power to Divisions 1 and 2 ESF 4

power distribution system. A third independent diesel generator provides power to the Division 3 power distribution system which powers the HPCS pump and associated equipment DC Power System - In the event that normal plant power sources become unavailable, the DC power system provides power for controls or components required for safe reactor shutdown. The ESF DC power system is divided into three divisions. Each division consists of two battery chargers (Division 3 only has one), a bank of batteries and appropriate distribution panels. In the 1

case of a complete loss of electrical power, the batteries can continue powering the DC system for a period of time.

Standbv Service Water (SSW) - The SSW system is designed to provide a reliable source of cooling for plant systems and components that are essential to safe shutdown of the plant. The SSW system supplies cooled water from the SSW cooling tower basin to various components and systems and then returns the water to the SSW cooling tower. The system consists of three different loops, each with a pump and associated valves and piping.

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l ENGINEERING REPORT NO. OONS-94 o0$4 PAE M W M l

REVISION O Varians HVAC Sv*mn - Equipment area cooling and ventilation systems are provided to maintain the local environment of specific areas at temperatures with design allowable operating ranges for electrical and mechanical components located within different areas The systems indvie the Standby Service Water Pump House Ventilation system; the Diesel Generator Room Ventilation system, the ESF Switchgear Room Coolers, the Switchgear and Battery Room HVAC system; the Control Room HVAC system; and, various ECCS Room Coolers. Those systems above that l

require a cooling water system rely upon Plant Service Water during normal operation but switch j'

to SSW following an accident. Many of these systems were not included in the GGNS IPE.

Because of the assumption of the need to maintain the shutdown condition for a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period, it was necessary to add these systems.

3.1.2.8 Equipment Identification and Selection i

GGNS engineering developed the Safe Shutdown Equipment List (SSEL) for the IPEEE including the requirement of USI A-45 " Shutdown Decay Heat Removal Requirements" and the seismic induced fire and flooding issue for the IPEEE. Redundancy, reliability, path independence, and highly successful operational sequences are attributes of the IPEEE SSEL for the four required safe shutdown function; reactivity control, reactor coolant pressure control, reactor coolant inventory control, and decay heat removal. The additional requirements imposed upon the IPEEE success paths (consideration of SBLOCA, consideration of containment isolation and cooling requirements, and USI A-45) were factored as additional requirements to the SSEL alternatives.

The objective of this task was to determine the optimum safe shutdown alternatives for achieving a safe shutdown condition during a seismic event at the SSE level. Optimum safe shutdown alternatives are defined as safe shutdown paths that not only meet the required safety functions to bring the plant to safe shutdown, but also are: a) reliable; b) seismically rugged at the SSE levels; and, c) involve operational sequences where operator actions are likely to be highly successful.

The equipment identified for seismic evaluation included:

Active mechanical and electrical equipment that operates or changes state to accomplish a safe shutdown function.

Active equipment in systems that support the operation ofidentified safe shutdown equipment (e.g., power supplies, control systems, cooling systems, lubrication systems).

Instrumentation needed to confirm that the four safe shutdown functions have been achieved and are being maintained.

Instrumentation needed to operate the safe shutdown equipment.

Tanks and heat exchangers used by systems on the SSEL.

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ENGINEERING REPORT NO. GGNS-94 0054 '

PAGE 27 OF $0 REVISION 0 The following equipment types were not identified for seismic evaluations:

Equipment that could operate but does not need to operate and that, upon loss of power, will fail in the desired position or state. This type of equipment is defined as passive for the SMA.

Passive equipment such as piping, filters, and electrical pere.non assemblies.

Note that although piping was not explicitly included in the SSEL, selected piping was examined during the walkdowns.

Self-actuated check valves and manual valves.

Major items of equipment in the nuclear steam supply system, their supports, and components mounted on or within this equipment such as the reactor pressure vessel, reactor fuel assemblies, reactor internals, control rods and their drive mechanisms, reactor coolant pumps, steam generators, pressurizers. and reactor coolant piping.

The performance of the SSEL identified 567 components which were the focus of the seismic walkdowns.

3.1.3 Seismic Walkdown This section describes the approach to the walkdown, the screen riteria used and the details of the walkdown.

3.1.3.1 Approach The key element of a reduced-scope evaluation is the plant walkdown. The approach used to perform the systems and element selection walkdown, and the seismic capability walkdown follows the recommendations of EPRI NP-6041. This includes the following parts.

Selection of the assessment team Pre-walkdown preparation Systems and element selection for walkdown Seismic capacity walkdown The assessment team, all part of the Seismic Review Team (SRT), was made up of eight members Six of the members were Entergy Operations Personnel, three of those possessing the qualification requirements of EPRI NP-6041. The remaining two members were outside consultants, also possesing the qualification requirements of EPRI NP-6041. The following is a listing of the SRT members, their affiliation and area of expertise.

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ENGINEERING REPORT NO. GGNS-94-0054

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PAGE28 OF $0 REVISION O Mr. Kyle Grillis Entergy Operations Operations Mr. Mark D. Locke Entergy Operations Seismic Capability Engineer Mr. Joseph D. Malara Entergy Operations Seismic Capability Engineer Mr. James Owens Entergy Operations L*w% and Operations Mr. Amir Shahkarami Entergy Operations Seismic Capability h6 Dr. John D. Stevenson Stevenson and Associates Seismic Capability Engineer Mr. Mike Sweeney Entergy Operations Operations Mr. George G. Thomas Stevenson and Associates Seismic Capability Engineer Prior to the walkdown, a detailed product plan was prepared, ielad% the detailed technical approach of each task and the interfaces between the team==hm and GGNS personnel. In -

preparation for the walkdown, data assembly and evaluations were performed to derme a technical baseline for the systems analysis and seismic screening walkdown. Two independent teams were used during the walkdown process.

Specific documentation assembled and evaluated prior to and during the walkdowns included:

[

GGNS Safe Shutdown Paths and Equipment Lists plant arrangement drawings sections of the GGNS UFSAR relating to the seismic criteria and licensing basis for the plant ground response spectra for the SSE floor response spectra and how they were generated a sample of construction details of the anchorage including drawings and specifications a sample of procurement and seismic testing specifications for equipment I

examples of calculations for seismic and anchorage qualifications design basis documents for the GGNS structures selected evaluations for block walls design calculations for a sample oflarge flat bottom tanks The structures at GGNS were screened generically. The drawings and analysis models were reviewed for details that might indicate seismic vulnerabilities in accordance with the requirements of a reduced scope SMA. The drawing and structural analysis reviews confirmed that consistent good practice in design detail and analysis was implemented at GGNS. Therefore, it was not necessary to review more than a small sample of the details of connections, reinforcement bar placement, construction joints, etc., to make the judgments on screening.

3.1.3.2 Screening Walkdown For the SMA at GGNS, rigorous statistically based sampling criteria were neither practical nor desirable. The SMA procedures and guidelines used were heavily reliant on the judgment of the highly experienced engineer and criteria for sampling in this plant likewise are modeled around this judgment.

H ENGINEERING REPORT NO. GGNS-94 0054

' PAGE 29 OF SO REVISION 0 There are two areas of a reduced scope SMA where sampling was applicable and used at GGNS.

They are; l) screening of structures and components, and 2) walkdowns. Issues that influence the sampling are; redundancy provided by multi-train systems, similarity in design and location of redundant trains, treatment of single failures, access to components dudng walkdowns, and systems interactions potential including fire and internal flood sources.

l The sampling approach described below are appropriate for modern plants of GGNS's vintage. The i

h=at review and walkdown verified that uniform practices in accordance with the plant design '

basis for construction, design and installation were implemented. Therefore, the sampling approach j

was used throughout the effort.

During the GGNS walkdown, it was confirmed as expected thst most items in a given equipment class were either identical or very similar. The plant documentation review and walkdown confirmed that the vast majority of equipment was manufactured, and installed as specified. The.

screening procedures used at GGNS for generic categories of equipment and structures contained caveats or inclusion rules that were checked during the ::alkdown. Since the equipment at GGNS was purchased and installed to similar codes and standards the SRT screened generic classes of-equipment on the basis of their relative ruggedness. The screening sampling size for identical or very similar equipment in a given class for caveats was one or greater for each walkdown team.

The screening size for very similar equipment ir. given class with identical or very similar anchorage was two or greater for each walkdown team. The incret. sed sample for anchorage is based on experience at other plants that anchorage installations are not always consistent. This is consistent with the guidance given in Appendix D of EPRI NP 6041-SL. A 100 percent " walk-by" of all equipment on the SSEL was employed to check for unique equipment details and for seismic interactions.

Distribution systems that were installed in bulk such as piping, cable trays, HVAC ducting, electrical conduit and instrument lines were screened generically aAer completion of a walkdown with verification that the distribution systems meet the inclusion rules. It was confirmed that the design and installation practice at GGNS are consistent, therefore the screening judgment was based upon a review of the general specifications and drawing for a single run of each generic class of distribution system, i.e., a sample size of one per generic class. As expected the review of the general specifications and drawings did not indicate significant differences in design and installations practice.

3.1.3.3 Walkdowns Two walkdowns were performed. An off-line outage walkdown was performed during the fall of 1993 (October 12,1993) that included equipment located in the drywell and on on-line walkdown was performW during the summer of 1993 (August 26-31,1993). The structures and distribution system review was performed during the course of the on-line walkdown.

ENGINEERING REPORT NO. GONS 94 0054 PAGE 30 OF SO REVISION 0 A typical day during the walkdown consisted of:

reviewing issues identified on previous days for determination as to whether the item is screened or an outlier planning the day's walkdown effort performing the walkdown 1

briefing GGNS Design Engineering on the day's progress 3.1.3A Seismic Analyis Results The seismic walkdowns found GGNS is seismically nigged and that there were no outliers affecting plant operability. However, there were several equipment items that could not be initially screened

^

Some of these items are candidater for voluntary design enhancements by Entergy, and some of these equipment items are potential outliers. Final disposition of these items have been d&==*ed and completed.

A complete listing of the items not initially screened during the walkdowns and the resolutions is included in the IPEEE Reduced Scope Seismic Marghs Assessment (SMA).0) 3.2 Conclusions of Seismic Analysis The conclusions of the seismic analysis is that Grand Gulf Nuclear Station is seismically rugged and that all components identified in the Safe Shutdown Path have adequately considered the seismic input.' All anchorage to these components was foi. d to be rugged.

Only one potential vulnerability to a seismic event was identified, which has Leen corrected.

4.0 INTERNAL FIRES ANALYSIS LATER.

5.0 HIGH WINDS. FLOODS. AND OTHERS 5.1 High Winds

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The Updated Final Safety Analph Report (UFSAR), Safety Evaluation Report (SER) and other i

pertinent design and licensing documents weit teviewed in order to fmd the site specific hazard data and licensing basis for high winds and tornadoes. The findmgs of this review were coir.pered to the criteria of the 1975 Standard Review Plan (SRP)W so that possible differences could be identified. To ensure that any possible significant changes to the reviewed docnents were i

included, recent weather data for the region was compared with data used to formulate the existing high winds hazard basis. Additionally, a review ofcurrent site dmwings and a walkdovm of the area was perfonned to determine if there are any potential vulnerabilities not included in the original design basis analysis (Reference - Engheering Report No.: GGNS-93-0048, Rev. 0)(18),

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l ENGINEERING REPORT NO. GGNS-94 0054 -

PAGE 31 OF SO

' REVISION 0 i

At Grand Gulf Nuclear Power Station (GGNS), Seismic Category I structures are designed for the extreme wind and tornado phenomena. The minhnam design wind velocity for these structures is 90 :nph at 30 feet above ground for a 100 year recurrence interval. This wind speed is the most conservative value based on SRP specified reference codes as wdl as site and regional meteorology data. A review of new meteorology data for the period aAer the original evaluation revealef. hat this existing parameter still represent the most conservative value. The procedures that were used l

to transform the wind velocity into pressure loadings on the structures and the===wiatad vertical distribution of the wind pressures and gust factors are in accordance with ASCE Paper # 3269(24) and ANSI Standard A58.l(25) and were acceptable as determined in the NRC review per section 3.3.1 of the SER. The tornado loadings are calculated on the basis of a maximum wind velocity of 360 mph which is the vector sum of a maximum peripheral rotational velocity of 290 mph and a translational velocity of 70 mph. The maximum design pressure drop is 3 psi with a maximum rated change of 2 psi /sec. The radius from the center of the tornado at which the maximum wind velocity occurs is 150 ft. These parameters conform to those given in Regulatory Guide 1.76(26). The methods employed to convert tornado loadings into Ibrces and to distribute them across the structures conform to the requirements in SRP Section 3.3.2.

The Structures, Systems and Components which need protection from externally generated missiles I

as required by SRP Section 3.5.2 are, in general, protected from the tornado missiles in SRP Section 3.5.1.4, due to their location in or behind miujle-proof structures. These structures were designed using methodology that predicts local and overall damage, meeting the intent of SRP 1

Section 3.5.3. However, several isolated components (e.g. Standby Service Water return lines) are not provided this type of protection. A probabilistic evaluation was performed to determine the total annual frequency of occurrence for missiles of the types described in SRP Section 3.5.1.4 in striking these vulnerable areas. This evaluation determined this frequency to be 0.77 X E-8/yr.

(Reference - Calc. CC-Qllll-94004, Rev. 0)(19). This low frequency of occurrence is well below the screening criteria in IPEEE.

In summary this review concludes that GGNS meets the intent / criteria set forth in the 1975 SRP for the High Winds and Tornado Hazard. Some concerns were discovered, but they were evaluated and found to be within the acceptance criteria. No changes to the protection for High Winds and l

Tornado were identified or required.

5.2 Floods 5.2.1 Hydrologic Conditions and Existing Mood Protection Grand Gulf Nuclear Station is on the east bank of the Mississippi River near river mile 406, approximately 25 mi. south of Vicksburg, Mississippi, and 6 mi. northwest of Port Gibson, Mississippi. The site is located in the water resources planning area No. 7 of the lower Mississippi River region. It is bounded on the west by the Mississippi River and on the east by loessial bluffs.

i At the plant site, the river flood plain is about 60 miles wide with elevations ranging from 55 to 75 ft. Mean Sea Level (MSL) Flow is confined to a width of about two to four miles by high bluffs on the east bank and man-made levees on the west bank.

ENGINEERING REPORT NO. GGNS-94 0054

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PAGE32 OF 50 REVISION O i

Ofimmediate rdevance to the plant site are two small steep streams. Stream A, north of one site, is perennial and drains Basin A with an area of 2.8 square miles. Stream B, south side of the site, is intermittent and drains Basin B with an area of 0.6 square miles. Both streams drain into Hamilton i

Lake located in the flood plain of the Mississippi River. Stream A, receives most ofits water from i

the watershed outside the plant area and has a 12-foot culvert under the access road to connect it to the flood plain. Stream B receives most ofits water from the site and has a 15-foot ' culvert under the access road to carry local floods and site drainage.

The plant yard has an average nominal elevation of 132.5 ft. Finished floor elevations for plant at-1 grade spaces is 133.0 ft. The plant yard is graded to direct runoff away from the buildings, and toward Streams A and B via a combination of drainage swales, ditches, and overland flow.

Flood seals are installed on eleven doors: OC313, and OCTS in 'the control building; ID301, ID308, ID309, ID310, and ID312 in the diesel generator building; and IM110, IMI11, 2M110 i

and 2M111 in the wandby service water (SSW) pump houses. Seals, penetration sleeves, toe plates, and curbs are installed in the SSW pump houses to prevent water from reaching safety related equipment.

5.2.2 Original Design Basis Evaluations The safety-related facilities, systems, and equipment are capable ofwithstanding the worst flooding caused by a combination of several hypothetical events. These events are: probable maximum flood of the Mississippi River coincident with wind generated waves; seismic failure of upstream dams coincident with the.U. S. Army Corps of Engineers design-project flood; ice flooding; probable maximum flood of the two small streams adjacent to the plant; and flooding of the site due to PMP rainfall on the site watershed. A detailed discussion of the external flooding analyses related to the existing GGNS design basis is presented in GGNS UFSAR, Section 2.4 and in Engineering Report GGNS-93-000104 These analyses can be briefly summarized as follows:

(a)

Evaluation of the Mississippi River, and adjacent streams A and B indicates that floodwaters associated with their flooding do not result in inundation of the power block structures. Therefore, GGNS may be considered a dry site as defined in Regulatory Guide 1.102, and hydrostatic loading of external walls and structures due to inundation are not applicable. As a result, the only structural concern related to external flooding, is the potential for failure ofroofing system due to ponding during heavy precipitation.

Rainfall that falls on the roof of on-site buildings is collected by roof drains and discharged directly into a subsurface drainage systenowhich is designed for the 100-year rainfall.

Safety related roofing systems are fitted with ovenlow scuppers. Water flowing over these 1

scuppers falls to the ground at the side of the building and then flows over the yard surface i

by natural drainage. Roof structures for the SSW pump rooms, as well as the auxiliary, centrol, and diesel generator buildings, were evaluated to ensure that during the PMP

ENGINEERING REPORT NO. GGNS-94 0054 PAGE 33 OF SO REVISION 0 rainfall the roofing systems will not fail, even if the depth of roof ponding is assumed to j

exceed parapet height.

(b)

Drainage oflocal intense precipitation is evaluated in accordance with the criteria set forth in Regulatory Guide 1.59. Based upon an average time of concentration in the yard of j

J approximately 30 minutes, the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> PMP estimates from HMR No. 33, and the temporal distribution obtained from EM-Il10-2-1411; a rainfall intensity of 16.4 inches per hour is i

used.

The Rational Formula is used to estimate flows for performance of a backwater analysis to determine water surface elevations near plant structures. Maximum calmia**A water levels near primary power block structures may be taken as 133.20 A. MSL on' the cast side of the power block and 133.25 A. MSL on the west side of the power block; and could exceed elevation 133'-0" for about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. As a result, flood barriers requiring 6" of freeboard are maintained on all of the at-grade penetration seals in the SSW pump houses, diesel generator building, and control building. Leakage through remaining at-grade power block openings has been estimated and determined to pose no threat to the safe operation of the plant.

J 5.2.3 Evaluation of Revised Hazards Due to Flooding of the Mississippi River l

The Mississippi River PMF, and the U. S. Army Corps of Engineers' Design Project Flood on which it is based, have not ir. creased since the GGNS became operational. The 2 year recurring wind speed at GGNS has also remained the same. Slightly higher levee elevations exist, and future i

levee elevations of 106 A. MSL have been approved by the U. S. Army Corps of Er.g*r----ns.

Resulting PMF water levels including wind wave effects would be approximately 116 A. Mht, which is still well below the elevation of the plant yard.

5.2.4 Review of Mood Hazards Due to Precipitation over the Site Watershed Applying the revised criteria, the maximum site rainfall intensity for use in the Rational Equation is approximately 28.2 inches of rain per hour. Water levels in the plant yard, are determined using flow models similar to those used in the design basis analysis. In some instances models were altered to consider less conservative flow widths and friction losses, comb'med weir and culvert flow, and partial by-pass of obstructions. Plant improvements were assumed as 'mdicated in section 7.3.2 For a detailed description of the methods by which site water levels are determined see Engineering Report GGNS-93-0001 and calculations CC-Q1Y13-93001 through CC-Q1Y13-93003. Superimposing the average wave height onto the calculated water surface elevations, and comparing to the As-Built flood level protection the following summary is obtained:

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f ENGINEERING REPORT NO. GGNS-94 0054 PAGE 34 OF 50 REVISION 0 l

l Barrier Too Elevation (A. MSL)

WSEL (A. MSL)

Freeboard Door OCT5 134.25 134.15 0.10 feet Door OC313 134.25 134.15 0.10 feet l

Door ID301 134.25 134.21 0.04 feet l

Door ID308 134.25 134.21 0.04 feet Door ID309 134.25 134.21 0.04 feet Door ID310 134.25 134.21 0.04 feet Door ID312 134.25 134.21 0.04 feet Door IMI10 134.25 134.07 0.18 feet Door IMill 134.25 134.07 0.18 feet 1

Door 2Mi10 134.25 133.61 0.64 feet Door 2 Mill 134.25 133.61 0.64 feet Other, SSW A 133.62 134.07

- 0.45 feet Other, SSW B 133.62 133.61 0.01 feet Therefore, the current level of protection at all structures except the SSW A pump house remain

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adequate for the new rainfall criteria. Protective barriers around penetrations inside the SSW A structure are sheltered from the wind, and are therefore still adequate using the revised criteria.

I However, the barrier crossing the exterior equipment hatch would require additional protection to i

guard against the revised PMP.

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Leakage into the power block through unprotected openings was estimated based upon the max' mum static water level calculated near the power block and assuming gross failure of non-safety related roof structures. Additionally, the entire six hour PMP can be expected to enter the incomplete decking on the 185 foot level of the former unit 2 auxiliary Mu% In this event, water entering the power block, would be expected to flood the control, turbine, radwaste, and former unit 2 buildings to an elevation of approximately 99.11 feet MSL All safety related equipment in the control building which is essential in attaining and maintaining a cold safe' shutdown is located above elevation 111 n. and the auxiliary building is watertight up to elevation 114 R. The design basis circulating water line break is expected to flood these spaces to a height of 108 A without functional degradation of any equipment essential to attaining and --W :-% a cold safe shutdown. Therefore, flHing of the power block to elevation 99.11 A. is bounded by the circulation water analysis, c d no ingact on the ability to attain and maintain a cold safe shutdown will result.

5.2.5 Effects of the New PMP Data on Roof Loading The control, auxiliary, and diesel generator buildings, as well as the SSW pump houses, have been waluated for stmetural adequacy of their roofing systems, and application of either the design basis rainfall data or the new rainfall data can be expected to have no impact on the stmetural adequacy of these structures. Funher, due to the presence of roof drains and overflows, calculations demonstrate that ponded roof depths do not exceed the level at which water could propagate into the buildings through doors, vents, or other openings. Water depths on the auxiliary building roof could exceed the sill elevations on doors I A502 and 1 A504 for a short time. However, these doors are secondary containment (airtight) boundaries fitted with gaskets similar to those making up the

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1 ENGINEERING REPORT NO. GGNS 94 0054 PAGE35 OF $0 REVISION 0

.i PMP door seal assemblies. Therefore, little if any leakage through these doors is expected. The enclosure building roofis not expected to fait during the PMP and ponded depths are not ---ded to be large enough to result in leakage through roof openings due to the presence of roof drains i

and overflows. However, the roo6ng system'is not adequate to withstand ponding up to the parapet height. The overflows alone will prevent roof stresses from exceeding material yield i

stresses. However, the roof drains alone would not be adequate to prevent roof failure. Therefore, the roof drainage system and roof overflows must be relied upon to prevent roof failure.

5.3 Transportation and Neady Facility Accidents The analysis for transportation and nearby facility accidents was performed using Standard Review Plan 2.2.1 & 2.2.2 " Identification of Potential Hazards in Site Vicinity" sections from NUREG 75-087(6), Potential external hazards or hazardous material may be a threat to plant safety if accidents l

involving nearby industrial, military, and transportation facilities and routes would constitute a i

design bases event. These facilities and routes include air, ground, and water trafEc, pipelines, and fixed manufacturing, processing, and storage facilities were reviewed for location and separation i

distance. This analysis was divided into the following sections.

Industrial and military facilities.

Transportation facilities and routes.

Mississippi River accidents.

Significant changes i

5.3.1 Industrial and Military Facilities There is no extensive industrial activity around the Grand Gulf site. There are no military installations, chemical or munition plants, stone quarries, or major gasoline-storage areas located i

within 5 miles of the plant. The nearest military facility is about 100 miles away. The nearest industrial facilities that have significant quanities of stored chemicals are in Port Gibson about 4.5 miles away and a 4-inch gas line is about 4.5 miles east of the site. Based on the separation distances of the chemical storage facility and gas line, the safety of the plant will not be affected.

There are no commercial airports within 10 miles of the site, and no major air routes near the site, i

a 5.3.2 Transportation Facilities and Routes There are two county roads near the site that carry local traffic. The nearest major highway is U.S.

61 which passes within 4.5 miles east-southeast of the site. The Natchez Trace Parkway is located about 6 miles southeast of the site. The nearest railroad carrying hazardous material is 30 miles south of the site. Based on the separation distances, we conclude that accidents along these routes j

will not affect the safe operation of the plant. A review of the 10-15-92 edition of the oil and gas map for Mississippi showed that no new pipe lines have been installed within 5 miles of the plant.

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ENGINEEIUNG REPORT NO. GGNS-94 0054 PAGE 36 OF 50 REVISION O 5.3.3 Mississippi River Accidents Mississippi River traffic pose no significant hazard to the plant. The Mississippi River passes 1.34 miles west of the site at its closest point. The plant is protected from the consequences of accidents on the river by distance and a 65 foot earthen east river bank. The effects of toxic clouds r@ias from barge accidents has also been analyzed and taken into account. The risk to control room operators is acceptably low due to the low concentration of the chemical upon reaching the control room or the low probability of an accident which falls within the guidelines established in Standard Review Plan section 2.2.3. There is no danger of ships or barges damaging intake structures or of corrosive liquids being drawn into the plant from an accident of the river since there are no intake i

structures.

5.3.4 Significant Changes A review was performed to determine if any significant changes have occurred from that reported in the UFSAR. The following is a.*iscussion of some of the findings. Highway 61 is in the process of becoming a 4-lane highway. The additional two lanes are being added on the east side of the existing highway. This activity required the relocation of the existing 4" natural gas transmission line to just west of the highway's right-a-way. Therefore, the closest distance of this highway to the plant site will not change. Additionally, the change from a two lane highway to a four lane highway should help reduce the accident rate on the highway. The Port Facility south of the plant has been built but at the present is only being used by local fishermen and there is no planned additional industrial development for the post at this time. Southern Cotton Oil (formally Port Gibson Oil Works) has replaced the two 12,000 gallon Hexan tanks with one 15,000 gallon Hexan tank which has a leakage monitoring system. The railroad line between Port Gibson and Vicksburg including the spur line to the plant has been removed.

5.3.5 Conclusion On the basis of the information provided the Grand Gulf facility is protected and can be operated with an acceptable degree of safety considering the activities at nearby transportation, industrial, and military facilities.

5.4 Others An assessment of the core damage risk from lightning initiators at Grand Gulf Nuclear Station was performed in support of the GGNS IPEEE effort, even though such an assessment was neither a requirement nor lightning an initiating event of special concern The Level 1 PSA model was used with updated GGNS-specific initiating event data, including recent lightning-initiated plant scrams to evaluate the core damage risk from lightning initiators (Reference 14). This report is in agreement with the conclusions made by the NRC, that the risk I

frorr. lightning-initiated severe accidents is insignificant, and there is no need to address lightning as a separate extemal event initiator in responding to the IPEEE generic letter supplement.

c.

ENGINEERING REPORT NO. OGNS-9441054 PAGE37 OF 50 REVISION 0 A review for other external events with potential severe accident vulnerability was performed as -

described in Section 2 of NUREG-1407. Of those items listed there is ro prevailing evidence that 1

would indicate that these events should be addressed in the IPEEE pus and there are no other known plant-unique external events that should be assessed.

6.0 LICENSEE PARTICIPATION AND INTERNAL REVIEW TEAM 6.1 IPEEE Program Organization Entergy Operations, Inc. (EOI) personnel were involved in all aspects of the IPEEE. Contractor expertise was used to supplement in-house capabilities. The vast majority of contractor work was performed on-site with EOI personnel working closely with the contractors.

This allowed i

technology transfer from the contractors and also ensured that the knowledge and insights gained from the analysis remained with the utility.

'f EOl personnel who performed the Seismic Analysis attended industry training on the EPRI Seismic Margin Methodology. Three individuals attended the EPRI SQUG training course and the EPRI IPEEE Seismic Add-on Course. All three of these individuals meet the qualification requirements of a " Seismic Capability Engineer" as described in EPRI NP-6041.

i 6.2 Composition ofIndependent Review Team Mr. Hany Johnson of Programmatic Solutions and Mr. Robert Budnitz ofFuture Resources Associates, Inc. performed the peer review for the IPEEE SMA at Grand Gulf Nuclear Station.

These contractors have expertise in the Nuclear Systems and Seismic Evaluations and are well recognized in the nuclearindustry.

Independent Peer Reviews for the remaining portions of the IPEEE evaluations were performed by EOI design engineering personnel from other nuclear sites as follows:

Portion Peer Reviewer Site External Flooding Murray Moser Arkansas Nuclear One (ANO)

Transportation & Nearby Roberrt Murillo Waterford 3 (WF3)

Facility Accidents High Winds & Tmnado Todd Reichardt Riverbend (RBS)

Assessment 6.3 Areas of Review and Major Commerts 6.3.1 Seismic Review

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Mr Johnson's review covered all seismic evaluations portions of the project and included a review of the Project Plan, the draft report, a visit to the site for a sample walkdown and a review of the documentation. Mr. Budnitz's review covered all aspects of the project and included a review of

ENGINEERING REPORr NO. OGNS-944054 PAGE 38 OF SO REVISION 0 the Project Plan, the SSEL development documentation, the draft report and c review of the documentation.

Mr Johnson and Mr. Budnitz concluded that there were no M in the IPEEE SMA effort at GGNS in the final report, the SSEL, the walkdown or dociunentation.

6.3.2 High Winds, Moods, and Others 1

Peer Review was performed for each major portion of the evaluation and focused on the individual engineering reports summarized in section 2.4. Copies of supporting calculations, UFSAR passages, and other reference material was provided to the reviewer for consideration during l

review of the report. Review considered the methodology used, the requirements of NUREG 1407, the requirements of the '75 SRP, and the requirements of other applicable regulatory and industry documentation.

The method of documenting Peer Review comments varied with each report. In some instarya comments were documented in correspondence between the reviewer and the preparer. In other instances, a summary of Peer Review comments was included as an attachment to the report reviewed. Engineering reports have been signed by the Peer Reviewer's to indicate the satisfactory incorporation of all comments.

6.4 Resolution of Comments Where applicable, Peer Review comments have been incorporated into the reports summarized in section 2.4. Similarly, where Peer Review comments affected the draft supporting calculations, the i

calculations were amended to address the comment. Engineering reports have been signed by the i

Peer Reviewer's to indicate the satisfactorf ncorporation of all comments.

i 7.0 PLANT IMPROVEMENTS AND UNIQUE SAFETY FEATURES This section defines vulnerability for each external event and then compares Grand Gulf Nuclear Station against the definition to determine if any vulnerabilities exist.

i 7.1 Seismic Analysis A vulnerability due to a Seismic event is defined as a component, identified as being needed in the SPLD, which is not capable of surviving the Review Level Earthquake (RLE). The seismic j

walkdowns found that the Grand Gulf Nuclear Station is seismically mgged and that all components in the SPLD adequately considered the Seismic input. All the SPLD equipment was screened out and the outliers were evaluated. There are no outliers requiring further evaluation.

5 It is therefore concluded the Grand Gulf Nuclear Station has no vulnerabilities with regards to Seismic events.

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' ENGINEERING REPORT NO. GGNS-944054 PAGE 39 OF SO.

J REVISION 0 7.2 Internal Fire Analysis (I.ATER)

~ 7.3 High Winds, Hoods, and Others 7.3.1 High Winds and Others A vulnerability to High Winds and Others External Event is defined as a plant non-conformance, with respect to the 1975 SRP, which has a significant contribution to GGNS's CDF. Hazards due to high winds and tomadoes meet the intent / criteria of the 1975 SRP. The threat to plant safety from transporation and nearby facility accidents has not been increased.

7.3.2 Moods GGNS is in compliance with criteria in the '75 Standard Review Plan. Significant changes to site features were not noted during IPEEE inspections. The design basis evaluations were last revised in 1992; and were performed in accordance with applicable sections of Regulatory Guide 1.59, Rev. 2(20), Regulatory Guide 1.102, Rev. l(21), and ANS 2.8/N170-1976(22). HMR 33 and EM-1110-2-1411 are used for determining PMP estimates and temporal distnktion. Complete stream blockage is assumed for Stream A. Partial blockage of Stream B is addressed by the restrictions imposed in GGNS Technical Specifications sections 3/4.7.10. With the exception of the enclosure building (designed to loose its roof and siding during design basis tornadoes and wind storms), all safety related structures can withstand roof ponding exceeding the parapet height. Roof drains and overflows are adequate to prevent ponding depths allowing propagating through roof doors, vents, and penetration sleeves. However, the following mechanisms and programmatic changes are being considered to prevent deterioration of site conditions from affecting the analysis.

Increase maintenance on drainage structures. Maintenance should include cleaning of culverts, concrete repair and removal of vegetation / debris which could obstruct flow. The portion of the south ditch near Culvert No.1 is currently being well maintained, but additional maintenance on the remaining drainage structures is wan anted.

Plant procedures 05-1-02-VI-l and 05-1-02-VI-2 currently require plant staff to insure that plant doors are closed during severe weather and in the event of plant flooding (Implicitly including former unit 2 doors). The only former unit 2 door explicitly required to be closed are the SSW B pump room doors. All other cited doors are Unit I doors. Since leakage through an open door could be substantial; revise one or both of these procedures to explicitly include at-grade former unit 2 doors.

Roof drains and overflows, particularly those on the enclosure building roof, should be periodically inspected to insure that they are not blocked.

Per NRC request, the new PMP estimates and distribution methods described in HMR's 51 and 52 were evaluated in accordance with applicable sections of Regulatory Guide 1.59, Regulatory Guide J

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ENGINEERING REPORT NO. GGNS-94 0054 PAGE 40 OF SO REVISION O 1.102, and ANS 2.8-1992. Appendix B to ANS 2.8-92 discusses the probability of occurrence of the PMP. Based upon portions applicable to the GGNS analysis, the combined probability of inundation of the GGNS site is between 5.0 x 10 4 and 5.0 x 10 -s.

Further, the combined probabilities associated with the coincident wind wave activity assumed in the evaluation, would effectively reduce this to between 2.1 x 10 -s and 2.1 x 10-10 Applying the new criteria, the bulk of the precipitation would occur over a shorter time frame, and markedly higher rainfall intensities would readt. As a result, the GGNS site is not WM to be completely protected against external flooding without making some site modifications. Evaluation i

reveals that the following site drainage / flood protection improvements would allow for adequate protection of the site against external flooding due to the revised criteria. However, they are not i

necessarily the only combination of potential changes for consideration. Given the small probability of occurance for the PMP, as described in the preceding paragraph, the relative cost and benifit for

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poteial improvemeds will be considered prior to implementation of any physical improvements.

Remove the wooden foot bridge crossing the northwest ditch near its upstream end.

Remove the 15" corrugated metal pipe located in the small auxiliary ditch parallel to the northwest ditch (at the same approximate location as the duct bank crossing the northwest ditch). Re-grade the area to provide a gradual transition between the yard upstream, and the auxiliary ditch.

]

Re-hang the security fence gates west of the control building to insure that approximately 5" of gap exists between the gate and the road.

i Grade down and remove the access road, the raised berm parallel to the access road, and curbs adjacent to the access road as necessary where they cross Culvert No.1, such that clevations above the culvert do not exceed 132.7 ft. MSL.

Replace the C8xil.5 channel forming the flood barrier across the SSW A equipment hatch opening with another member having a minimum depth of approximately 13".

l 8.0

SUMMARY

AND CONCLUSIONS (INCLUDING PROPOSED RESOLUTIONS OF USIs AND GIs)

EOI has performed a complete IPEEE, with the exception ofInternal Fires, for GGNS. Aralysis for Internal Fires will be submitted at a later date. For all other areas the intent of GL 28-20, Supplement 4 and NUREG 1407 has been met. EOI has expended significant resources develop

• g ~

l m

in-house capabilities in the performance of the IPEEE. The insights and knowledge gained in the process remains with the utility, l

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A e ems A.

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ENGINEERING REPORT NO. GGNS-94-0054 PAGE 41 OF 50 REVISION 0 8.1 Seismie Analysis h Seismic Analysis was performed using the EPRI Seismic Margins Methodology for Reduced Scope Plants. W Safe Shutdown Earthquake (SSE) ground response spectra armi corresponding i

in-structure response spectra were used as the Review Level Earthqake (RLW).

The conclusions of the seismic analysis is that Grand Gulf Nuclear Station is seismically rugged and that all components identified in the Safe Shutdown Path have adequately considered the aismic input. All anchorage to these components was found to be rugged.

Only one potential vulnerability to a Seismic event was identified which has been corrected During the review of pipe supports of Standby Service Water (SSW) piping in the Control Building, it was identified that the grouted condition of the penetration CP9A was not accounted for in the stress analysis of the piping systems. This is an exterior penetration at elevation 105', south wall of the Control Building. Four SSW (System P41) pipes were affected. hse pipes traverse from the Auxiliary Building into the Control Building and'are the supply and return lines to the Control l

Room AC units for both A & B loops of the SSW system. h as-found condition had the potential to induce significantly high seismic stresses into the piping between the buildings.

To correct the situation to meet design requirements., the grout was removed and a design change was issued to repair the penetration. h as-found grouted condition was evaluated for operability considerations and was detennined not to be an operability concern.

8.2 Fire Analysis:

(Later) 8.3 Hinth Winds. Floods. and Others Analysis i

Potential floodwater elevations at GGNS are controlled by application of the PMP over the site watershed. Three potential vulnerabilities to the design basis evaluation were identified during plant inspection, and corrected by means of the Material Non-Conformance Report process These challenges consisted of a panially crushed culvert, an abraded PMP door seal gasket, and l

conditions at a turbine building door which could allow more leakage than previously considered.

Several potential programmatic enhancements were noted for consideration as discussed in section 7.3.2. Engineering Report GGNS 91-0055 identifies, in general terms, those areas critical to PMP; and has been revised for the IPEEE to provide additional guidance in the selection of temporary lay-down areas to prevent critical drainage paths from being obstructed for any length of time.

J Additionally, it was determined that even though the probability of occurrence of the comb'med PMP event b well below 104, some additional vulnerability would result from the application of the revised PMP estimates in HMR 51 and HMR 52. Potential changes for consideration, are as described in section 7.3.2.

l h threat to plant safety from transportation and nearby facility accidents has not increased As l

reponed in section 5.3.4 of those changes that have occured, the threat to the plant has lessened. It a

p ENGINEERING REPORT NO. GGNS-94 0054 1

PAGE42 OF SO REVISION 0 is therefore concluded that activities at nearby transportion, industrial, and military facilities will not affect safe plant operation.

Hazards due to high winds and tornadoes meet the intent / criteria of the 1975 SRP. However, during the review it was determined that for some items credit for unit 2 structures as missile barriers, which would not be completed as originally designed, had been taken or the basis for protection was unclear. Therefore, a probabilistic evaluation was performed and determined that the frequency for a tornado generated missile striking these items was 0.77E-8/yr. This low frequency of occurrence is well below the screening criteria in IPEEE.

9.0 REFERENCES

1.

EPRI NP-6041-SL, "A Methodology for As===at of Nuclear Power Plant Seismic Margin Margin," Revision 1, Jack R. Benjamin and Associates, Inc. et. al., August 1991 2.

GGNS Individual Plant Examination Summary Reoott, Entergy Operations, Inc., Submitted via GNRO-92/00157, December,1991.

3.

Grand Gulf Nuclear Station Engineering Report for IPEEE Reduced Scope Seismic Margins Assessment (SMA), GGNS-94-0053, Revision O.

4.

Generic Letter 88-20, Supplement 4, United States Nuclear Regulatory Commission, June 28,1991.

5.

NUREG-1407, " Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities", United States Nuclear Regulatoiy Commission, June,1991.

6.

NUREG-75/087, " Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants", United States Nuclear Regulatory Commiedon, 1975.

7.

NUREG-0831, " Safety Evaluation Report Related to the Operation of Grand Gulf Nuclear Station, Units 1 and 2", United States Nuclear Regulatory Commission, September,1975.

8.

NUREG-0934, " Technical Specifications, Grand Gulf Nuclear Station, Unit No.1, Docket No. 50-416, Appendix A to License No. NPF-29", United States Nuclear regulatory Commission, October 1984. (As amended February 8,1994) 9.

Generic Letter 89-22, United States Nuclear Regulatory Commission, October 19,1989.

10.

Calculation CC-Q1Y13-93001, Rev. O, "PMF Hydrographs for Basins A and B (HMR 51 PMP Data)", August 11,1993.

11.

Calculation CC-Q1Y13-93002, Rev. O, " Backwater Analysis of External Flooding (HMR 51 PMP Data)", October 5,1994.

ENGINEERING REPORTNO GONS-944054 PAGE43 OF 50 REVISION 0 4

t 12.

Calculation CC-Q1Y13-93003, Rev. O, "In-L*=kar Analysis Due to External Flooding (HMR 51 PMP Data)", October 5,1994, 13.

Engineering Report GGNS-93-0001, Rev. O, " Individual Plant Examination for External Events (External Flooding)", December 5,1994.

14.

Engineering Report GGNS-93-0031, Rev. O, "An A=aaaamant of the Risk From Lightning Initiators", December 20,1993 15.

GGNS UFSAR 16.

NUREG/CR-5042 1

17.

Engineering Report GGNS-93-0017, Revision 0, " Selection of the Safe Shutdown Paths and Equipment For the GGNSD Seismic IPEEE 18 Engineering Report GGNS 0048, Revision 0, "High Wm' d and Tornado Assessment" 19 Calculation CC-Ql111-94004, Revision 0 20 Regulatory Guide 1.59, Rev. 2 21 Regulatory Guide 1.102, Rev. I 22 ANSI 2.8/N170-1976 i

23 NUREG /CR-4826 24

' ASCE Paper NO. 3269, " Wind Forces on Structures", Transactions of the American Society of Civil Engineers, Vol.126, Part II (1%1) 25 ANSI A58.1-1972, " Building Code Requirements for Minimum Design Loads in Buildings and other Structures", Committee A58.1, American National Standards Institute (1972) 26 Regulatory Guide 1.76, " Design Bases Tornado for Nuclect Power Plants", April 1974 l

ENGINEERING REPORT NO GGNS-94-0054 PAGE 44 OF 50 REVISION 0 Table 10.1 Front-line Systems FUNCTION PREFERRED PATH ALTERNATE PATH

• Reactivity Control CRD CRD Pressure Control SRVs SRVs in reliefmode in reliefmode for initial transient forinitial transient (Divi)

(Div2)

Inventory Control RCIC ADS (Div 2)

HPCS LPCI C l

SPMU A Decay Heat Removal RHR A in SPC RHR B in SDC (Hot Shutdown)

(Cold Shutdown)

• LOCA is not assumed for th 1

i ENGINEERING REPORTNO. GGNS-94 0054 PAGE 45 OF 50 REVISION O Table 10.2 FRONT LINE TO SUPPORT SYSTEM DEPENDENCY MATRIX

• Hrcs Rcx:

cRD ucs txa ssWi ras ads masDC masrc macs ecs cTur stuu sw RHR WATER VENT A

B C

X41E INJ A

B A

B A

B A

B A

B ESF AC DIVI X

X X

X X

X X

X X

X X

ESF AC Div H X

X X

X X

X X

X X

X X

X l

Esr AC Divm x

BOPAC X

X X

ESF DC DEVI X

X X

X X

X X

X

'X X

E8F DC DtV H X'

X X

X X

X X

X X

X Esr DCDiv m x

M)P DC X

i ssWTRAINA X

X X

X ssWTRAIN B X

X X

X X

X CCm X

TBCW X

t Psw CHILLED WTR CRC WTR X

INSTAR X10 8

X X -

X X

2 FAX 3 Rus HVAC X

X X

X X

X X

X.

X X

X X

7 l

Shaman h MVAC XI X

1 l

WS M bb @hb h M.

L

i ENGINEERING REPORT NO. GGNS-94-0054 PAGE 46 OF 50 REVISION 0 Table 10.3 SUPPORT SYSTEM TO SUPPORT SYSTEM DEPENDENCY MATRIX

• DOX SSW CCW TDCW PSW CHLD INST CIRC DO Rm SSW Puey ECCSRm STMInt AC Power DC Power DO WTit AIR Wilt HVAC House Vem HVAC HVAC IEM IHM I H A

B C

A B I

H HI ESF AC DIVI X

X X

X X

X 3

ESF AC DIV U X

X X

X X

X X X

ESF AC Divm X

X X

X X

BO.'AC X

X X

X X

X X

ESF DC DIVI X

X X

X ESF DC DIV H X

X X

X X

ESF DC DIV m X

X X

X BOP DC XII X12 XII X12 XI3 X

SSWTRAIN A X

X 4

SSW TRAIN B X

X X

X SSW TRAIN C X

X TBCW X

PSW X

X X

Cim 1Fn WTR X

INST AIR X

X DO Rm HVAC X

X X

3 SSW Pany A

X X'

House Ved B

X3 Swateurr & Bau Rm Cochng Divi X

X Div 2 X

X

• C,

- are M for the syssans in the cohann header.

1

ENGINEERING REPORT NO. GGNS-94-0054 PAGE 47 OF 50 REVISION O Notes for Tables 10.2 and 10.3

1. Delayed time depcedency. The RCIC pump will operate for 30 minutes after the steam leak detection signal is initiated. No isolation occurs during SBO due to loss of power to the timer.

i i

2. Delayed time dependency LPCS Pump will operate approximately 10 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without room cooling i

l

3. Train B pump l

l

4. SSW Train B is alternate source ofcooling water under certain conditions

5. Delayed time depers.ncy. SSW pumps will fail approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after loss ofHVAC.

6. Delayed time deperdency. SSW pumps will fail approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after loss ofHVAC. No -h +-' - y ifSSW A pump is l

not operating.

I

7. Deleted

8. Backup by accumulators.

9. Required for redundant actuation logic and Level 8 protection instmmentation.

10. Required for enhanced flow mode only.

11. DC power is required to start the pumps. However, the pumps are normally operating and DC power is not modeled.

12. DC power is required to start standby pump.

13. DC power is required to start the normally operating compressor. Therefore, DC power is not modeled.

-m.

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s i

ENGINEERING REPORT NO. GGNS-94-0054 PAGE48 OF 50 REVISION O 2

1 Figure 11.1 GGNS Success Path Logic Diagraan SAFE RN REACTOR REACTOR DECAY HEAT SHUTDOWN CONTROL PRESSURE BfVENTORY CONTROL RERIOVAL N:

CONTROL l

i neACWpiConE ecuttioncoouse MeHmussoune j

ccetE sment commeat nco o,,vg a sappuumecespact i

CCC1seBIKIIE CF pummum DEAT nemousL systems susve LONG SEISMIC se meter mooE suppneamen TEntA g

MAROM N,

=

==

SAFE SMUTDOWN EARTHQUAKE TWNMeNT F2 MOURS) seanceses cocussisoutop j

LostMuggunE magountpg47

=

l CooUINT

=

lleIsoW8L Sv5fgis

==

=

sumenon

=

Aos

=

=

l l

LWW N uses enma l

i l

w atsta mine

=emppemmene pum mesume aumene sur teamany imme to eymme pass e Mpce, I

nac numan a susnumammesensecer.

i GGNS PROPOSED SUCCESS PATH LOGIC DIAGRARI I

I

I ENGINEERING REPORT NO. GGNS-94-0054 PAGE 49 OF 50 REVISION 0 Figure 11.2 GGNS Perferred Success Path SAFE REACTMTY REACTOR REACTOR DECAY HEAT SHUTDOWN CONTROL PRESSURE INVENTORY CONTROL REMOVAL 1

FUNCTION:

CONTROL l

=

LONG TERM SAFE m

._ m sEieuse y woos us.croncone suransemanroot

- - mocaop SHUTDOWN h"

.,,oo MARosi auwwa sysTeu rou samt ' ~

,E F2 HOURS) isoLationcoatse e m Hou m l

l r-i i

6-l come erstav-l l

L--

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w. n,

-, cc

• SPRIU sugsted o ne spee h tukun tur Die Onnemunes mange Tenit i

se e asuuss Eur IICC musen.

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GONS IPEEE SEISMIC MARGIN ASSESSMENT SUCCESS PATH-PREFERRED l

t ENGINEERING REPORT NO. GGNS-94-0054 PAGE 50 OF 50 REVISION O Figure 11.3 GGNS Ahermate Success Path SAFE REACTIVITY REACTOR REACTOR DECAY HEAT SHUTDOWN CONTROL PRESSURE INVENTORY CONTROL REMOVAL FUNCTION:

CONTROL M M. 1 4@N M

M mesc nosovat s,svam EARTHQUAKE macent row n innes y

'918EB4MEudIB BIEf M Ef M EW8'Aur anSun asses abse to In suonMany Epen W GONS PEEE SEISMIC MARGIN ASSESSMENT SUCCESS PATH-ALTERNATE (ASSUMES NO LOCA)

o.

Attachment No.1 Engineering Report GGNS-94-0054 Page 1 of 3, Revision 0 SAFETY EVALUATION APPLICABILITY REVIEW FORM A)

Daenmant Evaluated: Rnoinaarinn Report GGNS-94-0054. Revinian 0 B)

Description of the Proposed Change: Pranuratian of a mimmary ranad for =>hmittal to the NRC documentino camalatian of the Individn=3 Plant Framination for Fvtarnal Evaa*= (IPRRR) a-cludiaa the evninntions for pI=a* fires. IPEE7 evalna*ians were raana=*ad from ameh n*il3*v in Gaaanic taar 33 20.

Supplamant 4: and are performad e =ccardance with the outdalinan oraeanted in NUREG 1407.

PRE-SCREENING Check the applicable boxes below. If any of the boxes are checked, neither a safety evaluation of,yuca,uity l

review nor a safety evaluation is necessary c4 steps C, and D may be skipped. The preparer and reviewer must sign at the bottom of the form.

The change is editorial only.

10CFIUO.54 applies to the change instead of 10CFR50.59.

An approved safety evaluation covering all aspects of this subject already exists. Reference SE#

The change, in its entirety, has been approved by the NRC.

Reference The change is an FSAR change that meets the exclusion criteria outlined in Site Directive G4.803 Safety Evaluation Annlicability Review If any of the following questions are answered "yes"..aen a full 50.59 Safety Evaluation must be completed.

C)

Does the proposed change or activity represent a change to the Technical Specifications?

YES Explain:

NO.2RL This report presents a summary of risk assessments===aci=*ad with seismic. flood.

hioh wind and linhinino evante as well as havards===^^iatad with -rby F=riliev and trananortation accidents. Evaluations are b==ad upon thnaa criteria set forth in the '75 Standard Review Plan. and other requirements set forth in NUREG 1407.

Any resultinn channes to the facility. procedures. UFSAR. Technical Specificatiora.

or other design basis documentation will receive the appropriate reviews. including 10CFR50.59 Safety Eyalnation. when the chanaan are = =Aa.

Rimilariv. all deviations from desian basis requirements identified durina the evaluation were Form 316.2, Revision 5 Page1

r y

e Attachment No.1 Faniaaaring Report GGNS-94-0054 q

Page 2 of 3, Revision 0 idaatinad in an anoropriate nonaar r==4 ranart. mad raemived the moorooriate e

reviews. includina 10CFR50.59 Safety Evain=*ian when the non-<- r-=== Hat identified.

D)

Does the proposed change or activity represent:

(1)

A change to the facility which alters, or has the potential to alter, the information, operation, function or ability to perform the function of a system, structure or component described in the SAR?

YES Explain:

NO.XX This report pra**nta a summary of risk====== manta===acintad with amiamic Saad 1

hiah wind _ and lich*nino evanta' as well as havards===aci=*ad with na= bv facili+v and trananor*=*ian accidants. Evaln=* ions are ba=ad upan *haa criteria set forth in the '75 Standard Review Plan and other raaniremanta set forth in NUREG 1407.

l Any raanitina chanaan to the facility. oracadares. UFSAR. Tachaie=1 St d" daan or other daeion b==is document =*ian will raemive the aceienri=** reviews. incindine 10CFR50.59' Safety Ev=ln=*ian_ when the chmanas are

==da.

Rim 14v. all

~

~

deviatians from daion h==ia raaniramaats idaatiGad durine the evain=+ian were idan*inad in an assicerinte non-caar manca raaart. mad received the.m,.usilate or reviews. includian 10CFR50.59 Safety Ev=1a=*ian when the nona-forrEnes was

~

identified.

(2)

A change to a procedure which alters, or has the potential to alter, a procedure described, outlined or summadzed in the SAR7 YES Explain:

NO XX This reooit praaa*= a summary of risk===== aman *====aci=*ad with =alamic anad.

hioh wind and liohinino evanta as well as hn/ards===aei=*ad with naarby facil;*v and transoortation accidents. Evaluations are based upon those criteda set forth in the '75 Standard Review Plan and other requirerrian*= net forth in NUREG 1407.

Any resulting changes to the facility. procedures. UFSAR. Technical Specifiesss.

or other daaien baats daenman*=* ion will raemive the maprooriate reviewa. includian 10CFR50.59 Safety Evain=*iort when the channen are mada.

Rim 11= tv. All deviations from da= ion h==in raaviramaa*= ideritinad durine the evain=+ian were idantined in an appropriate non-conformaare report =ad racalved the appropriate reviews. includian 10CFR50.59 Safety Evainatiart whan the non-conformance was identified.

Form 316.2, Revision 5 Page 2

-, = -

p

-0 Attachment No.1 Engineering Report GGNS-94-0054 Page 3 of 3, Revision 0 (3)

A test or experiment not described in the SAR or which requires that a

system be operated in an abnormal manner that is not described or previously analyzed in the SAR?

YES Explain:

NO.2RL This raaart pr_= = a anmmary of risk -

a==a==*==== art =*ad with =al=.ala anad r

hiah wind and liohtnino evante as well as hermeds===nelsead with aaarby r.aitiev and trananor*=tian medd*nts. Evain=*iaan are h==ad uaan thana criteria set forth in the '75 St=adard Review Plan =ad athar raaniramanen set forth in NUREG 1407.

Any **=* = arnarimenta or chanoas in svatam anar=*ian which raanl* form tha=a '

evain=+iana will receive the anoropriate reviews. laelndiaa 10CFR50.59 hfaev Evmin=* ion. whan the mndt PREPARER [

/-

/

N.

~ Name' Job title Date REVIEWERIMIIzM*

E-.6 I

/e-zs-N-Name

/

Job Title Date t

If the preparer performed an ap,1.cability review, the reviewer should check below to indicate by which means the independent review reached the same conclusions.

I Reviewed the applicability review documentation Completed an indanaadaat applicability review.

Performed a verbal review with the preparer.

j 1

l l

Form 316.2, Revision 5 Page 3 i

- -