NUREG/CR-4826 

UCID-20948 

Vol. 1 

Seismic Margin Review of the 

Maine Yankee Atomic Power Station 

Summary Report 

Prepared by P. G. Prassinos, R. C. Murray, G. E. Cummings 

Lawrence Livermore National Laboratory 

Prepared for 

U.S. Nuclear Regulatory 

Commission 
NOTICE 

This report was prepared as an account of work sponsored by an agency of the United States 

Government. Neither the United ·States Government nor any agency thereof, or any of their 

employees, makes any warranty, expressed or implied, or assumes any legal liability of re· 

sponsibility for any third party's use, or the results of such use, of any information, apparatus, 

product or process disclosed in this report, or represents that its use by such third party would 

not infringe privately owned rights. 

NOTICE 

Availability of Reference Materials Cited in NRC Publications 

Most documents cited in NRC publications will be available from one of the following sources: 

1. TheN RC Public Document Room, 1717 H Street, N.W. 

Washington, DC 20555 

2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, 

Washington, DC 20013· 7082 

3. The National Techn1cal Information Service, Springfield, VA 22161 

Although the listing that follows represents the majority of documents cited in NRC publications, 

it is not intended to be exhaustive. 

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection 

and Enforcement bulletins, circulars, information notices, inspection and investigation notices; 

Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and 

licensee documents and correspondence. 

The following documents in the NUREG series are available for purchase from the GPO Sales 

Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and 

NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of 

Federal Regulations, and Nuclear Regulatory Commission Issuances. 

Documents available from the National Technical Information Service include NUREG series 

reports and technical reports prepared by other federal agencies and reports prepared by the Atomic 

Energy Commission, forerunner agency to the Nuclear Regulatory Commission. 

Documents available from public and special technical libraries include all open literature items, 

such as books, journal and periodical articles, and transactions. Federal Register notices, federal and 

state legislation, and congressional reports can usually be obtained from these libraries. 

Documents such as theses, dissertations, foreign reports and translations, and nan-NRC conference 

proceedings are available for purchase from the organization sponsoring the publication cited. 

Single copies of NRC draft reports are available free, to the extent of supply, upon written request 

to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555. 

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process 

are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available 

there for reference use by the public. Codes and standards are usually copyrighted and may be 

purchased from the originating organization or, if they are American National Standards, from the 

American National Standards Institute, 1430 Broadway, New York, NY 10018. 
Seismic Margin Review of the 

NUREG/CR-4826 

UCID-20948 

Vol. 1 

Maine Yankee Atomic Power Station 

Summary Report 

Manuscript Completed: March 1987 

Date Published: March 1987 

Prepared by 

P. G. Prassinos, R. C. Murray, G. E. Cummings 

Lawrence Livermore National Laboratory 

7000 East Avenue 

Livermore, CA 94550 

Prepared for 

Division of Engineering Safety 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 

NRC FIN A0461 

',. 

ABSTRACT 

This Summary Report is the first of three volumes for the Seismic Margin 

Review of the Maine Yankee Atomic Power Station. Volume 2 is fhe Systems 

Analysis of the first trial seismic margin review. Volume 3 documents the 

results of the fragility screening for the review. The three volumes 

demonstrate how the seismic margin review guidance (NUREG/CR-4482) of the 

Nuclear Regulatory Commission (NRC) Seismic Design Margins Program can be 

applied. 

The overall objectives of the trial review are to assess the seismic margins 

of a particular pressurized water reactor, and to test the adequacy of this 

review approach, quantification techniques, and guidelines for performing the 

review. Results from the trial review will be used to revise the seismic 

margin methodology and guidelines so that the NRC and industry can readily 

apply them to assess the inherent quantitative seismic capacity of nuclear 

power plants. 

- i;; -

TABLE OF CONTENTS 

VOLUME 1. SUMMARY REPORT 

ABSTRACT •................................................................ iii 

LIST OF TABLES .. · ... ·.· ...................... · .•••.•••••.••••••••..••••••• -•• xi 

LIST OF FIGURES ..••.....•...........••...•••.•••.•..••••.•••.•.•• ~ .•••. xii 

p RE FAcE • • • • • • ' • • • • . • . • . • • • • • • • . • • • • • . •· • • • • • . . . . • • . . • . • • • . • . • • • . • . . • • • • X i i i 

EXECUTIVE SUMMARY •• ~ ••••••••••••••••••••••••••••••••••••••••••••••••••• XV 

1. INTRODUCTION ........................................................ l-1 

1.1 Organization -of the Report •..•••••.••..•.••••••..••.•.•••••.•. 1-4 

1.2 Scope of the Seismic Margin Review of·Maine Yankee •.•.•••.•••. 1-4 

2. SEISMIC MARGINS APPROACH ..••...•..•...•.•..•.••.••••••.•••••.•••••• 2-1 

2.1 Systems Screening .••..•....••.•••..•••••...•.••.•.•.•.••..•••• 2-2 

2.2 Fragility Screening •..••.•••.•.•..••.••••.••••••.•.••.••.•.•. ~2-4 

2.3 Approach for Performing Seismic Margin Reviews •••.• ~ •••.•••.•. 2-6 

2.4 Trial Guidelines for Seismic Margin Reviews .•..•.•••.•.•.•.•.. 2-7 

3. RESULTS OF SEISMIC MARGIN REVIEW ................................... 3-1 

3.1 Overall Results •.....••••••••.•....•••.•.•.•••..•••••••.•••••. 3-2 

3 . 2 Systems Res u 1 t s ... ~ ..•...• ~ •..•••••...•••.•... ~ ..•.•••••.•.•.• 3 -13 

3.3 Fragility Results ...•.•.••••..•.•.•.••.•.•... ~ ••••..•••.•••••. 3-17 

4. INSIGHTS AND LESSONS LEARNED •.•••.....••..•...•..•.••......••.•.•.• 4-1 

4.1 Approach and Methodology ....•..•.•.•.•••••.•.••••.•••••••••.•• 4-1 

4.2 Guidelines ............... ~.- ...... ~ .......................... ~.4-7 

4.3 Effort to Perform the Review .•...•....•..•...••.•••••••.••.••• 4-15 

4.4 Findings and Their Resolution •.•..•.••••.•.•••••••••••••.•• ~~.4-16 

5. CONCLUSIONS •.••••••••.••••..•.••••••••••••••.•.•.••.•.•••.•...•.••• 5-1 

5.1 Conclusions Regarding the Methodology, 

Guidelines, and Insights .•.••...•.•.••••.••....•...•...••.••.• 5-3 

5.2 Conclusion Regarding the Maine Yankee Licensing Issue •• ~ •..• ~~5-4 

6. RECOMMENDATIONS TO IMPROVE METHOOOLOGY ••.•....•••..•••...•••.•...•• G-1 

REFERENCES 

APPENDICES 

. ........................................................ R-1 

A Expert Panel Correspondence ....•.•...•.•.•.....•••..•••.....•. A-1 

B Peer Review Group Correspondence ..••.......••••...•••.•••••••. B-1 

C Utility Comments .....•........•...•...........•••••.•••.•.•••• C-1 

- v -
TABLE OF CONTENTS 

VOLUME 2. SYSTEMS ANALYSIS 

ABSTRACT ............................. , •••••••••••••••••••••••••••.•••.•• iii 

LIST 0 F .TABLES . •.........•......••••••••.••...•.........••••••...•••.•• v i i 

LIST OF FIGURES .......................................................... ix 

1. INTRODUCTION ....................................................... , 1-1 

2. METHODOLOGY •.•••.•••••.•••••.•.••••••••••••••. I ••••••••••••••••• • 2-1. 

2.1 General Approach ••.........•....•...••••... ~ •.....•........•.. 2-1 

2.2 Step 2- Initial Systems Review .....•..•.•......•••.•.••..•••. 2-1 

2.2.1 Gather System Information ....••.•.•....•...•••.•••.. ~.2-1 

2.2.2 Classify Front-Line Systems •.•.......•.....•.•.•...•.• 2-l 

2.2.3 Identify Group A Front-Line Components •••.••....••..• ~2-2 

2.2.4 Classify Support Systems ........•.....•....•••.••..••• 2-2 2.2~5 Identify Group A Support System Components ..•.•.•..... 2-2 

2.2.6 Identify Plant-Unique Features .•........•.•.•.•..••.•. 2-3 

2.2.7 Prepare for First Plant Walkdown .•..•.....••••.•...... 2-3 

2v3 Step 4- First Plant Walkdown •.•..•.•••....•.•..•••••.•...••.. 2-3 

2.3.1 Peer Review Group Meeting .•.......•...•..•.•.•••.•.... 2-3 

2.3.2 Plant Walkdown ..•••. ~ .......•.••......•..•••.•.•..•..• 2-3 

2.3.3 Plant Staff Discussions ..•......•..••..•.••.•.•....•.. 2-4 

2.4 Step 5- Systems Modeling •.......•..•..••.....•..•.•...•.•.•.. 2-4 

2.4.1 Review Event Trees .• ~ ....•...••.........•.• ~ .......... 2-4 

2.4.2 Develop Fault Trees •.•.•....•......••..•.••...•••• o •.. 2-5 

2.4.3 Develop Data Base ...••.•....•........•.....•.•.......• 2-7 

2.4.4 Determine System Cut Sets •....••...•......•...•....... 2-7 

2.5 Step 6- Second Plant Walkdown ....•....••..• ~ .•......•.....•.• 2-8 

2.6 Step 7 - Determine Minimal Cut Sets .•.••.•..•..•. 0 •••••••••••• 2-8 

2.6.1 Finalize Event Trees ..•..•.•....•.•. ~ •.•. ~ .•••...•.••• 2-8 

2.6.2 Finalize Fault Trees .•......•...•...•.•..•.•.•...•.••• 2-8 2~6.3 Link Fault Trees .....•..••......•....•........•..••... 2-8 

2.6.4 Determine Preliminary Boolean Equation •...•..•...••.•. 2-9 

2.6.5 Determine Final Boolean Equation •......•••• 0 •••••••••• 2-9 

3. SYSTEMS ANALYSIS .....•••• I ••••••••••••••••••••••.•••••••••••••••••• • 3-1 

3.1 System Identification .•......••..•.•.•..........•....•.•••.... 3-1 

3.1.1 Group A Systems •.•..........•.•..........•...•.......• 3-1 

3.1. 2 Systems Removed from Group A .......................... 3-1 

3.2 Systems Analysis of Front-Line Systems .........•....••... o •••• 3-3 

3.2.1 Auxiliary Feedwater System ............•..•............ 3-3 

3.2.2 High Pressure Safety Injection System •....•.........•. 3-5 

3.2.3 Power-Operated Relief Valves 

(Feed and Bleed) .........................•....•....•.. 3-8 

3.3 Systems Analysis of Support Systems ..........••...•.••.... ~ .•. 3-9 

3.3.1 Component Cooling Water Systems •...• ~ •......•.•....... 3-9 

- vi -
f, 

. ' 

• ~ '. I' 

TABLE OF CONTENTS (Cont.) 

VOLUME 2. SYSTEMS ANALYSIS 

3.3.2 Service Water System ................•••.......•.•..... 3-11 

3.3.3 Electric Power System •....•...........•.......•....•.• 3-13 

3.3.4 Actuation Systems .............•.•.••.•....•.•......... J-16 

3.4 Probability Calculations ................•.•..•..•...•••.••.... J-16 

4. ACCIDENT SEQUENCE ASSESSMENT AND RESULTS ..•.........•••..•......... 4-1 

4. 1 

4.2 

4.3 

4.4 

Event Trees ... .................................................. 4-1 

4.1.1 No LOCA Case ........................................... 4-1 

4.1. 2 Sma Tl LOCA Case ...•.•..•..........................••.• 4-2 

Core Damage Sequences .••.......•.•....•.•..•.....•..•••.•..... 4-3 

4.2.1 T1LP 1 ......................••.••...••••.•.•...••••..•. 4-3 

4.2.2 T1LD .•......•..•..••................•.••.•••.••.•....• 4-3 

4. 2. 3 s2 D ..•••.•..............•••....••.•.......•.....•.•.•. 4-4 

4.2.4 S2LP 2 .....•.....•.....••..............•••..........•.. 4-4 

4.2.5 s2LD ...••.........•...•.....•.....••.•.•.••........... 4-4 

Boolean Equations for No LOCA and LOCA Cases •.•...•••.•.••...• 4-4 

4.3.1 No LOCA Case •.....•.•.........•••••....•...•.••.•••••• 4-5 

4.3.2 Small LOCA Case ...........•.•.•.•••••••••.......•....• 4-6 

Plant-Level Boolean Equation .•....••.......................... 4-6 

5. ENGINEERING AND METHODOLOGY INSIGHTS ......................•.•••.... 5-1 

5.1 Engineering and Operational Insights ...........•........•...•. 5-1 

5.2 Insights on the Methodology and Execution •.•.. o .........•..... 5-2 

5.2.1 System Calssification and Screening Guidelines ••••.... 5-2 

5.2.2 Preparation for Walkdowns and Documentation .....•...•• 5-4 

5o2.3 Systems Analysis and Pruning Process ......• o .......... 5-4 

5.2.4 Minimal Cut Set Evaluation Process .•....•.......•.•... 5-5 

5.2o5 Schedule and Resources ..... o••o•••·····••o•oo····o····5-6 

REFERENCES 

APPENDIX 

· · · · · · · · · · · · · ... · ...................................... R-1 

A 

B 

c 

D 

E 

F 

G 

H 

I 

Identifiers and Symbols ... ooo••··························••o••A-1 

Auxiliary Feedwater System ...... oooo•········••o••············B-1 

High Pressure Safety Injection System .. o••····•••oo••••·······C-1 

Primary Pressure Relief System ......•.. o•o•o••o••··········••oD-1 

Primary Component Cooling Water System ........•••.....•.•...•• E-1 

Secondary Component Cooling Water System .... o ................. F-1 

Service Water System .....•.•••.•••• o••••o••···················G-1 

E 1 e c t ric Power System ... ~ .. o ..... o .••••..•..••..•••••.•.. 0 •• 0 • H -1 

Actuation Systems ......••...•.• o•••oo .•.. · .......•..••..••••..• I-1 

- vii -
: I ' 

TABLE OF CONTENTS 

VOLUME 3. FRAGILITY ANALYSIS 

A B S T RA C T • • • • • • • • • • • • • • • • • • • • • • • • • • • , • • • • • • • • • • • • • • • • • • • • • • • • • .. • • • • • • • • • i i i 

LIST OF TABLES~ .•.•..•....• ·.·~~ ••. ,; ..•......•...•...•.•....•••.••••••..• vii 

LIST OF FIGURES .......................................................... ix 

1. INTRODUCTION ..•....••••.....••...................•..•..•••...•.••• 1-1 

1.1 Background~ ..................................................... l-1 1~2 Objective of the Study ••.•.•.....•..••.•••..•.......•••.•••.•• 1-2 

1.3 Scope of the Study ............. ~ ............................... l-2 

1.4 Organization of the Report •..••.............•. ~ •. -~ ..•.•.••• ~.1-3 

2. REVIEW EARTHQUAKE LEVEL ••..•....•.•......•...•••••••..•.•...•.••••• 2-1 

2.1 Review of Available Seismic Hazard Studies .•••..........••.••• 2-1 

2.2 Implication of the Selected Review Earthquake Level •.•••••..•• 2-1 

2.2.1 Screening of Components .••.•.•.•.........••.•••. ~ ...•.• 2-1 

2.2.2 Estimation of Seismic Capacities •••.•...•..•••.•.••..•• 2-5 

2.3 Different Methods for Specifying Review Earthquake Level •..• ~.2-5 

3. DESCRIPTION OF PLANT STRUCTURES, SYSTEMS AND COMPONENTS ••••...•.•.• 3-1 

3.1 Maine Yankee Plant/Structures and Systems ..•.•..••..••..••..•• 3-l 

3.1.1 Structures .............................................. 3-1 

3.1.2 Systems ................................... · ............. 3-1 

3.2 Seismic Design Criteria ...••..........•....•••••••.•.••.••.•.• 3-2 

3.3 Availability of Plant Design Data .....•.•...•.•.•..•••..•.•.•• 3-2 

4. REVIEW OF PLANT INFORMATION AND WALKDOWN •.•••..•.•••••.••••••..•••• 4-1 

5. 

4.1 Initial Screening of Components ...•.•....•..•...••••.••.••.••• 4-1 

4.2 Review of Design-Analysis and Seismic Reevaluation Reports •.•. 4-4 

4.3 Plant Walkdown ...................•......•..................... 4-6 

4.3.1 Identification of Target Areas for First Walkdown .••••• 4-6 

4.3.2 Walkdown Procedures ••.•••.•.....••••••••. ~ .•..... ~ ..... 4-22 

4.3.2.1 Walkdown Team ....•••...••...•••••••. ~ •..•..•.• 4-22 4.3~2.2 Procedures for Structures and Equipment ••••... 4-23 

4.3.3 Walkdown Documentation ...•...••.•.•••.•.••••••.....•••. 4-29 

4.3.4 Walkdown Results ••.•••.•••.••.....•...•. ~ .......••••••• 4-30 

4.3.4.1 Walkdown Findings .•.•.••.•....••.•..•••••••••• 4-30 

4.3.4.2 Plant Unique Features~ ....••••..•.••••••....•• 4-56 

4.4 Maine Yankee Component Modifications .•....•.•.•...••...•.•.•.• 4-56 

EVALUATION OF SEISMIC CAPACITIES OF COMPONENTS AND PLANT ..•.••••••• 5-1 

5.1 Review of Structural Models .•.••..••.....•.•.•.•••••..•..••.•• 5-1 

5.2 Simplified Analysis and Use of Screening Tools ...• ~ •.•.•..•... 5-2 

5.3 Second Walkdown •.•..•...•.••.••••.•••.•.•.•.•.•••••.••.•..•••• 5-4 5~4 HCLPF Capacity of Components ....... ~-~ ....•..•.•...• ~ .•••.• ~.~5-4 

- viii -
TABLE OF CONTENTS (Cont.) 

VOLUME 3. FRAGILITY ANALYSIS 

Page 

5.4.1 Fragility Analysis Method .•..•••....•.•.......••....... 5-4 

5.4.2 

5.5 HCLPF 

5.5.1 

5.5.2 

5.5.3 

5.5.4 

5.4.1.1 Methodology •..•..•.....................•.•.•.. 5-4 

5.4.1.2 Steel Structures .•.•.•....•........•.•.......• 5-23 

5.4.1.3 Block Walls ....•...••.....••....•••........... 5-30 

5 • 4 . 1 . 4 F l at Bottom Storage Tanks ...............•.•... 5-3 2 

5.4.1.5 Inverter ....................................... 5-40 

5.4.1.6 Diesel Fuel Oil Day Tank Fragility 

Eva 1 uat ion ......•....•.......•.••...•......... 5-46 

5.4.1.7 Containment Spray Fans .•.•.•....•.........•... 5-52 

CDFM ·Method .. ........................................... 5-53 

5.4.2.1 Circulating Water Pumphouse ..•.....•.•........ 5-54 

5.4.2.2 Block Wall SB 35-3 ............•.........•.•.•. 5-54 

5.4.2.3 Refueling Water Storage Tank ............•...•• 5-55 

Capacity of Plant ...•..................••.....•...••.•.. 5-56 

Accident Sequences ..•....•.•.•.•.•........•....••.•.•.. 5-56 

Probabilistic Method .•...•..........•...•........••.••. 5-59 

Deterministic Method ....•.........•....•.....•......•.. 5-67 

Sensitivity Studies ..•.•..•.............••....•......• 5-68 

6. COMMENTS ON SEISMIC MARGIN REVIEW METHODOLOGY ......•........•.•.... 6-1 

7. 

6.1 Selection of Review Earthquake Level .......................... 6-1 

6.2 Use of Screening Guidelines ...•.•........................•.... 6-2 

6.2.1 Extent of Review Needed ....•.........•...••..•......... 6-2 

6.2.2 Additional Screening Guidelines ........•............... 6-2 

6.2.3 Difficulties of Reviewing Certain Components ........•.. 6-3 

6.2.3.1 Reactor Internals and CEDM .....•.......•.....• 6-3 

6.2.3.2 Instrumentation within Containment .•......•... 6-3 

6.3 Availability of Qualification Data ........••.................. 6-4 

6.4 Walkdown Procedures .......•......................•....•...•... 6-5 

6.5 Guidance on CDFM Capacity Calculation Procedures .............• 6-6 

6.6 Staffing Requirements and Schedule ...•..•....•................ 6-7 

6.6.1 Staffing Requirements ................•.•............... 6-7 

6.6.2 Schedule ................................................ 6-8 

6.7 Applicability of Methodology to Other Plants ............•..•.. 6-8 

RESULTS AND CONCLUSIONS .......•............................••.....• 7-1 

7.1 Screened Out Systems and Components ............•.............• 7-1 

7.2 HCLPF Capacities of Screened-in Components .................... 7-2 

7.3 HCLPF Capacity of Plant ..............................•........ 7-32 

7.4 Identification of Low Capacity Components ..................... 7-32 

7.5 Conclusions ................... , ........ , ...................... 7-32 

REFERENCES •............................................................. R-1 

- i X -
\ . :' 

APPENDIX A 

TABLE OF CONTENTS (Cont.) 

VOLUME 3. FRAGILITY ANALYSIS 

Maine Yankee Atomic Power Station Arrangement Drawings ...••...•.... A-1 

- X -
Table 3-1 

Table 3-2 

Table 3-3 

Table 3-4 

Table 3-5 

Table 3-6 

Table 3-7 

Table 3-8 

Table 4-1 

LIST OF TABLES 

f~ 

Component Seismic Fragility Parameters •....•...•...•.•.•.. 3-3 

Probabilities for Nonseismic Failures .•....•.....•...•..•. 3-4 

Summary of Plant Level HCLPF Capacities •..•.••••••...••.•• 3-9 

Front-line System vs Support System Dependency Matrix ••... 3-14 

Support System vs Support System Dependency Matrix ..••.... 3-15 

Minimal Cut Sets for Important Systems •.•........••..•.... 3-18 

Maine Yankee Structures and Block Walls ...•..•.•..••.•.•.• 3-21 

Maine Yankee Equipment List .......•.••....••••.•...•.•.... 3-23 

Cost Breakdown for Seismic Margins Trail Plant Review •..•. 4-11 

- xi 
LIST OF FIGURES 

Figure 2-1 Graphic rpresentation of the screening operations 

(Figure 2-6 from NUREG/CR-4482) •••.... ~ .....••....•....... 2-6 

Figure 2-2 Example systematic event trees •.................•.•....... 2-8 

Figure 2-3 Example of a simplified fault tree •......•.....•..•.....•. 2-11 

Figure 2-4 Fault tree for failure of RCS integrity .....•.•.•..••..... 2-13 

Figure 2-5 Example fragility curves for a structure ...•..•.......•.•• 2-16 

Figure 3-1 Fragility curves for small LOCA core damage •.•.........•.• J-7 

Figure 3-2 Seismic fragility curves for NO-LOCA core damage ••.•.....• J-8 

- Xi i -
PREFACE 

The seismic margin review of Maine Yankee Atomic Power Plant was performed for 

the U.S. Nuclear Regulatory Commission as part of a research program on 

quantification of seismic margins at nuclear power plants. The approach, methodology (NUREG/CR-4334), and guidelines (NUREG/CR-4482), were developed by the Expert Panel on the Quantification of Seismic Margins and its technical 

support personnel. Maine Yankee and Yankee Atomic provided necessary data and 

information to this margin review. The results were reviewed by a Peer Review 

Group and the NRC Seismic Design Margins Working Group. This report is a 

collective effort and presents the results of the seismic margins review of 

the Maine Yankee plant. 

AUTHORS OF THE REPORT AND ANALYSIS TEAM 

Volume 1. Summary Report 

Lawrence Livermore National Laboratory 

P. G. Prassinos 

R. C. Murray, Project Manager 

G. E. Cummings 

Volume 2. Systems Analysis 

Energy Incorporated 

D. L. Moore, Systems Analysis Team Leader 

D. M. Jones 

M. D. Quilici 

J. Young 

Volume 3. Fragility Analysis 

EQE Incorporated 

M. K. Ravindra, Fragility Analysis Team Leader 

M. J. Griffin 

G. S. Hardy 

P. S. Hashimoto 

U.S. NUCLEAR REGULATORY COMMISSION 

D. J. Guzy, Project Leader 

A. J. Murphy, Section Leader, Engineering Branch 

J. E. Richardson, Chief, Engineering Branch, Division of Engineering Safety 

- xiii -
·.· ... · 

NRC Seismic Design Margins Working Group 

N. Anderson, Cochairman 

J. Richardson, Cochairman 

G. Bagchi 

L. Beratan 

J. Chen 

N. Chokshi 

C. Grimes 

D. Guzy 

P. K. Niyogi 

L. Reiter 

UTILITY 

Maine Yankee Atomic Power Company 

S. V. Evans 

G. D. Whittier 

Yankee Atomic Electric Company 

P. L. Anderson 

W. E. Henries 

J. T. McCumber 

W. J. Metevia 

R. P. Kennedy, Consultant 

REVIEWERS 

Peer Review Group 

R. J. Budnitz (Chairman), Future Resources Associates 

M. P. Bohn, Sandia National Laboratories 

J. W. Reed, Jack R. Benjamin & Associates, Inc. 

J. Thomas, Duke Power Company L. A. Wyllie, H. J. Degenkolb & Associates 

Expert Panel on the Quantification of Seismic Margins 

R. J. Budnitz (Chairman), Future Resources Associates 

P. J. Amico, Applied Risk Technology Corporation 

C. A. Cornell, CAC, Inc. 

J. W. Reed, Jack R. Benjamin & Associates, Inc. 

M. Shinozuka, Columbia University 

- xiv -
EXECUTIVE SUMMARY 

The U.S. Nuclear Regulatory Commission is sponsoring a research program to 

develop and demonstrate a method for assessing safety margins at nuclear power plants with respect to seismic events. This program is called the Seismic 

Margins Program and was started in 1984 at the Lawrence Livermore National 

Laboratory, [Cummings et al., 1984]. The need for such research results from 

the changing perception of the seismic hazard in certain localities which 

could require reassessment of the adequacy of seismic design at some power 

plant sites. It is generally accepted that nuclear power plants are capable of withstanding earthquake motion substantially greater than their design basis, but methods are needed to systematically demonstrate this. An .. Expert Panel on the Quantification of Seismic Margins .. was formed to develop such a 

margins assessment method [Budnitz et al., 1985] and [Prassinos et al., 1986], 

and this report discusses the application of that method to the Maine Yankee 

Atomic Power Station, a Combustion Engineering, three-loop pressurized water 

reactor located near Wiscasset, Maine. 

The margin review process involves the screening of components based on their 

importance to safety and their seismic capacity. The products of the review 

are High Confidence of Low Probability of Failure (HCLPF) capacities for 

components, accident sequences and the plant. Mathematically, the HCLPF can 

be thought of as an estimate of the 5% failure probability point with 95% 

confidence. These HCLPF capacities, measured in terms of peak ground acceleration (pga), are compared to the peak ground acceleration predicted for 

the earthquake against which the plant is to be assessed, called the review 

level earthquake. This review level earthquake is chosen at some level above 

the design basis (safe shutdown earthquake, SSE) and a favorable comparison with the HCLPF capacity of the plant indicates that there is high confidence 

of a low probability of failure (core damage). The extent to which this plant 

HCLPF capacity is above the SSE level is a measure of the seismic margin of 

the plant. 

Systems analysis is used to determine those plant systems and components (including structures) that are important contributors to plant seismic safety and thus allow focusing of effort on components requiring a margin review. By studying previous seismic probabilistic risk assessments (PRA) of pressurized 

water reactors (PWR), it was found that only systems and components needed to 

assure reactor subcriticality and early emergency core-coolant injection needed to be considered. Therefore, only systems and components related to 

these two functions (with a few exceptions) were considered at Maine Yankee. 

Capacities of generic sets of components were estimated and given in NUREG/CR4334 based on estimated capacities in previous PRA•s, experience data gained from studying earthquake effects on industrial facilities and engineering 

analysis. These generic capacities were used to screen out from further 

consideration those components identified as important by systems analysis if 

the generic capacity was found to be higher than the review earthquake level 

for Maine Yankee. This review level earthquake was established by the NRC as 

- XV -
0.30g with a 50th percentile Newmark Hall spectra, as defined in NUREG/CR0098. To make sure that the generic capacity of all important components could be properly predicted, by using the tables found in NUREG/CR-4334, these 

components underwent review and inspection before being screened out. In 

addition, any potential system interaction or plant-unique features discovered 

were added to the list of review items. 

The components remaining after the systems and fragility screenings, plus the 

systems interaction and plant-unique global features, were then subjected to a 

margins quantification. Prior to this quantification, each remaining 

component was thoroughly inspected and studied, and systems models were 

developed to describe the possible seismic-initiated accident behavior of the 

plant. The quantification was accomplished by calculating the HCLPF 

capacities for each of these components using structural/mechanical analyses and then analyzing the minimal cut sets derived from the systems analysis using the rules of Boolean algebra and Discrete Probability Distribution 

methods to arrive at accident sequence, and plant level HCLPF capacities. Component HCLPF capacities were calculated for the important components remaining after screening, using the Fragility Analysis (FA) method. This 

method requires estimating the median failure capacity for the component, and 

its random and modeling uncertainties. Assuming a lognormal failure 

probability distribution, the HCLPF (5% failure with 95% confidence) capacity 

can be calculated. Checks on several key components were made using the 

Conservative Deterministic Failure Method (CDFM) which uses a deterministic, 

more design-oriented method of calculating component HCLPF's. Further work is 

underway to eros s-check the two methods. Random, test and rna i ntenance, and 

human error failure modes were also included in the analysis. 

MARGINS REVIEW OF MAINE YANKEE 

The s e i s m ic rna r g i n s rev i ew of M ai n e Yankee was an e i g h t-s t e p pro c es s wh i c h 

involved Maine Yankee, Yankee Atomic Electric, NRC, LLNL as project manager, and fragility and system analysis teams {EQE Inc. and Energy Incorporated, respectively). The first step was to establish the review level earthquake 

(0.30g). Steps 2 and 3 involved information gathering and preliminary analysis by the two teams leading to step 4, a plant inspection (called a 

walkdown). In step 5, following the plant inspection, event and fault trees 

were constructed including those components not screened out. In step 6, a 

second visit was made to the plant to recheck the components remaining which 

might require further analysis. In step 7, minimal cut sets leading to core 

damage were determined. In step 8, the component HCLPF capacities were 

finalized and HCLPF capacities for accident sequences and the plant calculated. This entire effort took about 3.0 man-years of effort. 

For Maine Yankee, two important accident sequences were identified. Both are 

initiated by a seismically induced loss of offsite power (LOSP) assumed to 

always occur at the review earthquake level. In one sequence there is a small 

loss-of-coolant accident (LOCA of 3/8 in. to 2 in. diameter equivalent area) 

assumed to occur because of seismically induced pipe breakage. HCLPF 

capacities for components which might cause other types of small LOCAs (pump seal or power-operated relief valve LOCAs) were sufficiently high so they could be screened out. The other accident sequence assumed no small LOCA. 

- xvi -
~ ' • I 

The small LOCA accident sequence involved seismic failures only and resulted 

in a plant HCLPF of 0.21g. Many plausible arguments, including the component screening table of NUREG/CR-4334, indicate that a small LOCA could be screened 

out for the review level earthquake considered. Since the analysts involved 

in this review could not get inside the Maine Yankee containment to inspect the small primary system piping, they chose not to screen small LOCA out. If 

they had, this sequence would not be considered and the plant HCLPF would be 

above 0.30g. 

The small LOCA accident sequence was composed of three singleton cut sets with 

the dominant contributor being failure of the Refueling Water Storage Tank 

(RWST) which had a HCLPF of 0.2lg. Other singletons in the sequence had 

HCLPF•s greater than 0.30g. 

The second accident sequence, LOSP with no small LOCA, involved no singletons but a number of doubletons, some combining seismic and nonseismic (random, 

test and maintenance, human error) failures. The most important doubleton is 

the Demineralized Water Storage Tank (DWST, HCLPF = 0.17g) and the Circulating 

Water Pump House (HCLPF > 0.30g). 

Nonseismic failures were found not to be important contributors. They made no 

contribution to the overall plant level HCLPF capacity. The most impor,tant nonseismic failure found was a common cause failure of the Auxiliary Feedwater 

System, caused by steam binding (median unavailability per demand of 1.2 x 

10-4). 

It should be noted that during the review process, important components were 

found for which a low HCLPF would result or insufficient data was available to 

determine the HCLPF. These were the lead-antimony station batteries, the 

station service transformers, a block wall near HVAC equipment, parts of the 

PCCW and SCCW air conditioning heat exchangers, and anchorage of the diesel 

generator day tanks. To reduce the uncertainty in their capacities, these 

components are being replaced or upgraded and the results of this review are 

based on the upgraded configurations. 

INSIGHTS AND LESSONS LEARNED 

In addition to the results already described, the Maine Yankee seismic margins review provided some lessons applicable to future reviews. The event/fault 

tree methods used provided a complete description of dominant contributors and 

considered all important systems. The fragility screening table (NUREG/CR4334) needs to be strengthened and more guidance given on how to select a 

review level earthquake. Also, more guidance is needed on how to use and 

compare the two methods used to determine component capacity (FA and CDFM) and 

how to combine seismic and nonseismic failures. 

The systems analysis effort should start early so that component screening and 

plant inspections can be done efficiently. Information in the plant Final 

Safety Analysis Report can be effectively used for this effort. Plant 

inspections (walkdowns) need to be carefully planned, taking into account 

auxi 1 iary systems such as the HVAC and actuation/control system, as well as 

- xvii 
important sys terns and components i dent i fi ed in the sys terns analysis. These 

walkdowns are essential to successful margin reviews. Specifically, they 

permit accurate data collection and allow identification of potential low 

capacity components. 

More guidance is needed on the consideration of the seismically initiated 

small LOCA. This initiating event turned out to be particularly important at 

Maine Yankee. It may be impossible to review and inspect all the small 

primary pressure boundary piping, and other screening methods need to be 

developed, e.g., inspection of similar piping outside of containment. 

Finally, there is a question about maintenance of hot shutdown for a specified 

period of time after an earthquake, e.g., 72 hours. The current methodology 

addresses the attainment of hot shutdown but not necessarily its 

maintenance. Review of previous seismic PRAs indicates that once hot shutdown 

is attained, the probability of maintaining it is large. Therefore, this 

issue may be of less importance. 

- xviii -
VOLUME 1. SUMMARY REPORT 

CHAPTER 1 

1. INTRODUCTION 

In 1984, the U.S. Nuclear Regulatory Commission ·(NRC) initiated the Seismic 

Design Margins Program (SMP) to address regulatory needs and a changing perception of the seismic hazard. The NRC formed the .. Expert Panel on the 

Quantification of Seismic Margins" and charged it to work closely with an inhouse NRC staff~ 11 Working Group on Seismic Design Margins:.'' to provide technical guidance on the assessment of seismic margins. The overall goal of 

the SMP is the development of a methodology and guidelines that can be readily used by the NRC and industry for assessing the inherent quantitative seismic 

capacity of nuclear power plants [Cummings et al., 1984]. 

The development of a soundly based, efficient and effective method for the 

assessment of how much margin actually exists in important components, 

systems, and the plant will serve to minimize the impact of changing regulatory requirements and 1 icensing actions as the estimates of seismic 

hazards change. In addition, a seismic margins assessment can provide a basis 

for confidence in the capacity of nuclear power plants and this methodology 

can be applied when questions arise about their seismic capacity. 

The most important regulatory need and the focus of the seismic margins effort 

is stated as follows: 

"There is a need to understand how much seismic margin exists at 

nuclear power plants. This seismic margin is to be expressed in 

terms of how much 1 arger must an earthquake be above the safe 

shutdown earthquake (SSE) before it compromises the safety of the 

plant ... 

The Expert Panel and its technical support personnel studied the available 

information on the quantification of seismic capacity of nuclear power plants and other industrial facilities. The results of several seismic probabilistic risk assessments of nuclear power plants were reviewed along with the behavior 

of industrial facilities during earthquakes. These studies were used to 

develop a margin review approach that involves both the screening of 

components based on their importance in preventing seismic core melt and their 

inherent seismic capacity. 

The seismic margins review approach has been documented in the report, .. An 

Approach to the Quantification of Seismic Margins in Nuclear Power Plants .. [Budnitz et al., 1985]. This document formed the basis for the development of 

guidelines for performing seismic margin reviews. These guidelines are given in ~~Recommendation to the Nuclear Regulatory Commission on Trial Guidelines 

for Seismic Margin Reviews of Nuclear Power Plants, Draft Report for Comment .. [Prassinos et al., 1986]. 

1-1 
c·,i J 

The performance of a trial plant review is needed to verify and test the 

review methodology and guidelines. For the trial review, the NRC negotiated 

with and selected the Maine Yankee plant. Once the trial plant was selected, 

the two analysis teams (Systems Team and Fragility Team) were chosen. These 

teams were selected based on their technical approach and team composition. 

In addition, a Peer Review Group was selected and charters were developed for 

both this group and the Expert Panel. The seismic margin review was then 

organized allowing for participation by the plant owner and its representative 

(Yankee Atomic Electric Company), the NRC Working Group on Seismic Design 

Margins, and the appropriate NRC program managers. 

The purpose of this report is to summarize the results of this first trial 

seismic margin review. The detailed results from this review can be found in 

Volumes 2 and 3 of this report. Volume 2, ~stems Analfsis, documents the 

results of the systems screening and analyses portion o the review, while 

Volume 3, Fragility Analysis, gives the component fragility screening and 

analyses results. 

The overall objectives of this trial review are to assess the seismic margins 

of a pressurized water reactor, the Maine Yankee Atomic Power Station, and to 

test the adequacy of the seismic margin review approach and quantification 

techniques, and the guidelines for performing these reviews. There are four 

related objectives of this study: 

o To demonstrate the use of the Expert Panel's approach (NUREG/CR-4334) 

and guidelines (NUREG/CR-4482) for seismic margin reviews. 

o To provide a basis for revising and upgrading the approach and 

guidelines. 

o To provide a benchmark for pos sib 1 e future seismic margin reviews, 

including an understanding of the level of effort in performing a 

seismic margin review. 

o To provide an assessment of the plant's capability to withstand a 

specific earthquake level greater than the SSE. 

The results from this trial review are intended to be used to revise the 

seismic margin methodology and guidelines so they are more prescriptive and 

ready for general use by the NRC and industry. 

The role of the Expert Panel during the Maine Yankee study was to review the 

procedures being used at an early stage of the process and to assure that the 

methods and techniques being employed are consistent with the Panel's guidance 

and are relevant for performing a seismic margins review. When the trial 

plant review is completed, the Panel will study the results, examine how the 

review was implemented, and evaluate the overall effort. The Expert Panel 

will be expected to appraise the overall usefulness of the seismic margin 

review approach and identify any limitations whether previously recognized or 

not. Appendix A defines the role of the Expert Panel for this review and 

contains minutes of the panel's telephone conference call. 

1-2 
~ ., ' ~ 

' : 

A Peer Review Group was selected and chartered to review the technical 

adequacy of this study including participation in the plant walkdowns. The 

objective of the Peer Review Group is to assure that the seismic margin review 

is executed in a fully competent and professional manner, uses appropriate methods, and follows the guidance established in NUREG/CR-4334 and NUREG/CR4482. At the conclusion of the Maine Yankee review, the Peer Review Group provided its best judgment with regard to both the review procedures and the 

technical competence of the reviews, based on its collective expert opinion [Anderson, 1986]. Peer Review Group correspondence including its charter, 

meeting minutes and summary report on this trial review are given in Appendix 

B. 

Appendix C contains comments received from the utility. 

1.1 Organization of the Report 

The remainder of this section gives a discussion of the scope of the effort. 

Chapter 2 presents a brief overview of the seismic margin review methodology and the guidelines as applied to this review. Chapter 3 provides the overall 

results of the Maine Yankee review, including a summary of the significant fragility and systems results. Insights and lessons learned are given in 

Chapter 4 followed by the conclusions in Chapter 5. Recommendations from this 

study are given in Chapter 6. 

1.2 Scope of the Seismic Ma!Jlin Review of Maine Yankee 

The Maine Yankee Atomic Power Station is a Combustion Engineering (CE) threeloop pressurized water reactor (PWR) located approximately 3.9 miles south of 

Wiscasset, Maine. The architect engineer for the plant was Stone and Webster 

Engineering Corporation. The Maine Yankee plant started commercial operation in 1972. Its present net electrical power output is 825 megawatts electric 

(2630 megawatts thermal). A brief description of the plant configuration, including structures and systems, is given in Volumes 2 and 3. A detailed 

description of the Maine Yankee plant is given in the FSAR [Maine Yankee, 

FSAR]. 

All seismic 11 Class 1 11 structures and components of the plant which are 

important to nuclear safety, and could affect the health and safety of the 

public, are designed based on a minimum horizontal ground acceleration of 

0.05g and a safe shutdown earthquake (SSE) horizontal acceleration of 0.1g [Maine Yankee, FSAR]. Damping at these acceleration levels is 2 percent and 5 

percent, respectively. Vertical acceleration is taken as two-thirds of the 

horizontal acceleration and is considered to act simultaneously with each 

horizontal component. 

The occurrence of two seismic events in the vicinity of the plant, one in 1979 

and the other in 1982, prompted Maine Yankee to upgrade the capability of the 

plant to withstand a potential seismic event in excess of the original design- basis event. Based on these upgrades and the inherent design capacity of the 

plant, Maine Yankee concluded that the plant structures, systems, and 

1-3 
components had sufficient strength to withstand a seismic event of at least 

0.2g with a Regulatory Guide 1.60 [USNRC, 1973] spectrum and still shut down 

without danger to the public health and safety [Miraglia, 1986]. 

To assure the NRC that the plant could withstand earthquake motion greater than the design basis, the utility agreed to participate in the trial seismic 

margin review of the Maine Yankee plant. For this review, it was agreed the 

seismic margin review earthquake 1 eve 1 wou 1 d be 0. 3g with a 50th percent i 1 e 

Newmark Hall Spectra defined in NUREG/CR-0098. Magnitude and duration 

criteria specified in NUREG/CR-4334 apply [Guzy, 1986], [Crutchfield, 1986]. 

The margins concept requires the HCLPF capacity to be associated with a 

defined response spectrum and a specified nonexceedance probability. 

The HCLPF capacity used for screening as well as the calculated HCLPF 

capacities for particular components not initially screened out and the final 

plant level HCLPF capacity are considered to be valid provided ground motion 

from any earthquake does not exceed the review earthquake level spectrum for 

more than 16% of the spectral frequencies within the range of interest. The 

review earthquake level spectrum is a spectral shape defined by the 50% 

exceedance spectrum specified in NUREG/CR-0098 and anchored at 0.3g pga for 

the initial screening. The seismic margin for the components and plant is 

referenced to this spectrum but anchored to the pga corresponding to the HCLPF 

capacity. 

This definition of spectra used to determine a HCLPF capacity does not in any 

way refer to the probability of occurrence of an earthquake. It is no more 

than an arbitrary spectrum used to define the HCLPF capacity that recognizes the dependency of a component capacity on the frequency content of the 

spectrum and not just the pga. 

The results of the seismic margin study are interpreted as follows. The HCLPF 

capacity of the structures, equipment and plant are conditional on the actual 

site-specific spectrum not exceeding the target spectrum; exceedance is 

defined as the event when 16 percent of the spectral ordinates exceed the 

target spectrum over the frequency range of interest. It is assumed that the 

spectrum peak-to-peak and earthquake direction variabilities are removed from 

the hazard analysis leading to the selection of the review earthquake. The 

review earthquake is specified by the same spectrum in two horizontal 

directions and 2/3 of the horizontal spectrum in the vertical direction. It 

is also assumed that the review earthquake level is specified as the higher of 

the response spesctra from the two orthogonal horizontal directions. 

1-4 
CHAPTER 2 

2. SEISMIC MARGINS APPROACH 

Insights gained from the results of seven published probabilistic risk 

assessments (PRAs) were used in the development of a screening approach that 

combined systems insights and fragility information to simplify the margin review process. This approach is directed at reviewing a specific plant at a 

selected earthquake acceleration level greater than the SSE. 

A general definition of "seismic margin" is stated below: 

11 Seismic margin is expressed in terms of the earthquake motion level 

that compromises plant safety, specifically leading to melting of 

the reactor core. In this context, margin needs to be defined for 

the whole plant. The margin concept also can be extended to any particular function or component ... 

The adopted measure of margin is the earthquake level for which there is a 

High Confidence, Low Probability of Failure (HCLPF). The HCLPF is a conservative representation of capacity and in simple terms corresponds to the 

earthquake level at which it is extremely unlikely that core damage will 

occur. Generallyt the median capacity is at least a factor of 2 greater than 

the HCLPF capacity and thus no proverbial 11 Cliff" or sudden failure is 

expected to occur immediately upon exceeding the HCLPF capacity. From the 

math emat i ca 1 perspective of a fa i 1 ure probabi 1 ity d i st r i but ion on capacity (fragility), the HCLPF capacity is approximately equal to a 95 percent confidence (probability) of not exceeding about a 5 percent probability of 

failure. HCLPF capacity for specific types of components are derived from a 

combination of engineering data, either test data or data from real earthquake experience, and engineering analysis. 

2.1 Systems Screening 

The review of the available seismic PRAs indicated some key trends and 

insights useful in developing seismic margin review criteria. Chapter 4 of 

[Budnitz et al., 1985] discusses the details of the screening approach. These 

trends and insights could only be obtained for pressurized water reactors 

{PWRs), for which there were seven available PRAs. 

The Expert Panel review indicated that systems screening must be performed at 

the functional level due to the diversity of plant designs and system configurations. This led to considering the general plant safety functions 

normally performed in PWRs: 

1. Reactor Subcriticality. 

2. Norma 1 Coo 1 down. 

3. Early Emergency Core Cooling (injection). 

4. Late Emergency Core Cooling (recirculation). 

5. Containment Heat Removal. 

6. Early Containment Overpressure Protection (injection). 

7. Late Containment Overpressure Protection (recirculation). 

2-1 
Examination of PRA results indicated that the dominant plant damage states to 

core melt for seismic events are generally early core melt with early containment failure. In addition, these plant damage states generally involve 

core melt induced by failure of the first three functions listed above 

followed by loss of containment integrity early in the accident progression. This insight led to the division of plant safety function into two groups: 

Group A - Functions 1-3. 

Group B -Functions 4-7. 

The systems screening approach is based on this insight. The dominant plant damage states are caused by failure of the Group A functions and these plant damage states are also characterized by functional failure of the Group B 

functions. Therefore, we consider core damage to occur whenever there is a 

failure of the systems that provide the initial shutdown of the nuclear 

reaction and cooling of the reactor core. These failures are followed by failure of the systems that provide the Group B functions so as to preclude the mitigation of consequences by providing containment protection and 

cooling. 

The normal cooldown function is not considered in our screening criteria 

because a loss of offsite power is assumed to occur in all seismic PRAs as a 

result of .. low capacity" switchyard components. 

The relationship between the Group A and Group B functions indicated that they 

are highly coupled so that the combined failure of Group A and success of 

Group B (or the combined success of Group A and failure of Group B) is 

virtually precluded since their weakest links are coupled. This insight indicates that only those systems and components needed to perform the Group A 

function must be considered in a seismic margins review. 

The Expert Panel also discovered that seismic core melt can occur due to the 

failure of specific unique features. This implies that it will always be 

necessary to perform some kind of plant walkdown to look for any unique features requiring a margin review. It must be emphasized that only major, unique features are of concern. Examples of these are the presence of a dam 

upstream of a plant or a free-standing stack that could damage surrounding buildings if it collapsed. 

The insights gained from our systems review of the PRAs and the development of 

a systems screening criteria to simplify the margin review process has 

resulted in the following conclusions: 

1. It is possible only to come to conclusions regarding the relative 

importance of plant systems and safety function for PWRs for which 

six plants were studied (one by two different methods). No 

function/systemic conclusions can be made about boiling water 

reactors (BWRs) without examination of additional PRAs. 

2. For PWRs, it is possib.le to categorize plant safety functions as 

belonging to one of two groups, one of which is important to the 

assessment of seismic margins and one of which is not. 

2-2 
"I·:~, 

3. The important group involves only two plant functions that must be 

considered for estimating seismic margin. These two functions are 

shutting down the nuclear chain reaction and providing cooling to the 

reactor core in the time period immediately following the seismic 

event. 

4. It is possible to reasonably estimate the seismic margin of the plant by performing a margins study only involving the analysis of the 

plant systems and structures which are required in order to perform those two safety functions. 

Using the systems screening criteria and combining it with fragility insights, 

we can establish functional/systemic guidelines for margin reviews. These 

guidelines show that it is possible to perform a reasonable seismic margin review by concentrating on those functions (and associated systems) which are 

required in the early part of the seismic event and eliminating from the 

review those functions which are not required until later. Further, depending 

on the level of earthquake for which it is desirable to define a margin, certain initiating events would not have to be considered (e.g., large 

LOCAs). This reduces the level of effort and scope of the analysis. 

Event trees need only be constructed up to the point of determining whether or 

not there is an early core damage. Fault trees would only have to be 

constructed for those front-1 ine systems (and their support systems) that 

appear on these abbreviated event trees. By combining a nonseismic failure 

probability and seismic fragility screening criteria with these systems models, it would only be necessary to include those components which have not 

been removed during the screening process. 

2.2 Fragility Screening 

The available fragility information that was reviewed and assessed is based 

primarily on the detailed analysis of nuclear power plants performed for 

PRAs. This PRA information was supplemented by recent systematic investigations of historic earthquakes. The information includes past earthquake performance data for eight classes of equipment obtained by the Seismic 

Qualification Utility Group (SQUG), and reviewed by the Senior Seismic Review 

and Advisory Panel (SSRAP). Work is ongoing to document the historic 

earthquake performance and qualification data for additional components such 

as piping, valve operators, penetrations, diesel generators, battery racks, 

and electrical equipment. This work is being conducted by SQUG, the Electric 

Power Research Institute, and the American Society of Civil Engineers Dynamic Analysis Committee [ASCE, 1986]. Additionally, the collective knowledge of 

the Expert Panel members with actual earthquake experience at nuclear and 

nonnuclear industrial facilities, performance test data, and other analyses 

were included in the assessment. 

These available sources of fragility information were used to arrive at 

conclusions about which components should be assessed from a seismic capacity standpoint. In making statements about the need for capacity assessments for 

each component, three ranges, stated in peak ground acceleration (pga), were 

2-3 
used: (1) less than 0.3g, (2) 0.3g to 0.5g, and (3) greater than 0.5g. Each 

type of nuclear power plant component was assessed to have a generic HCLPF 

capacity within one of these ranges. This resulted in an extensive table of 

components indicating at what earthquake level each component will require a 

margin review or be removed from the review process. This table is given in 

[Budnitz et al., 1985]. 

This categorization of components is based on the available information~ and 

the fragility screening resulting from the use of this table should only be 

performed with consideration of the caveats, limitations, and assumptions 

presented in Chapter 5 of [Budnitz et al., 1985]. 

During the process of assessing the HCLPF capacities for the various nuclear 

power plant components, an extensive fragility information base was developed 

from the available seismic PRAs. This fragility information base is available 

on a diskette for use on personal computers, and is documented in [Campbell et 

a 1., 1985]. 

2.3 ~proach for Performing Seismic Margin Reviews 

The combined insights gained on plant functions and component fragilities were 

used to develop an outline of an approach for performing seismic margin reviews. The review approach consists of eight steps, the first of which is 

the selection of an earthquake review level for which it is desirable to 

demonstrate margin. The eight steps developed in [Prassinos et al., 1986] are 

outlined later in this section. 

It is important to point out the assumptions and limitations of the approach. 

1. The systems screening part of the approach presently applies only to 

PWRs. 

2. The review approach focuses on earthquakes that could occur in the 

eastern part of the U.S., specifically east of the Rocky Mountains. 

3. The assessment of component HCLPF capacities is limited to earthquakes of less than a magnitude of about 6.5, which are characterized 

by 3 to 5 strong motion cycles with a total duration of 10 to 15 

seconds. 

4. The effects of undiscovered design and construction errors are not 

covered. 

5. Possible vulnerabilities in hydraulic systems associated with sensors 

and pneumatic systems are not fully covered. 

6. Electrical and control systems are incompletely covered because 

unrecoverable relay chatter and breaker trip is not adequately treated at this time. 

7. Evaluation of the effect of wear and aging on equipment function is 

not fully covered. 

2-4 
8. Possible adverse human responses caused by earthquake-induced stress 

are not explicitly covered. 

The first three limitations listed above are based on the data from PRAs and 

industrial facilities that were used in the development of this approach and 

present true limitations on the methodology but not on the Maine Yankee 

review. Some of the remaining items represent limitation on our knowledge of 

how to adequately address these issues, while others require considerable 

effort and are beyond the scope of this analysis. 

2.4 Trial Guidelines for Seismic Margin Reviews 

The objective of the seismic margin review guidelines is to provide guidance for determining whether a plant can resist with high confidence a specified 

earthquake level greater than the SSE. To accomplish this objective, analyses 

are performed on components, systems, and the plant to determine what the 

HCLPF capacity is so that it can be compared to the specified earthquake level. Plant failure is defined as the onset of core damage. 

A flow chart of the margin review process is shown in Figure 2-1 This process 

involves the screening of components based on their importance to plant safety and their seismic capacity. Inspection of Figure 2-1 indicates that Steps 2, 

5, and 7 are primarily concerned with plant safety functions and systems, and 

are performed by a team of systems analysts. Steps 3, 6, and 8 are mainly concerned with capacity assessment and are performed by a team of fragility 

analysts. Step 4 is performed by both teams of analysts. The entire process 

requires close cooperation and interaction between the two teams of analysts 

and the utility. 

The initial step in the review process is the selection of the margin review 

earthquake level. The margin review earthquake level selected for the Maine 

Yankee review was discussed in Section 1.2. 

In Step 2, plant information gathering, review, and analysis is performed to 

determine those plant systems and components (structures and equipment) that 

are important contributors to plant safety and thus allow focusing of the 

effort on the components requiring a margin review. Also performed during 

Step 2 is an identification of the relevant seismic initiating events and the 

development of preliminary event trees that describe the systematic behavior 

of the plant following these initiating events. 

The team of systems analysts review the plant information and determine those 

front-line systems that perform the two functions important to plant safety, 

reactor subcriticality and early emergency core cooling. Examples of these 

systems are the reactor scram, emergency boration, high pressure safety injection, and aux i 1 i ary feedwater sys terns. The support sys terns to these 

front-line systems are then determined. Examples of the support systems include electrical power, cooling, actuation, and control. 

2-5 
Gather information 

on systems and sort 

Group A functions. 

Use information 

on Table 2.3 and 

Ref. 1. 

. . . " .. . :: ...... .. .. . . . . . . . . • • ·2· ... • .•• • o I 

. . . . . . .. .. .. . . . 

Start 

Time axis 

Select an earthquake review level 

Gather information on 

~,....,...~~ the plant. Determine which 

broad classes or groups of 

components have HCLPF 

values greater than the review 

......,.~,...._..,. level. Possibly identify plantunique features. 

First plant walkdown: 

Concentrate on identification of problems. 

Emphasize systems interaction. Confirm 

applicability of screening tools. Complete 

identification of plant-unique features. 

KEY: Task is performed by: . . . . . . ...... . • •• •• • •• •• • • Revision of systems 

::. ·.: : • ·:: ·:: relationships established ..•.•. s, • ••·. 

·:. :· :.-:::::.. in Step 2. Develop fault 

~Fragility analysl 

OOsoth 

• • •• •• • ·:·: •••• trees and event trees. • • • • • 1111 ••• 

Second plant walkdown: 

Primarily fragility analyst for chec·ks. 

Collect specific data (size and other 

physical characteristics) or components 

requiring detailed analysis. 

Determine minimal 

cut sets for endpoint core melt. 

. . . . . . . . . t: .. . ·.·.:-·::G .. . . .. ........ :. . . .. . . . . . . ' .. • • • • ·•• 7 •• ·•• • • .... . . .. .. . . . . . . . . . . .. · .. ·.;. :: .. . ·. · .. . . . . . . . 

Margin 

assessment 

complete 

Finalize HCLPF value for 

components in final cut sets 

(components not screened out). 

Figure 2-1 Graphic representation of the screening operations. 

2-6 
Once these systems have been determined, the components that make up these 

front-line and support systems are listed. This list is shared with the 

fragility analysts and these components become the focus of the first plant 

walkdown. 

Using the identified front-line systems, preliminary systemic event trees are 

developed for seismic initiating events. An example of an event tree is shown 

in Figure 2-2. An event tree is a logic diagram that is used to determine the 

sequence of successes and failures of the systems of concern following a 

postulated initiating event. 

An illustration of ari accident sequence is shown in Figure 2-2 by the heavy dark line on the tree. This sequence assumes the initiation event (IE) has 

occurred followed by the failure of sysl, the failure of sys2, the success of 

sys3, and the failure of sys4. This accident sequence is represented by the 

Boolean expression: 

ASl =(IE) * (sysl) * (sys2) * ·(sys3) * (sys4), (1) 

where the system identifier (sysl, sys2, etc.) with the bar indicate system 

success and the others indicate system failure. This equation reads 11 the 

initiating event must occur and sysl must fail and sys2 must fail and sys3 

must succeed and sys.4 must fail 11 for AS1 to occur. The multiplication is the 

mat hem at i ca 1 repres entation of the 1 ogi ca 1 .. and 11 operator where each event 

must occur for the occurrence of the outcome. 

Based on the systems that fail and succeed in each accident sequence Boolean 

expression, it is determined whether that sequence of events leads to core 

damage (CD). If core damage is postulated to occur, we then need to determine 

the occurrence of each event within the accident sequence to determine the 

sequence outcome. Since we have assumed the initiating event always occurs 

given the occurrence of the margin review earthquake, the occurrence of the 

initiating event is taken as a certainty (probability= 1.0). 

Normally, system success is quantified by one minus the probability of system 

failure. However, where system failure is small, the occurrence of each 

system success is assumed to be unity (probability= 1.0}. This assumption is 

made for a s e i s m ic mar g i n rev i ew at a rev i ew earth qua k e 1 eve 1 of 0 . 3 g . For 

higher review levels, system success may need to be considered. 

A Boolean expression for the occurrence of each system failure is determined 

by fault tree analysis. Fault trees are used to determine the combinations of 

front-line and support system component failures that lead to system 

failure. For the Maine Yankee analysis, the component failures are 

represented by their fragility curves and nonseismic failure probabilities. Each accident sequence Boolean expression is then quantified by replacing each 

system failure with its Boolean expression and solving the resulting logical expression for the sequence HCLPF capacity. 

2-7 
Initiating SYS1 SYS2 SYS3 SYS4 event (IE) 

Success 

t 

Failure 

CD = Core damage 

Figure 2~2 Example systemic event trees 

2-8 

Status 

OK 

OK 

OK 

CD 

CD 
A more detailed discussion of systems analysis and event tree development is 

given in the !REP Procedures Guide [Carlson, 1983] and the PRA Procedures 

Guide [USNRC, 1983], 

For the Maine Yankee review, the initiating event was considered a seismic 

induced loss-of-offsite power (LOSP). An event tree was constructed that 

considered the systems that perform the reactor subcriticality and early 

emergency core cooling functions. 

In addition, the event tree considered the integrity of the reactor coolant 

systems (RCS). This event considered whether a loss-of-coolant accident 

(LOCA) would occur as a result of the earthquake along with the LOSP. Only a 

sma 11 LOCA was considered to occur. A 1 arge LOCA was not considered because 

large RCS piping and their supports have generic capacities above the review 

earthquake level and were screened out. The break size of the small LOCA was 

considered to have an area equivalent to a 3/8-to-2-in.-diameter pipe and 

requires the operation of the high pressure injection system for mitigation. 

The event trees developed in this step were preliminary. They were revised 

following the first plant walkdown and finalized after the second walkdown. 

If, however, the first walkdown uncovers reasons for including other systems, initiating events, or components within the analysis, these will be included. 

Concurrently, in Step 3, knowledge gathered from the plant and knowledge of 

the inherent capacity of components is used to sort the components developed in Step 2 into two groups, those with a generic HCLPF capacity larger than the 

review earthquake level and those that have smaller HCLPF capacity. 

For the Maine Yankee review, the culmination of steps 2 and 3 resulted in the 

identification of structures, block walls, equipment, and areas of the plant 

that need to be inspected and reviewed. In addition, the first walkdown was 

planned including organizing the walkdown teams, developing procedures for the 

review of the various components, developing data sheets for recording the 

findings, and making arrangements with the plant for the necessary health 

physics counting, badging and training. 

A first plant walkdown is performed in Step 4. This walkdown is performed to 

inspect the plant and confirm that the plant•s configuration is such that the 

rules developed for doing a margins assessment are applicable and that 

components can be screened out based on the generic evaluation. Assuming this 

is the case, the appropriate components are eliminated from further 

consideration. During this walkdown, any system iriteractions, system dependencies, and plant unique features will be identified along with 

confirmation of the accuracy of the system descriptions and configurations. 

During the first walkdown of the Maine Yankee plant, the analysis team members 

formed into groups and inspected components and plant areas that were 

identified during the previous steps. The Peer Review Group and NRC personnel also formed into groups to walk down the plant. The walkdowns were performed 

with various levels of detail depending on the requirements of a particular 

group. Arrangements were made to have team meetings at the beginning and end 

of each day. Meetings were a 1 so arranged with knowl edgeab 1 e p 1 ant per so nne 1 

to discuss details about the plant and the review. 

2-9 
Following the completion of Step 4, many of the components identified as 

belonging to the two important plant safety functions were screened out based 

on the inspection and their generic HCLPF capacities being larger than 0.3g. 

In addition, plant information that was gathered during this review was used 

to revise the plant models and perform a conservative evaluation of the 

remaining component HCLPF capacities. Those components that were evaluated to 

have a HCLPF capacity larger than 0.3g were also screened out. 

During Step 5, the information and understanding of the operation of the plant 

following Step 4 are used to review and revise, if necessary, the event trees 

developed in Step 3. Fault trees are developed during this step for systems 

that perform the two .safety functions (subcritical ity and early core cooling). 

A system fault tree is a logic diagram that models the various parallel and 

sequential combinations of faults that will result in the occurrence of the 

predefined undesired top event. For the seismic margins review, the faults 

are associated with component seismic failure capacities, human error, and 

other pertinent nonseismic failure events which can lead to system failure 

following an earthquake. A fault tree thus depicts the logical 

interrelationships of basic events that lead to the occurrence of the top 

event, system failure. 

A fault tree is a diagram of 11 gates., which serve to permit or inhibit the 

passage of fault logic up the tree. The gates show the relationships of 

events needed for the occurrence of a "higher" event. The "hi gheru event is 

the "output" of the gate; the "lower" events are the 11 inputs 11 to the gate. 

The type of gate denotes the relationship between the input events required 

for the output event. 

The two basic gates in a fault tree are the "OR" gate(+), and the "AND 11 

gate(*). The OR gate indicates that the occurrence of the output event will 

result from the occurrence of any of the input events. The output failure of 

an OR gate occurs if any of its inputs fail. The AND gate indicates that the 

output event will only occur if all of the input events occur. 

An example of a simplified fault tree is shown in Figure 2-3. This figure 

shows that the system (SYSl) will fail if both components (COMPl and COMP2) 

fail. COMPl fails if either basic event (BEl or BE2) occur, and COMP2 fails 

if both bas·ic events (BE3 and BE4) occur. The basic event represents the 

failure modes of individual components. 

The Boolean expression for the system failure, SYSl, is: 

SYSl = COMPl * COMP2, (2) 

where the mathematical representation of the logical AND operator is 

multiplication. 

The logical expressions for the component failures, COMPl and COMP2, are: 

COMPl = BEl + BE2, (3) 

2-10 
BE1 

COMP1 

SYS1 

and 

BE2 

SYS1 = COMP1 * COMP2 

COMP1 = BE1 + BE2 

COMP2 = BE3 * BE4 

BE3 

SYS1 = (BE1 + BE2) * (BE3 * BE4) 

SYS1 = BE1 * BE3 * BE4 

+ BE2 * BE3 * BE4 

COMP2 

and 

Figure 2-3 Example of a simplified fault tree 

2-11 

BE4 
COMP2 = BE3 * BE4, (4) 

where the mathematical representation of the OR operator is addition. 

The system Boolean expression in terms of the basic events is then derived by replacing the component failures with their respective logical expressions. The system Boolean becomes: 

or after expansion, 

SYSl = (BEl + BE2) * (BE3 * BE4), 

SYSI = (BEl * BE3 * BE4) 

+ (BE2 * BE3 * BE4). 

(5) 

(6) 

Each term in expression (6), separated by the addition operator, is called a 

"cut set." A minimal cut set represents the smallest number of events that 

will cause system failure if all the events fail. 

The above simplified example is intended to show the basic concepts of fault 

trees and systems analysis. Normally, the analysis of fault trees is very complex and requires the use of computers. Logical and Boolean expressions 

are usually very large and can contain thousands of terms that need to be 

reduced using Boolean algebra. 

A more detailed discussion of fault tree analysis is given in the Fault Tree 

Handbook [USNRC, 1981] along with (Carlson, 1983]. 

For the Maine Yankee review, fault trees were developed for the front-line and 

support system that perform the two important safety functions. For the RCS 

integrity event, a fault tree was developed that consisted of an OR gate containing three inputs, as shown in Figure 2-4. These inputs are small pipe 

ruptures, a failure in one of the power-operated relief valves (PORV) causing it to remain open, and a failure of a main coolant pump seal causing a small 

LOCA. During the review, it was possible to screen out the PORV failures and 

the pump seal LOCA because their components were estimated to have HCLPF 

capacities greater than the review earthquake level. 

The small pipe ruptures, however, could not be screened out. This was due to 

radioactivity concerns, because components within the Maine Yankee containment 

structure were unaccessible for review during the walkdowns. In particular, instrument impulse lines that form part of the RCS pressure boundary could not 

be inspected or reviewed. Therefore, these 1 ines could not be screened out 

and were assumed to be a source of a small LOCA at the Maine Yankee plant. This prompted the development of two event trees. One that considered a small 

LOCA concurrent with the LOSP and the other that considered the LOSP with no 

LOCA. 

2-12 
Small 

pipe 

break 

Failure of 

RCS integrity 

(small LOCA) 

Stuck 

open 

PORV 

Primary coolant 

pump , 

seal LOCA 

Figure 2-4 Fault tree for failure of RCS integrity. 

2-13 
At the completion of Step 5, the systems fault trees were "pruned .. by removing those components that had been screened out in the previous steps. Care was 

taken when pruning the fault trees so that the paths from the remaining lower 

level components were left intact. These lower level components represented the possible failures for the systems under consideration. 

A final plant walkdown is performed in Step 6. This walkdown is used to 

obtain additional specific information for determining the HCLPF capacities of 

the components that remain in the analysis. In addition, the systems models 

are verified for accuracy and any additional information needed to complete 

them is obtained. 

During the second walkdown of the Maine Yankee plant, the fragility analysts collected detailed information about those components for which a complete evaluation was necessary. This information included detailed measurements of 

anchorages, equipment supports, and structural details along with gaining an 

understanding of the general physical condition of equipment, their mechanical 

layouts including the peripheral components, and collecting other information 

needed to assess capacities. 

The systems analysts collected information needed to finalize their systems 

models. This included discussions with knowledgeable plant operations and 

maintenance personnel. The information collected included details concerning scheduled test and maintenance frequencies and durations for the components, and collecting plant data on equipment failures and repair times. 

During Step 7, the systems models developed in Steps 3 and 5, and finalized in 

Step 6, are analyzed to determine the Boolean expressions for the seismicinduced core damage accident sequences. This step involves the analysis of 

the event trees to determine the accident sequences that lead to seismic core 

damage and the analysis of the fault trees to determine the Boolean expression for each system failure. 

The development of accident sequence Boolean expressions follow in the same 

manner as the development of the systems Boolean expressions. For the 

accident sequence, the systems identifiers are replaced with their respective 

Boolean expressions. The sequence expression is reduced and the minimal cut 

sets are determined. The fragility curves and nonseismic failures for the 

basic events are then used to quantify the accident sequences. 

A plant level Boolean expression can be derived by logically combining all the 

core damage accident sequence Boolean expressions. The initiating events for 

these plant level accident sequences are assumed mutually exclusive. 

Therefore, the occurrence probability for each initiating event has to be 

determined before the accident sequences can be combined. 

This probability can be compared to the fraction of the time one initiating 

event occurs with respect to the occurrence of all the accident initiators. 

Since we consider a fraction for the occurrence of each accident sequence, we 

multiply accident sequences by a factor called a split fraction. The sum of 

all the split fractions is 1 since we assume that an initiating event always 

occurs. For example, during the SSMRP study of Zion, the small LOCA 

2-14 
"t: :· .... 

initiating event for a 0.3g earthquake was determined to occur about 1% of the 

time. 

The plant Boolean expression is then derived by multiplying each accident 

sequence by the appro p r i at e s p 1 i t fraction and 1 o g i c a 11 y comb i n i n g a 11 the 

sequences. For the Maine Yankee review, two p 1 ant 1 eve 1 ace i dent sequences 

were eventually developed. One for a LOSP initiator concurrent with small 

LOCA and the other for a LOSP initiator with no LOCA. 

The final step in the margin review process, Step 8, is to calculate the HCLPF 

capacities for the important low-capacity components, important. systems, 

accident sequences, and the plant. The HCLPF capacities are finalized for 

those components that appear in the single, double, and some low-capacity 

triple member cut set of the Boolean expression derived from the above systems 

analyses. These HCLPF calculations required detailed structural/mechanical 

analyses based on information gained in the previous steps. The fragility 

curves' for the components are then used to quantify the Boolean expressions 

for the system failures, accident sequences and the plant. 

There are two methods available to calculate the HCLPF capacity of 

components: the conservative deterministic failure method (CDFM) and the 

fragility analysis method (FA). For this trial review, the fragility method 

was used to calculate the HCLPF capacities for these components. This method 

was employed because the fragility analysis team has a detailed understanding 

of its application and use. In addition, the fragility method also allows the 

inclusion of nonseismic failures into the overall plant HCLPF and accidentsequence HCLPF calculations. 

For the FA method, a component•s fragility is represented by a simple model 

using three parameters: median capacity Am, and logarithmic standard 

deviations SR and S11 representing, respectively, randomness in the capacity 

and uncertarnty irr the median value. Using a double lognormal model, 

fragility curves 1 ike the one shown in Figure 2-5 are developed. The 

median, aR and Sy are estimated using design-analysis information, test data, 

earthquake exper ence data, and engineering judgment. The median capacity may 

be estimated as a product of an overall median safety factor times the SSE, 

where the overall safety factor is a product of a number of factors 

representing the conservatisms at different stages of analysis and design. 

When the scaling of response is not appropriate (e.g., soil sites), the median 

capacity is evaluated using median structural and equipment response 

parameters, median material properties and ductility factors, median capacity 

predictions, and realistic structural modeling and methods of analysis. 

The HCLPF capacity is expressed using this fragility model as: 

HCLPF = Am exp [-1.64 (SR + Su)J • 

The FA method allows the combination of nonseismic and seismic failures for 

the determination of plant HCLPF capacities. A more detailed explanation of 

calculating component HCLPF capacities can be found in [Kennedy and Ravindra, 

1984 J 0 

2-15 
.0 

ca 

.0 

0 

.:...,. 

c.. 

ca 

c: 

0 

-

0.8 

0.6 

0.4 

"0 

c: 0.2 

0 

u 

0.05 

95°/o confidence 

curve\ 

"-_Median fragility 

curve 

5°/o confidence 

curve 

0~--~~--~C-------~~~----~--------~--------~ 

0 0.17 0.3 0.4 0.8 1.2 1.6 2.0 

Peak ground acceleration (g) 

Figure 2-5 Example fragility curves for a structure. 

2-16 
The unavailability of the components (failure per demand) is determined by combining its random failure probability with its unavailability due to normal 

test and maintenance. Equipment unavailability also included both planned and 

unplanned maintenance and repair. The unavailability due to random failures 

considers the time between normal test and maintenance, or between scheduled 

plant outages. A component's unavailability due to human error is considered 

as a separate event. 

The component failures are combined following the rules of Boolean algebra and 

a discrete probability distribution (DPD) numerical procedure described by [Kaplan, 1981]. For this purpose, the component fragilities are discretized 

into a family of fragility curves. Each fragility curve is assumed to be 

completely described by the median capacity and the value of aR. The 

fragility curve is not truncated in either the lower or upper tail. 

The 11 Union" operation is performed for the first two components to obtain the 

composite fragility curves which are in turn combined with the next component 

fragility curves with either 11 Union11 or 11 intersection 11 operation in the 

Boolean expression. After each operation, the resulting fragility curves are 

condensed to keep the computation manageable. 

The details of the analysis are given in Volumes 2 and 3. 

2-17 
j 
'. ' ~' ' 

CHAPTER 3 

3. RESULTS OF SEISMIC MARGIN REVIEW 

An objective of this trial seismic margins review is to assess the capability 

of the Maine Yankee plant to withstand a specified earthquake level greater 

than the SSE. ·The results of this review consist of: 

o Boolean expressions for each seismic-induced core-damage accident 

sequence. 

o The dominant component for plant seismic safety. 

o An assessment of the HCLPF capacities for important components, 

accident sequences, and the plant. 

o Insight into the seismic behavior of the plant systems required to 

fulfill the safety functions of subcriticality and early emergency 

core cooling injection. 

The HCLPF capacities for the important components are 1 ogi ca lly combined as 

indicated by the Boolean expressions to estimate the HCLPF capacity for each 

core-damage accident sequence. Each of these accident-sequence HCLPF 

capacities represents a plant HCLPF capacity for the particular initiating 

event and plant systems response. 

A plant level Boolean expression can be derived by logically combining the 

accident sequences after they have been multiplied by their respective split 

fraction. , An overall plant HCLPF capacity can then be determined from the 

plant level Boolean expression. 

Section 3.1 presents a summary of the overall result of the Maine Yankee 

review. A surrvnary of the systems results, including a discussion of the 

accident sequences, is given in Section 3.2. Details of the systems analysis 

are in Volume 2. A summary of the component capacity assessment is given in 

Section 3.3. Details of the component capacity assessment are in Volume 3. 

3.1 Overall Results 

Following the review of plant information, a list of the components that make 

up the front-1 ine and support systems required to perform the plant safety 

functions was developed. This list is given in Volume 3 and provided the 

basis for the remainder of the review. After the plant walkdowns and 

subsequent ana 1 yses by the systems and fragi 1 it i es analysts, the 1 i st of 

components was reduced by screening out those components that had HCLPF values 

greater than 0.3g. For the remaining components, HCLPF capacities were 

calculated. The list of components for which a HCLPF capacity was calculated 

is given in Section 3.3. 

3-1 
;- .. -.. '

These remaining components were used in the development of the event trees and 

fault trees for the seismic-induced core-damage accident sequences as 

described in Volume 2. The event trees and fault trees were analyzed to 

determine Boolean expressions for each accident sequence that could lead to 

core damage. The component failures that are significant to these Booleans 

are given in Tables 3-1 and 3-2. Table 3-1 gives the seismic induced failures 

along with the fragility parameters used to quantify their HCLPF capacities. Table 3-2 gives the nonseismic failures and their unavailabilities. Note that 

the component items and nonseismic failure events are numbered consecutively; the missing numbers represent the items that were screened out in the final 

pruning of the event and fault trees. 

One component has been upgraded following the review during the two plant 

walkdowns. This component (number 4, Table 3-1) is a General Electric (GE) 

station service transformer (4160 V to 480 V) that supplies power to pumps and 

components needed to perform the seismic safety functions. 

This transformer was originally installed with a 11 floating" bus bar to limit 

the amount of noise on the system. Several years after its installation, GE 

performed seismic qualification testing on these types of transformers and 

made modifications that essentially resulted in securing the "floating .. bus 

bar and greatly increasing seismic capacity. Preliminary estimates of the 

transformer HCLPF capacity were approximately O.lg. More detailed 

calculations could have been performed, however, the utility decided to 

upgrade this component during its March 1987 outage. An analysis of this 

planned upgrade has shown an increase in the estimated HCLPF capacity for the 

transformer to 0.30g. 

Another component appearing in Table 3-1 with a HCLPF less than 0.3g is Number 

20, the seismic failure of the primary water storage tank (PWST). This tank 

provides an alternate supply of water to the auxiliary feedwater system. The 

PWST has a HCLPF of 0.27g. This tank does not appear in either of the core 

damage Boolean expressions. 

There is one type of component, not listed in Tables 3-1 or 3-2, for which 

there was insufficient data to determine the HCLPF capacity. This component is the lead-antimony type batteries used at the Maine Yankee plant. The four 

sets of Station Batteries (1, 2, 3, 4) have passed their electrical 

qualification testing. However, there is no seismic qualification or test 

data available on these types of batteries to estimate the seismic capacity of 

the aged lead-antimony plates within the battery casings. 

Systems analysis indicated that the important batteries to plant seismic 

safety are Station Batteries 1 and 3. Maine Yankee had intended to change out 

one of the station batteries during their plant refueling outage scheduled for 

March 1987. This analysis prompted them to replace both Station Batteries 1 

and 3. The remaining Station Batteries 2 and 4 have been scheduled for 

replacement during the 1988 outage. 

3-2 
Table 3-1 Component seismic fragility parameters. 

Item No. Item ~R 

4 Transformers 0.84 0.30 0.32 

(X507, X508) 

7 RWST 0.45 0.20 0.25 

(TK-4) 

8 DWST 0.36 0.20 0.26 

20 Circulating Water 0.69 0.24 0.27 

Pumphouse 

21* PWST 0.57 0.20 0.26 

(TK-16) 

* HCLPF less than 0.3g, but does not appear in the plant Boolean 

3-3 

· HCLPF 

Capac it y ( g ) 

0.30 

0.21 

0.17 

0.30 

0.27 

express ions. 
Table 3-2 Probabilities for nonseismic failures. 

Median 

Unavailability Error 

Item No. Description (per demand) Factor* 

10 Operator Failure to Close PCC B.OE-02 2 

Isol. Valves 

11 Random Failure of DG-1B 4.2E-02 5 

12 Random Failure of DG-1A 4.2E-02 5 

13 Operator Failure to Place AFW 1. 5E-01 2 

Pump Train B in Service Locally 

14 Nonseismic Common Cause 1. 6E-03 5 

Failure of DGS 

15 Nonseismic Common Cause 1. 2E-04 5 

Fai 1 ure of AFW 

16 Operator Failure to Refill DG B.OE-03 3 

Fuel Tanks by Opening Valve or 

Running P-33A,B 

17 Operator Failure to Place AFW 4. OE-02 3 

Pump Train B in Service from MCR 

22 Random Failure of the Turbine 3.0E-02 5 

Driven Aux Feedwater Pump 

*Error factor equals (95% Confidence Value/Median Value). 

:' 

3-4 
Seismic qualification and test data on the new batteries indicates that they have a HCLPF capacity greater than the review earthquake level. Subsequently, the station batteries were screened out and eliminated from further 

consideration. The new station batteries and racks should be inspected for 

proper installation according to Expert Panel guidance after they are in 

place. 

The capacity assessment of the new batteries being installed at the Maine 

Yankee plant assumes they possess the qualities of new equipment and are 

installed correctly. However, as with the analysis of all components during this seismic margin review, the effects of aging beyond this .. snapshot" of the 

plant are not considered. 

The analysis of the two event trees resulted in two Boolean expressions that 

lead to seismic-induced core damage. One of these expressions is the logical 

combination of three accident sequences that were initiated by the seismicinduced LOSP concurrent with a small LOCA. The other express ion is the 

logical combination of two accident sequences initiated by the seismic-induced 

LOSP without a small LOCA. These two Boolean expressions are given below: 

Small LOCA Core Damage 

(SL) [4 + 7 + 20]. 

No LOCA Core Damage 

= (LOSP) [(4 + 20) * (8 + 13 + 15 + 17 + 22) 

+ 8 * (14 + 16) + 15 * 7], 

where the numbers in the expressions correspond to the failure of the 

components given in Tables 3-1 and 3-2. Entries with designators 4, 7, 8, 20, 

are seismic-induced failures and given in Table 3-1. Entries with designators 

from 10-17 and 22 are nonseismic failures and given in Table 3-2. The missing numbers in Tables 3-1 and 3-2 are component screened out during the final 

screening. The terms ·SL and LOSP represent the small LOCA and loss of offsite 

power initiating events, respectively. In the above expression, the 11 +" 

notation denotes probabilistic addition (union) and the 11 * 11 denotes 

probabilistic multiplication (intersection). 

The small LOCA initiating event could be determined using the guidance given by the Expert Panel. This guidance indicates that for this review earthquake level, a walkdown of sample piping system should be conducted along with 

inspection of piping between buildings to look for possible problems such as 

weak spots, unanchored attached equipment, stiff sections between flexible 

termination points, and brittle connections. 

During the walkdown of the Maine Yankee plant, none of these problems were 

identified and piping was found not to be a problem. Therefore, the small 

LOCA initiating event could be screened out thus eliminating the small LOCA 

Boolean expression from the seismic margins review. However, the analysis 

3-5 
teams felt that eliminating the small LOCA from consideration was not 

conservative. The possibility of a small LOCA should not be screened out 

because the containment, which contains many small RCS pipes :t could not be 

inspected and reviewed as previously discussed. 

Impulse line failures were assumed to be the source of a small LOCA at Maine 

Yankee. This conservative assumption was required due to the tremendous 

number of hours which would be required to walk down each of these impulse lines and assess potential system interaction problems. These lines originate from the primary pressure boundary inside containment (i.e., RPV, steam 

generator, pressurizer, primary coolant loop piping, etc.) and are field 

routed to instrument racks inside containment. The amount of work required to 

demonstrate the seismic margin in each of these 1 ines plus the fact that 

walkdown of these lines would have to take place during a plant outage necessitated the assumption of a small LOCA as an initiating event. 

Inspection of the small LOCA core-damage Boolean expression indicates that the 

dominant components are the three singletons 4:t 7, and 20. The singleton 

component .with the lowest HCLPF capacity is the refueling water storage tank 

(RWST, number 7) with a HCLPF capacity of 0.21g. Failure of this tank results 

in no coolant being available for reactor vessel injection following a small 

LOCA. 

The other singleton components have HCLPF capacities of 0.30g. The capacity of the transformer (number 4) is estimated based on the proposed upgraded 

condition. The other singleton is the circulating water pumphouse (Number 

20). 

For the no LOCA core-damage Boolean expression, the occurrence of the LOSP 

initiating event is assumed a certainty (probability = 1.0) due to the low 

capacity of switchyard components. Therefore, the LOSP initiating event could 

be removed from the no LOCA Boolean expression. 

The HCLPF capacity for the no LOCA core damage Boolean expression is estimated 

to be greater than 0.30g. The higher capacity for this sequence, compared to 

the small LOCA sequence, is due to the absence of singletons and no low 

capacity doubletons in the expression. Although the DWST, i.e., component 8, 

with a HCLPF capacity of 0.17g appears in this sequence, its failure has to 

occur simultaneously with one of the higher capacity components, i.e., the 

transformer or the circulating water pumphouse. 

The most important nonseismic failure is number 15, a common cause failure of 

the auxiliary feedwater system caused by steam binding. This failure results 

in the inability to cool down the reactor coolant systems using the steam 

generators. This nonseismic failure is the most important because it appears in a majority of the doubleton cut sets. 

Figures 3-1 and 3-2 show plots of the small LOCA and LOSP core-damage fragility curves, respectively, in which the family of fragility curves is 

reduced to the 5%, 50%, and 95% confidence levels. · 

3-6 
~ ' ' ' 

1.0 

Q) Conf. .. 0.05 .-- :::s 

-·as 0.8 0.50 ..... 0.95 ..... 

0 

>. 

... 0.6 

:s 

ca 

.a 

0 

.. 

a. 0.4 

-ca 

c: 

0 

; 

·- 0.2 "'C 

c: 

0 

0 "t··' 

Peak ground acceleration (g) 

Figure 3-1 Fragility curves for small LOCA core damage 

3-7 
1.0 

Q) 

I... 

::3 

co 

..... 

Conf. 

0.8 0.05 

0.50 ..... 

0 0.95 

~ 

- 0.6 ·-.c 

ca 

.c 

0 

I... 

0.. 0.4 

-co 

s:::: 

0 

'"C - 0.2 

c 

0 

{) 

•,,• I 

I.'' 0.4 0.6 0.8 1.0 

Peak ground acceleration (g) 

Figure 3-2 Seismic Fragility Curves for No-LOCA Core Damage. 

3-8 
To account for the fact that we could not quantify the small LOCA initiating 

event, we can develop an overall plant core-damage Boolean expression by logically combining the two Boolean expressions accounting for the split fraction p between the small LOCA and LOSP initiating events. This overall 

plant core-damage Boolean expression is given below: 

Core Damage 

= p [small LOCA core damage] + 

( 1 - p ). [no LOCA core damage] 

In this case, the split fraction accounts for the fraction of the time a small 

LOCA will occur along with the LOSP event. 

When evaluating the HCLPF capacity for the small LOCA Boolean expression, there must be a consideration for the occurrence of the small LOCA initiating 

event. If the piping, whose failure will cause the small LOCA, has a HCLPF 

capacity l·ower than all the singleton components in this Boolean expression, the lowest capacity component (the RWST, number 7) will dominate the small 

LOCA core damage HCLPF capacity. If, however, the piping has a HCLPF capacity 

larger than the lowest capacity singleton, then the piping will dominate the 

small LOCA HCLPF capacity. 

Overall Plant Core Damage HCLPF Capacity 

As indicated above, to· account for the two Boolean expressions in the overall 

plant core-damage HCLPF capacity, a sensitivity calculation can be performed accounting for a variation in the split fraction between the two accidentsequence initiating events (small LOCA and no LOCA). For different assumed 

split fraction values, the overall plant core-damage HCLPF capacities were 

obtained as shown in Table 3-3. · 

The conclusion regarding the dominance of RWST failure in the HCLPF capacity estimation (displayed in the small LOCA accident sequence) is a function of 

the split fraction assumed. If the plant HCLPF capacity needs to be 

increased, it is not necessary to concentrate only on RWST. A walkdown and 

review of small impulse lines within the containment may be performed to 

estimate their fragilities in order to assign a realistic HCLPF capacity or 

split fraction. By this procedure, the plant HCLPF capacity may be shown to 

be higher without the necessity of any upgrading of the components. 

Effect of Nonseismic Failures 

The overall plant HCLPF capacity was calculated using the Boolean expressions for the core-damage accident sequences which contained both seismic (Table 3-

1) and nonseismic (Table 3-2) failures. Since this is a seismic margin review 

and the interest is only in the seismic capacity of the plant, one may choose 

to ignore the nonseismic failures in calculating the overall plant HCLPF 

capacity. In the small LOCA Boolean expression, there are no significant nonseismic failures. The effect of not including the nonseismic failures on 

the HCLPF capacity of the no LOCA Boolean had no effect on the plant HCLPF. 

3-9 
Table 3-3 Summary of plant level HCLPF capacities. 

Case Description HCLPF Capacity (g) 

1 Small LOCA 0.21 

- Independent Seismic Failures 

2 Small LOCA 0.21 

- Dependent Seismic Failures 

3 No LOCA ii: 0.30 

- Independent Seismic Failures with 

Nonseismic Failures 

4 No LOCA ~ 0.30 

- Independent Seismic Failures 

without Nonseismic Failures 

5 Core Damage 

- Split Fraction p - 0.01 ~ 0.30 

p = 0.10 0.28 

p = 0.50 0.23 

6 Small LOCA 

- RWST 0.26 

Reduced Fluid Level 

3-10 
Correlation Between Seismic Failures 

The above calculations were performed assuming perfect independence between 

seismic failures of different components; i.e., the seismic capacities are 

assumed to be statistically independent both in randomness and uncertainty. 

This' is a realistic assumption because the components involved in the coredamage Boolean expression are yard tanks (RWST, and DWST), transformer and 

circulating. water pumphouse. These are dissimilar items, their locations in 

the plant are different, and their dynamic characteristics are different. 

Hence, correlation in the seismic responses and capacities of these components 

is judged to be minimal. · 

It is realistic, however, to expect some correlation in the component 

failures. The assumption of perfect dependence in both uncertainty and 

randomness is an extreme case. Assumption of perfect , dependence iri the 

uncertainties of different component fragilities means that the median ground 

acceleration capacity of a component is known if the median ground 

acceleration capacity of another component is known. Since uncertainty arises 

from an insufficient understanding of structural material properties, 

approximate modeling of the structure, inaccuracies in the representation of 

mass and stiffness, and the use of engineering judgment in lieu of plant 

specific data, it is expected that all components will be affected to some 

degree by these uncertainties. Therefore, some probabilistic dependence 

between component median capacities may be expected. 

Dependence in the randomness arises from a common earthquake generating the 

responses in different components and common structural/material properties. 

Assumption of dependence in the randomness means that if the fragility 

(conditional probability of failure) of a component for a _given peak ground 

acceleration is known, the probability of failure of the other components is 

somewhat modified by that knowledge. 

The plant level fragility, and therefore, the HCLPF capacity depends on the 

degree of dependence in randomness and uncertainty between the component 

failures. However, the degree of dependence is difficult to estimate. One 

approach is to bound the core-damage HCLPF capacities by assuming perfect 

dependence as opposed to the case of perfect independence. This calculation 

was performed for the small LOCA core-damage Boolean expres~ion given above. 

The small LOCA core-damage HCLPF capacity is estimated to be 0. 21g, i.e., 

governed by the capacity of RWST. If the seismic failures were assumed to be 

perfectly dependent and the nonseismic failures were assumed to be perfectly 

independent. The plant level HCLPF is estimated to be 0.2lg. 

When the Boolean expression is dominated by singletons, the assumption of 

perfect independence is more severe than the assumption of perfect dependence 

between failures if the fragilities are approximately equal; if there is a 

single component with a very low capacity compared to the rest of the 

components in the Boolean expression consisting of singletons, both 

assumptions give about the same plant level HCLPF capacity. 

The question of dependence between failures is important when there are 

similar components experiencing common seismic excitation. The case in point 

3-11 
..... ·• 1, 

is the yard tanks (i.e., RWST, DWST, and PWST). By reviewing the cut sets, it 

was found that a tripleton cut set could lead to core damage. It is 

DWST * PWST * RWST. 

Although high dependence between the failures of tanks is possible and their 

seismic capacities are similar, the cut set should have a HCLPF capacity of 

0.27g. 

Deterministic Method 

This approach is based on the assumption that the HCLPF capacities of 

components are true .lower bound values. The HCLPF capacity of the plant is 

obtained directly by studying the Boolean expressions for small LOCA and no 

LOCA: 

Small LOCA Core Damage 

= 4 + 7 + 20 

No LOCA Core Damage 

(4 + 20) * (8 + 15 + 17 + 22} 

+ 8 * (14 + 16) + 7 * 15. 

For calculating the plant level HCLPF capacity in this method, the nonseismic 

failures are ignored. Also, the cut sets that includes low probability nonseismic failures are also omitted from this estimation. Therefore, the 

simplified no LOCA Boolean expression becomes: 

No LOCA Core Damage 

= (4 + 20) * 8. 

In the deterministic method, the HCLPF capacity of a "doubleton" cut set is 

represented by the higher of the two component HCLPF capacities. The HCLPF 

capacity of a 11 tripleton 11 cut set is represented by the highest of the three 

component HCLPF capacities. The HCLPF capacity of a union of singleton cut 

sets is estimated to be the lowest of all the component HCLPF capacities. Using this procedure, the HCLPF capacities of the core damage accident 

sequences are: 

HCLPF Capacity for Small LOCA Core Damage 

= min [0.30, 0.21, 0.30] 

= 0.21g. 

HCLPF Capacity for No LOCA Core Damage 

=max [min (0.30, 0.30), 0.17] 

3-12 
,-·' 

'",·. 

= rna x .[ 0 • 3 0 , 0. 17] 

= 0.30g. 

Note that the failure of the DWST, with a HCLPF capacity of 0.17g, is not 

governing the plant level HCLPF capacity because the DWST has to fail 

simultaneously with one of two nonseismic failures to lead 3to core damag3. Since these failures have median probabilities of 6 x 10- and 8 x 10-, 

respectively, it is appropriate to ignore the DWST failure in the plant level 

HCLPF capacity calculation. 

Sensitivity Studies 

Two sensitivity studies were performed to assess the effect of certain 

assumptions on the plant HCLPF capacity. These sensitivity studies addressed 

the following: 

o Effect of shearwall stiffness reduction on structure response and 

equipment seismic input. 

o Reduction of RWST fluid level. 

For the first sensitivity, on-going scale model testing at Los Alamos National 

Laboratory being sponsored by the NRC has indicated potential reductions in 

the stiffness of concrete shear walls of up to a factor of 4, from elastically calculated values due to cracking. This would imply a reduction in the 

elastic structure frequencies of up to 50%. These variations were considered 

at the suggestion of the Peer Review Group, however the research results are 

still preliminary. 

For no LOCA, the 4160/480-V transformer is the dominant contributor· to the 

plant HCLPF capacity. The transformer has a 12-13Hz fundamental frequency 

which corresponds to a spectral acceleration on the downward slope of the 

floor response spectra. A reduction of the building frequencies would result 

in a reduction of the seismic input to the transformer with a corresponding increase in its HCLPF capacity. As a further check on the effect of the 

building frequency shift, a ·similar evaluation was made on each of the 

components within the final Boolean expression. The potential building frequency reduction did not lower the HCLPF capacity for any of these 

components. Thus, for the purpose of the Maine Yankee seismic margin study, the shear wall stiffness reduction and resulting building frequency shift does 

not affect the plant seismic margin. 

For the second sensitivity study, the effect of reducing the fluid level 

within the RWST was investigated. A reduction of the fluid level to an 

arbritary height of 33 ft, which is approximately equal to 10% reduction from 

the current level of 37 ft, was assumed. This change leads to a reduction in 

the effective fluid weight, mass, centroid height, and overall tank seismic 

loads. The RWST HCLPF capacity is increased to 0.28g with a corresponding increase of the small LOCA HCLPF capacity to 0.26g. 

3-13 
3.2 Systems Results 

A review of the Maine Yankee plant information resulted in identifying five 

front-line systems that are required to fulfill the safety functions of 

reactor subcriticality and early ECC injection. These systems are: 

o The reactor protection system (RPS) including the control rod drive 

system used to shutdown the nuclear chain reaction in the core and the 

core internals through which the rods pass. 

o The boric acid transfer system used to provide emergency boration to 

the reactor system and also shut down the nuclear chain reaction in 

the core. 

o The high pressure safety injection (HPSI) system used to supply 

coolant to the primary system. 

o The auxiliary feedwater system used to cool down and depressurize the 

primary system through the steam generators. 

o The pressurizer power-operated relief valves (PORV) used in a feed and 

bleed mode to depressurize the primary system. 

The support systems to these front-line systems are shown in matrix form in 

Table 3-4. The support system versus support system matrix is shown in Table 

3-5. The components that make up the front-1 ine and support systems are 

identified and listed in Volume 3. 

During the review of the plant and as a result of discussions with Combustion 

Engineering (CE, the NSSS supplier), it was determined that the components 

that make up the RPS have HCLPF capacities larger than the review level 

earthquake level based on data and calculations performed by CE. These 

components included the control rod drive mechanisms, core internals, and the 

RPS actuation systems. Subsequently, the reactor subcriticality function was 

screened out and eliminated from further consideration. Screening out the 

reactor subcriticality function also eliminates the need for consideration of 

the boric acid transfer system. 

Systemic event trees were developed for (l) seismic-induced loss-of-offsite 

power (LOSP) concurrent with a small LOCA and (2) LOSP only initiating 

events. Fault trees were developed for the front-line and support systems 

using the identified seismic components that were not initially screened 

out. The fault trees incorporated nonseismic failures including random 

failure, common cause failure, and operator error. A discussion of the 

success criteria, assumptions, and bases for the fault trees is given in 

Volume 2. 

When modeling the primary component cooling system (PCCS), the support system 

which provides cooling to one train of required front-line equipment, it was 

discovered that a large portion of the system cools nonessential shutdown 

equipment which is not automatically isolated following an earthquake. 

3-14 
Table 3-4 Front-line system vs. support system dependency matrix. 

SUPPORT SYSTEMS 

AC Power DC Power ccw SIAS IA 

FRONT-LINE Bus 5 Bus 6 DC-1 DC-3 Train Train 

SYSTEMS Bus 7 Bus 8 DC-2 DC-4 PCC sec A B TK-25 

HPSI/ P-14A 

RECIRC to RCS X X X X 

P-14B 

to RCS X X X X 

FN-44A X X 

FN-44B X X 

AFW P-25A X X 

P-25B X 

P-25C X X 

PROVs PR-S-14 X 

PR-S-15 X 

NOTE: To determine the front-line system dependencies on the support systems, locate the front-line component in the first column and read across the 

row to find the support system dependencies. 

CCW =component cooling water 

SIAS = safety injection actuation system IA = instrument air 

3-15 
w 

I 

--' 

C) 

: \,., 

Table 3-5 Support system vs. support system dependency matrix. 

--------·-------~£ .f!!~~r. _________________ ----~ f~~~L--

... H~t.!!!L _!~!_~ -HQL~lHL!!!!_ ~-- __ !g~!.!!!! __ -~f~- Sll sus IIYAr. lA -- ----- -ra=--tii:-fH- Fi- --tK: 

SUPPORT SfSTEH 5 ' 1 8 I 2 l 4 lA 2l l 2 3 4 PCC SC£ A 8 2UA 2118 ll 12 IJR 

AC 4l60V 5 I I 1 I 

Po.,er Bus ' I I I I 

480V 1 I I 

Bus B I I 

uor I I 

Vital 2 

Bus J 

4 

Diesel lA X I X 

G~ne.utor 18 I X 

DC 125V I I I I 

Power Bus 2 I • I 

J I I I 

4 I I I 

ccw PCC I I I 

sec I 

s"s I I 

SIAS Channel l I 

I 

t l 

D I 

kt. SIS-A l 

SIS-8 

IIYAC FN-ZOA 

ffl-208 

fll-l1 

FU-ll I 

tlote: To deter•lne the support syste. dependencies Pn other support systeiiS, loute the support syste. In the first col~~~~~n. alld re.ad •cross the row to deter•ine dependencies on 

lht othtr suppurt systet~~s. 
Therefore, leakage in these components, if not isolated, could compromise the 

entire system. Much of this equipment, located in the PAB and containment, 

was not inspected during the walkdowns due to accessibility. This 

consideration prompted the development of postearthquake procedural guidance requiring the isolation of nonessential portions of the PCCS in the event of a 

major earthquake and combined plant trip. The nonessential portions of the 

system can be isolated using valves which are remotely operated from the 

control room and have HCLPF capacity greater than the review earthquake level. This isolation of nonessential components has no impact on seismic 

safety. Based on this procedure change, the only remaining failure to be 

considered is the operator error associated with not following the procedure and closing the valves (component No. 10 in Table 3-2). 

The event tree analysis resulted in five accident sequences that led to 

seismic-induced core damage. Three of these are small LOCA sequences and two 

are no-LOCA sequences. The three LOCA sequences are designated S2D, S2LD, and 

S2LP2, and given below: 

S2D = HPSI 

S2LD = HPSI * AFW 

S2LP2 = PPS-LOCA * AFW, 

where HPSI, AFW and PPS-LOCA represent the high pressure safety injection 

system, the auxiliary feedwater system, and the plant pressure protection 

system operating for a LOCA condition, respectively. 

For the seismic small LOCA, core damage will result if the HPSI system fails 

(S2D), or if both the HPSI and the AFW systems fail (S2LD), or if both the AFW 

system and power-operated relief valves (PORV) fail (S2LP2). For the LOCA 

case, both PORVs must fail since only one PORV is required to operate within 

approximately 30 minutes for feed and bleed. The block valve on each 

pressurizer relief line is included with its respective PORV. 

The combined LOCA Boolean expression is the logical summation of these three 

accident sequences. 

LOCA Core Damage = S2D + S2LD + S2LP2 = S2D + S2LP2 

The two no-LOCA accident sequences are designated TILD and TlLPl, and are 

given below: 

TlLD = HPSI * AFW 

TlLPl = AFW * PPS, 

where the PPS represents the p 1 ant pressure protection sys tern operating for 

the no-LOCA case. 

For the no-LOCA case, core damage will result at Maine Yankee if both the AFW 

and HPSI systems fail (TlLD) or if both the AFW and one PORV fail (TlLPl). 

Based on PRA results, for the no-LOCA case, only one PORV must fail since both 

PORVs must open within approximately 30 minutes for feed and bleed. 

3-17 
. ~ ... 

The systems failures were determined by fault tree analysis. The systems 

fault trees were developed and analyzed following an 11-step process. A brief 

description of the 11 steps is given in Volume 2. 

The analysis of the fault trees gives the minimal cut sets that lead to system 

failure. The single-failure (singleton) and double-failure (doubleton) cut 

sets for the systems given in the accident sequences are shown in Table 3-6. 

The cut sets shown contain either seismic-only failures or a combination of 

seismic and nonseismic failures. 

The HPSI system contains three seismic singletons and no seismic doubletons 

that lead to system failure. The AFW system contains no seismic singletons 

and 10 daub 1 et ons. The PPS sys tern cant a ins two singletons and no dou b 1 etons 

operating for both the small LOCA (PPS-LOCA) and no-LOCA (PPS) sequences. 

3.3 Fragility Results 

The safety system components identified by the systems analysts were reviewed 

and subjected to three different types of screening. 

1. Those components with a generic HCLPF value greater than the review 

earthquake level, as given in Table 2-1 in [Prass inos et al., 1986], 

were flagged for inspection during the first plant walkdown. The first 

plant walkdown confirmed that these components could be screened out 

based on their generic HCLPF values and the use of the screening 

table. Not all components with a generic HCLPF value greater than the 

review earthquake level were screened out after the first plant 

walkdown. 

2. The remaining components were examined following a detailed walkdown 

review of their seismic capacity during the second plant walkdown. If 

the seismic capacity was judged to be greater than the review 

earthquake level, they were screened out. 

3. The components remaining after the two screenings 1 isted above were 

screened based on a calculated HCLPF capacity. If their calculated 

HCLPF capacity was greater than the review earthquake level, they were 

screened out. Data needed to calculate the component HCLPF capacities 

were collected during both plant walkdowns, but specifically during the 

second plant walkdown . 

Both plant walkdowns were also performed to verify the configuration of the 

plant systems and to look for systems interaction and any plant unique 

features. 

The components remaining after the first screening and those requiring support 

systems and operator action to perform the safety functions were used in the 

development of the system fault trees. As the subsequent screenings were 

performed, the system fault trees were pruned to reflect the elimination of 

these components. 

3-18 
,.f .. 

Singletons 

Doubletons 

Singletons 

Doubletons 

Table 3-6 Minimal cut sets for important systems. 

High Pressure Safety Injection System (HPSI) 

1. 4160 V to 480 V station service transformer. 

2. Refueling water storage tank (RWST). 

3. Circulating water pumphouse. 

4. No seismic double failures. 

Auxiliary Feedwater System (AFW) 

1. No seismic single failures 

2. Primary water storage tank (PWST), and demineralized water 

storage tank (DWST). 

3. 4160 V to 480 V station service transformer, and the DWST. 

4. Circulating water pumphouse and th~ DWST. 

5. The DWST and human error, failure to open AFW pump trains A 

& 8 PWST isolation valves. 

6. 4160 V to 480 V station service transformer, and human 

error, failure to place AFW pump train 8 (P-258) in service 

from the control room. 

7. 4160 V to 480 V station service transformer, and random 

failure of AFW turbine driven pump P-25B. 

8. Circulating water pumphouse and random failure of AFW 

turbine driven pump P-258. 

9. Circulating water pumphouse and human error, failure to 

place AFW pump B (P-25B) in service from the control room. 

3-19 
Doubletons 

(Cont.) 

Singletons 

Doubletons 

Table 3-6 Minimal Cut sets for important systems. {Cont.) 

Auxiliary Feedwater System (AFW) (Cont.) 

10. DWST and human error, failure to refill diesel generator tanks by opening valves and/or running auxiliary fuel pumps. 

11. DWST and nonseismic common cause failure of the diesel 

generators. 

Plant Pressure Protection System (PPS) 

(small LOCA and no-LOCA) 

1. Circulating water pumphouse. 

2. 4160 V to 480 V station service transformer. 

3. No seismic double failures. 
The screening processes resulted in a reduced list of components for which a 

HCLPF capacity was calculated. This reduced list has been divided into two 

lists, one consisting of structures and block walls, and the other consisting of equipment. These two lists are given in Tables 3-7 and 3-8, respectively, along with the calculated HCLPF capacities. 

3-21 
. ' .. 

Table 3-7 Maine Yankee Structures and Block Walls· 

Str·ucture 

Circulating Water Pumphouse, steel portion above El 21'-0" 

Construction 

Structural steel framing and 

diagonal bracing, concrete slab 

HCLPF (g) Capacity 

0.3 

Service Building, El 39'-0" Structural steel framing and 0.38 

floor diagonal bracing, concrete slab 

Main Steam Valve House, 

interior steel structure 

Structural steel framing and >0.3 

diagonal bracing, metal grating 

Wall ID No. 

c 0.5-1 

c 20-1 

SB 21-17 

SB 21-18 

SB 35-1 

SB 35-2 

SB 35-3 

SB 35-4 

SB 35-7 

Group A Components HCLPF (g) and Lifelines Capacity· 

Pressurizer instrumentation 

(PT-104, LT-106) and tubing 

Main control board (aux feedwater 

system panels) 

Main control board (aux feedwater system 

panels), aux logic panels 

Battery groups 3 and 4, safety-related cable trays 

Battery groups 3 and 4, safety-related cable trays 

Battery groups 3 and 4, safety-related cable trays 

Battery groups 3 and 4 >0.3 

PCC Surge line, PCC temperature controller 

3-22 

>0.3 

>0.3 

>0.3 

>0.3 

>0.3 

>0.3 

>0.3 
TABLE 3-7 Maine Yankee structures and block walls (cont.) 

Group A Components HCLPF (g) Wall IO No. and Lifelines Capacity 

SB 45-1 125V DC distribution cabinets 1 to 4, 0.3 

battery group 2, inverters #1 and #2, 

Bus 8 

SB 45-2 Battery group 1, MCC SA, 480 V emergency >0~3 

switchgear 

SB 45-3 Battery group 1 >0.3 

VE 21-1, 2 Containment spray pumphouse Fans 44A >0.3 

and 448, filter 

VE 21-3, 4 sec line to penetration coolers >0.3 

3-23 
Table 3-8 Maine Yankee equipment list for margins review. 

Building 

and HCLPF (g) Equipment Item System Elevation Capacity 

TANKS 

Refueling Cavity Water HPSI Yd. + 20' 0.21 

Storage Tank TK-4· 

Primary Component PCC SB + 61 I >0.3 

Cooling Surge Tank TK-5 

Primary Water Storage AFW Yd. + 20' 0.27 

Tank TK-16 

Demineralized Water AFW Yd. + 20' 0.17 

Storage Tank TK-21 

Spray Chemical Addition HPSI Yd. + 20' >0.3 

TK-54 

Secondary Component sec SB + 70' >0.3 

Cooling Surge Tank TK-59 

Emergency Diesel Day OF AB + 21' 0.43 

Tank TK-62A, TK-628 

Diesel Compressed Air DG AB + 21' >0.3 

Tanks TK-76A1, TK-76A2, 

TK-76A3, TK-7681, 

TK-7682, TK-7683 

Diesel Starting Air DG A8 + 21' >0.3 

Receivers TK-76A4, 

TK-76A5, TK~76A6, 

TK-7684, TK-7685, 

TK-7686 

PUMPS 

Service Water Pump sw cw + 7' >0.3 

P-29A, P~29B, P29C, P-29D 

Containment Spray Pumps cs cs + 14' >0.3 

P-61A, P-618 

3-24 
Table 3-8 Maine Yankee equipment list for margins review. (Cont.) 

Building 

and HCLPF (g) Equipment Item System Elevation Capacity 

HEAT EXCHANGERS 

Residual Heat Removal PCC cs + 14' >0.3 

Heat Exchanger E-3A* 

Residual Heat Removal sec cs + 14' >0.3 

Heat Exchanger E-3B* 

Primary Component PCC TB + 21' >0.3 

Cooling Heat Exchangers E-4A, E-4B 

Secondary Component sec TB + 21' >0.3 

Cooling Heat Exchangers E-5A, E-5B 

Fuel Pool Heat Exchanger* PCC TB + 31' >0.3 

E-25 

CEDM Coolers* PCC RC + 46' >0.3 

E-53-1, E-53-2, E-53-3, E-53-4 

Reactor Containment* PCC RC + 46' >0.3 

Air Recirculation Coolers 

E-54-1, E-54-2, E-54-3, E-54-4, 

E-54-5, E-54-6 

Charging Pump Seal* sec PAB + 11' >0.3 

Leakage Cooler E-92A 

Charging Pump Seal* PCC PAB + 11' >0.3 

Leakage cooler E-928 .· ... ·,' 

*Component whose failure may breach critical system pressure boundary. 

3-25 
Table 3-8 Maine Yankee equipment list for margins review. (Cont.) 

Building 

and HCLPF (g) Equipment Item System Elevation Capacity 

ELECTRICAL DISTRIBUTION SYSTEMS 

4160-V Emergency Buses Elec SB + 46' >0.3 

Bus 5, 6, 7, 8 

480-V Emergency Motor· Elec SB + 46' >0.3 

Control Center MCC-7 A, SA 

480-V Emergency Motor Elec RMC + 21' >0.3 

Control Center MCC-7B, 8B 

480-V Emergency Motor Elec cs + 20' >0.3 

Control Center MCC-7B1, 8B1 

Station Battery No. 1, 2 Elec SB + 46' >0.3 

New Lead Calcium Batteries 

Station Battery No. 3, 4 Elec SB + 35' >0.3 

New Lead Calcium Batteries 

Battery Chargers BC-1, BC-2, Elec SB + 46' >0.3 

BC-3, BC-4 

Inverters INVR-1, INVR-2, Elec SB + 46' 0.82 

I NVR-3, I NVR-4 

Station Service Transformer Elec SB + 46 1 0.30 

X-507, X-608 (located adjacent 

to Bus 7 & 8) 

Diesel Generator Control Elec AB + 22' >0.3 '' Panel lA, 1B ~ ' . .. ' 

Main Control Board Elec SB + 21 1 >0.3 

120-V AC Vital Bus 1-4 

Electrical Control Board Elec SB + 21' >0.3 

DG-lA & 1B Start 1 & 2 

Circuits and Control 

Power 

3-26 
Table 3-8 Maine Yankee equipment list for margins review. (Cont.) 

Equipment Item System 

ELECTRICAL DISTRIBUTION SYSTEMS (Cont.) 

Auxiliary Logic Cabinets 

ESF Auxiliary Panels A & B 

Air Condition Control 

Panel ACCP 

Safety Parameter Display System Cabinets 

HVAC 

Elec 

El ec 

Elec 

Elec 

Computer Room Air Conditioner* SCC 

AC-lA 

Computer Room Air Conditioner* PCC 

AC-1B 

Lab Air Conditioner AC-2* SCC 

Containment Spray Fan* SCC 

FN-44A, FN-44B 

VALVES 

Power-Operated Relief 

Valve PR-S-14, PR-S-15 

Power-Operated Block 

Valve MOV PR-M-16, PR-M-17 

PORV 

PORV 

Building 

and 

Elevation 

SB + 21' 

SB + 21' 

SB + 21' 

SB + 21' 

SB + 39 I 

SB + 39' 

Unknown 

cs + 20' 

RC + 66' 

RC + 64 I 

HCLPF (g) Capacity 

>0.3 

>0.3 

>0.3 

>0.3 

0.38 

0.38 

0.37 

0.3 

>0.3 

>0.3 

*Component whose failure may breach critical system pressure boundary. 

\ 3-27 

CHAPTER 4 

4. INSIGHTS AND LESSONS LEARNED 

During the trial seismic margin review of the Maine Yankee plant, several 

lessons were learned about the various aspects of performing the review. In 

addition, many insights have been gained about the entire process of 

performing seismic margin reviews of nuclear power plants. 

Insights and lessons learned have been gained on: 

o Seismic margins approach and methodology including the quantification 

techniques. 

o The guidance for actually conducting seismic margins reviews 

including plant walkdowns. 

o The level of effort needed to perform such a review. 

o The identification and resolution of significant issues concerning the operation and licensing of the plant under review. 

The purpose of this section is to highlight these insights and lessons 

learned. 

4.1 Approach and Methodology (NUREG/CR-4334) 

This section presents a brief discussion of the insights and lessons learned 

that pertain to the seismic margins approach and methodology. We gained a 

better understanding of the areas listed below: · 

1. The seismic margins review methodology should be able to be applied by knowledgeable engineers in the areas of earthquake engineeringt 

component fragility analysis or systems analysis. They need to have 

some familiarity with these technologies. 

Guidance on performing seismic margin reviews needs to be prescrip- tive and descriptive enough to guide knowledgeable engineers so that 

an accurate determination of the plant's seismic margin can be made 

without allowing over-conservatism into the analysis. 

An independent review body should also be used to verify the 

application and accuracy of the review. 

2. The fragility screening table needs to be strengthened. The actual 

meaning {representation} of the cut-offs between the tab 1 e co 1 umns 

needs further explanation. In addition, more guidance in applying the table needs to be given. 

3. More guidance is needed on selecting, establishing, and defining the 

review earthquake level (Step 1 of the review). 

4-1 
There are a number of ways of specifying this review earthquake 

level: 

o Uniform hazard spectrum at a specified annual probability of 

exceedance and confidence. 

o Site specific response spectrum evaluated for the specific 

earthquake magnitude and epicentral distance ranges. 

o Standard spectrum such as the NUREG/CR-0098. 

o Use of peak acceleration or mean acceleration. 

o Peak ground acceleration without specifying any spectrum. 

Knowledge of the target spectrum allows the analysts to make a more 

de f i n it i ve s tat erne n t about the s e i s m i c capac it y of the p 1 ant • The 

review earthquake must be completely specified prior to any plant 

review, walkdowns or analysis. 

4. The seismic margin quantification techniques need to be better 

defined. There are two competing techniques for calculating the 

frag i 1 i ty of components, CDFM and FA. For this review, the FA 

technique was employed. However, the CDFM technique would be more 

easily applied if more prescriptive and better defined. More 

guidance is needed to use the CDFM technique. It is difficult to use 

as presently defined. 

More guidance is needed on the evaluation of plant level HCLPF 

capacity including the use of nonseismic failures and their 

significance to the results. More guidance is needed on methods for 

consideration of correlation between component failures. · 

5. More generic information is needed with respect to the safety 

function of reactor subcrit'ical ity. This function can be performed 

by e i the r the i n s e r t i on of the con t r o 1 rods , or by boron i n j e c t i on • 

Although the control rod drive mechanisms are screened out in Table 

5-l of NUREG/CR-4334 for review earthquake levels of 0.3g, the 

reactor internals are not screened out based on insufficient 

information. This meant that both the reactor internals and the 

boron injection system components were included in the information 

gathering process before and during the first walkdown. The boron 

injection system is fairly complex, with numerous components that are 

not initially screened out. Appreciable resources were expended in 

gathering information concerning the boron injection system. This 

was unnecessary after the seismic capacity review of the reactor 

internals. It would probably be more efficient in future seismic 

rna r g i n rev i ew s i f i n i t i a 1 effort i s p 1 aced i n v e r i f y i n g that the 

reactor internals have high capacity, and only look at alternate 

means of subcriticality if this is not the case. If necessary, this 

examination of alternate systems could be accomplished during the 

second walkdown. 

4-2 
6. Guidance in NUREG/CR-4334 and 4482, based on evaluation of previous 

PRAs for PWR plants, states that the emergency core cooling (early) function is included in Group A, while the emergency core cooling (late) function is in Group Not-A, and therefore screened out of the 

analysis. This screening is conditional on not finding any extremely 

gross plant-specific differences. 

The systems analysis team therefore included the initial switchover 

phase from emergency core cooling injection (early) to emergency core 

cooling recirculation (late) as a screening verification step in the 

first plant walkdown. While the guidelines are ambiguous, this 

screening verification included long-term area cooling for the 

recirculation systems. It was determined that the containment spray 

pump area cooling fans FN-44A and B, and a block wall near the fans, 

VE-21-1, could not be screened out based on the first walkdown. 

Based on the plant Boolean equation for small LOCA, both of these 

items were single failures resulting in core damage in the long term. 

The utility will make changes to these items to increase their 

capacity so that they do not impact overall plant capacity. Based on 

these findings, guidance in NUREG/CR-4334 and 4482 should be revised 

to insure that potential failures such as these are explicitly evaluated during a seismic margins review. 

7. For the no-LOCA case, the emergency core cooling (early) function is 

defined in NUREG/CR-4334 and 4482 as achievement of residual heat 

removal. The AFW or EFW system at Maine Yankee or other PWR plants will achieve this balance within the first hour. For most PWR 

plants, irrecoverable failure of the emergency ac power system (station blackout) will not prevent the turbine-driven AFW train from 

performing early residual heat removal, and therefore satisfy the 

emergency core cooling (early) function. However, in the longer term 

without ac power, the station batteries would be depleted, resulting in loss of instrumentation and AFW control power. Core damage could 

occur if de power is not restored, and manual control of the turbinedriven AFW train or other feedwater source fails. Based on the 

guidelines, this long-term failure of AFW was screened out of the 

analysis. If it were assumed that battery depletion and loss of 

instrumentation and control power results in loss of AFW and other 

feedwater sources, then the Boolean expression for the no-LOCA core 

damage case would be dominated by the seismic failures that result in 

station blackout: 

o Failure of the sec and PCC heat exchangers E-5A and 4B. 

o Failure of the station service transformers X-507 and 608. 

o Failure of the DG day tanks TK-62A and B. 

o Rupture of SCC and PCC because of chiller heat exchanger failure for the air conditioners AC-lA, 1B, and 2. 

o Structural failure of the circulating water pumphouse failing the SWS. 

4-3 
Explicit guidance on the treatment of these long-term battery 

depletion sequences would be helpful. 

8. The Expert Panel reports have not explicitly discussed the seismic 

capacities of steel structures. At Maine Yankee, the steel 

structures have capacities in excess of 0.30g. At the same time, 

they could not be screened out based on a walkdown review because of 

unusual connection details and structural arrangement. More guidance is needed on the treatment of steel structures. 

9. The availability of in-structure response spectra generated by 

current techniques can reduce the amount of effort necessary to 

quantify component fragilities since modification or regeneration .of 

responses was not necessary. Accurate floor spectra may not always be available for older plants. In such cases, it may be necessary to 

develop new dynamic models and perform dynamic analysis to define the 

seismic input to equipment. This will increase the amount of time 

and funding necessary to perform seismic margin studies. 

10. Detailed information on the Maine Yankee block walls was available. 

This data was very useful in conducting the trial plant review since 

it provided locations of all block walls in the plant and identified 

any safety-related components that could be affected by their 

failure. In particular, this latter set of information was valuable 

since it permitted quick identification of several Group A block 

walls. 

11. From our review of Maine Yankee, we identified an additional plant unique feature. This unique feature is cast iron service water 

p1p1ng. This cast iron piping at Maine Yankee was adequately supported and judged to have an acceptable capa~ity. 

4.2 Guidelines (NUREG/CR-4482) 

This section provides a brief discussion of the lessons learned and insights gained concerning the guidelines for conducting seismic margins reviews. 

1. Insights were gained about the planning and conduct of the physical plant walkdowns from our review of Maine Yankee. 

The most essential part of a plant walkdown is preparation and 

planning. All walkdown participants should help organize and plan 

the specifics of the plant visit which are dependent on the nature of 

the walkdown and the specific plant under review. The objectives of 

the walkdown should be clearly understood and arrangements for 

security, radiological safety, training, and plant operation should 

be considered. 

Walkdown groups consisting of three to five people with background 

and expertise in various engineering disciplines should utilize pre- developed inspection criteria and forms to collect information and 

data pertaining to the objectives of the plant review. The groups 

4-4 
should be organized with respect to the information and data that 

needs to be collected. Each group should contain a utility person 

familiar with the plant layout, and the location and operation of the 

systems and components under review. 

A clear understanding of what is going to be reviewed and inspected 

is also essential to the walkdown. A list of the systems, components 

and p 1 ant areas that need to be visited during the actua 1 wa 1 kdown 

should be developed. An itinerary will also facilitate the actual 

plant walkdown. 

Our i ng the p 1 ant visit, meetings s hou 1 d be arranged each day to 

organize the groups and prepare for the walkdowns. These meetings 

should allow time for discussion of findings at the end of each day 

and the p~eparation for the next day's effort. 

Walkdowns can be performed in many ways, from a quick "walkby" of the 

plant to a thorough "crawl through, 11 depending on the needs of each 

group, and the information and data collection requirements. The 

types of walkdowns and their progressions should be planned in 

advance. Tools for the collection and recording of the information 

and data will also help facilitate the walkdown. 

2. The identification of the front-line system components was relatively 

straightforward because of the detailed nature of the available 

information, and the sma 11 number of components. However, 

identification of support system components was more difficult. This 

is because these systems are generally more complex, have many more 

components and branches, and are not generally documented as well. 

In addition, the references concerning interfaces between the front1 ine systems and support systems are often ambiguous. Finally, the 

actual physical nature and location of some items, such as 

distribution cabinets and panels, is not shown on plant drawings or 

documentation. Based on this experience, there are two recommendations. First, when reviewing the plant information, emphasize the 

interfaces with sup port sys terns such as ac and de power, coo 1 ing 

water systems, HVAC systems, and instrument air systems. Document 

the ambiguities for later clarification. Second, plan to spend 

considerable effort tracing down these support system components 

during the first walkdown, and be prepared to make substantial 

revisions to the component list. 

3. Another insight concerns documentation and information transfer 

between the Systems Analysis Team and the Fragilities Team. Many of 

the components identified by the Systems Analysis Team for HCLPF 

screening or evaluation were selected because of the potential for 

component rupture to cause flow diversion and consequent system 

failure. The component itself was not needed to fulfill a safety 

function, but the integrity of the component pressure boundary had to 

be assured for overall system success. The common example was heat 

exchangers for nonessential equipment whose seismic rupture would 

fail a necessary cooling water system. Since it can make a 

4-5 
difference to the HCLPF assessment, the systems team must make a 

clear differentiation between components that are required to 

function for system success, and components that are required only to 

maintain pressure boundary integrity. 

4. There are a number of insights concerning the systems analysis and 

the fault tree pruning process which may be helpful to future seismic 

margin reviews. 

a. The procedure proposed to isolate the PCC 1 ines and components 

inside containment, and the automatic rupture isolation system on 

the sec greatly reduced the amount of components that had to be 

considered for HCLPF evaluation. Both the systems analysis and 

fragilities analysis efforts would have been larger, and 

eventually a containment walkdown might have been necessary. 

b. Early evaluation and screening out of potential recovery actions 

and alternate systems, such as the small positive displacement 

pump for core cooling injection, reduced the number of components 

that required systems and fragility evaluations. 

c. Being able to define all the components on one skid as one 

supercomponent, such as the DGs, reduces the systems analysis 

effort, but the evaluation for the fragility team may not be 

reduced. 

d. As the fault trees are developed, it is useful to keep a list of 

the failure modes which should be considered for each 

component. For example, the pump failure modes include fail to 

start, fail to run, and test or maintenance outage, as well as 

seismic failure, but the pump appears only once on the initial 

fault tree. This in format ion is needed 1 ater for the 

quantification process. 

e. In the initial trees, it is necessary to include all the 

components that require support systems, including those 

components such as motor-operated or air-operated va 1 ves that 

will likely be screened out later because of their high HCLPF and 

low nonseismic unavailability. Otherwise, if they are pruned 

from the initial trees, their dependency on support systems may 

be overlooked in the rest of the analysis. Also, since physical 

interactions between the component and structures, such as block 

walls or restraints, must be checked, it is better to include the 

component in the initial fault trees. 

f. In this project, those components and structures that were 

assigned a HCLPF of 11 >0.3g .. either because they fit the generic 

screening guidelines, ·or because a conservative fragility 

calculation was performed, were pruned from the fault trees 

before the Boolean equations were developed. The analysis team 

felt that their actual HCLPF would be well in excess of 0.3g, and 

that they wou 1 d not affect the p 1 ant s ei smi c capacity 

4-6 
calculation. This pruning process resulted in very manageable 

fault trees, relatively small Boolean equations, and satisfactory 

calculations for the plant HCLPF. It also enabled easier 

communication, understanding, and insight into the results than 

if less pruning had been performed, but provided more information 

on seismic margin than a more severely pruned Boolean equation. 

The balance appears to be satisfactory. 

g. Although a few seismic interactions, such as the possible impacts 

of potential seismically induced fires, were not evaluated, the 

potential for threaded firewater piping ruptures to damage 

equipment was reviewed. The DG control panels and distribution 

panels were located under firewater piping that could have 

considerable lateral movement. Upon investigation, however, it 

was determined that the piping was dry, and two signals would be 

required to fill the piping. The probability of inadvertent 

actuation was therefore negligible. The PCC and SCC pumps were 

also located under sprinkler nozzles, but the motor housings for 

these pumps were designed to prevent water from entering. 

Therefore, ruptures of firewater piping was not considered to 

impact seismic capacity. 

5. Minimal cut sets were developed at four stages in this project. 

Front-line system level cut sets using partially pruned fault trees, 

including their support systems, were developed just before the 

second wal kdown to provide some guidance to the fragility team. 

These pointed out some potentially important system minimal cut 

sets. Although plant or sequence level cut sets could have been of 

additional assistance, because the fault trees were still fairly 

large, the number of minima·! cut sets would have been large as 

well. The additional effort to develop plant level cut sets before 

the second walkdown is not judged to be an effective allocation of 

resources, but the effort to develop system cut sets is effective. 

6. The seismic systems models of the plant (event trees and fault trees) 

should be developed using best-estimate success criteria based on the 

FSAR and any relevant experience data. The FSAR analyses, in a 

sense, represents "high confidence .. because of the regulations upon 

which they are based. A best-estimate realistic success criteria 

beyond the FSAR should be used if there are data and/or calculations 

to support their use. 

7. Due to radioactivity concerns, critical components inside the Maine 

Yankee containment are accessible for review only at the time of a 

plant outage. Fortunately, Maine Yankee has maintained a relatively 

complete data file on components within the containment. 

Difficulties in reviewing plant components will be more severe for 

plants which do not maintain organized documentation on the 

components within containment. 

More guidance is needed on the consideration of the seismic small 

LOCA-initiating event. It may be impossible to review and inspect 

4-7 
the reactor coolant p1p1ng (primary pressure boundary) throughout the 

plant, in particular within the containmentt in order to screen out 

this possible accident initiator. For the Maine Yankee review, the 

small LOCA-accident sequence was explicitly considered in the systems analysis and was the controlling event sequence with regard to plant 

capacity. 

8. During the walkdowns, plant and utility personnel having expertise or 

specialized knowledge in particular fields were interviewed. 

Specific areas of expertise included: 

o Structural/mechanical. 

o Firewater systems. 

o Electrical. 

o HVAC. 

o Instrumentation and control. 

o Control room personnel. 

o Block walls. 

The insight gained from discussions with knowledgeable individuals 

proved to be very useful in assessing the overall state of the plant and resolving any particular seismic issues. 

9. Walkdown data sheets, while not noted in the review guidelines, were 

developed prior to the first walkdown. These data sheets served as a 

checklist of items to review during the walkdown and facilitated the 

gathering of information necessary for HCLPF evaluation. While no 

set format need be established for future margin reviews, the review 

guidelines should indicate that the use of data sheets is encouraged. 

10. Major distribution systems such as cable trays, piping, and ducting extend throughout the plant, there local configurations and support details can vary. Detailed walkdowns of these systems can be timeconsuming. More guidance in the seismic margins methodology on the 

level of walkdown for these systems would be useful. 

11. In the course of this study, five example components (i.e., refueling 

water storage tank, steel structure, diesel day tank, inverter and 

block wall) were analyzed using the two candidate methods (i.e., CDFM 

and FA). Our experience was that several judgmental decisions had to 

be made in arriving at the parameters of the CDFM method. In each 

case we were not sure whether we met the intent of the method, i.e., 

conservative estimation of the capacity, yet more liberal than ·the 

SRP requirements; in some cases, we may have been overly conservative 

as was pointed out by the Peer Review Group. The difficulties arise 

because of two factors: 

o The CDFM method has not been fully defined for all 

structures and equipment items. 

o The parameters of the CDFM method such as damping, material 

strength , s tat i c c a pa c it y e quat ions , s y s t em d u c t i 1 i t y , and 

4-8 
methods for floor spectra generation are not explicitly 

specified; even where they are specified they may be overly 

conservative. Also, the appropriate conservatism in the 

select ion of the CDFM parameters needs to be determined 

using calibration methods. 

We recommend that such a comparison study be performed. 

12. This plant review examined a PWR on a rock site with a review 

earthquake level of 0.30g. The methodology, the review guidelines, 

and the staffing requirements have not been verified for other 

conditions including BWRs, soil conditions, and a higher review 

earthquake. 

4.3 Effort to Perform the Review 

An objective of the trial seismic margin review is to gain an understanding of 

the level of effort, and thus cost, of performing seismic margins reviews of 

nuclear power plants. The trial review involved effort by these personnel: 

the Systems Analyst Team, the Fragility Analyst Team, the utility or plant 

operator, and the management of the overall review. 

NUREG/CR-4482 gives estimated staffing requirements for a seismic margin 

review; for a plant founded on rock with a review earthquake level of 0.30g 

pga, the estimate to perform the review is 2.5 staff-years. The actual effort 

expended in the present study is about 25% more than the Panel• s estimate 

since this was the first review conducted. 

Both analyst teams have divided their efforts into specific tasks. The tasks 

and the level of effort expended by both teams is given in Table 4-1. The 

Systems Analyst Team spent 1532 man-hours performing the review, and the 

Fragility Analyst Team spent 4176 man-hours. The combined effort was 

approximately 35 man-months. 

A breakdown of the uti 1 ity effort is given in Appendix C. They spent 1060 

man-hours for providing data, answering question, organizing walkdowns, and 

reviewing results. 

4.4 Findings and Their Resolution 

A number of issues became apparent during the Maine Yankee review. Primarily, 

these were findings concerning the seismic capacity of various components at 

the plant. While many of these issues were discussed in the result section 

(Section 3.0), this section is intended to briefly discuss all the seismic 

margins related findings and their resolution concerning the Maine Yankee 

review. A listing of each issue and brief discussion of the resolution are 

given below: 

1. Several types of analysis techniques can be employed to calculate the 

seismic capacity of free-standing tanks. For the Maine Yankee RWST 

these techniques resulted in the HCLPF capacity being greater than 

0.20g. The lowest of these values is shown in Table 3-2. 

4-9 
The Maine Yankee DWST, which is not a singleton to the seismic safety of the plant, was found to have a HCLPF capacity of 0.17g. 

4-10 
1.0 

2.0 

3.0 

4.0 

'· 

Table 4-1 Cost breakdown items for seismic margins 

trial plant review. 

Actual 

Fragilities Team 

Review Items Staff-Hr 

Collect Information on Design 

1.1 First Round Information 120 

1.2 Additional Specific Information 80 

1.3 Visit to Utility/AE/NSSS Vendor 96 

Review of Plant Information 

2.1 Review of Review Earthquake Level 40 

2.2 Initial Systems Review 

2.3 Identify Components for 

Group A Functions 

2.4 Perform Initial Screening of Components 176 

2.5 Review of Design-Analysis and Seismic 

Reevaluation Reports 216 

Plant Walkdowns 

3.1 Identify Target Areas for First 

Walkdown 240 

3.2 Perform First Walkdown 184 

3.3 Conduct Simplified Analysis 200 

3.4 First Walkdown Documentation 320 

3.5 Perform Second Walkdown 200 

Systems Modeling 

4.1 Develop Event and Fault Trees 

4.2 Derive Accident Sequences 4.3 Develop Boolean Expressions and 

Minimal Cut Sets 

4-11 

Actua 1 

Systems Team 

Staff-Hr 

20 

34 

8 

60 

36 

50 

50 

84 

36 

34 

26 

539 

70 

50 
5.0 

6.0 

7.0 

Table 4-1 Cost breakdown items for seismic margins 

trial plant review (Cont.). 

Actual 

Fragilities Team 

Review It ems Staff-Hr 

Seismic Margin Evaluation 

5.1 HCLPF Capacity of Components 

5. L 1 CDFM Method 280 

5.1.2 Fragility Analysis Method 800 

5.2 HCLPF Capacity of Plant 160 

Reporting 

6.1 Internal Review 180 

6.2 Letter and Final Reports 480 

Meetings 

7.1 Project Teams 224 

7.2 Peer Review Group 80 7~3 NRC/ACRS/Expert Panel 100 

4176 

4-12 

Actual 

Systems Team 

Staff-Hr 

10 

22 

253 

104 

22 

24 

1532 
2. There is no data or experience on the seismic behavior of aged leadantimony station batteries at the Maine Yankee plant. Therefore, 

there was no way to estimate their seismic capacity and a HCLPF 

value. 

Maine Yankee will have these station batteries replaced with leadca 1 c i urn batteries during the next two refue 1 i ng outages. Stat ion 

Batteries 1 and 3 will be replaced during the next refueling outage 

(March 1987) and Station Batteries 2 and 4 will be replaced during 

the 1988 outage. The HCLPF capacity of the station batteries was 

evaluated for the replacement units. 

3. Older General Electric station service transformers (4160 V to 480 V) 

were not tested or qualified for seismic loads and their seismic 

capacity is low. This component became the most dominant contributor 

to the seismic capacity of the plant prior to the upgrade. 

GE had recent seismic testing and qualification performed on its 

equipment transformer. This testing resulted in modifications. Base 

anchorage modifications are scheduled for the Maine ·Yankee 

transformers. 

4. Anchorages on the containment spray pump area fans FN-44A and B will 

be strengthened. Failure of these fans could have led to long term 

heat-up and failure of the containment spray pumps, with subsequent 

failure of high pressure safety recirculation if recovery actions 

were not effective. 

5. Block wall VE 21-1 will be strengthened to prevent its potential collapse from failing the containment spray area fans FN-44A and B 

discussed above. 

6. The anchorages of the chillers for the computer room air conditioners 

AC-1A and B and the laboratory air conditioner AC-2 will be 

strengthened. Failure of the heat exchangers on these chillers could 

have failed the pressure boundary integrity of the SCC and PCC, and 

resulted in core damage upon loss of component cooling water. 

7. A procedure is being developed to isolate nonessential PCC lines and 

heat exchangers following a large earthquake and receipt of a low 

level indication in the PCC surge tank. Although the PCC system was 

designed to seismic standards, the project team could not verify the 

capacity of all components. Isolating the PCC lines provides 

assurance that any potential small leakage of the pressure boundary will not fail the entire PCC system. 

8. An unanchored monitor in the main control room panel was anchored. 

Its impact on other components and the seismic capacity of the plant therefore did not have to be evaluated. · 

9. The emergency lights in the control room and throughout the plant 

were strapped and anchored. 

4-13 
10. A missing bolt on the anchorage of some level transmitters for the 

RWST was replaced. 

11. Loose pressurized gas cylinders, a welding machine, and some heavy 

parts near the containment spray pump area fans were moved or tied 

securely. 

12. Additional anchorage was added to both diesel generator day tanks. 

4-14 
... ' t ·~ 

CHAPTER 5 

5. CONCLUSIONS 

The trial seismic margins review of the Maine Yankee plant was conducted with 

the concerted effort of all parties involved. The analysis teams worked 

together closely and followed the guidance on performing the review to 

estimate the overall plant HCLPF and the HCLPF capacities for the accident 

sequences that lead to seismic-induced core damage. The Maine Yankee utility and Yankee Atomic Electric Company provided invaluable assistance in 

performing the review. Without their efforts, this review would have been 

much more difficult. The Expert Panel provided an initial review of the 

approach early in the project. The Peer Review Group provided guidance and a 

critical examination of the process and interim results at each stage of the 

review. The NRC assisted in the definition of the scope of the review and 

licensing issues involved. 

The conclusions for this trial review are divided into two sections: 

conclusions concerning the application of the seismic margins methodology, and 

guidelines including the numerical results and insights gained from this study 

are given in Section 5.1. The conclusions regarding Maine Yankee's licensing issue are given in Section 5.2. 

This seismic margin review has been performed with the following assumptions and 1 i mi tat ions: 

o The review earthquake level was specified by the NRC as the 

NUREG/CR-0098 median spectrum anchored to 0.3g. 

o The structural models and the in-structure response spectra generated by Maine Yankee have been judged to be adequate for the purposes of 

this margin review. 

o Since the Analysis Team could not perform the walkdown inside the 

containment, the seismic capacity of components inside the 

containment could not be determined. We could not confirm the 

absence of potential system interaction effects that may make the 

impulse lines inside the containment vulnerable to earthquakes and 

lead to a small LOCA. 

0 In keeping with the Expert Panel's philosophy, the screening of 

components was performed using conservative procedures. For the 

screened-in components, the seismic capacities have been calculated 

using conservative methods. In all cases, the factors contributing 

to the seismic margin and their variabilities are identified and 

quantified using procedures normally used within the 

state-of-the-art. 

o The HCLPF capacity of the plant has been determined based on the 

seismic capacities of components in their existing or proposed modified conditions. Maine Yankee has proposed that certain 

modifications or replacements would be made for station batteries, 

5-1 
transformer internal core/coil assembly anchorage, vibrationisolation supports for containment spray fans and air conditioners, 

anchorage of diesel day tank, and block wall near the containment 

spray fans. 

o The results of this seismic margins review represents the best 

estimate analysis of the components and the plant following the 

proposed modification as a result of this review. No effort was made 

to account for the effects of future aging. 

5.1 Conclusions Regarding the Methodology, Guidelines, and Insights 

The conclusions from the trial seismic margin review include: 

o The plant HCLPF capacity was determined to be 0.21g. This capacity is dominated by the small LOCA-accident sequence with the RWST being 

the dominant component. There was no effect on the plant HCLPF 

capacity when we considered the dependence between component fa i 1 ures. 

An arbitrary 10% reduction in RWST fluid level results in an increase 

in HCLPF capacity to 0.26g. 

Assuming an arbritrary ten percent (10%) probability of occurrence of 

the small LOCA-initiating event as compared to the occurrence of all 

the other possible initiating events in the plant, HCLPF capacity increases to 0.28g. 

o We found that careful plant walkdowns are essential to successful 

seismic margin reviews. 

o Insights and lessons learned (discussed in Chapter 4) concerning: 

The selection of the review earthquake level. 

The methodology and use of the review guidelines. 

Component qualification data. 

Plant walkdown procedures. 

Guidance on the CDFM HCLPF calculation procedure. 

o The maintenance of hot shutdown following a seismic-induced initiator 

was considered by performing a thorough walkdown review and analysis of the components needed to perform this function. This led to 

upgrading of Fans FN-44A, Band the adjacent block wall. 

o Important components and failure modes discovered during this review 

are: 

5-2 
. ) ) . . 

Station service transformers (4160 V to 480 V) require a review 

of their internal configuration. 

Lead-antimony station batteries may fail due to the failure of 

the plates within the battery casing. No data are available to 

estimate a HCLPF capacity for lead-antimony batteries. 

Consideration must be given to the location and possible systems interaction from threaded fire water piping. 

o Consideration must be given to modifications and upgrades identified 

during the seismic margins review process. 

o The small .LOCA-initiating event had to be considered for this review 

because of the difficulty in performing a walkdown inside the 

containment building. 

o The walkdown review of closed-loop component-cooling systems may require considerable effort. 

o The components that affect the reactor subcriticality function, in 

particular the control element drive mechanisms and the reactor 

internals, need to be considered early in the review. 

5.2 Conclusion Regarding the Maine Yankee Licensing Issue 

The Maine Yankee plant was found to be clean and well-maintained. There has 

been a concerted effort to seismically upgrade the plant consistent with newer 

plants. The Maine Yankee utility and its contractor, Yankee Atomic Electric 

Company, were very cooperative with this effort and responded on their own 

initiative to increase the seismic margin of the plant. 

A number of components were identified as having a low seismic capacity. A 

letter indicating that Maine Yankee will upgrade these components by the end 

of the next two refueling outages in March 1987 and 1988 is included in 

Appendix C. These components are listed below: 

o Important station service transformers (4160 V to 480 V) 

0 A block wall near the HVAC equipment (Fan 44A & B) needed to cool the 

containment spray pump enclosure. This enclosure houses the longterm cooling equipment. 

o The PCC and sec heat exchangers for both the computer room air 

conditioning compressors (chillers). 

o Station batteries 1 and. 3 will be replaced during the March 1987 

outage. Batteries 2 and 4 will be replaced in the 1988 outage. 

o Upgrading the anchorage of both diesel generator day tanks. 

5-3 
The plant HCLPF capacity after the planned upgrades was estimated to be 

0.21g. This HCLPF capacity is governed by the RWST and represents a 

conservative estimate of the seismic capacity of the plant. That is, given an 

earthquake producing this ground acceleration and specified spectral shape, 

tbere is high confidence (95%) that there is a low probability of core damage 

(occurring only approximately 5% of the time). 

It is recommended that an inspection and review of Maine Yankee be performed 

after the March 1987 outage to assure that the proposed ·modifications have 

been carried out and that their HCLPF capacity calculated in the present study 

are still applicable. 

5-4 
. . · . " 

CHAPTER 6 

6. RECOMMENDATIONS TO IMPROVE METHODOLOGY 

During the development of the seismic margins review methodology and 

guidelines, and as a result of performing this first trial seismic margins review, a number of recommendations can be made for further analysis and 

research that will improve the applicability and usability of this method for 

future reviews. These recommendations are listed below: 

o Need to revise and clarify the seismic margins methodology and 

guidelines based on lessons learned so they are more usable. 

o Need to make· a comparison of this methodology and its application with a similar effort being performed by EPRI. 

o Need a comparison study that addresses the CDFM and FA methods for 

calculating HCLPF capacities. 

o Need a study of the available methodologies for calculating the 

capacity of older tanks. 

o Need a trial seismic margins review of a BWR plant to test 

enhancement of the methodology presently underway that will address 

this type of plant. 

o Need testing of aged batteries to understand their seismic 

capacity. Additionally, to understand the internal plate failure 

mode for batteries in general • 

6-1 

REFERENCES 

American Society of Civil Engineers (ASCE), 11 The Effects of Earthquakes on 

Power and Industri·al Facilities and Implications for Nuclear Power Plant 

Design," Draft UCRL-93320 (1986). To be published by the ASCE. 

Memorandum from Newton Anderson, NRR Co-Chairman Seismic Margins Working 

Group, to Mr. James Richardson, RES Co-Chairman Seismic Margins Working Group, 

"Comments on the Seismic Margins Program and the Trial Plant Review," July 7, 

1986. 

Budnitz, R. J., P. J. Amico, C. A. Cornell, W. J. Hall, R. P. Kennedy, J. W. 

Reed, and M. Shinozuka, "An Approach to the Quantification of Seismic Margins 

in Nuclear Power Plants," NUREG/CR-4334, UCID-20444 (August 1985). · 

Campbell, R. D., M. K. Ravindra, A. Bhatia, and R. C. Murray, ••compilation of 

Fragility Information from Available Probabilistic· Assessments," Lawrence 

Livermore National Laboratory, Livermore, CA, UCID-20571 (September 1985). 

Carlson, D. D., "Interim Reliability Evaluation Program Procedure Guide," 

NUREG/CR-2728, SAND82-1100 (January 1983). 

Memorandum from Dennis M. Crutchfield, Assistant Director for Technical 

Support, Division of PWR Licensing-B, to Patrick M. Sears, Project Manager, 

Project Directorate No. 8, Division of PWR L icensing-B, 11 Recorrmended Spectrum 

for Maine Yankee Seismtc Design Program,"-May 7, 1986. 

Cummings, G. E., J. J. Johnson, and R. J. Budnitz, "NRC Seismic Design Margins 

Program Plan," UCID-20247 (October 1984). 

Memorandum from: D. Guzy, MSEB, DET, to NRC Seismic Design Margins Working 

Group, .. Maine Yankee Seismic Margins Review, .. May 12, 1986. 

Kennedy, R. P. and M. K. Ravindra, "Seismic Fragilities for Nuclear Power 

Plant Risk Studies, .. Nuclear Engineering and Design, Vol. 79, No. 1, pp. 47-68 

(May 1984). 

Maine Yankee Atomic Power Company, Final Safety Analysis Reeort (FSAR), 11 Section 2.5, Seismology ... 

Letter from Frank J. Miraglia, Director, PWR Project Directorate No. 8, 

Division of PWR Licensing-B, to Mr. J. B. Randazza, Executive Vice President, 

Maine Yankee Atomic Power Company, "Maine Yankee Participation in Seismic 

Design Margins Program, .. March 31, 1986. · 

Newmark, N. M. and W. J. Hall, "Development of Criteria for Review of Selected 

Nuclear Power Plants, .. NUREG/CR-0098 (May 1978). 

Prassinos, P. G., M. K. Ravindra, and J. B. Savy, "Recommendations to the 

Nuclear Regulatory Commission on Trial Guidelines for Seismic Margin Reviews 

of Nuclear Power Plants," NUREG/CR-4482, UCID-20579 (March 1986). 

R-1 
Oral Communication, Richardson, J. E., Office of Nuclear Regulatory Research, 

U.S. Nuclear Regulatory Commission, on the clarification of the seismic margin 

review earthquake level for the Maine Yankee trial review (December 11, 1986). 

U.S. Nuclear Regulatory Commission, "Design Response Spectra for Seismic 

Design of Nuclear Power Plants," Regulatory Guide 1.60 (December 1973). 

U.S. Nuclear Regulatory Commission, 11 Fault Tree Handbook, .. NUREG-0492 (January 

1981). 

U.S. Nuclear Regulatory Commission, "PRA Procedures Guide," NUREG-2300 

(January 1983). 

R-2 
APPENDIX A 

EXPERT PANEL CORRESPONDENCE 

',, •' 

A-1 
INTE~FACE BETWEEN THE EXPERT PANEL ON SEISMIC MARGINS 

AND THE TRIAL MARGIN REVIEW OF MAINE YANKEE 

R. J. Budnitz, Chairman 

The NRC 1s Expert Panel on Seismic Margins developed the original approach to 

performing seismic margin reviews (NUREG/CR-4334) and oversaw the development 

of interim guidance on how such a review should be carried out (NUREG/CR4482). It is important that the Expert Panel have confidence that the trial 

margin review is accomplished in a technically sound manner; this confidence 

will be assured by the Peer Review Group that is overseeing the trial margin 

study at Maine Yankee. 

The Expert Panel's role during the trial review will be very limited. 

Specifically, it is expected that the Expert Panel will review the approach 

being taken at an early stage of the Maine Yankee trial review, to assure 

itself that the methods and techniques being employed are consistent with the 

Panel 1s guidance and are relevant for performing a seismic margin review. 

This will involve a review of procedures prepared for the Maine Yankee review, 

discussion of these procedures on a conference call set up by the Chairman of 

the Expert Panel, and a follow-up letter stating the Panel members' opinions 

concerning the Maine Yankee review. 

After the early interaction, the Expert Panel will not be involved again until 

the trial review has been completed. At that time, the Panel will be convened 

so that it can study the results, examine how the trial review was 

implemented, interview the study team, and then evaluate the overall effort. 

The Expert Panel will be expected to re-examine its interim guidance and to 

revise it, if necessary, prior to issuance in final form. The Expert Panel 

will also be expected to evaluate the overall usefulness of the seismic 

margins review approach, any 1 i mi tat ions whe·:her previously recognized or not, 

and provide any other relevant comments to LLNL and the NRC. 

A-2 
Future ReJource.J AJJociate.J, Inc. 

2000 Center Street Suite 418 Bet·keley. CA 94704 415-528-5111 

11 July 1986 

TO: 

FROM: 

REF: 

Expert Panel on Seismic Design Margins (P. Amico, C.A. Cornell, 

J.Reed, M. Shinozuka) 

R.C. Murray, Lawrence Livermore National Laboratory 

R.J. Budnitz, Chairman of the Panel 

Minutes of the Telephone Conference Call Meeting of June 27, 1986 

This memorandum consists of the minutes of meeting of June 27, 1986 of the 

NRc•s .. Expert Panel on Seismic Design Margins", whose membership consists 

of the individuals listed above. The •meeting' was actually a conference 

telephone call arranged by Bob Murray of LLNL, who is the NRC/LLNL liaison 

for the Panel. In •attendance• were all five members of the Panel; Murray 

and P.G. Prassinos of LLNL;. David L. Moore of Energy. Incorporated; and 

M.K.Ravindra, P. Hashimoto, G. Hardy, and M. Griffin of EQE, Incorporated. 

The conference call began at 10:00 AM and ended at 12:40 PM, Pacific time. 

The objective of the meeting was to provide a forum for the Expert Panel to 

discuss progress to date on the trial seismic margin review that is now underway 

under NRC/LLNL sponsorship using the Maine Yankee plant as the trial plant. 

The review has been underway for 1!2 months, and this was an ideal time to 

obtain Panel comments and input as to the progress being made. As it turned 

out, there were several technical issues that had arisen in the course of 

the start-up phase of the trial review which benefitted from input from the 

Panel. 

The •bottom line• outcome of the meeting was that the Panel was satisfied with 

the way the trial review was being undertaken so far, with the exception of 

a few (important) comments that will be covered in the remainder of these 

minutes. It is NRC's intention that the Expert Panel be involved next at the 

stage when the trial review.has been completed, at which time the Panel will 

study the results, interact with the team of analysts, and then reevaluate 

the guidance that has been.provided in interim form in the Panel's NUREG publications on this subject, NUREG/CR-4334 and NUREG/CR-4482. (During the course 

of the conference call, the Panel estimated that about two person-weeks of 

effort would probably be required of each Panelist at this later time to carry 

out the review effectively, and this estimate does not count effort that may 

be needed to revise the interim guidelines.) 

1) The meeting began with Bob Murray's discussion of the progress to date of 

the analysis team, whose key participants were all on the telephone line: 

Pete Prassinos and Dave Moore representing the systems analysis team, and 

Bob Murray, Ravi Ravindra, and Ravindra•s EQE colleagues representing the 

fragilities team. Murray discussed the sechedule for the review, whose major 

A-3 
page 2 ---- Minutes of Expert Panel Meeting of June 27, 1986 

upcoming milestone will be the first plant walkdown on July 21-26, at the 

Maine Yankee plant. He discussed the interactions with the plant team, which 

interactions have been entirely favorable and cooperative so far; and the 

arrangements made for a •peer review group' to follow the progress of the 

trial review. Murray's introductory discussion set the stage for the rest 

of the conference call discussion. 

2) The Panel discussed the fact that the trial margins review at Maine Yankee 

is intimately tied up with an NRC licensing action. The Panel expressed regrets 

that this linkage was necessary, since the Panel believes that this trial 

review should be viewed in an important way as a research project whose outcome 

is by no means well understood until it has been accomplished. The Panel 

expressed a strong desire that NRC find a plant (expected to be a BWR) for the 

planned second trial review which does not have a licensing action tied to 

the seismic margin review. 

3) The Panel discussed the peer review arrangement for the trial review, and 

agreed that the set-up was consistent with the guidance provided in its earlier 

NUREG reports. The peer review group's charter is to assure the technical competence of the review, and to report on its findings to NRC and Maine Yankee. 

4) The Panel discussed at length the selection of the 'review level earthquake' 

(henceforth abbreviated RLEQ) by NRC, which s~lection is at the 0.30 g level, 

with a spectrum using the 50th percentile amplification factors from NUREG/CR-0095. 

A question arose as to how to cope with motions at higher frequencies than 

those in the more-or-less standard spectrum chosen. The Expert Panel reaffirmed 

its earlier position that a ~eoarate research project is needed to provide information to the Panel before the Panel can develop margin-review guidance on 

this issue, and that hioher-frequency motions are explicitly not included in 

the guidance already developed in the earlier NUREG reports. ---

5) Regarding the choice of the RLEQ, the Panel agreed that especially for a 

trial review like the one being undertaken at Maine Yankee, it would be preferable 

if the RLEQ were selected to be at a high enough level that the 'plant HCLPF 

value' can be affirmatively determined through the review, rather than having 

the review determine only that the plant HCLPF value is "at least as high as 

the RLEQ level ... The choice of 0.30 g for Maine Yankee's RLEQ does not obviously 

meet that criterion, although it may (after the fact) turn out that way. The 

Panel agreed to reconsider at a later date whether it is necessary to provide 

more detailed guidance on 'how to select the RLEQ'. 

6) The Expert Panel, after much discussion, agreed that it would be unfortunate 

if the fact that a given plant's HCLPF turned out to be less than the RLEQ were 

considered a •negative• outcome for the review. Considering the conservatisms 

embedded in the HCLPF idea, a plant•s median capacity would be expected to be 

considerably greater than its plant-level HCLPF value. 

7) The Panel discussed the meaning of its categorization in NUREG/CR-4334 and 

NUREG/CR·4482 of fragilities into three groupings, 11 below 0.30 g11 , 11 0.30 to 

0.50g", and "above 0.50g 11 • In its earlier reports, the Panel pointed out that 

these boundaries were not to be taken as being too precise: specifically, the 

Panel stated that the 0.30g boundary might just as easily have been a rough 

range from about 0.2:g to about 0.35g. While this Panel judgment still stands, 

A-4 
,· .... ' .. ,, 

page 3 ----Minutes of Expert Panel Meeting of June 27, 1986 

the calculated HCLPF values are not to be interpreted in the same way ..•. that 

is, if a plant-level HCLPF value is determined by the analysis to be, say, 0.30g the Panel believes that it is not correct to believe that the HCLPF value might just as easily have been anywhere in the range of, say, about 0.25g to about 0.35g. 

The conservatisms embedded in the 11 HCLPF" concept, through the concept of 'high confidence• and the concept of 'low probability of failure', should allow the 

regulatory decision-maker to cope with any technical decision without the additional 

•smearing' of placing an analyzed HCLPF value in a broad range like that just mentioned. The Panel was explicit and strong about its insistence on this point. 

8) The Expert Panel provided guidance to the margins review team that, even 

though categories or specific items of equipment are thought to possess quite high capacities compared to the RLEQ, it is still necessary to do at least some 

type of review of these items to confirm that they are not •outliers•. The criteria in the Panel's chapter 5 discussion (NUREG-CR-4334) are considered by the 

Panel to be 'generally conservative• but not 'absolutely conservative in every 

case.• Specifically, every screening decision must be confirmed somehow. 

9) The Panel expressed its strong disappointment that there has been no start to 

date on the separate research project that was recommended to compare the CDFM 

(conservative deterministic failure margin) method and the FA (fragility analysis) method for calculating capacity values. The Panel was told that in the Maine Yankee review Ravindra•s fragility team will do such a comparison on about a halfdozen items. This outcome, in which all of the capacities are to be determined 

using the FA method and the CDFM approach· used only as a •trial • or 'check', is 

certainly not what the Panel had envisioned when the guidelines were being developed. 

The trial review approach is not a substitute, in the Panel •s opinion, for a separate study to confirm the adequacy of the COFM method. The Panel expressed a 

strong need to go on record that the CDFM method needs to be confirmed. Although 

the Panel believes that the CDFM approach is ultimately the preferable method, it 

also believes that CDFM should not be used and relied on until confirmed. The Panel 

discussed possible dangers of two kinds: first, the CDFM method might be used 

prior to its being studied and confirmed; and/or second, it ~ight never be used 

because it is not a confirmed approach, thereby leaving margins analysts with no 

choice but to use the much more expensive and less desirable FA method. 

10) The Panel was told that it is very unlikely that Maine Yankee's reactor 

internals and control rod drives can be reviewed to obtain useful fragility information, because they are inaccessible and the vendor design/test information is unavailable. The Panel •s guidance was that in such a case it is 

preferable to do a detailed study of the back-up reactivity control sysems such 

as the borated water tanks and plumbing, rather than to leave the issue open, 

even though there is high confidence on a generic basis that the capacities of 

the reactor internals and CRDs are above the RLEQ selected for Maine Yankee. 

11) There was extensive discussion on the safe-shutdown end-point used by the 

Panel. The Panel reiterated that it used the traditional PRA approach: for 

transients a stable state is typically achieving hot shutdown plus holding it 

for 24 hours (sometimes 36 or 48), while for LOCAs a stable state is usually cold shutdown. The Panel did not explicitly use 'hot shutdown plus 72 hours•. 

The Panel reiterates its finding, based on the extensive PRA literature for PWRs, 

that it is very unlikely in a probabilistic sense that the systems being screened 

in the seismic margin method proposed would have a high HCLPF value while the 

systems needed for post-hot-shutdown heat removal would have a quite low HCLPF 

A-5 
page 4 ---- Minutes of Expert Panel Meeting of June 27, 1986 

value. Therefore, based on its study of the PRA literature the Panel has 

concluded that it is not necessary to do a detailed margin review that includes 

the systems supporting the function of post-hot-shutdown heat removal. 

12) Notwithstanding the above, the Panel reaffirms its earlier guidance that 

it is necessary to do a modest review of these and other systems to assure that 

there are not 'plant unique features' in the sense of the discussion in NUREG/ 

CR-4334 and 4482. 

13) The Panel also reaffirms its earlier approach to non-seismic-induced failures 

or unavailabilities, which approach is that these are to be combined with 

seismic-induced failures using an approximate probability-based cut-off criterion th~t relies on expert judgment (see NUREG/CR-4334). There is no precise numerical guidance on how to apply this probabilistic cut-off. 

14) Regarding seismic-initiated failures, the Panel was told that it would not 

be possible to enter the Maine Yankee containment during the upcoming review. 

Therefore it will not be possible to rule out small LOCAs inside containment 

caused by earthquakes like the RLEQ. The Panel agreed with Dave Moore (leader 

of the systems analysis subcontractor team) that an appropriate approach is to 

do the analysis two ways ----one way assuming that. a small LOCA inside containment does occur, and the other way assuming that it does not occur. Mitigating 

systems {injection pumps, etc.) will be studied to gain engineering insights as 

to the plant's possible vulnerabilities, if any. 

15) The Panel agreed that, at the RLEQ selected, the traditional large LOCA pipe 

breaks were not a problem, on a generic basis. The Panel was told that Maine 

Yankee has large loop valves in the primary piping which were placed there in the 

original design to isolate one of the main loops. These might be more vulnerable 

to the RLEQ than the primary piping itself. The Panel•s guidance was to treat 

these as 'plant unique featu~es' and to study them through design information if 

that was the only available approach. 

16, Panel discussed at length the difference between the Panel's and EPRI's 

approaches to the systems analysis. In a nutshell, EPRI's approach is to search 

for one 'success path' at the RLEQ, while the Panel's approach is to search for 

cut sets representing combinations of failures. The Panel reaffirmed its earlier 

conclusion that the cut-set approach is preferable since it provides more extensive engineering insights, but agreed to revisit this issue after both the EPRI 

trial review of Catawba and NRC's trial review of Maine Yankee are completed. 

17) Bob Murray reported to the Panel about the excellent cooperation being received from the EPRI margin study effort, which is doing a parallel review of 

Catawba. The Panel wishes to encourage the maximum cooperation, since both groups have already learned from each other and this will likely continue to be true. 

1g) The Panel was informed that the BWR systems study is now underway, which 

should enable the Panel to develop BWR guidance sometime next fiscal year. The 

Panel was urged by Bob Murray to give thought now to possible nominees of a BWR 

candidate for the second trial margin review. 

A-6 
page 5 ----Minutes of Expert Panel Meeting of June 27, 1986 

In conclusion, the Panel's main finding based on the conference call meeting of June 27 is that the trial margin review at Maine Yankee is proceeding appropriately, and that the Panel•s overall guidance is being followed, with 

the exception of our comment on the CDFM method (see item 9 above). The 

Panel is very pleases otherwise with the progress so far. 

Robert 

~~ J. Budnitz 

Chairman, Expert Panel 

A-7 
. •: 

Applied Risk Technology Corporation 

P.O. Box 1'75, Columbia, MD 21045 (301) 992-0525 

Date: 28 July 1986 

To: R~~~Mur Lawrence Livermore National Laboratory 

From: ~~· mico, Applied Risk Technology Corporation 

Subj: P.O. 9225705, Comments on Maine Yankee Trial Plant Review 

for Seismic Design Margins Program 

This memo constitutes the final deliverable for the subject 

purchase order, and is assumed to be sufficient documentation for 

the tasks performed. As you are aware, I reviewed the procedures 

de v e 1 o p e d by the NRC for t he M a in e Yankee t r i a 1 p 1 ant r e v iie w and 

partcipated in the conference call held on 27 June 1986. In 

general, my comments on the trial plant review are fully recorded 

on the tape recording of that meeting and adequately represented 

in the minutes of the meeting produced by R. J. Budnitz and dated 

11 July 1986. Insofar as this is the case, my comments will not 

be repeated in this report. However, I would like to make an 

addition/clarification to my position which was not included in 

the recording or minutes. 

First, regarding the cursory review of systems not part of the 

Group A functions (not-A systems), it should be clearly 

understood that the purpose of reviewing these systems is to 

identify any plant unique features which indicate that these 

systems are significantly more fragile with respect to earthquake 

than the Group-A systems. Under no circumstances is this to be 

construed as demonstrating that these not-A systems have a HCLPF 

greater than the RLEQ. By way of illustration, if a tank is 

located such that its rupture would cause failure of only not-A 

systems, this would be included in the review. Similarly, if for 

some reason there is similar equipment in both Group-A and not-A 

systems, but the not-A equipment has anchorages which are 

noticably inferior to the Group-A anchorages (easily recognizable 

during a walkthrough), this would be included in the review • 

However, if the above deficiencies existed for both the Group-A 

and not-A systems, the effect on the not-A systems would not be 

considered. Again, the purpose of the review is to determine 

only that the susceptibilities of the not-A systems to seismic 

events are generally of the same order as the susceptibilities of 

the Group-A systems, not whether they meet the HCLPF requirement. 

The review is very cursory in nature, and anything which cannot 

be clearly identified visually during the initial plant 

walkthrough is not to be pursued further. 

A-8 
page 2 - Comments on Maine Yankee Trial Plant Review 

Second, I wish to strongly express my disappointment with the 

decision that the expert panel would not constitute the peer 

review team, as was originally intended. In my opinion, the 

expert panel as a whole is much better suited to oversee the 

implementation of its review method and to assess its 

affectiveness and the required modifications than the group which 

has been selected as a peer review team. I wish to make clear 

that I will not "rubber stamp" the conclusions of the peer review 

team with regard to their assessment of the trial review process 

and guidelines, or blindly approve their suggested changes to the 

methodology based solely on their opinion of what changes should 

be mad e. Howe v e r , :i. t w i 11 o b v i o us 1 y be ext rem e 1 y d if f i c u ::,._ t for 

me to formulate an informed opinion on the validity of their peer 

review comments while not having any involvement in the trial 

plant review or its peer review process. The resolution to this 

dilemma is not clear. In the extreme case, it may be necessary 

for me to dissent from the peer review team's conclusions 

regarding the utility of the trial review guidelines. 

A-9 
. , 

Dr. R. Murray 

LLNL (L-197) 

P.O. Box 808 

Livermore, CA 94550 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, 0, C. 20555 

JUL iS ~ 

References: 1. 7/11/86 letter from R. Budnitz 11 Minutes of the Telephone Conference Ca 11 Meeting of June 27, 198611 • 

2. 6/18/86 memo from D. Guzy 11 Seismic Design Margins Program". 

Dear Dr. Murray: 

Reference 1 (enclosed) discusses the review of the Maine Yankee seismic 

margins review procedures by the Expert Panel on Seismic Design Margins. The 

Panel also discussed topics brought up by the NRC staff at our June 10, 1986 

meeting held here at the Nicholson Lane Building. I would like to have 

clarification of the following items from Reference 1. 

General: In Reference 2, I noted that I felt the review team's written 

procedures lacked depth, but that from the June lOth presentation it was clear 

to me that the team had definite ideas that would be finalized (and should be 

documented) as their review progresses. 

Please document whether or not the Expert Panel felt the review procedures had 

sufficient detail. If not, how did the panel make their judgment on the adequacy of the procedures (on subsequent verbal discussions?) and what should be 

added to future written procedures? 

Item 4 

This discussion seems to back away from a previous Expert Panel position that 

for high frequency, high acceleration, low to moderate magnitude earthquakes, 

their guidelines were conservative. Please clarify if the Panel's position has 

changed . 

Item 7 

Although this discussion addresses an issue raised at the June lOth meeting, 

J•m not sure it directly answers a specific question. That is, in performing 

a margins review, can we use the values in the first column of Table 2-1 of 

NUREG/CR-4482 and make a statement about the ability of the plant to withstand 

earthquake up to the .3g level? 

A-10 
Item 9 

I recommended that we delay the start of a separate study on CDFM vs. PRA 

fragility for the fo~lowing reasons: 

1. The end date for the proposed study was such that it could not be used 

in the Maine Yankee review. 

2. EQE's proposal presumably was going to use both methods and thus there 

would be redundancy if a separate study was run concurrently. 

3. FY 86 funding limitations wouldn't permit this study along with the other 

proposed activities. 

It's a particular sore point with me that after repeated requests for more 

detailed 189 work statements and review procedures, the extent of EQE's 

CDFM/PRA fragility comparison was not stated until I asked the question at 

our June 10 meeting. Now I see that the Panel feels EQE's use of the PRA 

fragility approach is "less desirable" and that the number of COFM examples is presumably inadequate to make any kind of a comparision. In light of these 

comments, I wonder if;stope of EQE's proposed use of the CDFM and FA methods 

was known and was a factor during LLNL's subcontractor selection process. 

It would be useful to know what kind of study the Panel had in mind for 

"confirming the CDFM method". It seems to me that building on the current 

Maine Yankee review would be the most efficient way to go. 

Enclosures: As stated 

cc: 

'J)Ct-v. G~ Dan Guzy 

Mechanical/Structural Engineering Branch 

Division of Engineering Technology 

A-ll 
t•t •.• 

Future Re.HJurceJ A~ociateJ, Inc. 

2000 Center Street Suite 418 Berkeley. CA 94704 415-526-5111 

Mr. Daniel J. Guzy 

Engineering Branch 

Division of Engineering Safety 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory .Commission 

Washington, D. C. 20555 

Dear Dan: 

September 26, 1986 

This letter is a reply to your memorandum of July 16, 1986 in which you raise several 

questions about the seismic margins program, both in general and in particular regarding 

the trial review at Maine Yankee. 

I will address the points one-by-one, going down your memorandum. Therefore, a copy of 

your two-page memo is attached to this letter for convenience. 

1. Your General Comment: The documentation of the NRC/LLNL review team's 

procedures has been insisted on by LLNL, and my understanding is that this 

documentation will be an integral part of the "deliverable" package that the review 

contractors will provide. The Expert Panel did not comment in detail about whether 

the procedures had sufficient detail, since at the time of the Panel's conference call 

of June 27 these procedures were still being developed. However, I am certain that 

the Expert Panel will comment on them later, because these procedures are intended 

to be a model of what should be followed by a review team implementing the "final" 

procedures to be finalized next year. 

2. Your "Item 4" Comment: The Expert Panel believes that a separate study is 

required of the issue of high-frequency, high-acceleration, low-to-moderatemagnitude earthquakes. The Panel believes that its current guidelines are 

conservative, but that a separate study is needed to reaffirm this, to clarify areas 

where the conservatism may be inadequate (or may be excessive), and to understand 

this issue better. 

3. Your "Item 7" Comment: This issue should be clear already, but I can clarify it 

again, by restating what the Expert Panel believes. The Panel believes that the first 

column should be used for nreview level earthquakes" up to 0 .. 30 g; the second 

column if the review level earthquake is between 0.30 and 0.50 g; and the third 

column for above 0.50 g. Suppose that the review-level earthquake is selected as, 

say, 0.29 g, so that the first column should be used. If, using the first column, a 

plant is found to have no Booleans (loosely, "accident sequences") that are 

A-12 
Mr. Daniel J. Guzy 

1.. 

-j- September 26, 1986 

vulnerable, then the conclusion is that the plant-level HCLPF is at least as large as 

the review level earthquake being reviewed against. If on the other hand, a plantlevel Boolean is identified, then the plant-level HCLPF can be quantified, for 

example at, say, 0.26 g (or whatever). 

4. Your "Item 9" Comment: The Expert Panel has continued to believe that a separate 

study is needed to exam1ne the relative merits of the CDFM and the PRA-fragility 

approaches to determining HCLPF values. The sooner this is begun, the sooner it 

will be completed. What is needed is a comparison of several different classes of 

equipment with several examples selected for each class. The first task is to 

establish a well-defined working definition of the CDFM method. Following this, for 

each specific equipment item being studied, the analysts should compute the HCLPF 

value using the CDFM method and compute a separate and independent HCLPF 

value using the fragility method. It should also be valuable, if possible, to use more 

than one group in the study, so as to illustrate any analyst-to-analyst differences in 

interpretation. 

The key aspect of the study is assuring that all assumptions made in each analysis 

for each equipment item are documented, so that differences between the HCLPF 

results may be traced and understood rather than merely stated. 

After this part of the study is completed, the analysts should be expected to make 

recommendations for any modifications in either the CDFM or the fragilities 

method, as needed to help resolve any differences found. 

The overall goal is to develop a method sufficiently "cut and dried" (that is, 

prescriptive) that it can be used routinely by the general engineering community for 

performing HCLPF calculations for seismic margins reviews. 

Your final comment is correct-building on the current Maine Yankee review may be the 

most efficient way to proceed on the CDFM comparison study at this stage. But that 

probably means starting as soon as we can, since getting a study in place probably will 

take a few months. 

I hope that this letter has answered your comments. If not, please let me know. 

RJB/sa 

Attachment 

c c. : iix p U" l P a. n l £_ 

R. 11 i).rrq~ ,~ LLN L 

Sincerely yours, 

~tJdg Robert J. Budnitz 

Chairman 

Expert Panel 

A-13 

APPENDIX B 

PEER REVIEW GROUP CORRESPONDENCE 

.. ' 

'. 

' ' 

B-1 
PEER REVIEW GROUP CHARTER 

R. J. Budnitz, Chairman 

The objective of the Peer Review Group is to assure that the trial seismic 

margins review is executed in a fully competent and professional manner, uses 

methods that are at the state-of-the-art, follows the guidance established in 

NUREG/CR-4334 and NUREG/CR-44S2, and takes cognizance of all relevant 

information. The sponsors of the stu~y (Lawrence Livermore National 

Laboratory for the NRC and Maine Yankee as the plant owner) desire to utilize 

the results of the study, and require the Peer Review Group's assurance that 

the study is technically sound. 

To accomplish its objective, the Peer Review Group will be provided full 

access to all materials, information, and methodologies that are inputs to and 

used by the study team. Access to the study team itself will occur through 

scheduled meetings to follow the study's progress. The Peer Review Group will 

also review draft reports and participate in walkdowns of the plant. Formal 

reporting and interface for the Peer Review Group will be through the Group's 

chairman to LLNL. 

8-2 
Future Re.Jource.J AJ.Jociate.J, I~re. 

2000 Center Street Suite 418 Berkeley. C:\ D470-t 415-526-5111 

4 June 1986 

Mr. J. Thomas 

Duke Power Company 

Design Engineering Department 

422 South Church Street 

P.O. Box 33139 

Charlotte, North Carolina 28242 

Dear Mr. Thomas: 

This letter is being written on behalf of Lawrence Livermore National 

Laboratory•s •seismic Margins Program... As you may know, I am the chairman 

of a newly-constituted "Peer Review Group 11 that will be assisting LLNL 

in carrying out a trial •seismic margins review• of the Maine Yankee plant 

in the coming months. I am grateful that you have agreed to participate 

as a member of this Peer Review Group, and I look forward to interacting 

with you. I know that Bob Murray of LLNL has given you more details. 

The purpose of this letter is to confirm the tentative dates of the first 

Peer Review Group meeting, to provide you with a tentative schedule for 

the rest of the Peer Review Group's activities, and to provide you with a 

copy of the Group•s charter. (The charter is attached to this letter.) 

The tentative schedule is as follows: 

1) The first Peer Review Group meeting will be held at the Maine Yankee 

site on July 21-22-23. The first day will be used for a Group 

meeting, and the next two days will involve a walkdown of the plant 

along with the margins review team and the utility personnel. 

Bob Murray of LLNL should provide you with travel details soon. 

2) In September, it will be necessar·y to devote 1 day (perhaps 2 days) 

to a review of the progress to that date; the study team will have 

provided their tentative findings and issues to us. There will not 

be a Peer Review Group meeting but we will discuss this by telephone. 

3) In November, there will be a 1-day Peer Review Group meeting to go 

over what has been accomplished to date. This meeting could go to 

a second day but I hope it will not. The site of this meeting and 

its exact date is not yet known. 

B-3 
J. Thomas, Duke Power Company 

page 2 --- 4 June 1986 

4) There will be a 1-2 day commitment in February, 1987 to review 

the final study report in draft form. This meeting will be at 

a site to be determined, but it will probably be he1d in 

Washington so that the Peer Review Group can interact with the 

relevant NRC staff members. 

The Peer Review Group reports formally to LLNL, and specifically to 

Dr. Robert C. Murray, who is the designated LLNL contact. However, 

our report will be a public document that will also be of use to the 

Maine Yankee group, other utilities, as well as EPRI and the NRC. The 

report format will be a letter from me as Chairman, reporting on the 

Group's findings. It has been agreed that there will be an opportunity 

for any Peer Review Group member to write a separate minority report 

if needed, but the expectation is that the Chairman 1 S letter can capture 

all of the review comments. 

I hope that this letter and the attached Charter can clarify your role 

as a member of the Peer Review Group. If you have any questions, I will 

be happy to discuss them with you. I look forward to meeting you at 

Maine Yankee on July 21. 

With warmest regards, 

Sincerely yours, 

£1./Jt:p' 

Robert J. Budnitz 

cc: R. C. ~1urray 

B-4 
·:. rE, .·"' 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D. C. 20555 

JUL ? 1986 

MEMORANDUM FOR: Mr. James Richardson, RES Co-Chainnan 

Seismic Margins Working Group 

FROM: 

SUBJECT: 

Newton Anderson, NRR Co-Chairman 

Seismic Margins Working Group 

COMMENTS ON THE NRC SEISMIC MARGINS 

PROGRAM AND TRIAL PLANT REVIEW 

I have recently received a copy of a letter from Dr. Budnitz, Chairman of the 

Seismic Margins Peer Review Group to members of that group. The letter 

transmitted a tentative schedule for Peer Review Group m~etings and a copy of 

the Peer Review Group Charter for the Maine Yankee Trial Review. 

I have some concerns about both the charter and the schedule. I believe that 

the Charter does not reflect what my understanding is of the Peer Review Group function. It should be clearly articulated that we expect the Peer Review 

Group to not only provide " ••• assurance that the study is technically sound .. , but to endorse the technical results. At the conclusion of the Haine Yankee 

review they should be expected to provide their best judgment with regard to 

both the review procedure and th~ seismic capability of Maine Yankee, based on 

their collective expert opinions and backed by their professional reputations. 

The other problem I have with the Charter is the statement on formal reporting. The Peer Review Group should not report through LLNL. They should report as an 

independent group, providing their results to both LLNL and NRC without any intermediaries. LLNL should provide administrative support to the group but I 

feel strongly they should maintain their independence. 

If my comments with regard to the Charter are accepted, then I feel that the 

proposed schedule for Peer Review Group meetings is inadequate. It appears to 

me that the limited number of meetings and allotted review time is not sufficient 

to allow the group to 11 Stake their professional opinions" on their assessment 

of the study results. The Peer Review Group members should decide what level 

of involvement they need based on what we expect of them. 

I would also like to express my concern about some problems I see with the 

relationship betwe~n the NRC/LLNL Seismic Margins Program and the EPRI Seismic 

Margins Program. My basic concern is that there is a significant difference in 

the approach and in the scope of review. We need to work hard to bring these 

programs together. My understanding is that EPRI sees the NRC program as a 

study to "develop a requirement 11 to do a margins review and that the EPRI 

program is to provide the utilities with a tool for perfonming plant specific reviews to meet that requirement. If this is a reasonable way to proceed, then 

we certainly should work with EPRI to ensure the two programs are compatible. I do not mean to imply that we are not coordinating with EPRI. I know we are; 

but 1 do not see these differences being resolved. 

B-5 
, ... ' 

•11···. 

- 2 -

I suggest that we hold a meeting with the NRC seismics margins working group in ~arly August for the purpose of discussing these issues and developing a 

plan for resolving them. 

1~~r~char;~n Seismic Margins Working Group 

cc: Seismic Margins Working Group Memb~r~ 

Or. Robert Budnitz 

or .... Garth" ·cunn1ngs. LLNL 

T. Spe1s B. Sheron 

R. Bosnak 

B-6 
2000 Center Street Suite 418 Bel'keley. CA 94704 415 .. 526-5111 

1 August 1986 

TO: U.S. Nuclear Regulatory Commission 

D. Guzy, Office of Nuclear Regulatory Research 

P. Sears, Office of Nuclear Reactor Regulation 

Maine Yankee .L\to1nic Power Company D. Whittier, Manager, Nuclear Engineering and Licensing 

FROM: Robert J. Budnitz, Chairman of Peer Review Group for the 

Maine Yankee Seismic Margin Review Study 

REF: MINUTES, FIRST MEETING OF THE PEER REVIEW GROUP 

These are minutes of the first meeting of the 11 Peer Review Group .. that is 

reviewing the technical competence of the .. Seismic Margin Review Study11 that 

is being undertaken on the Maine Yankee reactor plant under sponsorship of 

the U.S. Nuclear Regulatory Commission. 

The meeting was held on Monday, July 21, 1986 and continued on Tuesday and 

Wednesday, July 22 and 23. Attending besides the chairman were Dr. John Reed; 

Mr. Loring Wyllie; and Mr. James Thomas. Mr. Thomas' attendance was limited 

to only the morning of the first day. Absent was Dr. Michael Bohn, the fifth 

member of the Peer Review Group. 

Also attending were representatives of the NRC; the Maine Yankee staff and 

their Yankee Atomic Electric associates; Lawrence Livermore National Laboratory staff; LLNL's subcontractors performing the review, Energy Incorporated and 

EQE Inc.; and R. Kennedy, a consultant to Maine Yankee. The signup sheet 

for the opening session is attached as Attachment A; however, not all of these 

individuals participated in all of the meeting sessions. 

Attachment B shows the agenda for the first day's session. On the second day, 

a plant walkdown was done by most of the attendees in teams. of 4-6 each, preceded by a radiological protection briefing and followed by a de-briefing session 

in which technical issues raised during the walkdown were discussed. On the 

third day, another walkdown in the morning was followed by still another session 

in which technical issues were discussed. The Peer Review Group then adjourned its session just after noon, and the study team continued its discussions and 

its walkdown, which would last for the remainder of the week. 

The minutes of the meeting will be presented as numerically ordered topics, 

as follows: 

B-7 
Minutes of Peer Review Group, Meeting of 21-23 July 1986 

page 2 

1) The first session began with a briefing by Robert Murray of LLNL, who 

presented the background of the project. Murray discussed the purpose of the 

trial margin review, the development of the methodology by the NRC's "Expert Panel of Seismic Margins .. , the schedule of the review, its structure, and 

the relationships among the parties. During the course of Murray's presenta- tion an important comment by A. Thadani of NRC was made in which Thadani pointed 

out that any licensing action that might result from this study would only 

occur after its completion. NRC, according to Thadani, was awaiting the results 

because there are methodological issues as well as technical issues about the 

Maine Yankee plant that are being studied in this trial review. 

2) Robert Budnitz discussed the interactions that have occurred so far between 

this study effort and a parallel effort being undertaken by the Electric Power 

Research Institute, which is studying the seismic margin at Duke Power's Catawba 

station. Budnitz pointed out that recent discussions have assisted both study 

teams to understand in what ways their respective met~odologies are similar 

or different. Continuing interactions will be encouraged. 

3) James Thomas discussed the importance of interactions between this study and related technical work being done under the NRC•s A-46 program and under 

the industry's SQUG (Seismic Qualifications Utility Group) effort. Taking cog- nizance of these other efforts will be important for the study team. 

4) William Henries of Yankee Atomic then discussed the current status of the 

Maine Yankee plant in terms of its seismic capacity. He provided background 

on the history of the seismic design of the plant, recent utility-sponsored studies that provided information to assist them in understanding their plant, and recent actions taken to modify the plant's seismic performance. 

5) David Moore of Energy Incorporated, who is leading the systems-analysis 

team performing this review under LLNL subcontract, made a presentation that 

provided information on the approach being taken. Moore's viewgraphs are 

attached to these minutes as Attachment C. He pointed out that there are a 

few key issues that must be addressed, and asked for discussion and guidance from the group on how best to approach them. For example, because a walkdown 

inside the containment will not be possible, it will be necessary to make some 

assumptions about the presence of small LOCAs, and the approach being taken to 

this issue was covered. Also, the handling of the possible sequences in which 

control rod function might be compromised was discussed, and Moore•s review 

approach to reactor internals and boric-acid safety injection was covered. 

6) A lunch-time break was taken so that the Peer Review Group could go into 

executive session. After that executive session, the Peer Review Group suggested that minor modifications to the draft PRG charter should be made 

to reflect more accurately the actual approach being taken. The revised charter 

which was discussed by the PRG with those present is shown as Attachment D. 

The differences between tnis version and the earlier draft are a more explicit 

B-8 
\ ' : 

Minutes of Peer Review Group, Meeting of 21-23 July 1986 

page 3 

statement of the fact that the Peer Review Group is an independent group reporting to NRC and Maine Yankee, and a clarification of the Group's role 

vis-a-vis assuring that the methodology is being followed. 

7) M.K. Ravindra of EQE Inc. made a presentation in which he discussed his 

group's approach, as the fragilities subcontractor for this effort, in carrying 

out their work. His viewgraphs are attached as Attachment E. Technical topics covered included Ravin~ra•s approach to HCLPF determinations using the fragilities method and the CDFM (conservative deterministic fragilities method) for analyzing capacities. The handling of seismic capacities for those numerous components (valves, etc.) which can only be studies in a sampling way rather than in a 

100 %-analysis way was covered. Also, a discussion took place on how the study will examine those 'Group 8' components that require study to assure that there 

are not unusual issues involved in their capacities. 

8) The structure of the Final Report was covered, so that all parties present could understand how it will be structured, who will write which sections, and 

its schedule. The process whereby the Peer Review Group will be able to review 

a draft version of this report toward the end of the project was discussed. It 

is anticipated that there will .be a meeting of the participants to enable the 

Peer Reviewers to interact. 

9) The PRG's reporting was discussed. It was agreed here, as in earlier meetings, that a letter from the PRG Chairman would provide the Group's final comments 

on the study. Concurrence by the other members, with the opportunity for any individual PRG member to provide individual comments as a minority report if 

desired, would be the method used. 

10) An extensive discussion took place on how the systems analysis team would 

handle 'non-seismic induced failures• in their systems analysis, when a potential accident sequence might involve both these and seismic-induced failures. The 

Energy Incorporated study team will do an analysis which will incorporate these 

non·siesmic-induced failures where their presence will make a significant difference in the overall risk profile of the plant. 

11) The Monday session ended just after 5:00 PM. 

12) On Tuesday (July 22) a radiological briefing and a plant walkdwon in small 

groups took most of the day. A debriefing session from 4:30 to 7:00 PM ended 

the day. At that session, various technical issues uncovered during the day's walkdown were discussed, to allow the next day's walkdown to be more effective. 

An extensive disussion of the capacity of the DC batteries took place, along with discussions of the capacities and functions of several tanks. Some of 

the walkdown teams had not completed their entire first-pass tours, so this discussion was partly of the character of assuring that these teams looked at and 

studied plant features that some of the other groups had highlighted. 

B-9 
Minutes of Peer Review Group, Meeting of 21-23 July 1986 

page 4 

13) During this session late on Tuesday afternoon the role of the Peer 

Review Group in interacting with the study participants was discussed. The 

independence of the PRG must be assured, but it is also important that technical issues of concern to the PRG be provided to the study team rather than 

'kept to the end'. The handling of this aspect of the project was covered. 

14) On Wednesday morning (July 23)the plant walkdown continued, until 11:00 

when the meeting of the Peer Review Group continued for just over one hour. 

In this final session, PRG members discussed again the issue of assuring that 

independence for their work would be a fact as well as a perception. Also, 

the interactions between the NRC-sponsored study team and the utility staff 

were discussed, to assure that there would not be improper influence by the 

utility over the project outcome. 

15) The Peer Review Group•s final discussion and comments prior to adjournment 

were of the character that the study seems to be •on track' in a technical 

sesne so far, although of course this is a very preliminary observation, and was 

made by only three of the PRG members (Budnitz, Reed, Wyllie), the other two 

(Thomas, Bohn) being absent. 

16) The PRG meeting adjourned just after noon on Wednesday, 23 July. 

17) The next meeting of the Peer Review Group will be held on September 30, 

at a location in the San Francisco Bay area that will be identified by the LLNL 

team soon. The subsequent meeting of the PRG will be held in November in conjunction with the second plant walkdown. It wtll be held at the Maine Yankee 

plant site. While dates have not been firmly set, the dates of November 17-18-19 

were written down tentatively, and are being held by all participants pending further developments. 

18) These minutes, written by the PRG Chairman (Budnitz), are being sent to 

the other members for their review and comment, after which they will be made 

final. The draft version was not circulated outside the Peer Review Group. 

Attachment A: 

Attachment B: 

Attachment C: 

Attachment D: 

Attachment E: 

Robert J. Bud~fz, Chairman 

Attendance signup sheet, 7/21/86 

Agenda for first day•s session, as prepared by R. Murray of LLNL 

Viewgraphs for D. Moore•s presentation (Energy Incprporated) Revised charter for Peer Review Group Viewgraphs forM. K. Ravindra•s presentation (EQE Inc.) 

B-10 
Minutes of Peer Review Group, Meeting of 21-23 July 1986 

page 5 

ADDITIONAL ITEM: 

(This item was not in Budnitz•s notes for the meeting, but Reid and the others 

remembered that it was discussed ..... probably while Budnitz was out of the 

room. It is presented here because it is deemed important additional guidance from the Peer Review Group): 

19) The Peer Review Group•s interpretation of NRC•s requirement on the input 

spectrum to be used in this review is that a NUREG/CR-0098 spectrum, using 

50th percentile amplification factors,and anchored to 0.30 g zero period acceleration, is the proper ground motion to use. No variability in this 

input spectrum is to be used. Reference for NRC's guidance is the memorandum 

from D.M. Crutchfield to P.M.Sears, dated 7 May 1986. 

B-11 
I,,.' 

ATTACHr~ENT A 

·----------

_lit! ~---e_t~i'NJ";<I- _ _ bT"-tcL.~ J '

7

'/."( (J /-.!t~,t. ___ {'/! 5) )_·?: { ,'- ' _ 

A~t-ml<. --,;Ill>~' /'1/J< c @()!) fJ'fZ- '/SS3_ 

~8; Cv~M::~s -~~~l_ ______ _ __ ---· . (f,.;;_~22~ 4:'149 

-T/b\: --- l'i -~-1--- _ .-... _ _ ___ _____ . _ /t{l\) .'{72=03_01; 

__ Pel __ ~~~ .... rJ; _SJ.i_J\C ___ _ ________ __ _____ (i;ifZ7Z- ?/0_ 

-::r ·~ '-' _/J.., ~-YvJ _, _ .. P .. ':!.~k ~ ____ p9 ~!lr __ <;'?~---- _______ ( 7~4)_ 3? 3- ~ ~ I ~--

_Lor:.~ Wy/1•''!...- :Dejel\i4.1b.Az..soc.._ _ 0J.tr)3'1_2._-:-fa'1.S2.,-

--~0~o0j(~_ ··----~~--------- - (2l!_D_j_6__i:_84JZ. 

_fja_j),")_JYl.o~.!:~- -- -~l'l~)'-~C-c ---------· ------~~_(-zp_~)-~~!_--{)~_[Q___ 

~i(L: _ _Qu l~_,<:..)_ _ eN.E.U.'-1 :r:AL ----(~~)_3__)~ :_/;u~:J 

M_ .. J(_ ~-~VJ~~& r-,._, ~_C!(E._~_j_~ __________ L7_(:±)_8_~~ -_9_;{_2_~-. 

-----"'""'Del.~ Gt~:!7 _____________ _ JJ_ ~--.. ~~- 3_~1-et:J> .. -=?"'t_o __ ___ Jov __ Y9vJ'I~ .. _ .. C'':~G-'~~-YJi!~-. ________ 1~orS_)c!_"'$_~?Q8./2 ____ . 

_'13.Jl __ }~c~~'E .s ______ __ ~NKS£. ATD~a.c.._ _ ______ _f~L7) __ 8"_71ilB/c;o_ 

N,·lts.~ __ Choks~i .... __ .. ·-·- .N~_C______ _____ __3~L_..:_ _ _f.q_1,_- 83~7 

_l?£re. 1k.A$5.'l~.QS__ _ ______________ L._L.N L __ _ _______ ( 4..}_?_)_1~~~-~75_8_ __ _ 

~~-~-!J£4_-~~ -~~~-~ -- --- .. ~!f~--~~:~J -- .. c::;-~ ~~-

__ _/f_n1 Il t!l £ i cv 11{ 

?tl\R.(C..K -~G.~'S. 

J 0 'L P" c.. (\..- Mr \,. f;t,(" 

1/j I· ,/ . 

rNI!/IU'$ _f',~o~ __ 

-~tft=1 /vtc . YIJ}J!cF-e li7or .. ,, c 

~~1"-l~R 

'/A " lr..u- ~ ~ n,. )C._ 

NiL/ ~ji911f 

t:.are... :r.~ .. 

B-12 

~17- ?'~---·~. 

_____ __,Qj __ ':f 12. ~DC>b __ 

. -~ .. -- (;., (?.._~ B1 't- 9) o'=> _ 

o?o/ -88c<-7571 ·-· 

1 tc!\ -1t5" 2.. - 'l t. q~ 
9:00 

9:15 

9:45 

10:30 

10:45 

11:30 

12:00 

1:00 

5:00 

MoodaJ lleetiDC Aceoda 

Maine Yankee Atomic fbwer StatloD 

Starr Buildiag 

July 21, 1186 

Project Overview/R. c. Murray 

Yankee Atomic/Maine Yankee Plant Overview 

Systems Status/Energy Inc. 

Break 

Fragility Status/EQE, Inc. 

Open Discussion 

(collect Health Physic paperwork) 

Lunch (at plant) 

Peer Review Group Discussion/R. J. Budnitz 

Adjourn 

B-13 

ATTACHMENT B 
.. ' 

SYSTEMS ANALYSIS 

ENERG¥ INCORPORATED 

TEAM: 

DAVID MOORE 

JON YOUNG 

MARC QUILICI 

B-14 

ATTACHt~ENT C 
SYSTEMS ANALYSIS PROCEDURES 

STEP 2 - Initia1 Systems Review 

STEP 4 - First Plant Walkdown 

STEP 5 - System Modeling 

STEP 6 - Second P1ant Wa1kdown 

STEP 7 - System Modeling Analysis 

STEP 8 - Margin Evaluation of Components and Plant 

··· . .., 

B-15 
GROUP A SYSTEMS 

HPSI HIGH PRESSURE SAFETY INJECTION 

AFW AUXILIARY FEEDWATER(includes EMERGENCY FEEDWATER) 

ASOHR ALTERNAiE SHUTDOWN DECAY HEAT REMOVAL 

BAT BO~~c ~CID TRANSFER 

PPC PRIMARY PRESSURE.CONTROL 

SPC SECONDARY PRESSURE CONTROL 

ACP AC POWER 

DG DIESEL GENERATORS 

DCP DC POWER 

PCC PRIMARY COMPONENT COOLING WATER 

SCC SECONDARY COMPONENT COOLING WATER 

SWS SERVICE WATER SYSTEM 

ACTUATION (includes RPS, SIAS, and maybe RAS, CSAS, CIS) 

B-16 
SYSTEM: 

SAFETY FUNCTION: 

SYSTEM COMPONENTS: 

Tanks: 

Pumps: 

Heat Exchangers: 

SUPPORT SYSTEMS: 

AC Power: 

DC Power: 

A in 

HVAC: 

Pump CooUng: 

Actuation: 

HIGH ?~ESSURE SAFETY INJECTIO~ (HPSt) 

Injec: ~ora ted ·.water in to :he reactor ·tessel 

!mmediately afte:- a LCC.~. Also for feed and bleed, 

~ost-accident core cooling and additional shutdown 

capability curing rapid cooldown of RCS. 

?-!~A (N .. O.) 

P-! t:S fS} 

P-!t&S (Spare) 

E.-JA 

E-JB 

Re!ue!ing Cavity ':Vater Storage Tank 

C~1arging fHPS() Pump 

Charging (HPSI) Pump 

Charging (HPSI) Pump 

Residual Heat Exchanger 

Residual Heat Exchanger 

!: 160V Emergency Bus .S 

!L 160V Emergency Bus 6 

PCC 

sec 

SIAS 

B-17 

P-14A-2, 3 Lube Oil Pumps 

P-148-2, 3 

P-llJ-S-2, 3 

low :'ressurizer Pressure 

High ContainmP.nt Pressure 
r-< CV) 

- c.. "' a. 

-- s 0 \1 j "" -J 

r-1 1- f:'l{ lr1 

~ J~ 

- 17 N 

~~ 1 

~ I I! " I t 

\ I I I ,... .... ... . ~ 

"' nf. Yl "" Jl -- ;;: -- ..... - c{ io! ioi I I I 

..r;; 

4 !: 

~~ D 

- "" ~ ~c{ c:-e :r ';!"' - - <&.» 

<a I I t :::rll 

:< • 

"'~ I. I r r "'~ ~ ~ ~ • 

F- e- ..,.. Vl ... ~= 

- ~ ~ .L 

! ~~ 1::. 

- ~ 0 

·~ oi oi i ~ lr/ '-

~.:: ~f ~ UJ 1.!. t-1 I (j I <.J 

L--- z ___ _J 

c~ ...._ 

f'S 

k'J 

'() "' r2-.A e"' t UJ ~ci. c.;. C. ~ 1.!. !IJ - <C 0 Vl 

0 d) 

::r \1) 

ih d: ~ . ') J: 

~ V') - ~ Vl 

uJ 

r:C 1:. f- ~ 

t .J.. 

.... CJ ~ -....1.-

~ - Ul ~)"I:!.C ::t" .., ... It,! "'lt. ~ l 

.: - ""' ~ ; '::! '=>"'ll:c...: 

~ ¥.r-.,.; 'IIIII: S."'"" . r CC' V') 

B-18 
VALVE TABLE 

Otle rat ·1 ny 

Power Norma 1 Position Fail 

Valve Oescription (SOV) Position ( Actuat ·1 on) Posit. ion ----

~FN-A-101 Flow control to SG E-1-1 120VI\C 11\ 0 C( SLit) 0 

(12011\1) 

Afw ... J\ .. 201 Flow control to SG E-1-2 12f~VI\C 11\ 0 C(SLR) 0 

( 12Ul1H) 

AFW-A-301 Flow control to SG E-1-3 l20VI\C 1/\ () C(SLR) 0 

(1201Cl) 

AfW-A-338 Flow control isolation valve (1\FW-A-101} 120VI\C 31\ 0 C(SLR) 

(12051\) 

AFW-A-339 flow control isolation valve (1\FW-A-201) 120VI\C 31\ 0 C( SLR) c;o 

I (120511) ___. 

1.0 

AFW-A-340 Flow control isolation valve (AFW-A-301) l20VJ\C 31\ 0 C(SLR) 

( 1205C) 

BI\-J\-32 Doric acid VCT. isolation valve c C(SIASJ c 

(210Z) 

DA-A-00 Doric a~id VCT (bottom) isolation valve 0 c 

RI\-F-30 Boric acid flow control valve c C(SIAS) c 

(210Y} 

BA-M-36 Emergency boration isolation valve MCC 81\ c O(HAH) AI 

BA-M-37 Emergency borat1on isolation valve MCC 7A c O(HAN) AI 

CJt .. J\ .. 32 HPSl pump 8 discharge to charg1ng header OATT-1 0 C(SIAS) 

(255) 

CII-A-33 IIPSI pump 1\ discharge to charging header OATT -1 0 C(SIAS) (254) 

CC-2/1 
i~ li ,,: i I I ,I I II 

·oo: I I I I I ~ ~~~--~~--T---7--~~--~--~~--~--~~--~~--~ 

·~I I ! . I - (?~ I 

---

J-

. 

""' -r.r 

"' e.. c) <C -_... ( ~ IJ'J 

\ 0 rJ 

a- .. _·"'- n..· 

I I 

I I 

B-20 
I l ! l i I I i I i I I I I 1 

i ! l I I I I ! I I I I I I I --i-1 ~~~~--1--+--r---;.-.....; I I I I I I ' I I I I I I I I i I I I I ! l ! I I I ! 

~~ i I I I I I I I I I I I I I 

ll'll' I ' I I j I I I I I I I I ~! ;: I ! I I \ l I ! 

"'' •·ll ... Cl I : ·:': I I !. I1: I .I i I I II I I I· I I I ;: l I I . I . I : l ----1 • I ' ' I I ' . I I I 

I 1,/') I i ! I : ! I I I ; i I I I' I I a 1 ! i 1 i 1 ! : i ! i xl x 

' I .., i I i I I I I I~~ I I I I I I I I .I I I I I I ·I I . i I i i 

'li ~l I : I j I ! ! I I I I 

cJ I I I I I I 

I I l xl I I I 

I I I l I I J '1 I I I I I I ul""'t I I I I i I ~I I I I I ! I ! ! 

j~-! I ' I I I I xl I I I I .I I I I I I I "I I I I I I I i ! I 1 I 

! I I t:."". I I I I I I I I I I ! 

~i i i ~: I x: I I I;, I I xl I I I I I I I I ~I I xJ I I I I I l: • I I I I"""""' I ,- . I .... ! ! I : I ' I I 

~li1 I I I I I ><! ! I I I I I I I J 

, ... 

I I I ~t I I I I I I I I -r~1 "1 1 x 1 I x( I I I ~1 I I I I I I I xl I I I I lx I I I I i 

I •t I )( I I I I I I I I I I I I I I I I I I I I I I I I 'C -

r:: :i I I I I 

~.I I I I I I I I I I I I I I I I I i I I I I I I I I I I ~ ~ )( I I I I 

H I I I I I I I I I I I I I I I I I I I I I I I I I I l 

l I I i r:,J I I I I I I I I I 1 I I I I I I I I I I I I I I I ,,.. f 

c{i• I I I I I 

ail I t I I I I I I I I I I I I I I I I I I I I I I I I I I 

') I~..::"' 

1 o:l> I I P::!; 

c.J;P,.e-' I I I t I I 

I I I I I I I I I I I I I I I I I I I I I ~~ ,. I I I ! I i 

II;.! I I I I I I I I I I I )ll )ll ~ I xl I I I I I I 

I I I I I 

I I I I ..... I ; I I 

;. I 

I I I I 

I I I I I I I I I " I )£ I I I t I~ I I I I I I I I I 

IIJ! I I I' I I 

I 1 .. I I I I~ I I I I I I I I I I I xI I I I I I I. I I I I I I I I l..t' 1>4c::: i I I 

~ I I I 

!r. .. I I f x! I I I I I I I I I J I >(I I I I I I I I I I I I I ·i I I • .., I I fc: I I I I I I 

H,.l-;jr-! .. 1-<IN!..,I-:-Ici,.i_.l,..i..,l,.l lvl I' lie "'II I I 

I 

I 

I 

I ~-;. /; f1 ~ 1 ~! ~ 1 ~ • ~ : ~ ~ ~! ~ 1 ~ l '"'! ..; j .; v: ~ v! c: :01 u a 1 ~ ~ I 

~ ;., ! c:tJ ' ::::. I ca. : c:C j :> : ~ l > : I ~ ; &. : ,:;:.. ::. I p. ! I c/'1 I ~ I I ~ "' I 

"'k i ~ , 1 ~ ·, 1 I ':. I - I 1 ~ ~ ~ ! 

. ~""' 

t , 1 ,., 

,.J , " t • 1 a ]l to\ 

; 1 .:: !j 1 

I ~ ':" I / : . 'to - ,....., c ' ""' I - ~ t.! v l i fi. It:'"'".. I ::: L ""' ' · "' e ... t: , • .., 

B-21 
1 n-::; ..... .). 

RPS 

I . I 

:---------

1 

I 

I 

I 

I 

~i\'S - BAT 

\ 

<.~tArw I R:s Intes. I HPSI/F&a 

·----

EFW/AF\' SRVs ~ Open/Reclos HPSI 

rI 

Seismic Margin Event Tr@es (?reli~inary) 

B-22 

OK 

OK 

co 

OK 

CD 

TransferA 'ilr-'S 

Status 

OK 

OK 

co 

co 

CD 
PEER REVIEW GROUP CHARTER 

(21 July 1936) 

ATTACHMENT 0 

The objective of the Peer Review Group is to assure that the trial seismic 

margins review, following the guidance established in NUREG/CR-4334 and 

NUREG/CR-4482, is executed in a fully competent and professional manner, uses 

methods that are at the state-of-the-art, and takes cognizance of all relevant 

information. The sponsors of the study (Lawrence Livermore National Laboratory for the NRC and Maine Yankee as the plant owner) desire to utilize the 

results of the study, and require the Peer Review Group•s assurance that the 

study is technically sound. 

To accomplish its objective, the Peer Review Group will be provided full access 

to all materials, information, and methodologies that are inputs to and used 

by the study team. Access to the study team itself will occur through scheduled meetings to follow the study•s progress. The Peer Review Group will also 

review draft reports and participate in walkdowns of the plant. A formal report 

for the Peer Review Group will be made by the Group•s chairman to NRC and Maine 

Yankee. It is understood that the Peer Review Group's report will be a public 

document. 

B-23 
ATTACHf~ENT E 

NRC SEISMIC MARGINS PROGRAM 

TRIAL PLANT REVIEW 

FRAGILITY ASPECTS 

STATUS REPORT 

PRESENTED BY 

M.K. Ravindra 

G.S. Hardy 

P .S. Hashimoto 

S.W. Swan 

EQE Incorporated 

Newport Beach. CA 

PRESENTED TO 

PEER REVIEW GROUP 

JULY 21, 1986 

B-24 
OUTLINE 

• Initial Screening 

• Identify Target Areas for Walkdown 

• Walkdown Procedures 

• Documentation 

• Outstanding Issues 

B-25 
INITIAL SCREENING 

• Categorize Maine Yankee Components Into Generic 

Component Categories Identified by Panel 

• Pre-screen Components In or Out Based on Panels 

Recommended Guidelines 

• Identify Areas of Concentrated Effort for Each 

Generic Component Class 

B-26 
MAINE YANKEE SEISMIC MARGINS REVIEW 

SYSTEM: HIGH PRESSURE SAFETY INJECTION (HPSI) 

COMPONENT 

TK-4 Refueling Cavity Water Storage 

Tank 

P-14A Charging {HPSI) Pump 

P·14B Charging (HPSI) Pump 

P-145 Charging (HPSI) Pump 

E-3A Residual Heat Exchanger 

E-3B Residual Heat Exchanger 

Bus 5 4160V Emergency Bus 

Bus 6 4160V Emergency Bus 

P-14A-2,3 Lube Oil Pumps (Pump cooling) 

P-14B-2,3 Lube Oi 1 Pumps {Pump cooling) 

P-145-2,3 Lube Oi 1 Pumps (Pump cooling) 

Actuations 

INITIAL SCREENING 

11i OUT 

X 

X 

X 

X 2 

X 2 

X 3 

X 1 

X 1 

X 1 

X 1 

X 1 

X 1 

1 Anchorage must be inspected during walkdown a~d verified adequate. 

2 Cabinet anchorage & attached component anchorage must be inspected during walkdown and verified adequate. 

3 Actual components yet to be identified. 

B-27 
MAINE YANKEE SEISMIC MARGINS REVIEW 

SYSTEM: Auxiliary Feed Water (AFW) 

COMPONENT 

TK-21 

P-25A 

P-25B 

P-25C 

T-1 

Bus 5 

Bus 6 

E-86A 

E·868 

E-86C 

Actuations 

Demineralized Water Storage Tank 

Emergency Feed Pump 

Emergency Feed Pump 

Emergency Feed Pump 

Turbine for P-258 (Powered from 

Main Steam) 

4160V Emergency Bus 

4160V Emergency Bus 

Oil Cooler 

Oil Cooler 

Oil Cooler 

Instrumentation 

INITIAL SCREENING 

lN OUT 

X 

X 

X 2 

X 2 

X 

X 

X 

X 3 

X 3 

X 1 

X 1 

X 1 

I Anchorage must be inspected during walkdown and verified adequate. 

2 Cabinet anchorage & attached component anchorage must be inspected during walkdown and verified adequate. 

3 Actual components yet to be identified. 

B-28 
r ',·, 

+ •• ·: ' .~ 

IDENTIFY TARGET AREAS FOR WALKDOWN 

• Locate All Components for Walkdown on Plant 

Layout Drawings 

• Identify Buildings and Areas Requiring Access for 

Walkdown 

• Based on Initial Screening-Identify Specific 

Component Areas for Concentrated Review, ie 

(Anchorage, Lateral Restraints, etc.) 

• Develop walkdown Data Sheets for Each Generic 

Component Class Defining Areas of Concentrated 

Walkdown Effort 

B-29 
WALKDOWN PROCEDURES 

• Perform Walkdown for Identified Components 

• Address Areas Identified in Initial Screening 

Requiring Concentrated Review 

• Confirm Screening Criteria of Panel is Satisfied for 

Each Component 

• Identify Areas That May Require Additional Review 

B-30 
STRUCTURES SCREENING PROCEDURE 

Component Screening Comments 

Structures Out Structures housing Group A components 

are tyP.icaiiR categorized as Class I 

in the FSA . Class I structures were desi~ned for the O.lg hlflothetical 

eart quake using the A I 318-63 and 

AISC codes. Any gross structural 

deficiencies will tie Identified by review of the design drawings and 

walkdown. 

Structure Out Class I structures are either cast 

Impact intetral with each other or are separated by t ree inch gaps. This will be 

confirmed by walkdown. 

Block Walls In A comprehensive screening procedure has been developed. · 

Yard tanks In Information from drawings will be 

supplemented by walkdown. 

Soil Liquefaction Out Structures and yard tanks are founded 

on rock. 

Control and In Presence and adetuacy of safety wiring I• Battery Room will be confirmed y walkdown 

Ceilings 

Dams, Levees, In Nearby dike will be reviewed in 

and Dikes walkdown 

B-31 
I 

WALLS SUPPORTING 

GROUP A 

COMPONENTS 

SCREENED IN ~~ 

BLOCK WALL EVALUATION PROCEDURE 

REVIEW AVAILABLE INFORMATION. 

IDENTIFY ALL BLOCK WALLS USING 

MAINE YANKEE SUMMARY TABLE. 

LOCATE BLOCKWALLS ON EQUIPMENT LAYOUT DRAWINGS. 

I 

WALLS WHOSE 

COLLAPSE. 

COULD RESULT 

IN IMPACT 

ONTO GROUP 

A COMPONENTS 

l WALKDOWN. J 

I 

WALLS SUPPORTING 

LIFELINES (I.E. CABLE TRAYS, PIPING, ETC.) 

WILL IMPACT ARE THESE 

CAUSE DAMAGE LIFELINES 

TO GROUP A PART OF 

COMPONENTS? GROUP A 

SYSTEMS? 

BY SYSTEMS 

ANALYST 

WALLS WHOSE 

COLLAPSE 

COULD RESULT 

IN IMPACT 

ONTO LIFELINES 

(I.E. CABLE · 

TRAYS, PIPING, 

ETC.) 

I 

ARE THESE 

LIFELINES 

PART OF 

GROUP A 

SYSTEMS? 

BY SYSTEMS 

ANALYST 

y I N 

\SCREENED INI !SCREENED OOTI ~ I I N ~W~I L~L~I~M!"!!"'PA~C'!!!!!'IT SCREEN ED OUT 

I 

CAUSE 

DAMAGE TO 

GROUP A 

LIFELINES? 

y I N 

' !SCREENED INI JSCREENED OU] 

y N 

r 1 - - 1 

!SCREENED INj ISCREENED OUTj 

PERFORM FRAGILITY ANALYSIS 

TO CALCULATE HCLPP. 

B-32 
~· 1 I 

,I,,,··, 

DOCUMENTATION 

• Review of Design Calculation/Qualification Test 

Reports 

• Conclusions of Previous Studies 

• Walkdown D~ta Sheet 

• Walkdown Photographs 

• HCLPF Calculations· 

B-33 
OUTSTANDING ISSUES 

• Lack of Qualification Data 

• Extent of Review for Components Identified as C 

• CDFM Method Requires Site Specific Spectrum 

• Reactor Internals 

•' .. 

B-34 
ENGINEERING. PlANNING AND MANAGEMENT CONSULTANTS SHEET NO _I_I __ 

8227-09 JOB. llNl Seismic Margin Review ev DATE 

.JOSNO. ------------------ ---- ---

CLIENT _L_l_NL __ SUBJECT _.:...Pl.:..:a:.:.n:.:t~W:.:.:a:..:.l.:.::.kd=.:o:.:.:w:.:..:.n....:D:.::a:..:t.::.a~S:.:..:h:..:::e=-et=--------- CHK'D DATE---

FACILITY: Maine Yankee Atomic Power Station 

COMPONENT: ------------------------

line Number & Diameter ---------------

LOC~ION: Building • ~~~~~~~~-~---~~~~~ 

Elevation • ------------------- Room • -------------------

1.0 CONCENTRATED AREAS OF REVIEW: 

1.1 Pipe Flexibility: 

• Piping run = -----------

• Piping to equipment • ----------- • Building penetrations = -----------

Note: Provide details on building penetrations if low capacity is observed. Indicate dimensions and details in Section 4 

sheet 2. 

Photograph Roll No. __ Frame No.s __ _ 

1.2 Pipe Condition: 

• Corrosion • ----------- • Brittle connections = -----------

• Cast iron • -----------

1.3 Support Details: 

• Type of anchor point Directional 

Full restraint • --------------------- • Spans between supports = ------------ • Support anchorage details- • -----------

Photograph Roll No. __ Frame No.s __ _ 

1.4 Joints and Connections: 

• Threaded 

• Socket welded = 

= -----------------------

--------------------- B-35 Photograph Roll No. __ Frame No. s __ _ 
ENGINEERtNG. PlANNII\IG AND MANAGEMENT CONSULTANTS 

SHEET NO. _2_/ __ 

JOB NO. 8227 .. 09 JOB LLNL Seismic Margin Review BV --- DATE __ _ 

CLIENT llNl SUBJECT Plant Wa 1 kdown Data Sheet CHK'D DATE---

3.0 ADDITIONAL COMMENTS OR OBSERVATIONS: 

4.0 Sketch details of building penetrations if low capacity is observed. 

B-36 
ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS SHEET NO _I_/ __ 

8227-09 JOB LLNL Seismic Margin Review sv ___ DATE __ _ 

JOBNQ______ ---------------~---------------- . 

CLIENT __ ll_N_L_ SUBJECT _ _!.P-!.l~an~t~W~a l~k~d~ow~n!.-...!:!.:Da~t::..!:!a~S:!!!hJ.!!e~e..!c.t _______ CHK'D DATE---

FACILITY: Maine Yankee Atomic Power Station 

COMPONENT: ----------------------------

LOCATION: Building = ----~----~~~~~--~----------~ 

Elevation = -------------------

1.0 COMPONENT DATA: 

Plant ID Number = ------------------------

Manufacturer = ------------------------ Model = ------------------------ Function = ------------------------

Photograph {overall) Roll No. ____ Frame No.s ____ _ 

2.0 AREAS REQUIRING DETAILED REVIEW: 

2.1 Battery Rack Anchorage: 

Number and size of anchor bolts -= ------------

Type of anchor bolts = ------------ Description of foundation = ------------

Photograph Roll No. ___ Frame No.s -----

Note: Provide a sketch of anchorage plan with dimensions and indicate 

any foundation deficiencies observed in space provided under 

Section 4 sheet 2. 

2.2 Battery Rack Support: Type of support Overall dimensions 

Lateral Restraints 

Type of member connections 

= 

= 

= 

= 

-------------------

------------------------

--------------------

----------------

Photograph Roll No. ___ Frame No.s -----

Note: Provide a sketch of the rack indicating lateral restraints and 

dimensions in Section 5 sheet 3. 

2.3 Batteries: 

Type of batteries 

Description of ba~tery spacers 

B-37 Photograph Roll No. --- Frame No.s 
ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS 

SHEET NO. _2_/ __ 

8227·09 LLNL Seismic Margin Review DATE 

JOB NO JOB---------=----:-------- BY··--- ---

CLIENT _L_L_N_L _ SUBJECT _ _:_P..:...l a:..:.n.:.:t:._:::Wa:....:l:...::k.=.do::..:w::.:n:...._:::;Da:..:t:..=a~S::.:h.:.:e:.=e.=..t ------ CHK'D DATE---

3.0 ADDITIONAL COMMENTS OR OBSERVATIONS: 

4.0 Sketch anchorage plan with dimensions and note any foundation deficiencies 

observed. 

B-38 
ENGINEERING, PlANNING AND MANAGEMENT CONSULTANTS 

SHEET NO. _3 ...... /__ 

JOB NO 8227-09 JOB __ -=.;ll=.N:.:l:._::.;Se::...:i:..:s:.:.:.::m..:..:i c::..._:.:M:.:.ar!..igoz...:i..:.:n_:R:..:.:e::..::v_:...i.::::..ew:..:.__ _____ BY·--- DATE __ _ 

CLIENT LLNL SUBJECT Plant Wal kdown Oat a Sheet CHK'D DATE---

5.0 Sketch battery rack indicating lateral restraints and dimensions. 

B-39 
ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS SHEET NO. ...,~li.J-/ __ 

JOB NO. 8227-09 JOB __ ....!La.!.L..!!N..._L _.S!.S&e..l...lj siUimui.i.o..c _a:M111a.~..:rgw..iUJnL....DJRe::.Jivu.i ~:.~ew~~~-.-_____ BY--- DATE---

CLIENT llNl SUBJECT Plant Wal kdewn Data Sheet CHK'D DATE---

FACILITY: Maine Yankee Atomic Power Station 

COMPONENT: --__.;...-------------------------

LOCATION: Building = ~~~~~~~~~~~~~~~~~~ 

Elevation = -------------------

1.0 COMPONENT DATA: 

Plant ID Number=------------------------

Manufacturer = ------------------------ Model • ------------------------ Function • ------------------------

Photograph (overall) Roll No. ____ Frame No.s -----

( 2.0 AREAS REQUIRING DETAILED REVIEW: 

2.1 Anchorage: Number and size of anchor bolts • -.-.---------- Type of anchor bolts 

Description of foundation • ------------

Photograph Roll No. ___ Frame No.s --~--

Note: Provide a sketch of anchorage plan with dimensions and indicate 

any foundation deficiencies observed in space provided below. 

B-40 
ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS 

SHEET NO. _2_f __ 

JOBNO. 8227-09 JOB LLNL Seismic .Margin Review BY--- DATE __ _ 

CLIENT LLNL suBJECT Plant Wa 1 kdown Data Sheet CHK'D DATE---

3.0 ADDITIONAL COMMENTS OR OBSERVATIONS: 

Note any system interactions. 

l. 

B-41 
2000 Center Street Suite 418 Ber·keley. C.-'\ 04704 415-526-5111 

14 October 1986 

TO: 

FROM: 

REF: 

u.s. Nuclear Regulatory Commission 

D. Guzy, Office of Nuclear Regulatory Research 

P. Sears, Office of Nuclear Reactor Regulation 

Maine Yankee Atomic Power Plant 

D. Whittier, Manager, Nuclear Engineering and 

Licensing 

Robert J. Budnitz, Chairman of Peer Review Group for 

the Maine Yankee Seismic Margin Review Study 

MINUTES AND REPORT, SECOND PEER REVIEW GROUP MEETING 

This is the report of the second meeting of the "Peer Review 

Group" that is reviewing the technical competence of the "Seismic 

Margin Review Study" that is being undertaken on the Maine Yankee 

reactor plant under sponsorship of the u.s. Nuclear Regulatory 

Commission. 

The meeting was held at the San Francisco Airport Clarion Hotel 

on Tuesday, September· 30, 1986. Attending besides the chairman 

were all four of the other Peer Review Group members: Michael 

Bohn; John Reed; James Thomas; and Loring Wyllie. 

Also attending were representatives of the NRC; the Maine Yankee 

staff and their Yankee Atomic Electric associates; Lawrence 

Livermore ·National Laboratory. staff; LLNL' s subcontractors 

performing the review (EI International and EQE, Inc.); and R. 

Kennedy, a consultant to Maine Yankee. The sign-up sheet for the 

day 1 s session is attached as Attachment A. All attendees 

attended essentially the entire meeting. 

Attachment B shows the agenda for the day's session. The meeting 

began at 9:00 AM and ended at 6:00 PM. The order of the agenda 

was followed quite closely, although. some of the times were 

different. All agenda topics and presentations were given, and 

most of the day was used for technical discussion of aspects of 

the on-going margins review study. 

The minutes of the meeting will be presented as numerically 

ordered topics, as follows, with commentary included: 

B-42 
1) R. Murray of LLNL led the introductory session, which 

consisted mainly of a discussion concerning the future schedule 

for the review effort. The tentative schedule, as distributed by 

Murray, is shown as Attachment C. The schedule is judged to be 

satisfactory, in the sense that the project is currently 1 on 

schedule' and the participants believe that it can remain on 

schedule according to the schedule shown in Attachment c. The 

followinq specific dates have been decided on for the schedule: 

January 15, 1987: first full draft report due from 

contractors, for limited 

distribution only to Peer Review 

Group, Maine Yankee, LLNL, 2 or 3 

·NRC staff 

January 22, 1987: next meeting of Peer Review Group, 

in San Francisco; attendance by 

invitation of PRG only 

February 12. 1987: second draft due, to be distributed 

to wider distribution including 

broadly in NRC 

FeblJlary 19, 1987: meeting in Washington with NRC inhouse "Working Group on Seismic 

Margins". 

It was also decided that another Peer Review Group meetfnq would 

be held sometime after the February time period, to study the 

final version of the report to be prepared after the February 

comments are in. 

2) Daniel Guzy of NRC gave a brief summary of the July meeting 

of the NRC "Working Group on seismic Margins". He discussed the 

Working Group' s current thinking, incl udinq the selection of a 

BWR plant as the subject of a possible second trial margins 

review study. Newton Anderson of NRC, who is co-chairman of the 

NRC working Group, provided additional comments. This part of 

the meeting was mainly for information purposes and elicited very 

little discussion. 

3) M. Ravindra of EQE, Inc. gave an overview presentation of the 

work and preliminary findings of his group, who are the 

subcontractors doing the fragilities analysis of Maine Yankee. 

He discussed a list of key items that will be examined in detail, 

and another list of items that the fragilities team has already 

2 

B-43 
analyzed. This presentation was the subject of much technical 

discussion on several items of equipment, although a few key 

items were put off until more detailed discussions scheduled for 

the afternoon. Ravindra discussed various data that the utility 

had furnished to assist the fragilities evaluation, and also 

discussed his group's use of various experience data in their 

analyses. It was emphasized that the Peer Review Group will need 

access to all data that will be relied on in these evaluations, 

in order to review its applicability. 

Among the technical topics covered in Ravindra 1 s presentation 

were service water piping; heat exchanger supports; valves with 

extended operators; HVAC fans and blower supports; cable tray 

motion and a possible interaction with valve operators; cable 

trays themselves, including Maine Yankee pull tests; steel frame 

buildings (specifically the pump house); and several others. 

Maine Yankee agreed to provide the summary report to the Peer 

Review Group which presents the results of the pull tests 

conducted on concrete inserts. 

The main thrust of this part of the session was to familiarize 

the Peer Review Group with the approach being taken by the EQE 

team. 

4) David Moore of EI International next presented an overview of 

the systems analysis work that his firm is doing. He discussed 

the first walkdown, and a few tentative 'lessons learned 1 that 

may assist others in preparing for a first walkdown, such as 

doing more detailed preparatory study of HVAC and actuation 

systems. He emphasized that the first round of systems analysis 

is well under way, with seven fault trees complete and a few more 

under development. His group is now going back to answer 

specific questions that have arisen, in preparation for the 

second walkdown. The data sources being used were discussed, and 

it was again emphasized by the Peer Review Group that access to 

those data sources will be needed. Also, the method of combining 

seismic and non-seismic failures was covered. 

Moore's group will be doing some system cut sets soon, to obtain 

early guidance for EQE on what items seem to be more important. 

However, they do not plan to do sequence cut sets until after the 

second walkdown, as is called for in the NUREG/CR-4334 and 4482 

guidance. 

There was discussion of certain specific items, and the Peer 

Review Group was provided with some detailed fault trees for 

their study. 

3 

B-44 
5) Robert Kassawara of EPRI next gave a presentation covering 

EPRI's on-going seismic margins review project at Catawba. This 

review effort is on a schedule not very different from the 

schedule for NRC's review at Maine Yankee: initial report due 

early in 1987, final report due a few months later. 

While this presentation was mainly intended for information, it 

generated significant discussion concerning the differences 

between EPRI's and NRC's approaches. EPRI 1 s approach is emphasizing the CDFM method for HCLPF determinations, while NRC's 

approach has not yet settled on one or another method, although 

it favors the CDFM method if it can be studied enough. EPRI's 

approach also uses a 'success path' method for the systems 

analysis rather than a fault-tree;event-tree method. There was 

much discussion about how to cope with small LOCAs inside 

containment, which perhaps cannot be walked down and analyzed 

well. EPRI 's analysis at Catawba is studying some re·lay chatter, 

but their analysis is limited to a very few relays on their 

chosen success path; NRC's approach is not considering relay 

chatter for the time being, because further research is needed. 

The general flavor of this discussion was that both the EPRI 

review at Catawba and NRC's review at Maine Yankee will be able 

to learn much from each other, and continuing cooperation and 

coordination will be encouraged. 

6) Philip Hashimoto of EOE, Inc. gave a long and detailed 

presentation about three specific technical items that EQE is 

analyzing: the RWST, the pump house, and one block wall. He 

presented both CDFM and fragilities analyses of HCLPF values, and 

his presentation provided a vehicle for extensive discussion 

about the methodology used, the data relied on, approximations 

introduced, and whether EQE's analytical approaches were generic 

or only specific to the item being analyzed. 

It was pointed out in the discussion that in the analysis 

approach used for the RWST, the deflection compatibility between 

anchor bolt and water resistance modes is not provided. This 

needs to be carefully investigated before the capacities from 

these two modes are combined. 

In the course of this discussion, much detail was covered that 

will not be discussed here. The principal thrust of the Peer 

Review Group's comments were that there is a need for careful 

study of both the CDFM and fragilities methods, since the 

• results' for HCLPF seem to depend greatly on assumptions made 

and data chosen. One key aspect was the Peer Review Group • s 

observation that experience data, if available, play a key part 

of the underlying approach to HCLPF analysis. It was requested 

that realistic resistance modes be analyzed rather than using 

4 

B-45 
design assumptions which are often overly conservative and do not 

represent the realistic response. 

7) Gregory Hardy of EOE. Inc. gave a detailed presentation about 

two other items being analyzed, the diesel day tank and one 

inverter located in the switchgear room. This discussion 

followed the same tenor as the discussion of Hashimoto 1 s work 

just earlier. There was extensive interaction between the Peer 

Review Group and the analyst team, and much detailed discussion 

of data bases and assumptions,. Again, discussion of experience 

data played a key role. 

8) Greg Hardy then gave a briefer presentation, for information 

purposes only, about progress in tracking down fragilities 

information on three key issues: the lead-antimony batteries, 

the reactor internals and CRDMs, and the fire water threaded ~- In the battery case, he has been unsuccessful so far in 

identifying relevant test or experience data. Therefore, it is 

the Peer Review Group's understanding that analysis will be 

performed with and without these batteries. For the other two 

cases, the information now in hand, either due to configuration 

or fragility aspects, should be adequate for the purposes of this 

margins analysis. Concerning the internals and CRDMs, the 

comment was made that Combustion Engineering • s cooperation has 

been outstanding. 

9) David Moore of EI, in reply to Peer Review Group questions, 

discussed how his analysis group is coping with Group B 

functions, especially those items needed to support long-term 

heat removal. He provided a rationale for his approach, which 

will be the subject of later review by the Peer Review Group. He 

then discussed in more detail the treatment of non-seismic 

failures, with an emphasis on those that might compromise both 

redundant trains of some function. The question of how to 

combine these failures with seismic-induced failures in a HCLPF 

analysis was covered, but not resolved. What approach to take 

remains an open question, and the NRC Expert Panel's guidance on 

this subject is inadequate. 

The Peer Review Group discussed a problem with incorporating nonseismic failures after an earthquake, if their incorporation can 

change the HCLPF level found. The problem arises if the way they 

affect post-earthquake plant response at the SSE level is 

identical to the way they affect plant response near the HCLPF 

level or near the margin-review-earthquake level; in this case, 

their actual effect on 1 plant seismic margin • is minor. This 

issue will require more discussion in the future. 

5 

B-46 
10) There was further discussion of how the comparison between 

CDFM and fragilities methods for HCLPF determination will be 

accomplished. It is recognized that the EQE effort in this 

project will not extend to CDFM-fragilities methodological 

comparisons beyond those few presented to the Peer Review Group 

at this meeting. There is a continuing need for a more thorough 

way to address this issue, which the Peer Review Group will 

undoubtedly comment on further later on. It is likely that the 

NRC Expert Panel .may need to give this issue much more careful 

thought after this Maine Yankee review has been completed. 

11) John Reed asked a question of the NRC staff whose answer was 

not f\llly covered, since the NRC staff hazard experts were not 

present. The question concerned exactly what was intended by NRC 

staff in their choice of the margin review earthquake being used 

in the Maine Yankee analysis. After some discussion, it was 

decided that it would be assumed that the median NUREG/CR-0098 

spectrum anchored to o. 30 g ZPA represented a uniform hazard 

spectrum at the 84 % confidence level. NRC staff will attempt to 

obtain some clarification on this issue soon. 

12) James Thomas raised a general question of how systems 

interactions aspects should be dealt with in the analysis, such 

as when a valve may hit a support during strong earthquake 

motion. Some generic guidance is needed to provide analysts with 

an acceptable approach, which should" be usable in most cases 

without doing a detailed and expensive calculation. This aspect 

was not resolved in the meeting, but will be given· further 

thought. 

OYERVIEW COMMENT: 

It is the broad consensus of the Peer Review Group that the Maine 

Yankee t~l margins review project is being accomplished so far 

with acceptable technical competence. In the course of its 

review, the Review Group discussed several technical issues tha·t 

have been subjects of continuing difficulty in the analysis, and 

in some cases these difficulties may not be resolved during this 

trial margins review project. The Peer Review Group recognizes 

that one key objective of this trial review is to uncover such 

issues, especially methodological issues or issues of inadequate 

data. Its overview comment is that the analysis team, including 

6 

B-47 
LLNL, EI International, and EQE experts, is carrying out a fully 

acceptable analysis so far, within the constraints of the project 

scope and subject to the comments in the detailed discussion 

above. 

The Peer Review Group looks forward to further interactions as 

the analysis proceeds through a second walkdown and then to final 

analysis and documentation of the project results. 

That completes this report and minutes. 

/)~A£-!/1 / 

Robert J. BUdnitz 

Chairman, Peer Review Group 

7 

B-48 
9/'i,o(8'b 

·pc;.ei2 i2C5VIf_u/ ~1TG. 

?o S r ·-;-,o ,-J 

L1 SIJ i2cjN i~~'i!./PD 8 

A) rt-G/fl t:. J 

'Sa"' dt;7, ,J:.c. t{ s 

e-q,z: . E. 1.. ~ ~,... r:~ 

~E-J:.-,.(..,. 

E Q E" l V\-C.. 

LLYVL 

EPit\ 

D~.~<a. f OLLJ<J...r 

E!['l?T 

Jfr~r:f. Ar>rh. tDtv.//. (Vt~e [a.~~<:~ ~'"4~ (},~ 

YA"')c.u.. A-M""\·... t:.l c. c.~"~ Co 

f. l,j~ -:::,.J ( 

LLAJL 

IJ \J... C ... 

,, 

B-49 

ATTACHMENT A 

3ot- t/-12- }too'-. 

3u/ -f'f?- 7~ 9 ~ 

:;z,~- 2'1'6 -~'12'? 

-<;/J.,- <;/ 9 ~-- ~ ~ ZI'Z:> 

].. t?t; jt6s-'l-t'Jc:J ~0 

11'\- ~52..- qz.'}'1 

/14 .... 8s -z- 9 29.9 

..ttt(-- ~ad-..- o3o s 

.it Ll" i.fS ... l..11S 

?D 'f 3/3 - 1,16/ ;;:).,_ 

4-1 S c ·"-_;::., '} S :_,· lr. 

7/t/, 77?. 2/c.J 

;.tJ 7 - t;;z.3 - .!.S 2..( 

c..•, -8-.1..-9100 

1/ 
ATTACHMENT B 

AGENDA 

Seismic Margins Program 

Peer Review Group Meeting 

San Francisco Airport Clarion Hotel 

September 30, 1986 

~:00 a.m. Introductions Budnitz/Murray 

9:10a.m. Briefing on NRC Working Group Meeting Guzy I Anderson 

9:30a.m. Walkdown Summary Ravindra 

10:00 a.m. Systems Summary Moore 

11:00 a.m. Briefing on EPRI Program Kassawara 

12:00 noon Lunch 

1:00 p.m. Fragility Summary 

HCLPF Comparisons Hardy /Hashimoto 

Tank Methodology Hashimoto 

3:00p.m. Important Open Issues 

Batteries Hardy 

Internals/CRDM Hardy 

Threaded Fire Water Piping Hardy 

5:00p.m. Report Outline Murray 

Second Plant W alkdown 

Schedule 

6:00p.m. Adjourn 

' .. 

B-50 
September 23, 1986 

September 30, 1986 

October 1, 1986 

October 22, 1986 

November 10, 1986 

November 17-19, 1986 

November 18-19, 1986 

December 8, 1986 

December 10-12, 1986 

...... January 1987 

-February 1987 

ATTACHMENT .C 

SEISMIC MARGINS PROGRAM 

Tentative SChedule 

Federal Express mailing to Peer Review Group 

Peer Review Group meeting at San Francisco Airport Clarion 

Hotel 

Analysis Team meeting at San Francisco Airport Clarion 

Hotel 

Analysis Team meeting at Energy Incorporated, Kent, 

Washington 

Second pre-walkdown status summary Federal Express 

package 

Analysis Team second plant walkdown at Maine Yankee. 

Peer Review Group meeting and additional walkdown at 

Maine Yankee 

Preliminary plant HCLPF available 

Possible exchange meeting with EPRI and Analysis Team 

meeting (most participants will be at the Symposium on 

Current Issues Related to Nuclear Power Plant Structures, 

Equipment and Piping to be held at North Carolina State, 

December 10-12, 1986). 

Briefing at the NRC on program and results 

Draft (ina! report submitted to the NRC 

B-51 
Future RuourceJ A-Moci.ateJ, lne. 

5 December 1986 

TO: 

FROM: 

REF: 

u.s. Nuclear Regulatory Commission 

p. Guzy, Office of Nuclear Regulatory Research 

P. Sears, Office of Nuclear Reactor Regulation 

Maine Yankee Atomic Power Plant 

D. Whittier, Manager, Nuclear Engineering and 

Licensing 

Robert J. Budnitz, Chairman of Peer Review Group for 

the Maine Yankee Seismic Margin Review Study 

MINUTES AND REPORT, THIRD PEER REVIEW GROUP MEETING 

This is the report of the third meeting of the "Peer Review 

Group" that is reviewing the technical competence of the "Seismic 

Margin Review Study" that is being undertaken on the Maine Yankee 

reactor plant under sponsorship of the U.s. Nuclear Regulatory 

Commission. 

The meeting was held at the Maine Yankee site on TUesday and 

Wednesday. November 18-19, 1986. Attending besides the chairman 

were all four of the other Peer Review Group members: Michael 

Bohn; John Reed; James Thomas; and Loring Wyllie. 

Also attending were representatives of the NRC; the Maine Yankee 

staff and their Yankee Atomic Electric associates; Lawrence 

Livermore National Laboratory staff; LLNL's subcontractors performing the review (EI International and EQE, Inc.); and R. 

Kennedy, a consultant to Maine Yankee. The sign-up sheet for the 

first day's session is attached as Attachment A. All attendees 

attended essentially the entire meeting. D. Whittier of Maine 

Yankee, whose name is not on the sign-up sheet, attended the 

second day's session. 

Attachment B shows the agenda for the meeting, which lasted all 

day Tuesday and half of Wednesday. The Tuesday meeting began at 

9:00 AM and ended at 5:30 PM. Wednesday's meeting began at 8:30 

AM and ended about noon. The order of the agenda was followed 

closely, although some of the times were different. All agenda 

topics and presentations were given. Most of the meeting time 

B-52 
·'. 

' . ' 

was used for technical discussion of aspects of the on-going 

margins review study. In addition, on Tuesday afternoon all 

attendees participated in a walk-down of the plant to study 

particular aspects of the design relevant to its seismic margin. 

The minutes of the meeting will be presented as numerically 

ordered topics, as follows, with commentary included: 

1) R. Murray of LLNL led the introductory session, which consisted mainly of a discussion concerning the future schedule for 

the review effort. The schedule is judged to be satisfactory, in 

the sense that the project is currently •on schedule' and the 

participants believe that it can remain on schedule. The following specific dates have been decided on for the schedule: 

December 11, 1986: meeting between NRC margins team and 

EPRI margins study team at Raleigh, NC 

January 15, 1987: first full draft report due from 

contractors, for limited distribution only to Peer Review Group, 

Maine Yankee, LLNL, 2 or 3 NRC 

staff 

January 22, 1987: next meeting of Peer Review Group, 

in San Francisco; attendance by 

invitation of PRG only 

February 12, 1987: second draft due, to be distributed 

to wider distribution including 

broadly in NRC 

February 19. 1987: meeting in Washington with NRC inhouse "Working Group on Seismic 

Margins". 

Another Peer Review Group meeting is planned for sometime after 

the February meeting, to study the final version of the report to 

be prepared after the February comments are in. 

Members of the Peer Review Group requested permission to visit 

the offices of the EQE and EI subcontractors toward the end of 

the current phase of the project, in order to review ongoing work 

prior to the completion of the subcontractors' reports. It was 

agreed that a visit to EI could occur any time after early December, but that visits to EQE should not be held until early January. A meeting on the fragilities is scheduled for January 6 at 

2 

B-53 
EQE. The Peer Review Group wishes to emphasize the importance of 

obtaining access to various analyses and calculations as early as 

is feasible, to allow for optimum opportunity for review. 

2) M. Ravindra of EOE, Inc. gave an overview presentation of the 

work of his group, who are the subcontractors doing the fragilities analysis of Maine Yankee. He presented slides that included 

preliminary tables on HCLPF values for various components and 

structures, which elicited much discussion from the Peer Review 

Group. The question of how best to document and present these 

results was covered in some detail. 

The issue of .how to treat dependencies was discussed, especially 

in regards to the CDFM approach to determining HCLPF values. 

(For the fragilities approach, the methodology for treating 

dependencies is more straightforward.) 

Discussion about how to standardize the CDFM approach was lively. 

It was agreed that th~s issue, including comparison of the CDFM 

and FA approaches, would be a major topic in the Peer Review 

Group's final review work for the Maine Yankee project. 

Some of this discussion covered specific items of equipment at 

Maine Yankee, which the Peer Review Group agreed would be concentrated on during their afternoon walkdown of the plant. 

It was emphasized that the Peer Review Group will need access to 

all data that will be relied on in these evaluations, in order to 

review its applicability. This especially includes various 

aspects of the earthquake experience data base. In this regard, 

G. Hardy of EQE agreed to send the recent SQUG report on about 20 

classes of equipment to J. Reed. 

3) David Moore of EI International next presented an overview of 

the systems analysis work that his firm is doing. He discussed 

the preliminary system 'min cut sets• that have been developed, 

including a few interesting failure modes that are still being 

investigated in detail. 

Specific technical topics covered in Moore's presentation included load sequencing, starting the turbine-driven emergency 

feedwater pump after an earthquake, and the reactivity-control 

systems. 

The group discussed EI 's process for pruning the trees during 

their development. Moore agreed that his team would publish the 

entire un-pruned trees, for later use, along with the pruned 

trees that will be used in the quantification process. 

3 

B-54 
An extensive discussion occurred about how to combine non-seismic-induced failures with seismic-induced failures in the systems analysis. This discussion covered many aspects of this issue. 

The guidance given to the analysis team is as follows: 

The systems analysis team should identify all 

non-seismic failures that, in combination with a 

seismic failure, produce the undesired end-point 

(core damage) with a frequency above a certain 

cutoff. (The cutoff that EI is now using will 

include non-seismic failures whose contingent failure probability is above about 0.01 for 

failure of a single train, and above about 0.001 

for failure of two trains or of a full safety function.) The key seismic contributors thus 

identified will then require determination of 

HCLPF values by the fragilities team. However, 

the analysis will DQt quantify an overall HCLPF 

plant value that includes these non-seismic 

failures. Instead, the combinations of seismic 

failures and non-seismic failures thus identified 

will be documented and discussed, so that decision-makers can be aware of their existence and 

the HCLPF values associated with the seismic 

failures identified. This documentation should 

be done in such a fashion that a complete HCLPF 

re-evaluation that includes the non-seismic 

failures is possible. 

The EI group's main work in recent weeks has been development of 

the system fault trees, which were distributed in preliminary form to the Peer Review Group. EI's main work in the coming period will be to finalize these, based in part of the new information being gathered at the current walkdown. 

4) An extensive discussion took place about how to analyze Maine 

Yankee's batteries. The issue came up because the fragilities 

team has been unable to develop any defensible technical basis 

for calculating fragilities for the current lead-antimony batteries. The MY plant will be changing out 2 of the 4 battery racks during their upcoming shutdown in March, 1987, and may change out the other 2 battery racks at a later shutdown in a 

future year. Therefore, the question arose as to which configu- ration should be analyzed in this study. The Peer Review Group's guidance is that the study team should analyze both the configu- ration in which only 2 of the 4 racks will be changed, and the 

ultimate configuration in which all 4 will be changed. If the 

4 

B-55 
technical basis for developing fragilities for the existing batteries cannot be developed, this fact should be stated in the 

report. 

5) Daniel Guzy of NRC gave a brief summary of the meeting the 

previous week of the NRC "Working Group on seismic Margins". He 

discussed the Working Group's current thinking, including the 

selection of a BWR plant as the subject of a possible second 

trial margins review study. The NRC's current thinking is that 

they may decide to perform the planned BWR trial margin review in 

full collaboratipn with EPRI (same plant, same study). 

6) The topic of how the margin review earthquake was selected by 

NRC was then discussed. John Reed agreed to draft a paragraph for these minutes on this subject, which paragraph is as follows: 

Dan Guzy discussed the question of what confidence and 

probability level the margin earthquake represented (i.e., NUREG-0098 rock median spectrum anchored to 0.3g peak ground acceleration). It was explained by Guzy that 

the NRC did not use probability concepts to establish the 

margin earthquake requirements, but rather adopted the 

NUREG-0098 spectrum shape to be consistent with a 

previous licensing decision made for Maine Yankee. John 

Reed requested as a minimum that the NRC agree that the 

margin earthquake input represents a uniform hazard 

spectrum over the entire frequency range. This 

understanding is consistent with the fragility method 

being used by the analysis team in developing HCLPF 

values for structures and equipment. In addition, it 

allows the NRC to make confidence and probability 

statements in the future after the margin analysis is 

completed, if they so desire. 

7) The Peer Review Group went on an extensive walkdown of the 

plant on the afternoon of Tuesday, November 18. Items walked 

down were .those that were of special interest to the reviewers, 

based on either their importance to the Maine Yankee margins 

analysis or their intrinsic methodological interest. .After this 

walkdown, the Peer Reviewers and the analysis team reconvened to 

discuss what was observed. Interaction with the Maine Yankee 

staff and their consultant (R. Kennedy) was important during this 

session. 

5 

B-56 
._.,. 

QVERVIEW COMKENTS: 

1) The main overview comment here is identical to that made 

after the last Peer Review Group meeting. Specifically, it is 

the broad consensus of the Peer Review Group that the Maine 

Yankee trial margins reyiew project is being accomplished so far 

with acceptable technical competence. Numerous technical and 

methodological issues have come up, and all have been coped with 

well by the analysis team. The cooperation of Maine Yankee and 

Yankee Atomic personnel has been excellent so far. 

2) Although PRA-type information is being developed during the 

trial margins review, the Peer Review Group wishes to emphasize 

that it is probably not usable for full-scope PRA analysis per se 

without significant additional effort. 

3) The Peer Review Group also wishes to emphasize that it cannot 

be expected to endorse the results of the trial Maine Yankee 

review. If all goes well, the Peer Review Group should find 

itself able to endorse the methodology used, and should also be 

able to comment on some specific technical aspects of the study. 

However, an endorsement of the study results would require a 

level of review effort much greater than can be expected from our 

PRG. 

To summarize, the Peer Review Group looks forward to further 

interactions as the project proceeds to final analysis and documentation of the project results. 

That completes this report and minutes. 

Robe'!::ut:f 

Chairman, Peer Review Group 

Attachments: A = sign-up sheet 

B = meeting agenda 

6 

B-57 
.· ' ,! 

A fTltJIINCE) J'f ~N -u. ~ JI-IUJ 

AI N/Jii,Jl y/AJJ<ll 

~ LLNL 

I MtC.C,~~.el f 5atu.. $fa .....lui tV..,. (p J~ /, 

' . 

1/~t/n. " 

/lUII/IJ;IK 

(ftJj)2. 6" -J/ 1/ - . ( ~~~-) 4:22- 4 2-4:!1-

(;sl;s-) f'~6, - -'It:; 2 ~ 

,- --

: ---L..::.M....::..::·.....:..k...::;_:__. R~o...:....;:;..v.;..:_, "Yl_d.....;_nv;.,..__ ___ _.:;:..E......;;;]QL-...;€:.;;;...__ _______ 6..:::...;.._1-=4J:___.;_e_s ;;t - C? ~.tt 

i~-----"'J)~a-"'_._G_o_1..7~----·- tv~~----· 3o/ 'ftJ- ?·K'q '7_ ______ _ 

~1 ~::r~ ·~ /)/- ~'t'l- 7_2.-L.Q ______ _ 

_ ').-17- fiZ..3- 3~/~ 

G.,.,- B1~- CO\~~--

7/c}- 772-li&L _____ _ 

41) '<61--fL 

tfl:)- 3 '12-l.oCfS _L_ 

1117- 27il-f'ltJo 

2tJt::,/ 'i~'t-t'tJ~-- 1 

?ct,t- 373- YiJJ st, 

~ _ ___l:j]L£~H c . cJ./oksH/ !tit?.' jt/fl{J... ________ .. ..EJ_~-7 __ 

-~.___:_~..:...:..;O\'C:....::...:.~_,__ ______ ---'6:::rJ..Q...A..:£ ....... _.______ 11~- F-5"~ -q;J~-------

~-~c..l\~i..L l. (:,g,~~\~ .£,0 £, -···---- ___ ]__..\=i_j)EX-~~-. _ 

1l4 <i.S.2._~aa_ ---

- .. ____________ . ·-·--

8-58 -· ---· .. ··-··-----

-------- ----·----
AGEN-DA--

Peer Review Group Meeting 

November 18-19, 1986 

Maine Yankee 

Tuesday, November 18, 1986 

9:00 a.m. 

9:15 a.m. 

10:00 a.m. 

10:45 a.m. 

11:00 a.m. 

12:00 noon 

1:00 p.m. 

5:00p.m. 

Introduction 

Status of Fragility Evaluation 

Status of Systems Analysis 

Open Items 

Peer Review Group Discussion 

Lunch 

Plant Tours and Discussions with Plant Personnel 

Adjourn 

Wednesday, November 19z 1986 

8:00a.m. Plant Tours and Discussions 

10:00 a.m. Peer Review Group Wrap-up Meeting 

12:00 noon Adjourn 

B-59 
Future ReJourceJ AttMociateJ, Inc. 

2000 Center Street Suite 418 Berkeley, CA 94704 415-526-5111 

27 February 1987 

TO: U.S. Nuclear Regulatory Commission 

FROM: 

REF: 

D. Guzy, Office of Nuclear Regulatory Research 

P. Sears, Office of Nuclear Reactor Regulation 

Maine Yankee Atomic Power Plant 

D. Whittier, Manager, Nuclear Engineering and Licensing 

Robert J. Budnitz, Chairman of the Peer Review Group 

FINAL REPORT: PEER REVIEW GROUP for the MAINE YANKEE 

TRIAL REVIEW OF THE NRC SEISMIC MARGINS METHODOLOGY 

This is the final report of the "Peer Review Group" that is reviewing the 

technical competence of the "Seismic Margin Review Study11 that is being 

undertaken on the Maine Yankee reactor plant under sponsorship of the U.S. 

Nuclear Regulatory Commission. 

The Peer Review Group has five members: 

Robert J. Budnitz (chairman) 

Michael P. Bohn 

John W. Reed 

James Thomas 

Loring Wyllie, Jr. 

The PRG is pleased to report that the five members are all in agreement with 

this report, and all endorse it fully. 

B-60 
I. Peer Review Group Charter 

At its first meeting on July 21, 1986 the Peer Review Group adopted the 

following charter, which was also agreed to by all project participants, 

including NRC and the utility participants: 

The objective of the Peer Review Group is to assure that the trial 

seismic margins review, following the guidance established in NUREG/CR4334 and NUREG/CR-4482, is executed in a fully competent and professional manner, uses methods that are at the state-of -the-art, and takes 

cognizance of all relevant information. The sponsors of the study 

(Lawrence Livermore National Laboratory for the NRC and Maine Yankee 

as the plant owner) desire to utilize the results of this study, and require 

the Peer Review Group's assurance that the study is technically sound. 

To accomplish its objective, the Peer Review Group will be provided full 

access to all materials, information, and methodologies that are inputs to 

and used by the study team. Access to the study team itself will occur 

through scheduled meetings to follow the study's progress. The Peer 

Review Group will also review draft reports and participate in walkdowns 

of the plant. A formal report for the Peer Review Group will be made 

by the Group's chairman to NRC and Maine Yankee. It is understood 

that the Peer Review Group's report will be a public document. 

II. How the Review Was Accomplished 

The Peer Review Group met on five occasions during the course of this 

project. Attendance was perfect except that at the first meeting one member 

could not attend and another's attendance was cut short. The five meetings 

were: 

1. July 21-22-23, 1986, at the Maine Yankee site. During this meeting 

the PRG had the opportunity to walk down the plant, and to meet 

with all project participants. During this meeting the PRG's charter 

was discussed and clarified. 

2. September 30, 1986, near San Francisco Airport. At this meeting the 

PRG received presentations and documentation from the project 

participants, and asked questions to clarify technical issues. 

3. November 18-19, 1986, at the Maine Yankee site. During this meeting 

the PRG had a second opportunity to walk down the plant, to meet 

with all project participants, to review documentation, and to ask 

questions. 

4. January 22, 1987, at Loring Wyllie's office in San Francisco. During 

this meeting the PRG met alone with no other attendees, discussed its 

conclusions about the study, discussed an early draft version of the 

2 

B-61 
project report, formulated verbal comments on this early draft that 

were then relayed to the project team, and laid out the work needed 

for the PRG to arrive at its final conclusions. 

5. February 19, 1987, at NRC's Bethesda offices. The PRG met alone 

briefly in the morning, spent the main part of this day in a large 

meeting with the entire team of project participants as well as NRC 

and utility representatives, and then met again at the end of the day 

to finalize its comments as embodied in this letter. 

Besides these meetings, each of the PRG members had access to considerable 

information during the course of the project. Technical material was 

distributed at each of the first three meetings by the project participants 

(including material from Lawrence Livermore National Laboratory, from their 

subcontractors at Energy Incorporated and EQE Incorporated, and from the 

utility). This material enabled the PRG to remain fully informed and up-todate on project progress, technical problems encountered, and proposed 

resolutions of various issues. 

In addition, PRG members met individually on a few occasions with project 

participants to allow interaction on technical issues, and of course the 

participation of PRG members in both walkdowns of Maine Yankee provided 

hands-on interactions with the project. 

Copies of the first version of the project's draft final report were distributed 

to the PRG only six days prior to its fourth meeting. About a month later, 

copies of the second version of the final report were distributed to the PRG 

only 3 to 6 days prior to its fifth and final meeting. In both cases, there 

was not enough time for the PRG to undertake as thorough a review as would 

have been possible if more time had been available. However, the PRG 

believes that there was enough time to allow the PRG to reach conclusions on 

each subject in its charter. This is primarily due to the opportunity for PRG 

interaction with the project team extensively throughout the project, including in the last several weeks by telephone, personal interaction, and study of 

preliminary rna terial. 

III. PRG Findings 

The PRG charter requires findings on several topics. These findings form the 

substance of our review, and are presented here through discussion of the 

following five topics: 

1) The PRG's access to technical information and to the project team been 

fully adequate to allow it to reach useful conclusions on each topic in the 

charter. The PRG members believe that their interaction with the project 

team has been significantly more extensive than is typical of reviews of this 

kind. Access to technical information has been complete and timely, within 

the obvious constraints of a very tight schedule toward the end of the 

project. 

3 

B-62 
2) The project team followed the guidance established in NUREG/CR-4334 

and NUREG/CRw4482 as fully as could be expected. In several areas, the 

guidance provided in the two NUREG reports was found to be incomplete or 

inadequate, and in each case the PRG believes that the project team successfully overcame the specific issues involved. We are grateful that the PRG 

was consulted on the major issues, a few of which are discussed below. 

3) The project team executed the study in a fully competent and professional 

manner. The PRG believes that the professional competence of the project 

team is outstanding, and that the execution of the project has also been 

outstanding. This is all the more remarkable when one considers the various 

constraints on the project (such as the fact that this has been a research 

project which was unfortunately carried out as an NRC licensing action, 

which led to the problem that various uncertain researchable questions took 

on a licensing and therefore a financial aspect). 

There were numerous issues within the project about which professional 

disagreements occurred between various participants, or between participants 

and the PRG. All were resolved in a fully professional manner, with open 

discussion and honest effort to find the best approach. 

4) The project team used state-of-the-art methods. In fact, during the 

project a few advances were made in the state-of -the-art. Among them were 

the approach taken to combining seismic and non-seismic failures in the 

systems analysis; the issue of how to reconcile the CDFM and fragilities 

approaches to calculating the HCLPF values of components; and the way the 

systemsNanalysis team was able to use pruned fault trees to develop PRAwtype 

systems cut sets with considerable savings of effort. 

5) The proiect team seems to have taken cognizance of all relevant information. In fact, much information not previously published, or available only in 

draft or incomplete form, was used by both the systems analysts and the 

fragilities analysts. Information from the Maine Yankee and Yankee Atomic 

Electric contacts was well used. 

IV. Other PRG Comments 

A number of technical issues arose in the course of the project that deserve 

special comment by the Peer Review Group. These topics are the following, 

discussed in the next paragraphs: 

4 

B-63 
. ... ' ·~:' 

A. selection of the review level earthquake 

B. combining seismic and non-seismic failures 

C. CDFM method vs. fragilities method for HCLPF analysis 

D. level of expertise required by a margins review team 

E. earthquake experience and test data base 

F. treatment of relay chatter 

G. correlations among earthquake-induced failures 

H. comments on the screening methods used. 

A. Review level earthquake: Throughout the Maine Yankee trial review, 

there has been confusion as to the precise interpretation of the review-level 

earthquake. The PRG believes that the confusion has now been cleared up 

(in large part through PRG interaction with the project team and with NRC), 

and that the analysis performed for Maine Yankee has correctly utilized a 

consistent review-level earthquake. However, the PRG believes it essential 

that more explicit guidance be provided as to what technical features must be 

established for the review level earthquake, and how it is to be used. 

(Presumably, the provision of this explicit guidance should be a task for the 

NRC Expert Panel, after interaction with the NRC staff and after seeking 

advice from industry representatives.) A second comment on this subject, 

which has been given verbally to the study team, is that the write-up in the 

draft report version we have studied in mid-February is still inadequate in its 

description of both how the review level earthquake for Maine Yankee is 

characterized and how it was used. 

B. Combining seismic and non-seismic failures. The methodology used in the 

Maine Yankee review for combining seismic and non-seismic failures is not a 

rigorous methodology, in the sense that the final plant-level HCLPF value for 

Maine Yankee depends on factors that are not completely specified in the 

guidance. For the Maine Yankee review project, the final HCLPF values for 

both LOCA and non-LOCA cases have essentially no dependence on this 

inconsistency, but in a general case there can be differences. The PRG urges 

that the methodology for this aspect of seismic margins reviews be developed 

fully (presumably, again, by the NRC Expert Panel) and documented for 

general use. It is important that the way NRC will use the results of a 

margins review be explicitly considered in developing guidance on this issue. 

It is also important in this regard to note that there needs to be clarification 

of how a "seismic margin" is to be interpreted in terms of being greater than 

the SSE level, especially if non-seismic failures affect the HCLPF value but, 

of course, do not affect the SSE level to which the HCLPF value could in 

some minds be compared . 

C. CDFM vs. fragilities method for HCLPF analysis. The PRG endorses the 

recommendation in the Maine Yankee report that further research is needed 

to develop the CDFM (conservative deterministic failure method) approach to 

analyzing HCLPF values, so that it can be applied routinely. We have learned 

much from the comparisons in the current project between HCLPF values 

calculated for a few components by both the CDFM and fragilities approaches. 

In particular, we have learned that the CDFM method as currently set down 

has too much latitude for the analyst and therefore does not provide robust 

HCLPF values. We urge that this situation be resolved by further research. 

5 

B-64 
D. Level of expertise needed. Based on our interactions with the Maine 

Yankee trial review, the PRG beUeves that a review team must have certain 

attributes in order to carry out a successful seismic margins review. A utility 

whose choice of a review team does not take these attributes into account 

will be less than fully successful in applying the methodology. The main 

attribute for the fragilities analysis team is that it must have experience in 

plant walkdowns, in how to focus on the critical components, and in realistically analyzing seismic capacities; otherwise the margins analysis will be 

burdened by unnecessarily conservative results beyond those already embodied 

in the HCLPF approa'ch. The systems analysis team must be familiar with 

PRA methods, and will be substantially strengthened by prior experience in 

PRA analysis of external initiators. Of course, it is not necessary for the 

analysis team to possess expertise at the level that was needed to develop the 

margins methodology in the first place. 

E. Earthquake experience and test data base. In the current Maine Yankee 

review, considerable reliance has been placed on the earthquake experience 

and test data base for various components. The draft report we have 

reviewed contains citations of much of this data base. Even though some 

members of the review team are among those who have developed the data 

base, its interpretation for the purposes of this study has been problema tical 

in a few situations because some of the data base is not yet publicly a vailable. For other analysis teams in the future, this issue will be also exist. 

The ref ore, it is essen tia 1 in future studies that the data base relied on be 

thoroughly documented, and that its interpretation by the analysis team be 

set down in detail in future written reports. Guidance to future analysts 

along these lines must be strengthened (presumably, again, by the Expert 

Panel) to assure that future reviews are carried out acceptably. 

In addition to the documentation aspect, it is important that the guidance be 

more explicit regarding the type and configuration details for components 

covered by the screening guidance, because some analysts in future routine 

applications may not be as familiar with the underlying raw experience data 

as were the analysts performing this Maine Yankee trial review. 

F. Treatment of relav chatter. Following the guidance of NUREG/CR-4334, 

the issue of relay chatter and its recovery by operating crews has not been 

covered in the Maine Yankee review. The PRG wishes to point out that 

considerable progress in understanding relay chatter has recently occurred, 

and that the guidance for future analysts may be able to be strengthened in 

this aspect. (This is presumably another task for the NRC Expert Panel, with 

input from electrical circuitry and systems analysis experts). 

G. Correlations among earthquake-induced failures. The handling of 

correlations among earthquake-induced failures is a methodological problem. 

These correlations can exist in several areas, including the fragilities, the 

responses, and the systems aspects. Although there are methodological 

weaknesses in the approach that was used, they do not affect the plant-level 

HCLPF results at Maine Yankee, and the review team's treatment is therefore 

acceptable. However, there remains a general methodological question as to 

how these correlations are to be treated, which the current guidance in 

NUREG/CR-4334 and 4482 does not cover well. The problem is that, while 

6 

B-65 
there are exact methods for handling these correlations in a full-scope and 

realistic seismic PRA study, their treatment in the approximate HCLPF 

analysis requires some explicit guidance not yet a vail able. This area is in 

need of methodological development before the HCLPF methodology can be 

considered fully acceptable. In addition, the guidance needs to be more 

explicit concerning when correlations should be assumed to be I 00 % --- and 

hence treated implicitly in the basic event definitions in the fault trees·--

and when they should be treated explicitly in the numerical evaluations of the 

cut sets. 

H. Comments on the screening methods used. The PRG has carefully 

studied the methods used by the Maine Yankee analysis team in its several 

screening tasks during the study. (This has been one of the most important 

reasons why PRG interaction during this trial review has been so _valuable.) 

The approach taken, which employed an iterative process between systemsbased screening and fragility-based screening, seems to have been very 

successful in reducing the amount of analysis work required while retaining 

the key items for final study. The PRG wishes to affirm its endorsement of 

this iterative and interactive approach. It also wishes to endorse the original 

guidance (in NUREG/CR-4334 and 4482) on how important a walkdown-type 

screening step is even for those components that are thought be screenable 

out based on generic guidance in the NUREGs. Only by such walkdown work 

(which is not necessarily a 100 % walkdown of all components) can there be 

high assurance that specific components do in fact fit within the generic 

categories for which the generic guidance is applicable. 

I. Additional comments on the screening guidance: There are a few other 

areas where the Peer Review Group believes that the screening guidance 

should be improved. Among these are the approach to screening or analyzing 

batteries, especially aged batteries; the approach to walking down and 

analyzing fragilities for small LOCAs~ in recognition of the fact that a 

thorough walkdown of all small lines in the primary pressure boundary may be 

prohibitively difficult, so that a screening approach may be needed; the 

treatment and documentation of system successes in the event trees; and 

how to treat component interfaces and 'super-components' such as multiple 

items on a single skid. 

V. Level of Peer Review Group Effort 

The PRG has been asked to comment on whether the level of review effort 

expended by the Group has been adequate; and more generally on how much 

effort should be devoted to peer review of seismic margins studies in the 

future. 

I) On the first issue, the PRG is unanimous in believing that the peer 

review effort expended on this trial review has been a necessary part of 

accomplishing the project's objectives effectively. This project was exploring 

a new methodology, determining what are its strengths and weaknesses, and 

performing the research work in the context of a licensing action. This 

combination of ingredients would have been much more difficult to accomplish 

well without the continuing interactions that the project team had with our 

7 

B-66 
Peer Review Group. These continuing interactions allowed for the timely 

resolution of several technical issues that arose during the project. If there 

had been no PRG, or if the peer review activity had only been a review of 

the final report once completed, we believe that the success of this project 

in accomplishing all of its objectives would probably not have been as 

outstanding. We also believe that all of the project participants agree with 

this conclusion. 

2) On the second issue (concerning peer review of seismic margin studies in 

the future), our PRG members are divided. Here we are considering the 

situation in which the margin review methodology has become relatively 

"standardized" and· a review is being undertaken by a utility, on either its 

own initiative or NRC's initiative. 

(i) Some PRG members believe that carrying out a successful seismic 

margin study requires the interaction of a group of outside reviewers, 

experts in the disciplines involved, to provide technical review as a way of 

assuring the validity of the study results. The rationale for this position is 

that it will be difficult for a utility to perform such a review competently, 

especially for routine applications of the methodology in which review 

teams are composed of engineers who may not be truly expert already in 

the technical aspects of these subjects. 

(ii) Other members of the PRG believe that, once the methodology has 

been developed into a more-or-less standard methodology, the necessity for 

peer review will be no different than it is for any other engineering study 

performed by utilities on their plants. In this view, the internal review 

procedures that utilities already use to assure the technical validity of their 

'internally·supported analyses should be just as adequate in this arena as 

they are in the numerous other engineering arenas already part of reactor 

safety analysis. 

(iii) There is yet a third view among the PRG members that can be 

summarized as follows: outside peer review is not an essential attribute of 

any routine utility-supported seismic margins study, but its presence will 

surely enhance the study•s credibility with NRC, and hence the NRC's need 

for its own separate review may be less. Therefore, outside peer review 

should be considered carefully by any sponsoring utility, especially if the 

margins review team itself is not fully familiar with all of the technical 

aspects of performing these studies. 

This division of views among the PRG members on the need for routine 

outside peer review does not conflict with the unanimity of views on the 

usefulness of the peer review exercise for the current project, which was a 

trial (research) project. In particular, the value of ongoing interactions 

throughout he course of the trial project cannot be ov·eremphasized. 

8 

B-67 
VI. Final Remarks 

In conclusion, it is our pleasure to acknowledge the outstanding cooperation 

and assistance of all J)arties who participated in this trial margins review. 

The PRO found excellent cooperation from all five parties to the review 

effort: the NRC staff; the utility team from Maine Yankee and Yankee 

Atomic Electric; the Lawrence Livermore staff; and LLNL's subcontractor 

teams from EQE Inc. and Energy Incorporated. The PRO believes not only 

that a project like. this should be technically competent (it was) and timely 

within schedule and budget (it was). but also enjoyable for the participants, 

including both project team and reviewers. For making this last aspect come 

true, we thank one and all. 

Robert J. Budnitz 

Chairman, Peer Review Group 

9 

B-68 
APPENDIX C 

UTILITY COMMENTS 

C-1 
·.' 

ATOmiC POWER COmPADH • 1 671 WORCESTER ROAD 

ENGINEERING OFFICE FRAMINGHAM, MASSACHUSETTS 01 701 

617-872-8100 

MAINE YANKEE SEISMIC DESIGN MARGIN PROGRAM 

UTILITY PERSPECTIVE 

Prepared By: W. E. Henries 

C-2 

Lead Mechanical Engineer 

J. T. McCumber 

Project Engineer 

December 1986 
>· 

I. PURPOSE 

This report provides a description of the Utility role in a Seismic 

Design Margins Program. This description will assist potential generic 

implementation by documenting our experience with the methods utilized, costs 

incurred, and impact on normal Utility operations. 

-1-

5210R 

C-3 
'.' 

: .. -~·. 

II. BACKGROUND 

Maine Yankee has been involved with the issue of demonstrating seismic 

margins for several years, with particular focus generated by the 1982 

New Brunswick earthquake. 

In addition to identifying the inherent conservatism in Maine Yankee's 

seismic design, a series of voluntary reviews and walkdowns were initiated in 

1982 to find any potenti'al weak areas. These plant reviews were performed by 

known experts in the fields of structural and seismic design. Several cost 

beneficial upgrades, primarily associated with equipment anchorage, were 

identified which if made would enhance seismic adequacy. The equipment 

anchorage and supports upgraded by Maine Yankee are listed in Table 1. 

Maine Yankee and Yankee Atomic Electric Company have been very active 

in the efforts of the Seismic Qualification Utility Group (SQUG). The 

equipment at Maine Yankee has been favorably compared to the SQUG experience 

data base as it was developed. 

When assessing the applicability of the Maine Yankee experience to the 

generic Seismic Design Margins Program, it should be recognized that Maine 

Yankee may not represent the typical state of seismic adequacy for a pl~nt of 

its age. This is due to upgrades associated with the seismic review and 

enhancement program described above. 

-2-

52 lOR 

C-4 
III. INFORMATION RETRIEVAL/ANALYSIS SUPPORT 

The primary role of the Utility in the Seismic Design Margins Program 

was to provide a complete and accurate description of the operation and 

construction of the Maine Yankee plant. It was first necessary to provide a 

description of the safety-related systems, their functions, and 

interdependencies. The Analysis Team (AT) could then properly categorize them 

as "Group A" or "Not Group A" Systems (Reference 1). In addition to the 

systems information needed for the Group A screening, it was also necessary to 

provide seismic design d~tails to the AT to aid in their seismic capacity 

determinations. A summary of the information provided to the AT is contained 

in Table 2. 

Once the AT systems modeling and fragilities analysis began, the 

Utility role evolved into a question response mode. Specific system operation 

or equipment anchorage/design questions were forwarded to the Utility and the 

requested answers or design details were returned to the AT. Many of the 

questions were readily answered but, occasionally, specific details were not 

available and required more extensive review. Examples of the latter included 

reactor internals design details and anchorage details for original plant 

equipment which was not within the scope of the previous seismic review 

programs. 

In order to obtain the necessary reactor internals and control rod 

drive information, a meeting with the NSSS vendor was arranged and attended by 

representatives of both the Utility and the AT. The meeting was a success, 

and the information required to perform the fragility analyses was obtained. 

Similarly, when original installation information was not readily available, 

searches through microfilm archives, or actual field measurements were 

required to answer the AT questions. The availability of data is a major 

factor in determining the Utility workload and associated costs. 

The final role played by the Utility was to review the systems models 

and assumptions developed by the AT. This effort involves close scrutiny 

during the entire development process to insure the accuracy of the model and 

-3-

52 lOR 

C-5 
',, 

'I 

its assumptions. It is important to prevent misconceptions or easily upgraded 

"weak links" from entering into the models. 

Following this approach, Maine Yankee agreed to implement 

post-earthquake procedural guidance and to make minor anchorage upgrades to 

increase the confidence level of the system models and equipment fragilities. 

The total Utility effort was accomplished without major impact to 

normal operations. As discussed in Section V, slightly more than 1,000 

man-hours were dedicated to the program over a ten-month period. Given 

advance planning and an absence of a refueling outage, permitted the 

engineering hours shown on Table 4 to be provided by the existing engineering 

personnel. 

-4-

52 lOR 

C-6 
IV. PLANT WALKDOWNS 

General 

The Seismic Design Margins Program included two walkdowns of the Maine 

Yankee plant. The first walkdown, lasting a week, was performed early in the 

program to obtain: 

1. A firsthand understanding of Maine Yankee's structures, systems, 

and equipment, 

2. Equipment anchorage details, and 

3. Assurance that plant-specific problem areas or seismic interactions 

were not a concern. 

The final plant walkdown, lasting three days, was performed near the 

end of the program following the completion of preliminary fault tree and 

fragility analyses. The objectives of the final plant walkdown were to: 

1. Gather additional information on systems and equipment needed to 

complete the analyses, 

2. Obtain equipment operating and maintenance information, 

3. Insure the validity of system interaction assumptions, and 

4. Resolve open issues. 

The walkdowns were each scheduled in conjunction with Peer Review Group 

meetings, so in addition to the AT, the walkdowns included the Peer Review 

Group (PRG), NRC, and Lawrence Livermore (LLNL) personnel. The specific 

interests of each group had to be carefully considered in the walkdown 

planning. 

-5-

5210R 

C-7 
Walkdown Planning 

In addition to the actual walkdown activities, PRG meetings, plant 

access training, body counting, getting dosimetry, and miscellaneous 

information gathering meetings had to be scheduled within the walkdown time 

period. Considering all of the above, a productive walkdown required careful 

preparation and planning. 

Coordination with the AT well before the actual walkdowns was necessary 

to determine the areas of the plant and specific equipment of interest. This 

advanced notice allowed the Utility Walkdown Coordinator to familiarize 

himself and other escorts with the exact areas of the plant to be visited, as 

well as the system's function and anchorage details of the equipment to be 

inspected. Pre-walkdowns, marked-up drawings and equipment lists were all 

helpful tools in assuring efficient use of the limited time available. 

Other preparations which facilitated plant walkdown activities included: 

1. Sending out all of the required health physics forms with 

instructions to allow walkdown members to arrive at the site with 

the necessary paperwork completed. 

2. Scheduling body counting during the PRG meetings (approximately 15 

minutes for each person - took turns leaving the meeting). 

3. Notifying key plant personnel of the walkdown schedule and assuring 

their availability, as needed, 

4. Reserving meeting rooms and work rooms, and 

5. Obtaining necessary approvals for camera passes. 

Walkdown Logistics 

The plant walkdowns were made up of a large group of people (around 

20), of varied interests, and conducted in a limited amount of time in a 

limited access environment. Therefore, complex logistics were to be expected. 

-6- C-8 

52 lOR 
... ! .: 

Plant security regulations limited the number of unbadged people which 

could be escorted throughout the plant by one escort to five. This meant that 

four or more prepared escorts were necessary at any time. Also, it was 

important to match the escort's expertise to the group; i.e., systems engineer 

for a group seeking plant operations information and a structural/mechanical 

engine~r for a group looking at anchorages. 

The objectives of the walkdown members are generalized below: 

Analysis Team 

Peer Review Group 

NRC and LLNL 

Gather data on equipment throughout the plant. 

Obtain information from plant operations, 

engineering, and maintenance personnel. 

Obtain a good overview of the plant. 

Inspect equipment of specific interest to them. 

Knowledge of AT activities. 

Obtain general overview of plant. 

General knowledge of AT activities 

(programmatic). 

A preplanned tour of the plant was arranged at the beginning of the 

initial walkdown for all of the walkdown members. The list of plant areas and 

equipment which the AT wanted to see was used to develop the tour. This 

assured that everyone involved received a good overview of the plant and had a 

chance to inspect the equipment of interest. 

Subsequent plant tours were arranged to facilitate the specific 

information which had to be gathered. During the initial walkdown, the PRG 

could only stay three days, so an effort was made to meet their plant walkdown 

needs during that time. Otherwise, the priority was to assure that the AT 

obtained information needed. 

Meetings were scheduled each evening with all walkdown participants to 

discuss progress and potential problem areas. Also discussed, was the 

schedule of the next days activities and information which may be needed. 

-7-

52.10R 

C-9 
, .. ··.f.· 

This allowed time for preparation that night, if required. These meetings 

were extremely productive, and while they made for a long day, are highly 

recommended. 

Plant Impact 

As a result of the preplanning associated with the walkdowns, the 

impact on normal plant operations was minimized. Additional steps which were 

taken to lessen the impact included: 

52 lOR 

1. Staggering the time which various walkdown groups pass ·through HP 

to minimize their processing time and to avoid congestion in the 

change area. 

2. Keeping the number of people in the Control Room to a minimum, 

preferably doing Control Room work during the evening shift. 

3. Using simulator and simulator staff to answer operations questions 

where possible. 

4. Letting operations or security personnel know in advance of locked 

areas which must be accessed. 

5. Letting HP personnel know in advance when access to a high 

radiation area is required so they may arrange proper support. 

6. Having radiation work permits made out in advance. 

-8-

C-10 
V. COST 

The Utility costs to support the Seismic Design Margins Program are 

summarized in Tables 3 and 4. Several observations regarding these costs are 

worth noting. 

First, the majority of man-hours were spent by Mechanical Engineering 

since that discipline has historically been responsible for seismic review 

programs at Maine Yankee; thus, the bulk of data retrieval and program 

management fell to them. Systems Engineering also required considerable 

man-hours. Their input was vital in identifying the required systems and 

components and assuring plant unique operational features were accurately 

modeled. The Licensing man-hours and Consultant costs are attributable to the 

fact that this pilot research program also had licensing implications for the 

Utility. 

Secondly, the relatively high travel costs and miscellaneous expenses 

are attributable to the fact that the AT consultants are all from the West 

Coast, while the subject plant is on the East. Additionally, the tight 

program schedule requirements required frequent overnight mailings and phone 

calls. 

Finally, the total program effort may vary for other utilities for two 

reasons. Reduction in the data retrieval time for the Seismic Design Margin 

Program was greatly aided by the walkdown documentation and equipment 

anchorage upgrades previously performed at Maine Yankee. Since 1982, three 

separate plant walkdowns have been conducted at Maine Yankee as part of its 

voluntary seismic review program. As a result of these walkdowns, many 

equipment anchorages and masonry walls have been strengthened beyond FSAR 

requirements. These upgrades, and the relatively recent design analyses, 

allowed easy information retrieval for the Utility and fragility screening for 

the AT. On the other hand, the time involved with retrieving original design 

data for those components not within the scope of the previous walkdowns was 

significant. The inability to readily retrieve old (16 to 18 years) design 

records or to obtain as-built details due to radiation or operational 

-9-

52 lOR 

C-11 
constraints prevented the Utility from being as responsive to AT requests as 

we would have liked. Other utilities should thus estimate their man-hours 

accordingly. 

-10-

52 lOR 

C-12 
VI. IMPRESSIONS/RECOMMENDATIONS 

The concept of using plant walkdowns and experience screening to 

evaluate seismic margins is excellent and long overdue. The value obtained by 

eliminating those components shown by experience not to be seismically fragile 

is enormous. Just as importantt the walkdown methodology provides a thorough 

systems interaction review. This program is a major step in developing a cost 

effective method of identifying plant-specific seismic margins andt perhaps 

more importantt easily fixed seismic "weak links." 

While we praise the overall thrust of the program, several 

recommendations are provided below which may improve future reviews. 

52 lOR 

1. Separate the seismic review from the licensing area. This change 

will allow closer Utility/AT cooperation without the glare of the 

"public meeting" requirements. 

2. Define the review earthquake with greater precision. The ongoing 

concerns regarding the confidence level of the ground response 

spectral shape and ZPA resulted in too much wasted effort. 

3. Precisely define the need for long-term safe shutdown (Group B 

functions) capability. Considerable effort was expended on these 

components which we believed were excluded from the review scope 

per Reference 1. 

4. Provide better field guidance on the use of the SQUG experience 

data base (i.e., MOV/AOV interactiont cable trays). 

5. Develop improved methods of determining/defining "high confidence" 

levels for components such as yard tanks. As demonstrated by our 

experience, the significance of yard tank capacity is high and the 

analytical methods for determining actual capacity seem 

unnecessarily conservative. 

6. Decide if nonseismic failures should be included in the HCLPF 

determination andt if so, how. 

-11-

C-13 
VII. REFERENCES 

1. NUREG/CR-4482, "Recommendations to the NRC on Trial Guidelines for 

Seismic Margin Reviews of Nuclear Power Plants," by P. G. Prassinos, 

M. K. Ravindra, and J. B. Savy, March 1986. 

-12-

52 lOR 

C-14 
52 lOR 

TABLE 1 

Maine Yankee Equipment Anchorage and Support Upgrades 

Emergency Buses 5, 6, 7, and 8 

Battery Chargers 1, 2, 3, and 4 

Batt~ry Inverters 1, 2, 3, 4, and 5 

Fl~or Mounted Diesel Generator Control Panels 

Main Control Board 

Control Room Auxiliary Cabinets 

Electrical Control Board 

Radiation Monitoring System 

Heat Tracing Cabinet 

Air Conditioning Control Panel 

EHC Panel 

Reactor Regulating System 

Feedwater Regulating System 

Vibration and Loose Parts Monitoring Cabinet 

Reactor Protective System 

Core Loading Panel 

Radiation Monitoring System 

Meteorological Survey Cabinet 

Safety-Related Instrument Racks 

Masonry Wall Reinforcement 

Service Water Piping Support 

-13-

C-15 
.. ' , . 

. ,. 

TABLE 2 

Summary of Information Provided for Maine Yankee 

Seismic Design Margins Review 

o Maine Yankee FSAR 

o Ground and Floor Response Spectra for 0.18g NUREG/CR-0098 an 0.2g 

Regulatory Guide 1.60 Earthquakes 

o Approximately 15 Specifications 

o Approximately 40. Calculations (or portions thereof) 

o Approximately 550 Drawings 

o Emergency Operating Procedures 

o Pump NPSH Curves 

o "Tagging" and Modification Control Procedures 

o Maine Yankee Technical Specifications 

o Maine Yankee Training Manuals 

o Equipment Testing Frequencies and Procedures 

o Abnormal Operating Procedure for Flooding of the Circulating Water Pump 

House 

o Abnormal Operating Procedure for Earthquakes 

o Reports on: 

52 lOR 

Class I Structures Dynamic Modeling 

Components Required for Accident Mitigation and Safe Shutdown 

Reactor Protective System 

Appendix R Alternate Shutdown System 

Maine Yankee Masonry Walls 

1983 Seismic Walkdown and Modifications 

Seismic Review Program Summary 

Maine Yankee Seismic Hazard Analysis (UHS) 

Conservative Seismic Capacities of Maine Yankee's Reactor Containment 

Evaluation of LOCA-Related Loadings on the Reactor Coolant System 

Expansion Anchor Information and IEB 79-02 Anchor Bolt Testing 

-14-

C-16 
TABLE 3 

Maine Yankee Seismic Margins Program Costs* 

Man-Hours 

Mechanical Engineering 

Systems Engineering 

I&C Engineering 

Electrical Engineering 

Nuclear Engineering** 

Licensing 

Plant Engineering/Operations 

Consultant (Includes Travel) 

Travel 

Miscellaneous 

(Drawings, Xerox, Mail, Phone) 

TOTAL*** 

* Does not include cost of anchorage upgrades. 

** Includes PRA review. 

585 

195 

15 

45 

75 

100 

45 

*** Does not include costs to review final report. 

-15-

52 lOR 

C-17 

Dollars 

$21,000 

11,400 

1,525 

!!!33.225 
52 lOR 

Month 

May 

June 

July 

August 

September 

October 

November 

December 

January 

February 

TOTAL 

TABLE 4 

Monthly Man-Hour Breakdown 

-16-

C-18 

Man-Hours 

67 

50 

170 

154 

88 

159" 

171 

66 

100 

___TI_ 

~ 
rebruary 26, 1987 

NN-87-22 

01rtctor of Kucltar Reactor Regu1at1on Untttd St&tes Nu~ltar Rf9ul•tory Comm\ss1on 

Washington, o. c. 20555 

Attentton~ Mr. Ash~ c. Thadant, D'rector 

PHR Project D1rtctor&te 18 

Olv1s1on of L1cens1ng 

Rtferencts~ (a) L\cense Ho. OPR-36 <Docket Ho. 50-309) (~) HRC to M.Y. tatter dated 

(e) H.Y. to HRC lttttr dated 

Subject: M&1ne Yankee Se1stnic Mar·g\ns Pro.qra.t~ 

Gtn t l enen: 

~VOUSTA., MAINE t'4S36 

{~7') 123-35..2 t 

for the p'st e19ht eonths. Hltne Yankee hi$ been worktno with the NRC 

staff and 'ts consultants on a se1sm1c •arg1ns program. This progr~. vas 

des1gned to d~nstrate the feasibility of the Margin assess~ent met~logy. 

and to reto,ve HRC staff eoncerns about the adeQuacy of Ma\ne Yankee's se1s~,c 

des1gn. The program has been ca.~leted and a draft report issued. 

The progrlll has deaonstrated that the Naine Yan~ee plant po~sesses seistJaic 

ruggedness well bt~ond the plant's Otig,nal design. Tht progra~ also · fdtntif\ed several area$ Yhere the se\s~tc ruggedness of the p~ant could be 

effectively tnhtnted. 

The ch&ngts that we have decided to make along with the sthedu1ed c~let1on dttes are listed bt1ow. 

Dfscr1ptton of 

Change 

• Dttstl futl day tank anchorage upgr4de ' Co~trot Room cooler an(hOr~ge uP9rade 

• Heldt~g tart/;as bottle t,taown 

• Securtty l1ght,ng tledo'tfn 

• Mlint control board alara tttdo~n 

• Strengthen blockwa11 VE 21-J 

• Upgrade anchors for fans 44A & B 

• Inst&ll internal anchors tor transformers 507 & 608 

• Repl&ct safaty class bttttrlts I & 3 

• Rep1aee $lftty class battertes 2 • 4 

• Refueltng water storag• ttnk anchorage upgrade 

C-19 

Schedule Completion 

Date 

done 

done 

done 

done 

do-ne 

1987 OUtaoe 

1981 OUttge 

1987 Outage 

l987 OUtage 1988 Outage 1988 Outa9e 
.· i •'. 

·Untted States Huclear Rtgu1ttory to~M\s11on 

Attentton: ~r. Asho~ c. Thad'n,, D,rtctor Page Two W«-87-~Z 

Tht drtft setsm1c ~rgfns report tndtcates that the Ha1nt Yankee plant could w\thstand st1sm,c events wtl1 ,n excess or tht plant's orfginal design b«sfs wtthout end&ngering publ1c health and safety. The report al'o \nd1cates 

that tht ""' t' ng cOttl)Ontnt 1 s the P'tfue1 t ng water storage ta~ (RHST) vi th ' 

htgn eonf1dtnee of low probabil,ty of fa,1ure Chclpf) of O.Ztg. As 1ndtcated 

tn the fortiQfng table, we pl&n to upgrade t~e seisMic ruggedness of th1s 

tank. by ,aprov1ng 1t$ anchorage, no later then the 1988 refueltng ootage. He 

are ~ak1ng an effort to cOMplete all or part of thtse upgrades during the f987 

rtfutl\ng OYtagt. ~lthough the a~chorage upgr,dts have not been flna,1y des1gntd, wt exoect to upgr~de the tank's seic~lc rugg~dnets to a hclpf value 

tn the rtnge of 0.27g. 

Ht wou'ld apprtctate your c.or~currenct that c:CMRPietion of tharst coan\tttd aod1t1cat1ons resolves this lonq standing 1\cens!ng concern so that we .ay a~ply our att1ntion and rtsources to other sartty enhancing efforts. 

cr:.t/bjp 

£nclosurt 

cc: Dr. fholas E. Murley 

Mr. Pit SH.rs 

Mr. CorneJ,us f, Holdtn 

149Zl/GDN 

Very tru J y your$~· 

~IH£ YAH)(EE AT<JtlC P()fER C()(PAHY 

-&19-/lo/~ C. 0. Whittter, Mlnager Nuclear Eng1neer1n; and Lt(tns~ng 

C-20 
NRC FORM 331 

12·8•1 

U.S. NUCLEAR REGULATORY COMMISSION 

NRCM 1102, 

:1201,3202 BIBLIOGRAPHIC DATA SHEET 

SEE INSTRUCTIONS ON THE REVERSE. 

2. TITLE ANO SUBTITLE 

Seismic Margin Review of the Maine Yankee Atomic Power 

Station . 

Vol. 1. Summary Report 

5. AUTHORISl 

P. G. Prassinos, R. C. Murray, and G. E. Cummings 

7 PERFORMING OI'!;:;ANIZA T ION NAME AND MAILING A00R E:SS (lnc•ual!! i'ro Codf) 

Lawrence Livermore National Laboratory 

P.O. Box 808~ L-197 

Livermore, CA 94550 

10 SPONSORING ORGANIZATION NAME .II."ND MAILING ADDRESS (lnciUoo .£,0 Coot~J 

Division of Engineering Safety Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 

12. SUPPLEMENTARY NOTES 

13. ABSTRACT 1200 wrJrd• or I"JJJ 

I. REPORT NUMIER 1Au•f1nlltil by TIDC . .J(Jd Vol. No, t1 llllfl 

NUREG/CR-4826 

UCID-20948 

Vo 1. 1 3. LEAVE BLANK 

4 DATE REPORT COMPLETED 

MONTH YEAR 

('1a rch 1987 ,; C.:.TE REPORT ,;:;SUED 

YEAR 

March 1987 8 PRQJECT.'T•'.SI(/WOR~. uNIT 1\olJMBEA 

') Fir.. OR GRAhiT NUMBER 

NRC FIN No. A-0461 

'lo fYPE OF "'E?ORT 

Technical o ?ERIOO COVERED lloctus.,e cJdti!!SJ 

June 1986 - March 1987 

This Summary Report is the first of three volumes for the Seismic Margin Review of the 

Maine Yankee Atomic Power Stati'on. Volume is the Systems Analysis of the first trial 

seismic margin review. Volume 3·documents the results of the fragility screening for the 

review. The three volumes demonstrate how the seismic margin review guidance (NUREG/CR4482) of the Nuclear Regulatory Commissi'on (NRC) Seismic Design Margins Program can be 

applied. 

The overall objectives of the trial review are to assess the seismic margins of a partic- ular pressurized water reactor, and to test the adequacy of this review approach, 

quantification techniques, and guidelines for performing the review. Results from the 

trial review will be used to revise the seismic margin methodology and guidelines so that 

the NRC and industry can readily apply them to assess the inherent quantitative seismic 

capacity of nuclear power plants. 

14 DOCUMENT ANALYSIS d ~EYWOAOSIOESCRIPTORS 

Seismic Margins 

b. IOENT!FIEI'ISIOPEN.£NDED TEAMS 

~ U. 5, COV(RNmE:NT PRINTING orr ICt :1997- 131-682:60035 

15 AVAILA81Li':'Y 

STATEME!><T 

Unlimited 

16 SECURITY C;..ASSIFICATION 

Unclassified 

Unclassified 11. NUMBER OF PAGES 

18 PRICE 
'\",\ 

'• '' 

•'' 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300 

SPECIAl. FOURTH·CI.ASS RATE 

POSTAGE & FEES PAID 

USNRC 

WASH. D.C. 

PERMIT No. G·67 
NUREG/CR-4826 

UCID-20948 

Vol. 2 

Seismic Margin Review of the 

Maine Yankee Atomic Power Station 

Systems Analysis 

Manuscript Completed: February 1987 

Date Published: March 1987 

Prepared by 

D. L. Moore, D. M. Jones, M. D. Quilici, J. Young 

Energy Incorporated 

Kent, WA 98031 

Under Contract to: 

Lawrence Livermore National Laboratory 

7000 East Avenue 

Livermore, CA 94550 

Prepared for 

Division of Engineering Safety 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 

NRC FIN A0461 

ABSTRACT 

This Systems Analysis is the second of three volumes for the Seismic 

Margin Review of the Maine Yankee Atomic Power Station. Volume 1 is the 

Summary Report of the first trial seismic margin review. Volume 3, Fragility 

Analysis, documents the results of the fragility screening for the review. 

The three volumes are part of the Seismic Margins Program initiated in 1984 by 

the Nuclear Regulatory Commission (NRC) to quantify seismic margins at nuclear 

power plants. 

The overall objectives of the trial review are to assess the seismic 

margins of a particular pressurized water reactor, and to test the adequacy of 

this'review approach, quantification techniques, and guidelines for performing 

the review. Results from the trial review~will be used to revise the seismic 

margin methodology and guidelines so that the NRC and industry can readily 

apply them to assess the inherent quantitative seismic capacity of nuclear 

po,-ier pl!:lnts. 

i ; ; 
:•; 
TABLE OF CONTENTS 

Section 

ABSTRACT 

Page 

iii 

LIST OF TABLES ••••••••••••••••••••••••••••• o • • • • • • • • • • • • • • vii 

LIST OF FIGURES • • • • • • . . • • • • • • . • • • • • • • • • • • • • • • • • • • • • • • • • • • • ix 

1. INTRODUCTION .. • • .. • • • • • • • • • • • • • .. • • • • • • • • • • • • • • • • • • • • • • • • • • 1-1 

2. METHODOLOGY • • • • • • • • • • • • • • • • • • • • • • . • . • • • • • • • • • • • . • . . • • • • 2-1 

2.1 General Approach. • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 2-1 

2.2 Step 2 -Initial Systems Review • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 2-1 

2.2.1 Gather System Information • • • • • • • • • • • • • • • • • • • • • • • • • • • 2-1 

2.2.2 

2.2.3 

2.2.4 

2.2.5 

Classify Front-Line Systems •••••••••••••••••••••••••• 

Identify Group A Front-Line Components ••••••••••••••••• 

Classify Support Systems •••••••••••••••••••••••••••• 

Identify Group A Support System Components ••••••••••••• 

2-1 

2-2 

2-2 

2-2 

2.2.6 Identify Plant-Unique Features • • • • . • • . • • • • • • • • • • • • • • • • 2-3 

2.2.7 Prepare for First Plant Walkdown • • • • . • • • • • • • • • • • • • • • • • 2-3 

2.3 Step 4 - First Plant Walkdown • • . • • • • • • • • • • • • • • • • • • • • • • • • • • • • 2-3 

2.3.1 Peer Review Group Meeting • • • • • • • • • • • • • • • • • • • • • • • • • . 2-3 

2.3.2 Plant \Valkdown • • • . • • . • • • • • • • . • • • • • • • • • • • • • • • • • • • • 2-3 

2.3.3 Plant Staff Discussions . • • . • . • • • • • • • • • . . . • • . • . . . • • • • . 2-4 

2.4 Step 5 - Systems Modeling. • . . • • • . • • . • • • • • • • • • • • • • • • • • • • • • • • 2-4 

2.5 

2.6 

2.4.1 Review Event Trees • • • • • • • • • . . . • • • • • • • . • . . . • • • • • • • • 2-4 

2.4.2 Develop Fault Trees . • . . • • • • • . • • . . • • • • • . . • • . • • • • • • • • 2-5 

2.4.3 Develop Data Base • • • • • • • • • • • • • . • • • • • • • • • • • • • • • • • • • 2-7 

2.4.4 Determine System Cut Sets ••.•••••••••••••••••.•..••• 

Step 6 -Second Plant Walkdown •••••••••••••••••••••.••••••• 

Step 7 - Determine Minimal Cut Sets ••...••••••••••••.••••••.•• 

2.6.1 Finalize Event Trees •••••..•.••.•••••••••••••••••••. 

2.6.2 Finalize Fault Trees ••••..•....•••••••.•••..•••••••• 

2.6.3 Llnk Fault Trees •...•..•..••••••••••.•••..••••...•. 

2.6.4 

2.6.5 

Determine Preliminary Boolean Equation ....••••••••••..• 

Determine Final Boolean Equation •••••••••••••••......• 

v 

2-7 

2-8 

2-8 

2-8 

2-8 

2-8 

2-9 

2-9 
i' 

TABLE OF CONTENTS (Cont'd) 

Section 

3. SYSTE.'viS ANALYSIS •••.••••••••••••••••.•••••••••••..•••••.• 

3.1 System ldentifica tion ............................... 0 ..... · •• 

3.1.1 Group A Systems .................................. . 

3.1.2 Systems Removed from Group A ••• o .................... .. 

Page 

3-1 

3-1 

3-1 

3-2 

3 .. 2 Systems Analysis of Front-Line Systems •.••..•••••••••• o • • • • • • • 3-3 

3.2.1 Auxiliary Feedwater System •••••••••••••• o........... 3-3 

3.2.2 

3.2.3 

High Pressure Safety Injection System .................... o 

Power-Operated Relief Valves 

(Feed and Bleed) •....•.•..••.•••.•..••.••....••...• 

3-5 

3-8 

3.3 Systems Analysis of Support Systems ••••••••••••• 0 •••• 0 • • • • • • • 3-9 

3.3.1 Component Cooling Water Systems • • • • • • • • • • • • . • • • • • • • • 3-9 

3.3.2 

3.3.3 

3.3.4 

Service \Vater System ............................... . 

Electric Power System •••••••••••• o ••••••••••••••••• 

Actuation Systems ................................ o • 

3.4 Probability Calculations ••••••••••••••••••••••••••••••••••• 

3-11 

3-13 

3-16 

3-16 

4. ACCIDENT SEQUENCE ASSESSMENT AND RESULTS • • . . • • • • • • • • • • • • • 4-1 

4.1 Event Trees ........................................... . 

4.2 

4-.3 

4-.1.1 

4- .. 1.2 

No LOCA Case . ................................... . 

Small LOCA Case . ................................. . 

Core Damage Sequences ................................. ., ••• 

4.2.1 

4.2.2 

4.2.3 

4.2.4-

4.2.5 

T 1 LP 1 .......................................... . 

T 1 L D ........................................... . 

s 2o .............................................. . 

S2LP2 .......................................... . 

S 2LD ........................................... . 

Boolean Equations for No LOCA and LOCA Cases ................. . 

4.3.1 No LOCA Case . .................................. . 

4.3.2 Small LOCA Case ....................... . . . . . . . ..... 

4--1 

4--1 

4--2 

4--3 

4-3 

4-3 

4--4-

4-4 

4--4 

4-4 

4-5 

4-6 

4.4 P !ant-Level Boolean Equation . . • • • • • . . • • • • • • • • • • • . • • • • • .. • • . . 4-6 

vi 
TABLE OF CONTENTS (Cont'd) 

5. ENGINEERING AND METHODOLOGY INSIGHTS • . • • • • • • • • • • • • • • • . • . . 5-l 

5.1 Engineering and Operational Insights • • • • . • • • • • • • • • • • • • • • • • • • • • 5-l 

5.2 Insights on the Methodology and Execution • • • • • • • • • • • • • • • • • • • • • • 5-2 

5.2.1 System Classification and Screening 

5.2.2 

5.2.3 

5.2.4 

5.2.5 

REFERENCES 

APPENDIX 

Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-2 

Preparation for Walkdowns and Documentation .............. . 

Systems Analysis and Pruning Process ••••••••.•••••••••• 

Minimal Cut Set Evaluation Process ••••••••••••••••••••• 

Schedule and Resources ••••••••••.••••.••••••••••••• 

................................................ 

5-4 

5-4 

5-5 

5-6 

R-1 

A Identifiers and Symbols • • • • • • • • • • • • • • • • • . . • • • • • • • • • • • • • • • • • A-1 

B · Auxiliary F eedwa ter System • .. • . • • • • • • • • • • • • • • • • • • • • • .. . • • • • • B-1 

C High Pressure Safety Injection System . • • • • • • • • • .. • • • • • • • • • • • • • • C-1 

D Primary Pressure Relief System • • • • • • . • • • • • • • . . . . • • • • • • • • • • • D-1 

E Primary Component Cooling Water System ••••• ·• • • • • . . . • • • • • • • • • E-1 

F Secondary Component Cooling Water System . . • • . • • • • • • • • • • . • • • • F-1 

G Service Water System. • • • • • • • • . . • • • • . • • • • . .. • • • . • • • • • . . • • • • G-1 

H Electric Power System • • • • • • • • • • • . • • • • • • • • • • • • • • • • • • • • • • • • H-1 

Actuation Systems • • • • . • • • • • • • • • . • .. . • • • . • . • . • • • • • • • • • • • • • 1-1 

vii 
.•,'' "\ 

2-1 

3-1 

3-2 

3-3 

3-4 

3-5 

3-6 

3-7 

3-8 

3-9 

3-10 

3-11 

3-12 

3-13 

3-14 

3-15 

4-1 

4-2 

4-3 

4-4 

4-5 

4-6 

4-7 

5-1 

A-1 

A-2 

LIST OF TABLES 

Definition of plant safety functions •.•.•••••••••••.••.......• o o • 

Front-line system vs support system dependency 

matrix ................................................. . 

Support system vs support system dependency matrix • , •••• o ••••••••• 

AF\V screening overview table ••.••• o ••••••••••••••••••••••• o •• 

Standard footnotes for screening overview tables •••••••••• o •••••••• 

AFW system cut sets ••••••••••• o ••••••••••• o ••• o ••••• 0 o •••.• 

HPSI/CSPPCL screening overview table • o ••••••••••• o •••••••• o • o • 

HPSI system cut sets ••• o • 0 •••• o •••••••• o • o ••••••••••••• o ••• 

PORVs screening overview table •.• o • o •••••••••••••••• o • o •••••• 

PORV (no LOCA) cut sets •••••••••••••••••••••••• o ••••••••••• 

PORV (small LOCA) cut sets .................................. . 

PCC screening overview table •• o •••••••••••••••••••••••••••••• 

sec screening overview table ••••.••••••••••••• 0 •••••••••••••• 

SWS screening overview table ••.••••••••••.••..••••.•••••••••. 

Electric power screening overview table •••.••••••.••••.••••••.•• 

Nonseismic event failure probability calculations ••••••.••••.••••.•• 

Basic event descriptions •••...•••...•••.•••••••.•••..••..•••• 

Sequence T 1 LP 1 cut sets .••••.•. , ••..•.•.•••••••.•••• , ••.•••. 

Sequence T 1 LD cut sets •••••••••.•.•.••••••••••••....•••. , •. 

Sequence S 2o cut sets •••.•••••••••.•••.•...••••••••••.•.• , •• 

Sequence s2LP 2 cut sets •••••••.•..•••••••.••••.••••.•..•• 

Sequence S 2L D cut sets •••.••••..•....••••••••••••.•.•... 

Boolean equations for no LOCA and small LOCA accident 

2-10 

3-19 

3-20 

3-21 

3-22 

3-23 

3-24 

3-26 

3-27 

~-28 

3-29 

3-30 

3-32 

3-33 

3-34 

3-37 

4-7 

4-8 

4-9 

4-10 

4-11 

4-12 

sequences . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-13 

Systems analysis resource expenditure ..•..•••••.•.••••••....•..• 

Fault tree symbols ................... , .••••••.•..•.•....•... 

Fault tree event identifiers .•••.•..•.••...•••••••..•••.•.••.•• 

viii 

5-7 

A-1 

A-2 
;-

1''.'1 

B-1 

B-2 

B-3 

C-1 

C-2 

C-3 

D-1 

E-1 

E-2 

E-3 

E-4 

F-1 

F-2 

F-3 

F-4 

G-1 

G-2 

H-1 

H-2 

H-3 

H-4 

I-1 

LIST OF TABLES (Cont'd) 

Auxiliary feedwater (AFW) ••••••••••••••.••••.•..••.•••••• 0 •• 

A F \V valve tab 1 e . . . . . . . . . . . . . . . . a • • • • • • • • • • • • • • • • • .. • • • • • • • • 

AF\V cooling requirements •••• 0 ••••• o •••• o ••••••••• o ••••••••• 

High pressure safety injection (HPSI) and containment 

spray pump area cooling (CSPPCL) ••••• 0 o ••• o ••• o o ••••••••••• o 0 

HPSI/CSPPCL valve table .•••••• o ••••••••••••••••• o •••••••••• 

HPSI cooling requirements •••.•••••••••••• o 0 ••••••••••••••••• 

Primary pressure relief system (PPS) ..•••••. o • o ••••• o ••• o •••••• o 

Primary component cooling (PCC) •••••••.••••••••••.•• o • o •••••• 

PCC valve table. o •• o • o ••••••• 

PCC cooling requirements ••••••. 

PCC cooling loads ........................................ . 

Secondary component cooling (SCC) ..••.•••. o ••••••••••••••••••• 

sec valve table .. t1 •••••• e e • e ••• • •••••••••••• e e •• e ••••••• II e 

sec cooling requirements .••••••.••.•••••. 0 ••••••••••••••••• 0 

sec cooling loads . ........................................ . 

Service water system (SWS) •• o 

SWS cooling requirements .... 

AC power 

DC power ..... 

Diesel generators (DG) 

DG cooling requirements •..•.•• . . . . . . . . . . . . ' .. 

B-1 

B-2 

B-3 

C-1 

C-3 

C-5 

D-1 

E-1 

E-2 

E-3 

E-4 

F-1 

F-2 

F-3 

F-it 

G-1 

G-2 

H-1 

H-2 

H-3 

H-it 

Actuation systems . . . • . . . . . . • • . . . . . . • • • • • . . . . . . . . . . . . . . . • . • I-1 

ix 
LIST OF FIGURES 

2-1 Graphic representation of the screening operations 

(Figure 2-6 from NUREG/CR-4482) • • • • • • • • • .. • • • • • • • • • • • • • .. • • • • • 2-11 

3-1 Auxiliary feed water • • • • • • • • • • • • • • • • • • • • • • • • • • . • • • • . • • • • • • • • 3-39 

3-2 High pressure safety injection • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 3-40 

3-3 Primary pressure relief • • • • . • • • • • • • • • • .. • • • • • • • • • • • • • • • • • • • • • • 3-41 

3-4 Primary component cooling water • • • . • • . • • • • • • • • • • • • • • • • • • • • • • • 3-42 

3-5 Secondary component cooling water. • • • • • • • • • • • • • • • • • • • • .. • • • • • • • 3-44 

3-6 Service water system • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • .. • • • • • • • • 3-46 

3-7 AC power system • • • • • • • • • • • . • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 3-47 

3-8 DC power system • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 3-48 

3-9 Diesel generators • • • . • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • j"-.49 

4-1 Seismic event, LOOP event tree • • • • • • • • • • • • • • • • • • • • • • • • • • • .. .. .. • 4-14 

4-2 Seismic event, LOOP concurrent with small LOCA 

event tree • • • • • • . . • • • . • • • • • • • • . • • • • • • • • . • • • • • • • • • • • • • • • • • 4-15 

B-1 Auxiliary feedwater system fault tree ••••••••••••••••••••••••••• B-4 

B-2 Auxiliary feedwater system fault tree, pruned and 

merged . . . . . . . . . . . . . • . . . . . . . . . . . . . . . . . . . . . . • . . . • . . . . . . .. . B-9 

C-1 High pressure safety injection system fault tree • • • • • . • • • • • • • . • • • • • • C-6 

C-2 High pressure safety injection system fault tree, 

pruned and merged . • . • • • • • • . . . • • • • • • • • • • • • • • • . • . • • • • • • • • • • • C-16 

D-1 Power-operated relief valve fault tree •••.•••.••••••••.•••••••••• D-2 

' ' ", ~' D-2 Power-operated relief valve fault tree (no LOCA), 

pruned and merged . . • • • • . • . . • • • • . • • • • • • • • • • • • • • • • • • • • • • • • • . D-4 

D-3 Power-operated relief valve fault tree (small LOCA), 

pruned and merged • • • . • . . . . . . • • • . . . . . . • . . • • • . • • • • • • • • . • • • • • D-7 

E--1 Primary component cooling system fault tree •..••.•••.•••••••••••• E-6 

X 
LIST OF FIGURES (Cont'd) 

F -1 Secondary component cooling system fault tree • . . • • . . • • . • . . . . • • . • • F -6 

G-1 Service water system fault tree. • . • . • • • • • . • . . . • . . . . • • • • • • . • • . • • G-3 

H-1 AC power (MCC 7 A) fault tree • • • • • . . • • • • . • • . • . • • . • • • . . • • • • • . . H-5 

H-2 AC power (MCC 8A) fault tree • • • . . • • • • • • • . • . . • • . . • • • • • • . • . • • • H-6 

H-3 Diesel generator lA fault tree. • • • • • • • • • • • • . • • . • • • • • • • . • • • • • • • • H-7 

I-1 Safety injection actuation system (Train A) fault 

tree. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I-2 

' . 

xi 
.... • .. ', " 
',I 

1. INTRODUCTION 

This report documents the approach, results, and insights of the systems analysis tasks 

for the Seismic ~·1argin Review of the l\·1aine Yankee Atomic Power Station. This project 

is part of the Seismic Design Margins Program performed by Lawrence Livermore 

National Laboratory for the Nuclear Regulatory Commission. The purpose of a seismic 

margin review is to determine the capability of a plant to resist with high confidence an 

earthquake level greater than the design basis Safe Shutdown Earthquake (SSE). This 

seismic capacity is expressed as the earthquake acceleration level for which there is a 

high confidence of a low probability of failure (HCLPF value). 

The objectives of this seismic margin review of the Maine Yankee Atomic Power Station 

are to: 

• Provide an assessment of an actual plant's capability to withstand a 

specific earthquake level greater than the SSE. 

• Demonstrate the use of the Expert Panel's approach (NUREG/CR4334) and guidelines (NUREG/CR-4482) for seismic margin review. 

• Provide a basis for upgrading the Expert Panel's approach and 

guidelines, if improvements are needed. 

• Provide a benchmark for possible future seismic margin reviews, 

including cost and time resources. 

The overall approach for this seismic margin review, as detailed in NUREG/CR-4482 

(Prasslnos et al., 1986), was to: 

• Select a review level earthquake. 

• Determine the systems, components, and structures that are 

important to plant seismic capacity by applying a successive set of 

screening filters based on system function, nonselsmlc unavailability, 

and seismic fragility. 

• Use plant walkdowns to gather information for the screening process, 

and for the determination of fragilities and plant seismic capacity. 

• Model the system and plant response to the review level earthquake 

through fault trees and event trees, and develop a Boolean equation 

for the endpoint of core damage. 

• Calculate the plant seismic capacity (HCLPF) using the Boolean 

equation and fragility information. 

This report is Volume 2 of a three-volume report documenting the project results. 

Volume 1 provides the overall project results, and Volume 3 provides the fragility and 

HCLPF analyses. Volume 2 provides the systems screening analysis, the logic models of 

the plant and systems, and the plant-level Boolean equation for core damage. Chapter 2 

describes the methodology used in these systems analysis tasks. Chapter 3 provides the 

1-l 
results of the screening assessment and discusses each of the systems, including the 

detailed system fault tree models. Chapter 4- presents the plant event trees, and the 

Boolean equations. The resulting minimal cut sets are also discussed. Comments and 

conclusions on the methodology and seismic margin review process are presented in 

Chapter 5. References are given at the end of the report.. The appendices contain 

definitions of abbreviations and symbols for the fault trees and basic events, and the 

actual system fault trees. 

1-2 
2. METHODOLOGY 

2.1 General Approach 

The approach for the Maine Yankee seismic margins review followed the same eight 

steps outlined in Chapter 2 of NUREG/CR-44-82 (Prassinos et al., 1986), which are 

graphically represented in Figure 2-1 (Figure 2-6 of NUREG/CR-4482). The systems 

analysis tasks involved Steps 2, 4, 5, 6, and 7. The extensive guidance on these steps in 

NUREG/CR-4482 will not be repeated here. Rather, this discussion will focus on the 

plant specific steps and considerations for the Maine Yankee review and any departure 

from the methodology described in NUREG/CR-4482. 

2.2 Step 2 - Initial Systems Review 

The initial systems review consisted of seven tasks: 

• Gather system information, 

• Classify front-line systems (Group A or Group NOT -A), 

• Identify Group A front-line system components, 

• Classify support systems (Group A or Group NOT -A), 

• Identify Group A support system components, 

• Identify any Group A plant unique features, 

• Prepare for first plant walkdown. 

2.2.1 Gather System Information 

An initial information request was made to Yankee A tomlc personnel. The information 

sources received included the Maine Yankee FSAR, systems training manuals, piping and 

instrumentation diagrams, electrical/actuation system schematics, Technical 

Specifications, emergency and abnormal operating procedures, surveillance, tagging and 

temporary modifications control procedures, and various structural specifications and 

calculations. From these sources, and from answers to specific questions asked 

throughout the review, the systems review was conducted. 

2.2.2 Classify Front-Line Systems 

For each initiating event to be considered, the Maine Yankee front-line systems were 

classified as Group A or Group NOT -A according to the safety functions listed in Table 

2-1 (Table 2-2 of NUREG/CR-4482). For a PWR, the Group A safety functions are 

reactor subcri ticality, normal cool down, and emergency core cooling (early). Based on 

guidance in NUREG/CR-4482, a seismic induced loss of offsite power (LOOP) is assumed 

for a margins review. With a LOOP, the normal cooldown systems would not be 

available, and are therefore not considered further. In addition, an initiating event which 

includes a small LOCA due to the seismic event, or a nonseismic event such as a reactor 

coolant pump seal LOCA, or a safety /relief valve stuck open was considered. Based on 

guidance from NUREG/CR-44-82 medium and large LOCAs were not considered. 

The impact of a containment isolation failure during no LOCA and small LOCA events 

was considered. It was deemed to be not important to core damage frequency for a 

LOOP transient, although it could affect the accident source term and offsite dose. For 

2-1 
a small LOCA, failure of containment isolation could affect recirculation by an eventual 

loss of inventory in the containment sump, but this would only occur in the very long 

term. Given an emergency source of ac power, recovery actions such as refilling the 

R WST or manually isolating containment are likely. Therefore, potential impacts of 

containment isolation failure were not considered further .. 

From an understanding of Group A systems and chosen initiating events, preliminary 

event trees were developed to show the potential accident sequences leading from an 

initiating seismic event to core damage. Guidance in NUREG/CR-4482 suggested that 

event trees and schematics from the NRC sponsored Accident Sequence Evaluation 

Program (ASEP) could be used. However, the ASEP event trees and system schematics 

were not available for Maine Yankee, and the ASEP generic trees were too nonspecific 

for use in this analysis. Therefore, methods for event tree development from the !REP 

procedures guide (Carlson, 1983) and the PRA procedures guide (NRC, 1983) were used. 

2.2.3 Identify Group A Front-Line Components 

Once the Group A front-line systems were determined, the components within these 

systems were identified. For each component, its function, necessary support systems 

such as power or cooling, normal status (operating, standby, open, closed), and failure 

mode (if applicable) were noted. Components of Group A systems that were not required 

to support the safety function were generally screened out. However, components which 

did not support the Group A function of the system but whose failure could cause the 

failure of a necessary component were identified. A primary example is a heat 

exchanger that is not required to remove heat during an accident, but is required to 

maintain its pressure boundary integrity in order to assure system success. Components 

located on potential flow diversion paths from the main piping were also noted. Flow 

diversion paths less than one-third the size of the main piping or with flow restriction 

orifices were screened out only for open systems. For closed systems, all diversion paths 

were considered. Simplified schematics for the systems and components were then 

developed for use in the first plant walkdown. More components were shown on the 

simplified system schematics than would be evaluated, such as manual valves, check 

valves and recirculation lines. These components were included because they were 

needed to understand the system operation. All questions which arose and assumptions 

made were recorded for discussion during the walkdown. 

2.2.4 Classify Support Systems 

Those systems which provide a required support function for the operation of a Group A 

front-line system were classified as Group A. In turn, some of the Group A support 

systems required the functions of other support systems, and these other support systems 

were also classified as Group A. From the components identified in the prior step, a 

front-line system versus support system dependency matrix was developed. 

2.2.5 Identify Group A Support System Components 

In a manner similar to that described for Group A front-line systems, the Group A 

support system components were identified. Those support system components which do 

not support the Group A function of the system were screened out, unless their failure 

could cause the failure of a necessary component. The resulting list of support system 

components was used in developing a support system versus support system dependency 

matrix. Simplified schematics were also developed for each Group A support system. 

2-2 
2.2.6 Identify Plant-Unique Features 

Any features (systems or components) unique to the Maine Yankee plant that could 

potentially affect the seismic margin of the plant were identified to be considered along 

with the Group A front-line and support systems. An example of a plant-unique feature 

is the dam and fire water pond at Maine Yankee. 

2.2.7 Prepare for First Plant Walkdown 

In preparation for the first walkdown, several items were submitted to the fragilities 

analysis team as the system documentation became available. These items included the 

Group A component lists, the simplified system schematics, the preliminary event trees 

and the lists of the questions and assumptions to be discussed. The lists of components 

designated each component as required for Group A system operation, or as required for 

pressure boundary integrity only. Some of the system assumptions and bases listed were 

based on knowledge of other PRAs and required verification for Maine Yankee. At this 

point some alternate systems (such as the alternate shutdown decay heat removal 

Appendix R system) were still being evaluated and were included on the preliminary 

event trees prior to being screened out. 

2.3 Step 4 - First Plant Walkdown 

The first plant walkdown provided an opportunity to accomplish three tasks: 

• Hold a Peer Review Group meeting 

• Conduct the plant walkdown 

• Hold discussions with plant staff 

Each of these tasks contributed to the systems analysis. Documentation of the first 

walkdown included marked up schematics, revisions to system component lists, answers 

to questions, verification of assumptions, and new system information. 

2.3.1 Peer Review Group Meeting 

During the Peer Rev lew Group meeting, clarification on the treatment of several open 

items was sought. These items included whether or not to evaluate the boric acid 

transfer system for the reactor subcriticallty function, long-term room cooling for 

pumps, and the lni tial switchover to recirculation from high pressure safety injection. 

2.3.2 Plant Walkdown 

Several systems analysis tasks were to be accomplished in the first walkdown. Most 

important of these were: 

• Verification of screening completed thus far using a checkoff list. 

• Identification of the relationships between component layout and the 

timing for local recovery actions. 

• Location of some components, such as the diesel generator start 

circuits and load shed/ sequencers. 

2-3 
\. 

• Identifica tlon of the impact of potential failures of the system 

components. 

• Identification of possible physical interactions between Group A 

system components and other items (block walls, fire water). 

• Walkdown of second.ary component cooling water piping to check 

integrity and potential interactions. 

In performing the walkdown it was useful to have the plant arrangement drawings in 

addition to the simplified system schematics. As the walkdown was conducted with the 

analysis team members divided into two groups, it became important for the groups to 

consult at the end of each day to compare notes and prepare for the next day. 

2.3.3 Plant Staff Discussions 

Discussions held with members of the plant staff were to gain additional system 

information and answers to the questions existing at that point. Information on valve 

failure positions, normal positions of manual valves, unidentified instrument air and 

power supplies, and operating procedures would be used in completing the systems 

doc;:umentation. Also, simplified system schematics were verified and plant-unique 

features discussed. 

2.4 Step 5 - Systems Modeling 

To complete the systems modeling task it was necessary to: 

• Review, update, and docume'nt the event trees and their success 

criteria. 

· .. Develop and document fault trees for the front-line systems which 

are included in the event tree sequences, and for their required 

support systems. 

• Develop a data base cons1st1ng of probability cutoffs for screening, 

and data for calculating component unavailabilities and human error 

probabilities. 

• Determine the minimal cut sets for the front-line systems (including 

support system faults) to verify the fault tree logic and identify 

critical items to review during the second plant walkdown. 

2.4.1 Review Event Trees 

During the first plant walkdown, several insights were gained which were used in revising 

the preliminary event trees. The types of revisions are illustrated by the following: 

• The recirculation function of the containment spray pump? (which 

provide water from the containment sump through the residual heat 

removal (RHR) heat exchangers to the high pressure safety injection 

(HPSI) pumps for long-term core cooling) were included. Specifically, 

the fans, ducting and dampers which provide room cooling for the 

containment spray pump area were added to the HPSI system fault 

tree. Thus, failure of room cooling would eventually fall HPSI in the 

recirculation mode. 

2-4 
;~ ~ •• f ,. :~ 

• The alternate shutdown decay heat removal (ASDHR) system was 

screened out of the analysis. The main control room panels were 

judged to be seismically sound, precluding the need for use of the 

alternate shutdown panel (ASP). Also, operation of the turbine ... dri ven 

auxiliary feedwater (AFW) pump is required for ASDHR, and is 

already considered in the AFW system. Finally, the auxiliary charging 

pump is judged too small to replace the HPSI pumps for feed and 

bleed or core inventory makeup. 

• The evaluation of the boric acid transfer (BAT) system was delayed 

and eventually screened out, because the reactor internals and control 

rod drive system were found to have high seismic capacity, thus 

eliminating the need for BAT. 

• An anticipated transient without a SCRAM (ATWS) was not evaluated 

because the reactor internals and control rod drive system were found 

to have high seismic capacity. 

Two separate event trees were developed, with different initiating events. One tree 

evaluates mitigating system success following a seismic event causing loss of offsite 

power (T 1 ). The second tree addresses system success following a seismic event which 

causes loss of off site power, concurrent with a small LOCA (5 2). 

The final event trees are discussed in greater detail in Section 4.1. 

2.4.2 Develop Fault Trees 

Once the functions, success criteria, and support system interfaces for the Group A 

systems were identified, system fault trees were developed using standard techniques. 

For each system, potential flow diversion paths were analyzed. A path was excluded if it 

met at least one of the following criteria, which are based on probabilistic arguments. 

• There is a normally closed automatic valve isolating the line. 

• 

Exception: Include the path if interlocks exist which may prevent the 

valve from closing, and these interlocks are connected to a 

component or system which is modeled, or the interlock and valve are 

powered by opposite buses. 

There are two normally open valves on the line which will 

automatically close. 

Exception: Include the path if the line up to the second valve has a 

potentially low seismic capacity (i.e., contains a heat exchanger or is 

not seismic class 1 ). 

• The line or flow restriction orifice is less than one-third the diameter 

of the main pipe. 

Exception: Include the path if it is a closed system. However, the 

ability of the operator to isolate small leaks will also be taken into 

consideration. 

During this process, additional questions and issues were raised. A list of these questions 

was compiled for referral to the plant staff. It was necessary to include some diversion 

paths until information was received which would allow screening of the path. 

2-5 
The components and paths identified by the above steps were used in developing the 

system fault tree logic. The fault tree symbols and event identifiers used are found in 

Appendix A. The first set of front-line and support system fault trees developed include 

every system component identified (unless it was screened out as a diversion path 

component). These fault trees would later be pruned using failure probability cutoffs. 

As seismic and nonseismic failure modes were not differentiated at this point, a specific 

component failure mode was not identified. Rather, an "XX" was used as a generic 

failure mode identifier. Support system faults are added to the tree as developed 

events. The support system fault trees would later be merged into the front-line system 

fault trees at these points. Fall ures of support systems were included at each component 

level rather than being combined at the top of the fault tree. Check valves were not 

included in the fault tree logic based on their low unavailability. Manual valves normally 

in the correct position were also not included in the fault tree logic because of their low 

probability of being in an incorrect position. However, if a manual valve could be in an 

incorrect position and also fail all trains of a redundant safety system, then its 

probability of incorrect positioning was evaluated further to determine if it was below 

the screening probability cutoff. 

Once the initial system fault trees were completed, preliminary front-line and support 

system cut sets (first, second, and third order) were obtained using Micro-PRANK. These 

cut sets were analyzed to verify the logic of the fault tree. 

Following verification of the fault tree logic, probability cutoffs described in the 

following section were used to prune the fault trees. Seismic, random, common cause, 

test and maintenance, and human error failures were considered for each component. 

Along with seismic failures of specific components, seismic failures of adjacent 

components, walls or buildings which may fail a component of interest were also 

considered. For seismic events screened in, the failure mode identifier "EQ" was used. 

The same event names were used for several component failures if the seismic failure of 

the components was highly correlated. For example, CCW-HTX-EQ-4B5A represents the 

seismic failure of the PCC and the SCC coolers since they are identical and located 

adjacent to each other. All components with seismic HCLPFs greater than 0.3g, or 

non seismic failure probabill ties below the probability cutoffs were pruned from the fault 

tree. The support system inputs for these components remained on the tree. It took 

several passes to prune the fault trees as fragility calculations were finalized and system 

spatial dependencies were identified in the second walkdown. 

Also pruned from the fault trees were any components which are isolated by valves 

removed from the tree in the step described above. This eliminated flow diversion paths 

which would be successfully isolated following the initiating event. Support system 

interfaces for these pruned diversion path components were also removed from the 

tree. The separation of instrumentation, racks, and impulse lines which make up the 

actuation systems included in the fault trees was also checked to determine if a single 

physical seismic interaction could fail multiple actuation channels. If there was 

sufficient separation (no more than one channel is failed by a physical seismic failure), 

and the transmitters and racks were of high seismic capacity, then actua tlon system 

faults were pruned from the fault trees. 

Upon completion of the pruning process, the fault tree "OR" gates with zero or one 

remaining inputs were collapsed into their output gates. The "AND" gates which had one 

or more inputs removed were deleted from the tree. The pruned support system fault 

trees were then merged with the pruned front-line system fault trees. Developed events 

on the trees became the top gate of the support system tree to be merged at that point. 

2-6 
.. 

The final r.esult of the fault tree development process was a fault tree for each front-line 

system included in the event tree sequences. Each fault tree includes basic and 

undeveloped events representing the seismic and nonseismic failures of the front-line and 

support system components which were not screened out. 

2.4-.3 Develop Data Base 

A data base containing probability cutoffs and nonseismic failure data was developed to 

aid in pruning the fault trees. Seismic failure data from the fragilities team which was 

required in the component screening was included in the data base. 

Generic component unavailability and failure rate data from ASEP was used to calculate 

random nonseismic failures. Beta factors from EPRI NP-3967 (Fleming, 1985), with 

supplements from ASEP and other PRAs were used to develop nonseismic common cause 

unavailabilities for components. A list of components for which information on testing 

and maintenance is required was compiled for verification on the second walkdown. All 

of the components were adequately represented by generic ASEP test and maintenance 

unavailabilities. To determine probabilities for human errors and recovery, time based 

data from the IREP NUREG/CR-2787 (Kolb, 1982) with some guidance from ASEP, was 

used. 

In screening system components, seismic failures for those components with a HCLPF 

capacity calculated to be greater than 0.3g were screened out. Random, common cause, 

test and maintenance, and human error failures with probabilities less than the following 

guidelines were also screened out. 

• 0.01 if the failure leads to the loss of only one train in one system. 

• 0.001 if the failure leads to the loss of all trains in one system. 

• 0.001 if the failure leads to the loss of one train in multiple systems. 

• Exception: If the failure probability is greater than the cutoff, but 

has a high probablli ty of recovery, making the combined probability 

less than the cutoff, the component failure may be screened out. 

Component unavailabilities, human error probabilities, and nonseismic common cause 

failure probabilities are presented for the components of interest in Chapter 3. 

2.4-.4- Determine System Cut Sets 

The final front-line system fault trees obtained from the steps described in Section 2.4-.2 

(pruned and merged with the support system fault trees) were analyzed to determine the 

first, second, and third order cut sets for the system. This minimal cut set evaluation 

before the second walkdown was not included in the NUREG/CR-4482 steps, but proved 

to be very useful. Micro-PRANK, a minimal cut set routine included in the personal 

computer based Fault Tree Workstation, was used for this analysis. The resulting cut 

sets aided in recognizing some critical items (especially unexpected single faults). These 

items were identified as warranting a special look in the second walkdown, and discussion 

with the plant staff. Among the items listed for further evaluation during the second 

walkdown were the new procedure for isolating portions of the PCC, the station 

transformers, and the containment spray pump area fans. 

2-7 
2.5 Steo 6 -Second Plant Walkdown 

A second plant walkdown was performed to look at those items that had been added to 

the component list since the first plant walkdown. This included items such as block 

walls which may lm pact the major system components. Also, any questions which arose 

during the fault tree development task were discussed with plant personnel and, if 

necessary, a second examination given to the subject items. Plant-specific test and 

maintenance (planned and unplanned) data for important components (pumps and diesel 

generators primarily) was also gathered. 

The second plant walkdown also provided an opportunity to discuss the seismic review 

methods and any issues which had arisen with the Peer Review Group. It was also a 

chance to support the fragility analysis team in their efforts to calculate component 

capacities, and to obtain a preliminary estimate of which components may be screened 

out due to sufficient seismic capacity. 

2.6 Step 7 - Determine Minimal Cut Sets 

In order to obtain. the minimal cut sets and Boolean equations for the two event trees (no 

LOCA and small LOCA) the following steps were completed: 

• Rev lew and finalize the event trees. 

• Review and finalize the fault trees. 

• Link the fault trees according to the event tree sequences. 

• Obtain a preliminary Boolean equation for each event tree. 

• Repeat these steps to incorporate the complete fragility analysis 

results to obtain the final Boolean equations. 

2.6.1 Finalize Event Trees 

The event trees produced in Step 5 were reviewed to determine consistency with the 

additional information gathered during the second plant walkdown. 

2.6.2 Finalize Fault Trees 

The pruned and merged front-line system fault trees developed in Step 5 were also 

reviewed to incorporate the answers to the questions discussed during the second 

walkdown, the test and maintenance data, and the preliminary fragility analysis results. 

Screening overview tables were developed to trace the status of each system 

component. For each item, the tables list the component name, its screening status (in 

or out, for seismic and nonseismlc failures), the reason for its screening status, the 

applicable event name, and the component unavailability or seismic capacity value used 

for comparison with the probability and HCLPF cutoffs. 

These finalized pruned fault trees were used in determining the Boolean equations. As 

discussed in NUREG/CR-4482, it is possible to determine preliminary Booleans prior to 

pruning the system fault trees, but this would result in roughly an order of magnitude 

more cut sets. The method of using pruned fault trees was chosen because 'it is more 

efficient. 

2.6.3 Link Fault Trees 

The fault trees were first linked to form sequence level fault trees, and then to form 

event tree level fault trees. The pruned front-line system faui t trees which had been 

2-8 
merged with the required pruned support system fault trees were linked in combinations 

which represent the various event tree sequences. For example, if failure of auxiliary 

feed water along with failure of the high pressure safety injection system would result in 

core damage, these two system fault trees were linked by an "AND" gate to create a 

larger fault tree which represents this core damage sequence. 

Once fault trees were developed for each sequence, the sequence level trees were linked 

to create a fault tree which represents an entire event tree. For example, given an 

initiating event, if there are three sequences, of which any one will lead to core damage, 

the fault trees for these three sequences were linked by an "OR" gate. These linked fault 

trees were then analyzed to determine the Boolean equation for each event tree .. 

2.6.4 Determine Preliminary Boolean Equation 

Analysis of the linked fault trees was performed in two stages to determine the Boolean 

equations. A first analysis was performed with the analysis routine Micro-PRANK. The 

linked sequence level fault trees were analyzed separately for each sequence. 

Once cut sets for each sequence were obtained, they were combined and reduced by hand 

to develop the cut sets for an entire event tree. In doing this, it was important to ensure 

that only minimal cut sets were included for the event tree level result. Since it was 

possible that an event that occurs as a single fault for one sequence appeared in a double 

fault for another sequence, when these sequence cut sets were combined the double 

faults containing that event were deleted. 

After event tree level cut sets were determined in this manner, they were transferred to 

the fragility analysis team for preliminary plant HCLPF determination. Later, the large 

event treee level fault trees were analyzed using VAX SETS to obtain more detailed 

Boolean equations. These equations were used to verify the results obtained from MicroPRANK. Because this is a seismic review, only the cut sets which include at least one 

seismic event are of interest. Therefore, those cust sets that contained only nonseismic 

events (e.g., random, human error) were deleted from the Boolean equations. 

2.6.5 Determine Final Boolean Equation 

When the component fragility analysis calculations had been finalized, those results were 

. incorporated in the plant level Boolean equations. To do so, the steps described in 

Sections 2.6.1 through 2.6.4 were repeated. This resulted in a change in screening status 

for several components, which in turn altered the final Boolean. The final Boolean 

equations were used by the fragility analysis team in determining the final plant 

capacity. 

2-9 
Table 2-1 .. Definition of plant safety functions. 

(Table 2-2 from NUREG/CR-4482) 

IDENTIFICATION OF SAFETY FUNC~IONS 

1. Reactor Subc::::- iticali tv - shutting down the nuclear reaction such that 

the only heat being generated is decay heat. 

2, Normal Cooldown - providing cooling to the reactor core through the use 

of the normal power conversion system, normally defined as the main 

steam, turbine bypass, condenser, condensate, and main feedwater 

subsystems .. 

3. Emeraencv Core Coolino {Earlv) - providing cooling to the reactor core 

in the early (transient) phase of an event sequence by the use of one 

or more emergency systems designed for this purpose. The exact timing 

of "early" is somewhat plant specific and sequence dependent. However, 

for our purposes it can be deemed to be the time period during which 

these systems are initially called upon to operate. 

4. Emeraencv Core Coolino {Late) - providing cooling to the reactor core 

in the late (stabilized) phase of an event sequence by the use of one 

or more emergency systems designed for this purpose. In context with 

the above definition of "early", for our purposes "late" can be deemed 

to begin with the switchover to recirculation (for LOCAs) or with the 

achievement of residual heat removal conditions (for transients). 

5. Containment Heat Removal - removing heat from the containment to the 

ultimate heat sink during the late (stabilized) phase of an event 

sequence by the use of one or more safety systems designed for this 

purpose. 

6. Containment Overnressure Protection (Earlv) - controlling the buildup 

of pressure in the containment caused by the evolution of steam by 

condensing this steam during the early phase of an event sequence by 

using one or more safety systems designed for this purpose. "Early" in 

the context of containment functions is not the same as "early" for 

core cooling. In this case "early" is deemed to be the time period 

commencing when this function is required.. after the beginning of core 

melt when these systems are operating in the injection mode. 

7. Containment OverPressure Protection {tate) -controlling the buildup of 

pressure in the containment caused by the evolution of steam by 

condensing this steam during the late phase of an event sequence using 

one or more safety systems designed for this purpose. In the context 

of the previous definition, "late" in this case is deemed to start when 

these systems are operating in the recirculation mode. 

IDENTIFICATION OF THE FUNCTIONAL GROUPS FOR PWRS AND BWRS. 

PWR 

Group 

Group 

BWR 

Group 

Group 

A: 

NOT-A: 

A: 

NOT-A: 

Functions 1,2,3 

Functions 4,5,6,7 +All plant functions not related 

to Safety 

Functions 1,2,3,4,5,6, 7 

All plant functions not related to Safety 

2-10 
Gather information 

on systems and sort 

Group A functions. 

Use information 

on Table 2.3 and 

Ref. 1. 

Start 

Time axis 

Select an earthquake review level 

Gather information on 

the plant. Determine which 

Interaction broad classes or groups of 

----- components have HCLPF 

values greater than the review 

level. Possibly identify plantunique features. 

First plant walkdown: 

Concentrate on identification of problems. 

Emphasize systems interaction. Confirm 

applicability of screening tools. Complete 

identification of plant-unique features. 

KEY: Task is performed by: ·.::·::·· ·:::~: • Revision of systems D Systems Analyst :::: ~: ~·: :·.: :·: relationships established 

·.: •: •• ::. •• :. in Step 2. Develop fault 

·. · : < ·: • · · -: · trees and event trees. 

~ Fragility Analyst 

.. . . . . 

rl.i.E Both 

Determine minimal 

cut sets tor endpoint core melt. 

Figure 2-1. 

: :. ·: =·: :~ : ··~ ~ .. :.: .: :: .. ·: ·. : ... · .. : .. .. . .. . .... . . ·.··=.·7··:··.·. 

:·:·.:=:··.::= .. ·· .. . . . . .. . .. . . · ... . . . . .. . 

Margin 

assessment 

complete . 

Second plant walkdown: 

Primarily fragility analyst for checks. 

Collect speciiic data (size and other 

physical characteristics) of components 

requiring detailed analysis. 

Finalize HCLPF value for 

components in final cut sets 

(components not screened out). 

Graphic Representation of the Screening Operations 

(Figure 2-6 from NUREG/CR-4482). 

2-11 
.., . . ~ .. ' 
I' I ~ . ~' • 

3. SYSTEMS ANALYSIS 

This chapter presents the results of the systems modeling and quantification steps. 

Section 3.1 discusses the identification of Group A systems, both front-line and support. 

It also presents the basis for screening out potential Group A systems from further 

evaluation. Each Group A front-line system is then described in Section 3.2, and the 

fault trees and screening evaluations are presented. Section 3.3 provides similar 

information for the Group A support systems. The failure probability and unavailability 

bases for the components that could not be screened out immediately is presented in 

Section 3.4. 

3.1 Svstem Identification 

The purpose of the system identification and classification was to determine which plant 

systems are potentially available to bring the plant to a safe shutdown following a 

seismic event. As discussed in Chapter 2, those front-line systems (with the necessary 

support systems) whose failure will result in failure of reactor subcriticality or loss of 

emergency core cooling (early) are designated as "Group A" systems. Unless there are 

plant unique features or other reasons for inclusion, all other systems are designated as 

"Group Not-A" systems. Failures or successes of the Group Not-A systems are not 

evaluated. 

3.1.1 Group A Systems 

Those systems identified as Group A which perform the reactor subcriticality and early 

core cooling functions are the reactor protection system (control rods and reactor 

internals), the auxiliary feedwater system, the primary pressure relief system (poweroperated relief valves for feed and bleed), and the high pressure safety injection system. 

Based on the fragility analysis of the reactor internals and the generic HCLPF for the 

control rod mechanisms, the seismic capacity of this reactor subcritlcality system was 

found to be adequate at the review earthquake level. Nonseismic failures of the system 

are below the probability screening cutoffs. Due to these findings, fault trees of the 

reactor internals and control rod mechanisms were not developed, and reactor 

subcri ticality systems would not impact the plant HCLPF calculation. 

Upon loss of the main feedwater system (due to loss of offsite power) and a 

corresponding drop in steam generator level, the auxiliary feedwater (AFW) system 

supplies water from the demineralized water storage tank or the primary water storage 

tank to the steam generators for decay heat removal. A minimum level must be 

maintained in at least one of the three steam generators for successful heat removal. 

The power-operated relief valves (PORV), which with the safety relief valves comprise 

the primary pressure relief system, are used for bleeding the pressurizer in the reactor 

coolant system (RCS) during feed and bleed core heat removal. The PORVs are actuated 

by the operator for this function. If the seismic event is accompanied by a small LOCA, 

opening of only one PORV would provide sufficient feed and bleed capability. Otherwise, 

both PORVs must be opened. 

The high pressure safety injection (HPSI) system is also required for feed and bleed 

capability, and for core cooling during a LOCA. The HPSI system provides borated water 

3-1 
from the refueling water storage tank to the reactor coolant system loops for makeup 

during feed and bleed, and core cooling. Also included in the HPSI analysis are the fans 

which provide cooling for the containment spray pump area, and the valves which are 

required to operate to achieve recirculation. Although recirculation is not a Group A 

function, it is necessary to ensure the long-term operating ability of the containment 

spray pumps for emergency core cooling during recirculation. This departure from the 

NUREG/CR-4334 and 4482 guidelines is discussed in Chapter 5. 

The necessary support systems identified for the Group A front-line systems are electric 

power (ac, de, and the emergency diesel generators), component cooling water (primary 

and secondary), service water, and actuation (primarily the safety injection actuation 

system, SIAS). Some air accumulators are also· required, but are included with the frontline systems they support. The front-line to support system relationships are shown in 

the dependency matrix in Table .3-l. The support system - support system dependencies 

are shown in Table 3-2 .. 

3.1.2 Systems Removed from Group A 

Some systems which might be called upon to perform a required Group A function are the 

boric add transfer system, the low pressure safety injection system, the safety injection 

tanks, the secondary pressure control system, and the alternate shutdown decay heat 

removal system. This section discusses the basis for not evaluatin~ these systems in 

detail. 

The boric acid transfer (BAT) system provides emergency boration from the boric acid 

storage tank to the R~S in the event of an A TWS or loss of shutdown margin. As the 

probability of failure of the reactor internals due to the review level earthquake or 

nonseismic event is insignificant, emergency operation of the BAT system is not likely to 

be required for prevention of core damage, and was excluded from further consideration. 

The low pressure safety injection (LPSI) system and the safety injection tanks (SIT) 

provide sufficient borated water to flood and cool the core following a medium or large 

LOCA. As discussed in Chapter 2, only a small LOCA is postulated as concurrent with 

the seismic event. Therefore, the LPSI and SIT systems would not be required for core 

cooling. The feed and bleed actions by the POR Vs and the HPSI system provide 

sufficient core cooling for transients or a small LOCA. Also, the LPSI pumps are not 

used in the recirculation mode at Maine Yankee. 

The secondary pressure control system provides overpressure protection for the steam 

generators and main steam system piping by means of safety relief valves and an 

atmospheric steam dump valve. The safety relief valves are expected to open for 

pressure relief, and it is not of major concern if they should stick open, although the RCS 

would rapidly depressurize. If the HPSI system should fail, the atmospheric dump valve 

may be used to depressurize and cool the secondary system which in turn would cool the 

primary side, enabling the use of the LPSI system for depressurization. However, with 

the reactor coolant pumps made inoperable by loss of offsite power, there is no assurance 

that sufficient depressurization would occur (Fletcher, 1981). The atmospheric dump 

valve may also be used for depressurization to enable the use of a low pressure water 

supply (such as fire water) to feed the steam generators in the event of AFW £allure. 

However, the procedure for this operation depends on use of the AFW system at the start 

of the procedure for initial cooldown. Due to these usage limitations and dependencies, 

the secondary pressure control system was not evaluated further. 

3-2 
"... ' ~~' 

"o.'>., • I: ~ •. 

The alternate shutdown decay heat removal (ASDHR) system provides an alternate means 

of controlling and monitoring a plant shutdown, and was designed to meet the Appendix R 

requirements for fire mitigation. The alternate shutdown panel (ASP) provides a remote 

monitor and control location, but as the main control room (MCR) panels were found to 

be of sufficient seismic capacity, use of the ASP is not necessary. The ASDHR system 

utilizes the turbine-driven AFW pump and the auxiliary charging pump to achieve 

shutdown. The turbine-driven pump is analyzed with the Group A AFW system, and the 

auxiliary charging pump is of too small a capacity to replace the HPSI pumps for feed 

and bleed or coolant makeup operations following a small LOCA. Due to these 

limitations of the ASDHR system, it is not analyzed further as a Group A system. 

3.2 Systems Analysis of Front-Line Svstems 

Information on the Group A front-line and support systems includes the Final Safety 

Analysis Report (FSAR), system piping and instrument diagrams (P&IDs), standard and 

abnormal operating procedures (SOPs and AOPs), system training manuals, and system 

analyses and calculations. Maine Yankee and Yankee Atomic personnel are a major 

source of information for questions and clarification. 

Major i terns covered in the following system analysis discussion include the function of 

the system, success criteria, system components, boundaries of the system, support 

system interfaces, operator and recovery actions, necessary assumptions, bases for fault 

tree development, fault tree logic, failure probabilities of events, fault tree pruning and 

merging logic, and system cut sets. 

3.2.1 Auxiliary Feedwater System 

The auxiliary feedwater system is used to maintain a m1n1mum level in the steam 

generators for decay heat removal, following loss of the main feedwater system. For 

system success, operation of at least one of the three AFW pumps is required to maintain 

a minimum water level in at least one of the three steam generators. The AFW system 

consists of the demineralized water storage tank (DWST), three pumps (two motor-driven 

and one turbine-driven), and associated flow control and isolation valves. Although the 

motor-driven pump trains at Maine Yankee are termed emergency feedwater, they will 

be classified as part of the auxiliary feedwater system in this report. A complete list of 

AFW components, valves and cooling requirements is included in Appendix B. 

The two motor-driven pumps (located in the auxiliary feed pump house, el. 21 ft) start 

automatically on a steam generator low level signal. Upon loss of offsite power, the 

diesel generator load sequencers restart the motor-driven pumps after a 20-second time 

delay. If necessary, the turbine-driven pump (located in the steam and feed pump valve 

area, el. 21 ft) is placed in service by the operator by aligning the steam inlet valves to 

the turbine drive. Flow to each steam generator is regulated by an air-operated flow 

control valve (located in the auxiliary feed pump area, el. 23 ft) which is paired with an 

air-operated isolation valve. A schematic of the AFW system is shown in Figure 3-1. 

Support systems required for AFW operation are 4160-V ac emergency power, 125-V de 

power, instrument air (from various accumulators), and main steam to drive the turbine. 

The following items were used as the basis for development of the fault tree logic: 

• An alternate water supply for the AFW system is the primary water 

storage tank (PWST). However, there is no check valve on the line 

between where the PWST feed joins the suction for the turbine-driven 

3-3 
pump and the DWST, making it possible for the P\VST to be drained 

through a ruptured DWST. Therefore, the P\VST is considered an 

alternate supply only for the motor-driven pumps. 

• A potential source of makeup to the DWST is from condensate 

makeup. As this makeup is supplied by gravity feed to the DWST, it 

could also be drained by a DWST rupture. Therefore, this source of 

makeup is not included in the analysis. 

• Sufficient separation of the steam generator instrumentation exists, 

such that a single physical seismic failure would not fail the complete 

AFW instrumentation and actuation systems. Based on the high 

seismic capacity of the racks and transmitters, the instrumentation is 

not analyzed further in the analysis. 

• The turbine-driven AFW pump may be placed in service from the main 

control room or locally. Since the diagnostic portion of these two 

actions are related, this dependency is explicitly modeled in the fault 

·tree. If the operator fails to place the pump in service from the 

control room, it is not likely that he will do so locally. 

• Check valve failures, random valve and MOP failures, and valve 

failure due to plugging, testing or maintenance are of low 

probability. Therefore, these failures are screened out of the 

analysis. 

• Because the AFW flow control and isolation valves fail open upon loss 

of instrument air or solenoid power, the related air accumulators and 

power supplies are not included in the fault tree. 

• Because the AFW pump oil coolers are considered an integral part of 

the pumps, they are not shown as separate components in the fault 

tree. 

• Several recovery actions are not included in the fault tree. These 

actions in.clude opening the manual flow control bypass valves to 

allow flow, the use of fire water to maintain water level in the steam 

generators, and the potential recovery of the main feedwater 

system. Inclusion of these recovery actions is judged not to impact 

the plant HCLPF significantly. · 

• The recirculation lines from the AFW pumps to the DWST are not 

included as potential flow diversion paths. Each recirculation line 

contains a flow restriction orifice and is less than the screening value 

of one-third the size of the discharge line. 

• The lines off the AFW pump discharge to the main feedwater header 

are not included as potential flow diversion paths. Each line contains 

a normally closed manual valve .. 

• The line containing the chemical feed tank (TK-89) and pump (P-115) 

is not included as a potential flow diversion path. The line is less than 

one-third the size of the AFW pump discharge line and contains a 

normally closed manual valve. 

3-4 
''' 

• Of the valves which supply steam to the turbine-driven pump, two 

(MS-P-168 and MS-T-163) are air-operated and receive air from the 

same accumulator (TK-25). The third (MS-A-173) is mechanically 

operated and is locked open. 

The basic fault tree for the AFW system is shown in Appendix B. This fault tree includes 

the AFW components, which have not been screened out at this poiht, and their required 

support system inputs.. The faults for each component are nonspecific as to the type of 

failure. Specific failures (e.g., seismic, random, common cause) were added .as each 

component was analyzed. Linear sections of the system are represented by pipe 

segments {PS-1, PS-2, etc.). The top event is failure of the AFW system, this failure 

occurs through common cause failure of the AFW, failure .of the main control room 

panels, or failure within the system to maintain a minimum level within any one of the 

steam generators. This latter event is caused by lack of flow through all three pairs of 

control and isolation valves, which in turn may be caused by lack of flow from all three 

AFW pumps. 

From this basic fault tree, screening criteria described in the methodology chapter were 

applied to each component in order to prune the fault. tree. These screening results are 

summarized in Table 3-3. The footnotes for Table 3-3, and for all the other screening 

tables, are listed on Table 3-4. Once the AFW fault tree was pruned, lt was merged with 

the necessary support system fault trees. (Construction and pruning of the support 

system trees is discussed in Section 3.3.) The completed AFW system fault tree is 

presented ln Appendix B. 

Analysis of the pruned, merged fault tree is performed to determine the AFW system cut 

sets. A validity check of these cut sets ensures that the fault tree logic is correct. The 

minimal cut sets (first and second order) for the AFW system are listed ln Table 3-5. For 

ranking purposes, this analysis assumes that the failure probability for each seismic event 

is 1.0. The HCLPF for these events will be developed by the fragility analysis team. The 

calculations for the failure probabilities of the nonseismic events are shown in Section 

3.4. 

The only singlet for the AFW system is AFW-CCF-FC-AFW, nonseismic common cause 

failure of the pumps or air-operated valves. Of the 23 double faults, the first 19 contain 

one or more seismic events. Seismic failure of the DWST will fail the turbine-driven 

AFW pump and the primary water supply for the motor-driven pumps. Failure of the 

PWST fails the motor-driven pump backup water supply. Seismic failure of the 

transformers, PCC/SCC coolers, air conditioner chillers, circulation water pump house, 

DG day tanks, or failure to refill the DG fuel tanks will result in loss of power to the 

motor-driven pumps, as will common cause failure of the DGs. 

3.2.2 Hlgh Pressure Safety Injection System 

The high pressure safety injection system is used to supply makeup to the reactor coolant 

system for post-accident core cooling and during feed and bleed. For system success 

injection of borated water from the refueling water storage tank (R WST) to at least one 

RCS loop by operation of at least one HPSI pump is required. The HPSI system consists 

of the R WST, three motor-driven pumps (one is an installed spare), two pairs of R \VST 

isolation valves, two pairs of pump discharge valves, and three pairs of injection isolation 

valves (one pair per RCS loop). Complete lists of the HPSI components, valves and 

cooling requirements are included in Appendix C. 

3-5 
Also included \Vith the HPSI system are the two fans that provide cooling for the 

containment spray pump area. Although the containment spray pumps are used for 

recirculation, which is not a Group A function, it is necessary to ensure the long-term 

availability of these pumps. In part, this is accomplished by maintaining a cool operating 

environment. This departure from the NUREG/CR-4482 screening guidelines is discussed 

in Chapter 5. 

A schematic of the HPSI system is shown in Figure 3-2. One HPSI pump (P-14A or B, 

located in the primary auxiliary building (PAB), el. 21 ft), is normally operating as a 

charging pump, the other (standby) pump is automatically started by a safety injection 

actuation signal (SIAS). The third pump (P-145, also in the PAB) is a spare that must be 

placed in service manually. Upon loss of offsite power, the diesel generator load 

sequencers restart the two in-service pumps. The pump suction valves on the R WST 

discharge open on a SIAS. One motor-operated valve of each pump discharge valve pair 

(HSI-M-41 and 42, located in the PAB, el. 23ft) is opened upon a SIAS. 

Support systems required for HPSI and containment spray pump area fan operation are 

4160- and 480-V ac emergency power, 125-V de power, primary and secondary component 

cooling, and safety injection actuation trains A and B. 

The following items provided the basis for the fault tree logic: 

• Failure of the spray chemical addition tank (SCAT), which is 

connected to and located adjacent to the R WST, may lead to failure 

of the interconnecting line and thus drain the R WST. 

• Sufficient separation of the SIAS instrumentation exists, such that 

seismic failure of a rack or instrument would not fail the entire 

actuation system. Also, the operator is capable of initiating the 

actuation system. Therefore, SIAS instrumentation is not included 

further in the analysis. 

• Control power for the motor-operated valves is assumed to be 

transformed off the bus which provides the valve motive power. 

• Check valve failures, random valve failures, and valve failure due to 

plugging, testing or maintenance are of low probability. Therefore, 

these failures were not included in the analysis. 

• Pipe ruptures were included only if screened in during the walkdowns. 

• Because the recirculation mode is not a Group A function, flow from 

the residual heat removal heat exchangers to the HPSI pump suction 

header is not included in the fault tree. 

• Several recovery actions are not included in the fault tree. These 

actions include opening the pump discharge valves which do not open 

on a SIAS, and placing the spare pump in service by opening manual 

valves and racking in the breaker. Their inclusion would not 

significantly impact the plant HCLPF. 

• Block wall VE 21-1, 2 may fail the containment spray pump area fans 

should it collapse. 

3-6 
..... ' ~ ~. . .. ~ 

. ~ . 

• The recirculation lines from the HPSI pumps to the seal water heat 

exchanger and volume control tank are not included as potential flow 

diversion paths. Each line contains a restriction orifice and is less 

than the screening criteria of one-third the size of the discharge line. 

• The lines from the boric acid transfer pump discharge, volume control 

tank and RHR heat exchangers to the HPSI pump suction header are 

not included as potential flow diversion paths. Each line contains a 

check valve to prevent backflow, and either a normally closed motoroperated valve, or a MOV which closes upon a SIAS. 

• The line from the P-14-A suction to the suction of the auxiliary 

charging pump (P-7), and the line from the R WST to P-7 are not 

included as potential flow diversion paths. Each line contains a 

motor-operated valve, and the P-7 discharge line contains a normally 

closed manual valve. 

• The lines from the HPSI pump discharge to the charging header are 

not included as potential flow diversion paths. Each line contains an 

air-operated valve (CH-A-32 and 33) which closes on a SIAS. The two 

lines combine and then split into three lines which go to the charging 

header, loop fill header, and seal water heater. The line to the 

charging header contains an MOV (CH-F-.38) which closes on a SIAS. 

The line to the loop fill header contains a normally closed flow 

control valve (CH-F-70). Downstream of the seal water heater is an 

isolation valve (SL-P-3) which also closes on a SIAS. As CH-A-32 and 

CH-A-33 are seismically sound, they will successfully isolate these 

lines and the regenerative heat exchanger and seal water filter. 

The basic fault tree for the HPSI system is shown in Appendix C. This fault tree includes 

the HPSI components which were not yet screened out and their required support system 

inputs. The type of failure for the components is not specified. Specific failures 

(seismic, random, etc.) were added as each component was analyzed. The top event of 

the tree is loss of HPSI (short-term cooling) or recirculation (long-term cooling). This 

event may occur through failure of the main control room panels, failure of both spray 

pump area fans, or failure of the HPSI system. The latter failure is caused by a system 

common cause failure, or by a failure within the system which prevents injection to the 

RCS system. Loss of injection is characterized by lack of flow to any of the three RCS 

loops, which may be caused by loss of both HPSI pumps. There are several basic events 

involved in each pump and valve failure. 

From the basic fault tree, the screening methods outlined in Chapter 2 were applied to 

each component in order to prune the tree. The results of this screening process are 

detailed in Table 3-6. Once the front-line system tree was pruned, it was merged with 

the pertinent support system fault trees (described in Section 3.3). The final HPSI fault 

tree is shown in Appendix C. 

Once pruned and merged with its support systems, the HPSI fault tree was analyzed to 

determine the system cut sets. These cut sets were used to verify the logic of the fault 

tree. The minimal cut sets and event descriptions for the HPSI system are listed in Table 

3-7. For ranking purposes, seismic failures were assumed at this point to have a 

probability of 1.0. HCLPFs were later assigned by the fragility analysis team. Failure 

probabilities for the nonseismic events have been calculated in Section 3.4-. 

3-7 
'~I '-

.Of the nine single faults for the HPSI system, six are seismic events. Seismic failure of 

the 4160- to 480-V transformers, the PCC/SCC coolers, the air conditioner chillers, the 

circulation water pump house, or the DG day tanks; or failure to refill the DG fuel tanks 

or common cause DG failure leads to loss of power to both HPSI pumps and both spray 

pump area fans. HPSI is also lost upon the seismic failure of the R WST or common cause 

failure. There are two doublets, both of which consist of only nonseismic events. 

3.2.3 Power-Operated Relief Valves (Feed and Bleed) 

The power-operated relief valves (POR Vs) are opened by the operator for feed and 

bleed. This action must take place within approximately 30 minutes of the initiating 

seismic event and loss of AFW. There are two sets of criteria for POR V system 

success. The first requires that both PORVs must be opened if a LOCA does not 

accompany the seismic event. The second requires that at least one POR V must be 

opened if a small LOCA occurs with the seismic event. In addition to the two PORVs, 

there is a motor-operated block valve for each PORV that must be open. A list of the 

system components is given in Appendix D. 

The primary pressure relief system that includes the POR Vs is represented in Figure 

3-3. The POR Vs (located at el. 6.5 ft ln the reactor containment) are opened remotely 

from the main control room for feed and bleed. The motor-operated isolation valves are 

normally open, but allow isolation of a PORV for maintenance or if it fails to reseat. At 

least one isolation valve must be open at all times. The only support system required for 

POR V operation is 480-V ac emergency power. 

The following items provided the basis for the development of the fault tree logic: 

• Control power for the PORVs and block valves is transformed off the 

bus which provides the valve motive power. 

• The automatic pressure relief function of the POR Vs and the safety 

relief valves is not of interest in this application of the POR Vs (feed 

and bleed). 

• Random, testing or maintenance probablli ties of failure for the 

POR V s and block valves are low. These events were not included in 

the analysis. 

• The pressurizer, pressurizer quench tank, quench tank cooler and 

quench tank pumps were not included in the fault tree analysis. A 

seismic failure (rupture) of any of these items will not hinder the feed 

and bleed function. The safety relief valves were also excluded from 

the fault tree. 

• If the control switch for an isolation valve is in the "OFF" or "CLOSE" 

position, it must also be opened by the operator for feed and bleed. 

The unpruned fault tree for the POR Vs is shown in Appendix D. Included are failures for 

each of the system components and the required support systems. The events in this tree 

do not specify the types of failures, these will be added as the tree is pruned. The top 

event shows failure to support feed and bleed. This is caused by a common cause failure 

of the POR Vs and block valves, by failure of the main control room panels, or by a 

system failure that prevents flow through both POR Vs (for the case with a small 

LOCA). For the case in which no LOCA has occurred, lack of flow through only one 

3-8 
PORV will cause system failure. There are several events shown which may lead to loss 

of flow through a POR V. 

From the fault tree described above, the screening methods from Chapter 2 were used to 

prune the tree. The results of applying these screening criteria to the system 

components are found in Table 3-8. The pruned PORV fault tree was merged with the 

necessary support system fault trees (developed in Section 3.3). The complete PORV 

fault trees are found in Appendix D. 

The final PORV fault trees were analyzed to determine the system cut sets. The cut sets 

were then used to check the tree logic. Table 3-9 lists the minimal cut sets for the 

system when no LOCA has occurred. Table 3-10 lists the cut sets for the case when a 

small LOCA has occurred. For minimal cut set evaluation and ranking, the failure 

probabilities of the seismic events were assumed to be 1.0 for this analysis. HCLPFs will 

be assigned later by the fragility analysis team. The nonseismic probabilities were 

calculated as previously described. 

For the case with no LOCA, any event which causes failure of one PORV is a single 

fault. This turns out to be all twelve events included in the fault tree, which includes 

five seismic failures. In addition to the faults described below, random failures of the 

DGs and failure to isolate portions of the PCC are included. There are no double or 

triple faults for this case. 

For the small LOCA case, only those events which lead to failure of both PORVs make up 

the single-order cut sets. There are nine of these faults, five of which are seismic. 

Seismic failure of the station transformers, PCC/SCC coolers,· air conditioner chillers, 

circulation water pump house, DG day tanks, failure to refill the DG fuel tanks or DG 

common cause failure causes loss of power to both PORVs. Common cause failure of the 

PORV s, or opera tor failure to actuate feed and bleed, will also fail the system. There 

are two system double faults, both of which include only nonseismic events. Both faults 

lead to PORV failure as a result of support system failures. There are no system triplets. 

3.3 Svstems Analysis of Support Systems 

The support systems which ate required for operation of the Group A front-line systems 

described in Section 3.2 are primary and secondary component cooling water, service 

water, electric power and actuation. The analysis of these systems is similar to that for 

the front-line systems. 

3.3.1 Component Cooling Water Systems 

Component cooling water consists of the primary component cooling water (PCC) system 

and the secondary component cooling water (SCC) system. Pec and SCC provide the 

cooling required by plant equipment for normal operation, and decay heat removal during 

cooldown or accidents. The PCC and sec are redundant systems, in that PeC will 

provide cooling for one train of a front-line system and sec will provide cooling for the 

other train. Within the PCC and SCC systems, operation of at least one pump and one 

cooler are required for system success. Each system consists of a surge tank, two motordriven pumps, two coolers (heat exchangers) and valves for isolating nonessential portions 

of the system. Complete lists of the components, valves, cooling requirements and 

cooling loads are in Appendix E for the PCC, and Appendix F for the SCC. 

Schematics of the PCC and SCC systems are shown in Figures 3-4 and 3-5 respectively. 

One pump in each system (located in the turbine building at el. 21 ft) is normally 

3-9 
'. t • r • . ~, 

' ' ,•' ~. 

operating. The standby pump will start automatically on a supply header low pressure 

signal. Upon loss of offsite power, the diesel generator load sequencer will start the 

pump which had been operating in each system, after a 10-second delay. If all four 

pumps were operating, only the preferred pumps (P-9A and P-lOA) will be restarted. If a 

preferred pump fails to start, the alternate must be started manually. In each system, 

one cooler (also located in the turbine building at el. 21 ft) is normally in service. The 

cooler in standby has its cooling water outlet valve closed. As there must be flow 

through a cooler for system success, the cooler bypass valves (PCC-T -20 and SCC-T -23) 

must be closed. The cooler bypass and isolation valves are linked by a single operator 

and fail to the full cooling position. The cooling water flows from the coolers to the nonisolated loads and is returned to the pump suction headers. 

Support systems required for component cooling water availability are 4160-V ac 

emergency power, 125-V de power, and service water. 

The following items provided the basis for the development of the fault tree logic: 

• A new procedure change will instruct the operator to isolate portions 

of the PCC system following an earthquake if the PCC surge tank low 

level alarm is annunciated. This is accomplished by closing valves 

PCC-M-90, PCC-M-150, PCC-M-219, and PCC-A-268. It is assumed 

that failure or inability to close these valves will result in loss of 

PCC, due to a potential breach of the system pressure boundary. 

Only those loads which are not isolated were included in the PCC 

fault tree. 

• The nonseismic loads in the sec system are isolated by valves 

SCC-A-460 and SCC-A-461, which close automatically on a suction 

header low pressure signal, indicating a breach in the system pressure 

boundary. It is assumed that failure of these valves to close will 

result in loss of SCC. (Although it is on a return line, SCC-A-461 

must close as there is no check valve to prevent backflow through the 

line.) Only those loads which are not isolated were included in the 

sec system fault tree. 

• All four pump motors are equipped with drip-proof shields. It is 

judged that these shields will provide protection from potential spray 

from the fire water lines located over the pumps if the lines break. 

• It is assumed that control power for the motor-operated valves is 

transformed off the same bus which provides the valve motive power. 

• Check valve failures, random valve failures, and failure due to 

plugging, testing or maintenance are of low probability and are 

excluded from the analysis. 

• The instrument air supply for the sec isolation valves is not modeled 

beyond the accumulator (TI<-11 0) which is seismically designed. The 

accumulator inlet line contains two check valves to prevent 

depressurization, and TK-11 0 contains enough air to reposition the 

valves once and hold them closed for 24 hours. 

• The chemical additive tank and supply line are not included as a 

potential flow diversion path in either system. The supply and return 

3-10 
lines each contain a normally closed manual valve. (The PCC P&ID 

incorrectly shows these valves as normally open.) 

• The gland leak-off tank (TK-93, 94), pump (P-112, 113), and filter 

(FL-69, 70) are not included as a potential flow diversion path in 

either system. Only a small amount of leak-off flow is present in 

com pari son to cooling water flow. 

• The surge tank vent, overflow and waste lines are not included as a 

potential flow diversion path in either system. The vent and overflow 

lines contain an automatic valve, and the waste line contains a 

normally closed manual valve. 

• The recovery action of placing the standby cooler in service by 

opening the manual outlet valve is not included in the system fault 

trees. This action would not significantly change the plant HCLPF. 

• The failure of the SCC line to the penetration coolers by the collapse 

of block wall VE 21-3, 4 is included in the SCC system fault tree. 

The basic fault trees for the PCC and SCC systems are shown in Appendices E and F, 

respectively. These trees include all system components and nonisolated cooling loads, 

along with the required support system inputs. As in the front-line system fault trees, 

the types of failures were not specified. The top event of each tree represents loss of 

cooling to the required safety system components. This event may occur by common 

cause failure of the cooling water pumps, failure of a nonisolated cooling load (breach in 

system pressure boundary), failure of the required system isolation to occur, or failure 

within the system to provide flow. The last event may be caused by lack of flow through 

the in-service cooler, or by loss of both system pumps. There are several events on the 

tree which may lead to these faults. 

The screening criteria described in Chapter 2 were applied to the cooling water system 

components ln order to prune the PCC and SCC fault trees. The results of this screening 

are summarized in Tables 3-11 and 3-12 for PCC and sec, respectively. The failure 

probability calculations and results are detailed in Section 3.4. Although the PCC and 

sec coolers and the air conditioner chillers are located in separate systems, there is only 

one event for cooler seismic failure (CCW -HTX-EQ-4B5A) and one for chiller seismic 

failure (CCW-ACX-EQ-CHILL). This is because the coolers are located in the same area, 

and are identical components. The pruned cooling water fault trees were then merged 

into the front-line system trees. Therefore, the analysis of the merged front-line fault 

trees includes failures due to loss of PCC and SCC. 

3.3.2 Service Water System 

The service water (SW) system is not directly a support system for the Group A front-line 

systems, but is a support system for the support systems required by Group A systems. It 

provides cooling for the PCC and SCC systems. Operation of at least one SW pump is 

required for system success. For PCC success there must be SW flow to the PCC cooler 

in service, and for SCC success there must be SW flow to the SCC cooler in service. The 

service water system consists of four traveling screens, four motor-driven pumps, and 

manual valves to align service water flow to the PCC and SCC coolers. Lists of the SW 

components and cooling requirements are in Appendix G. 

3-11 
, ... ' ; ' ~ 

A drawing of the SW system is shown in Figure 3-6. Two of the pumps (located in the 

circulating water pump house, el. 7 ft) are normally operating, the two standby pumps 

must be placed in service manually when necessary. Upon loss of offsite power all four 

pumps receive a start signal from the diesel generator load sequencer, the pumps are 

interlocked so only one in each pair will run to prevent overloading (i.e., if P-29A starts, 

P-29e will trip off). One PCC and one SCC cooler are normally in service. The PCC and 

sec coolers in standby have their SW outlet valves closed. Service water flows from the 

PCC and sec coolers to the seal pit. Support systems required for SW availability are 

480-V ac emergency power and 125-V de power. 

The following items provided the basis for the development of the fault tree logic: 

• The capacity of the traveling screens is large enough that it is· 

assumed to be unlikely for all four to be blocked badly enough to 

choke all four SW pumps. The screens are designed for the circulating 

water system, with much larger flows than the SW system. Also, the 

screens are heavily used only a few times a year, therefore they have 

been excluded from the analysis. 

• The loss of all four SW pumps due to high water level in the pump 

house is judged to be improbable. An alarm is sounded at a 3-inch 

water level, and the circulating water pumps trip off at a 10-inch 

water level. As the SW pump motors are mounted above the pumps, a 

10-inch water level is not threatening. 

• All valves on the main SW lines, except for the standby cooler outlets 

and pump discharge header cross-tie, are assumed to be normally 

open. 

• Check valve failures and plugging of manual valves are of low 

probability and excluded from the fault tree. 

• The lines for the screen wash systems are not included as potential 

flow diversion paths, as they are less than one-third the size of the 

pump discharge header. Also the screen wash system and traveLing 

screen motors are interlocked, with neither powered from an 

emergency bus. Thus upon loss of offsite power the screen wash 

system is inoperable. 

• The mussel control pump and discharge line are not included as a 

potential flow diversion path, as it is normally isolated and only used 

in special operations. 

• All sample lines, pumps and collection tanks are excluded as potential 

flow diversion paths. These lines are considerably less than one-third 

the size of the main SW lines. 

• Recovery actions, such as placing a standby cooler in service, are not 

included in the fault tree. This does not significantly impact the 

HCLPF analysis. 

• Collapse of the circulating water pump house will fail all four service 

water pumps. 

3-12 
The service water system fault tree is shown in Appendix G. This tree includes 

nonspecific faults for each of the SW components, along with the required support system 

interfaces. The top event, loss of service water flow through the PCC/SCC coolers, is 

caused by common cause failure of the SW pumps, or by a system failure which prevents 

flow to the coolers. The latter would be caused by isolation of all four coolers, or failure 

of all four service water pumps. 

Screening techniques from Chapter 2 were used to prune the SW system fault tree. The 

results of this screening process are listed in Table 3-13. Of all the front-line and 

support systems analyzed, the service wate~ system is unique in that all system 

component failures (seismic and nonseismic) were screened out, leaving only the power 

inputs and circulating water pump house failure. When support system fault trees were 

merged with the front-line system trees, the power inputs were pruned to eliminate 

circular logic (SW fails PCC failing DG-lA failing Bus 7 failing SW pump A). Therefore 

the only fault from the service water system in the front-line system cut sets is collapse 

of the pump house. 

3.3.3 Electric Power System 

The electric power system consists of the emergency ac power (ACP) system, the de 

power (DCP) system, and the ansi te electric power (OEP) system (i.e., the diesel 

generators). The OEP system provides power to the emergency buses of the ACP system 

upon loss of offsi te power. The ACP system provides operating power to the plant 

equipment, feeds the DCP system battery chargers, and provides 120-V ac power to plant 

instrumentation. The DCP system provides 125-V de power to the switchgear, vital bus 

inverters, instrumentation and controls. The OEP system consists of two diesel 

generators (DGs), each with a fuel oil supply system, air starting system, and distribution 

and control panels. The ACP system consists of two 4160-V ac emergency buses, two 

4160 to 480-V station transformers, two 480-V ac emergency buses, six 480-V ac 

emergency motor control centers (MCCs), and four 120-V ac inverters. The DCP system 

consists of four 125-V de buses, each with a station battery and battery charger, and four 

distribution panels. Complete lists of the ACP and DCP system components are included 

in Appendix H. The OEP system components and cooling requirements are also in 

Appendix H. 

Simplified schematics of the ACP, DCP, and OEP systems are shown in Figures 3-7, 3-8, 

and 3-9. Both DGs (located in the turbine building auxiliary bay, el. 22 ft) are normally 

in standby. A DG is automatically started by a one-second loss of voltage on its 

associated emergency bus (Bus 5 for DG-lA, Bus 6 for DG-lB). A DG will also start on a 

10-second low voltage condition in its associated bus, concurrent with a SIAS (this case 

was not included in the analysis). A separate fault tree is constructed for each DG and 

each MCC (MCC 7B and 7Bl, and MCC 8B and 8B 1 are treated as the same MCC as they 

are connected by a tie without a breaker). The 4160-V ac buses, 480-V ac buses, and 125-

V de buses are inc! uded within the MCC fault tree logic. The 120-V ac buses and 

inverters were excluded as they supply instrumentation and actuation systems which 

were screened out. Also, for solenoid-operated valves requiring 120-V ac power, the 

batteries which feed the inverters have been shown as the inputs to the front-line 

systems. Support systems for the electric power system are PCC and SCC for the DG 

coolers. 

The following items provided the basis for developing the fault tree logic: 

• Cooling for the switchgear, cable tray, and battery rooms is provided 

by fans FN-31 and FN-32. Upon loss of offsite power, ac power loads 

3-13 
,•' ·, 

are shed and only some reloaded by the load sequencer. It is assumed 

that there is sufficient load reduction to not require switchgear 

cooling for a long period of time, so the fans were removed from the 

fault tree. 

• The diesel generator room exhaust fans must be operating, and the air 

intake and exhaust dampers open. It is assumed the dampers are 

powered off the same bus as the DG room fan. 

• Because the operator must make a periodic check of the DG day tank 

fuel level, it is judged that the tank will not be allowed to drain 

through a broken vent line without some preventive action taken. 

• It is judged that flooding of the diesel rooms is of low probability. 

The curbs at the room entrances protect against external flooding 

from the Turbine Building and the threaded fire water piping in the 

rooms is normally dry. Heat and smoke are required to actuate the 

spray of aqueous foam. 

• Replacement of the current lead-antimony station batteries with 

lead-calcium batteries will sufficiently raise their seismic capacity. 

• Relay chatter is not included in the analysis based on the guidance in 

NUREG/CR-4334- and 4-482. 

• Nonseismic circuit breaker failure is of low probabill ty and excluded 

from the analysis. 

• The bus cross-tie breakers are physically constructed to prevent 

closing unless one of the bus normal feed breakers are open. 

Therefore no failures were postulated for spurious closure of the 

cross-tie breaker resulting in the failure of both buses. 

• The tie from the 480-V ac buses to the inverters is a synchronizing 

tie, which ensures that the 120-V ac vital system is at the same 

frequency as the rest of the ACP system. The only power supply to 

the inverters is the batteries. 

• Although there is a procedure for feeding a 125-V de bus wl th an 

alternate battery (e.g., battery 3 to feed DC-1), this is a recovery 

action and is not included in the fault tree. Because the new 

batteries have high HCLPFs, this action is not significant to the plant 

HCLPF. 

• Although the 480-V ac MCCs provide power to the battery chargers, 

this relationship ls not shown on the fault tree as it would create 

circular logic (MCC 7 A falls BC-1 failing DC-1 falling Bus 7 SWGR 

failing MCC 7 A). For the purpose of starting the DGs, the required 

de power is only available from the station batteries. 

• The logic of the diesel generator load shed/sequencers is not 

developed in the fault trees. As long as the relays of the sequencer 

are not physically damaged, the sequencer may be reset from the 

main control board (MCB) or from local panels near the DGs. Once 

3-14 
reset, loads on the sequencer may be started from the MCB. The 

relays are located in the main electrical panel in the control room. 

The diesel engine, generator, blower which cools the generator, air 

starting units, integral fuel tank and pumps, integral cooling water 

pumps, and lubricating oil system are treated as a single component in 

the fault trees (DG lA and DG 1 B). 

• The external DG fuel oil systems are represented separately in the 

fault trees. Each system consists of a day tank, an auxiliary fuel oil 

transfer pump and an underground fuel oil storage tank. 

• The external DG air starting tanks are represented separately in the 

fault trees. For each DG there are two air receiver banks, consisting 

of three compressed air tanks each. One bank is sufficient to start a 

DG. 

• The auxiliary boiler fuel oil supply lines from the auxiliary fuel oil 

transfer pumps are not included as potential flow diversion paths. 

Each line contains a normally closed air-operated isolation valve 

which fails closed. 

• Refilling the integral fuel tanks by opening the manual valves is 

included in the fault trees as this is an operating procedure performed 

approximately every 2 hours. Refilling the day tanks by placing the 

oil pumps in service is also included in the trees. 

• Some recovery actions are not included in the fault trees. This 

includes opening the manual valves between air receiver banks, and 

opening the manual valves which allow fire water to supply the DG 

coolers. 

• The DG-lB cooler SCC outlet valve, SCC-T -305, is physically held 

open to provide continual flow through the cooler. 

The fault trees for MCC 7 A, Bus 7, Bus 5, and de Bus 1, for MCC &A, Bus 8, Bus 6 and de 

Bus 3, and for DG-lA are shown in Appendix H. The trees for MCC 7B, MCC 8B, and 

DG-lB are analogous to these. The top events of the MCC trees are failure of the MCC, 

due to faults within the MCC or failure of the 480-V bus. Loss of the 480-V bus is due to 

faults within the bus, failure of the station transformer or failure of the 4160-V bus. 

Loss of the 4160-V bus is due to faults within the bus, failure of the DG, failure of the 

load sequencer, insufficient switchgear cooling or loss of the de bus. Failure of the de 

bus is caused by faults within the bus, failure of the battery or failure of the battery 

charger. 

The top event of the DG fault trees is loss of power output from the diesel generator, 

due to failure of the output or bus cross-tie breakers, failure of the DG control and 

distribution panels, loss of de power required to start the DG, or failure of the DG. 

Failure of the DG is caused by faults within the diesel generator, failure of the fuel oil 

supply, failure of the DG start signal (caused by loss of bus voltage), failure of the air 

starting banks, failure of the room fan or dampers, or insufficient cooling of the engine. 

The screening criteria were applied to the electric power system components to prune 

the fault trees. The results of this process are summarized in Table 3-14. The failure 

3-15 
probability calculations and results used are detailed in Section 3.4. The pruned electric 

power system fault trees were then merged with the front-line system trees in order to 

determine the front-line system minimal cut sets. Faults from the electric power 

systems which will lead to front-line system failure are included in the front-line system 

cut set tables in Section 3.2. 

3.3.4 Actuation Systems 

Various actuation systems are required to initiate engineered safeguards systems, these 

include the safety injection actuation system (SIAS), the containment isolation system 

(CIS), the containment spray actuation system (CSAS), the recirculation actuation system 

(RAS), and the reactor protection system (RPS). The AFW actuation system was included 

with the AFW system. Of these, SIAS is the only system required by the evaluated Group 

A systems. The SIAS consists of two trains both of which are actuated by the same four 

instrument channels (2/ 4 logic). These channels consist of instruments which measure 

pressurizer pressure and containment pressure. SIAS ls actuated on low-low pressurizer 

pressure or high containment pressure. A list of instrumentation for the actuation 

systems is shown in Appendix I. 

Loss of power wl thin a logic channel results in a channel trip. Loss of power in the 

actuation logic leads to an actuation signal. Loss of power also results in the disenabling 

of the SIAS block. For these reasons power inputs were not included in the SIAS fault 

trees. As sufficient separation of the instrument channels exists, the system was not 

modeled down to the sensor and relay level (undeveloped events). 

The fault tree for SIAS Train A is shown in Appendix I. The fault tree for Train B is 

analogous to this. Operator actions to unblock and initiate SIAS are included in the 

trees. Since there is sufficient separation of the required instrumentation and there is 

operator ability to actuate the system, automatic actuation of SIAS is not judged to be 

an issue. Therefore, the SIAS inputs were pruned from the front-line and support system 

fault trees and the SIAS fault trees were not further developed. 

3.4 Probability Calculations 

The data for the probability calculations of the nonseismic common cause, random, test 

and maintenance, and operator failures are shown in Table 3-15. Unless otherwise 

stated, mean beta factors from EPRI NP-3967 (Fleming, et al., 1985) and mean 

unavailabili ties and failure rates from ASEP were used in the calculations. The results 

were later converted to median values based on the following: 

Error 

Factor 

3 

5 

10 

Median Mean 

Conversion 

Mean to Median 

Conversion 

1.26 

1.6 

2.66 

The following items were used as the basis for the calculations: 

• The automatic valves ln the AFW system are air-operated, for which 

a beta factor was not found. These valves fail open on loss of air or 

3-16 
\ ··:, 

power so the probability of common cause failure was judged to be 

below the screening cutoff value. 

• Steam binding of the three AFW pumps was the major common cause 

failure mode considered. 

• The HPSI common cause failure calculation does not include the spare 

pump, P-145. Common cause failure of the three pairs of injection 

valves is below the cutoff value. Common cause failure of the five 

remaining valve pairs (R WST outlet, RHR heat exchanger isolation, 

charging line isolation, pump discharge, and recirculation sump 

valves) was included in the calculation. 

• The time used in determining HPSI pump fail to run (FTR) (12 hours) 

is based on the expectation that after this time the LPSI pumps would 

be placed in sevice, or the auxiliary charging pump could be used. 

• A beta factor of 0.08 was used for the POR Vs. The beta factor for a 

PWR safety relief valve is 0.07, and for an MOV it is 0.08. 

• The DG auxiliary fuel oil pumps are small (low head), but are located 

in a relatively harsh environment. Therefore a beta factor of 0.11 

was used. The probability of failure to run for these pumps is 

insignificant. 

• The failure to start figure for random failure of the DGs is taken 

from ORNL data (Battle, 1985). The testing and maintenance 

calculation reflects the estimate by Maine Yankee personnel of six 

days of T &M time per year. 

• To place the turbine-driven AFW pump in service the operator must 

open the steam supply valves and monitor the flow. The time allowed 

to do this is roughly 45 minutes. It is assumed that after 20 to 30 

minutes of attempting to start the pump from the main control room 

(MCR) he would attempt to do so locally. The human error 

probability (HEP) for starting the pump from the control room is 0.05. 

• The operator may attempt to start the turbine-driven AFW pump 

locally after a hardware or actuation/power failure. It is assumed to 

take about 10 minutes to diagnose failure of the motor-driven pump, 

10 minutes to attempt starting the TDP from the MCR, and 10 

minutes to get to the TDP to start it locally. The operator then has 

roughly 10 minutes left to start the pump. The HEP for 5 to 10 

minutes is 0.25, and for 10 to 20 minutes is 0.10; the average is 0.18. 

• To align the PWST to the motor-driven AFW pumps, the pumps must 

be tripped to prevent cavitation (5 to 10 minutes), and the manual 

valves from the PWST opened and the pumps restared before the SGs 

empty (30 to 40 minutes). Isolation of the DWST is not required due 

to the placement of check valves. The HEP for tripping the pumps is 

0.25, and for opening the valves and restarting the pumps it is 0.03, 

for a total of 0.28. 

3-17 
• The operator actions for feed and bleed are guided by the ERGs and 

FRGs. Various studies show the response time to be 15 to 45 minutes. 

Because no plant specific study was available, the 20 to 30 minute 

HEP was used (0.05). This valve is higher than most other feed and 

bleed estimates (ASEP gives approximately 0.0 1), but the operators 

will first attempt to start the AFW pumps, and it is an earthquake 

situation. 

• The new procedure for PCC isolation calls for the operator to close 

four valves from the MCR in the event of an earthquake with low 

level alarm on the PCC surge tank. If there are PCC ruptures lnslde 

containment, the operator may not have much time to perform this 

action, and indications other than the surge tank level alarm may not 

be checked or available. A 10- to 20-mlnute time period, with a 0.1 

HEP, is used. Recovery is not considered likely. 

• Appoxlmately every 2 hours the operator must refill the DG integral 

fuel tanks, and occasionally refill the day tanks from the underground 

storage tanks. The steps for thls are Included In the loss of offslte 

power procedures and are simple, but must be performed local to the 

DGs. Failure to do so would cause the DG to stop, making it 

necessary to reprime and restart. Because the procedure ls not 

difficult to perform, and the failure is recoverable, a probability of 

0.01 is used. 

3-18 
Table 3-1. Front-line system vs support system dependency matrix. 

SUPPORT SYSTEMS 

AC Power DC Power ccw SIAS IA 

FRONT-LINE Bus 5 Bus 6 DC-1 DC-3 Train Train 

SYSTEMS Bus 7 Bus 8 DC-2 DC-4 PCC sec A B TK-25 

HPSI/ 

RECIRC 

P-14A 

to RCS X X X X 

P-14B 

to RCS X X X X 

FN-44A X X 

FN-44B X X 

AFW 

P-25A X X 

P-25B X 

P-25C X X 

PORVsPR-S-14 X 

PR-S-15 X 

Note: To determine the front-line system dependencies on thP. support systems, , ..... locate the front-line component in the first column and read across the 

row to find the support system dependencies. 

3-19 
w 

I 

N 

0 

SUPPORT SYSTEM 

AC 4160V 5 

Power Bus 6 

480V 7 

Bus 8 

120V 1 

'lit a 1 2 

Bus 3 

4 

Diesel lA 

Generator lB 

[)C 125V 1 

Pov1er Sus 2 

3 

4 

ccw PCC 

sec 

51-iS 

Sl.\5 Channel A 

B 

c 

0 

Act. SIS-A 

SIS-B 

BVAC FN-20A 

fii-20B 

F/l-31 

Fll-32 

Tahle 3-?.. Support system vs support systPm depP.ndency matrix. 

--------------~------------~f -~9~~r __________________________ ------ ~f: ~~0~~!' ______ 

~!~~_Ru~- _1~~!'._!!~!-. __ l!Q~-~H!L~_us __ OG _ __ !g§L~~~-- -----

5 6 7 8 1 2 3 4 lA 2A 1 2 3 4 

X X 

X X 

X X 

X 

X 

X 

X 

X 

X 

X 

X 

X 

X X 

X 

X 

X 

_f:~!!-

PCC sec 

X 

SW S lAS IIVAf. IA 

X 

X 

------- -.=-~ttf_:---~FN- -·-rK·: 

A B 20A 21l8 31 32 110 

X X 

X X 

X X 

X X 

X 

X 

X 

X 

X 

X 

X 

X 

X 

rlote: To determine the support system dependencies on other support systems, locate the support system 1n the first column, and read across the row to determine dependenc1es on 

the other support systems. 
Component 

AFW-A-101 

AFW-A-201 

AFW-A-301 

AFW-A-338 

AFH-A-339 

AFW-A-340 

E-86A 

E-868 

E-86C 

MS-A-162 

MS-A-173 

MS-M-161 

MS-M-255 w MS-P-168 ~ MS-T-163 ....... 

P-25A 

P-258 

P-25C 

T-1 

TK-16 

TK-21 

TK-25 (Accum.} 

TK-111 (Accum.) 

TK-123 (Accum.) 

System 

. ,. 

Table 3-3. AFW screening overview table. 

Screening Logic 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-Integral to P-25A 

Out-Integral to P-258 

Out Integral to P-25C 

Out-Group Not-A (Atm. Steam Dump) 

Out-High HCLPF 

Out-Group Not-A (Atm. Steam Dump} 

Out-Group Not-A (Aux. Steam) 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

In-Random (High HCLPF) 

Out-High HCLPF 

Out-Considered with TOP P-258 

In-Low HCLPF 

In-Low HCLPF 

In-Unknown HCLPF 

Out-AOVs fail open on loss of air 

Out-AOVs fail open on loss of air 

In-Common Cause failure for pumps and AOVs 

Event 1 

AFW-AOV-XX-101 

AFW-AOV-XX-201 

AFW-AOV-XX-301 

AFW-AOV-XX-338 

AFW-AOV-XX-339 

AFW-AOV-XX-340 

AFW-AOV-XX-A173 

AFW-PCV-XX-P168 

AFW-PCV-XX-T163 

AFW-MOP-XX-PTRNA 

AFW-TDP-LF-P258 

AFW-MDP-XX-PTRNC 

AFW-TNK-EQ-PWST 

AFW-TNK-EO-DWST 

AFW-TNK-EQ-TK25 

AFW-CCF-FC-AFW 

Value2 

>0. 3g , >0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.5g 

8.0E-2 

>0.5g 

0.27 

0.17 

NA (>0.3g) 

2.0E-4 

NA indicates that a HCLPF was not available at the time of screening, and the component remained in the analysis. The 

value in parenthesis is the HCLPF that was available later. 

1, 2 -- See Table 3-4 for standard footnotes on the screening overview tables. 
,._\, 

Table 3-lf. Standard footnotes for screening overview tables. 

Footnote 

Number 

1 Events using the XX cause code are shown i~ the basic system fault trees. Events 

using other cause codes (EQ, FC, etc.) are shown in the pruned system fault trees. 

Components with no event listed were screened out without being added to a fault 

tree. Appendix A contains the definitions of symbols and abbreviations. 

2 Values shown are the preliminary HCLPF capacity for seismic failures, or the 

calculated probability of failure for nonselsmic failures. 

3 Components which make up the various actuation systems, which were screened out 

based on separation and operator ability to actuate manually. These systems 

included: 

SIAS-A, B (SIS-ACT -FA-TRMA, B) 

Steam Generator Level (RPS-ACT -FA-SGLEV) 

PCC Pump Discharge Pressure (PCC-PST -FA-ACTPP) 

SCC Pump Discharge Pressure (SCC-PST -FA-ACTPP) 

SCC Header Isolation Pressure (SCC-PST -F A-17 50A, B) 

DG Load Shed/Sequencer (OEP-DGN-VF-A, BLDSQ) 

DG Initiation- Bus Voltage (OEP-ACT -FA-BUS5, 6) 

NA Indicates that a HCLPF was not available at the time of screening, and the 

component remained in the analysis. The value in parentheses is the HCLPF that 

was available later. 

3-22 
, '.. ~ . 

.... ' 

Table 3-5. AFW system cut sets. 

Ranked Single Faults 

AFW-CCF-FC-AFW 

Ranked Double Faults 

AFW-TNK-EQ-PWST 

ACP-TFM-EQ-57X68 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BK W-EQ-CIRC 

OEP-TNK-EQ-TK62X 

AFW-XHE-FO-EFWXX 

AFW-XHE-FO-TRBMC 

ACP-TFM-EQ-57X68 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK 62X 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SW 5-BK W -EQ-CIRC 

OEP-TNK-EQ-TK 62X 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

OEP-XHE-FO-FUEL 

AFW-XHE-FO-TRBMC 

OEP-CCF-FC-DGN 

OEP-CCF-FC-DGN 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

ACP-TFM-EQ-57X68 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-XHE-FO-TRBMC 

AFW -XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF -P25B 

OEP-XHE-FO-FUEL 

AFW-TDP-LF -P25B 

AFW-XHE-FO-TRBMC 

Nonseismic 

Probabili tv 

2.0E-O 4-

Nonseismic 

Probability 

SF 

SF 

SF 

SF 

SF 

SF 

2.8E-O 1 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

l.OE-02 

2.6E-03 

5.0E-04 

5.0E-04-

1.3E-04-

1.3E-04 

SF indicates that the cut set events are seismic failures, and their HCLPFs will be 

provided by the fragility analysis team. 

3-23 
w 

I 

N 

-f:o 

'<· 

Table 3-6. HPSI/CSPPCL screeni~g overview table. -~ 

Component Screening logic Eventl 

CH-A-32 ___ --- ---lJut--High- HClPF HPI-AOV-XX-CHA32 

CH-A-33 Out-High HCLPF HPI-AOV-XX-CHA33 

CH-F-38 

CH-M-1 

CH-M-87 

CS-M-1 

CS-M-2 

CS-M-91 

CS-M-92 

E-34 

E-67 

E-96 

Fl-348 

FN-44A 

FN-44B 

HSI-M-11 

HSI-M-12 

H SI -M-21 

HSI-M-22 

H SI -M-31 

HSI-M-32 

HSI -M-41 

H SI -M-42 

HSI-M-40 

HSI-M-43 

HSI-M-50 

HSI -M-51 

HSI-M-54 

HSI -M-55 

Out-Isolated by CH-A-32, 33 

Out-Path ruled out as flow diversion 

Out-Path ruled out as flow diversion 

Out-Group Not-A (Recirculation) 

Out-Group Not-A (Recirculation) 

Out-Group Not-A (Recirculation) 

Out-Group Not-A (Recirculation) 

Out-Path ruled out as flow diversion 

Out-Isolated by CH-A-32, 33 

Out-Isolated by CH-A-32, 33 

Out-Isolated by CH-A-32, 33 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-Recovery only 

Out-Recovery only 

Out-High HCLPF 

Out-High HCLPF 

Out-Group Not-A (Recirculation) 

Out-Group Not-A (Recirculation) 

HPI-MOV-XX-CHF38 

HPI-HTX-XX-REGEN 

HPI-HTX-XX-SLWTR 

HPI-FlT-XX-SlWTR 

C SS-FAN-X X-FN44A 

CSS-FAN-XX-FN44B 

HPI-MOV-XX-MV11 

HPI-MOV-XX-MV12 

HPI-MOV-XX-MV21 

HPI-MOV-XX-MV22 

HPI-MOV-XX-MV31 

HPI-MOV-XX-MV32 

HPI-MOV-XX-MV41 

HPI-MOV-XX-MV42 

HPI-MOV-XX-MV50 

HPI-MOV-XX-MV51 

Value2 

>0~3g 

>0.3g 

>0.3g 

>O. 3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 
w 

I 

N 

l.n 

Component 

LSI -t4-40 

LSI-M-41 

P-14A 

P-148 

P-14S 

P-61A 

P-618 

P-61S 

SL-P-3 

TK-4 

TK-54 

VE 21-1, 2 

System 

Screening Logic 

Out-Group Not-A (LPSI) 

Out-Group Not-A (LPS1) 

Out-High HCLPF 

Out-High HCLPF 

Out-Spare, recovery only 

Table 3-6 (Cont'd) 

Out-Group Not-A (Recirculation) 

Out-Group Not-A_. (Recirculation) 

Out-Group Not-A (Recirculation) 

Out-Isolated by CH-A-32, 33 

In-Low HCLPF 

Out-High HCLPF 

Out-Although failure impacts FN-44A, B, 

upgraded wall is of sufficient capacity 

In-Common cause failure for pumps & MOVs 

Event1 

HPI-MDP-XX-PTRNA 

HPI-MDP-XX-PTRNB 

HPI-MOV-XX-SLP3 

HPI-TNK-EO-RWST 

HPI-TNK-XX-SCAT 

HPI-CCF-FC-HPSI 

1, 2 - See Table 3-4 for standard footnotes on the screening overview tables. 

Value2 

>0.5g 

>0.5g 

0.2lg 

0.59g 

>0.5g 

2.1E-3 
PCC-XHE-FO-ISOL 

OEP-PSF -FC-DG lB 

Table 3-7. HPSI system cut sets. 

Ranked Single Faults 

ACP-TFM-EQ-57X68 

HPI-TNK-EQ-R WST 

CC \V -HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

5\VS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK62X 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

OEP-CCF -FC-HPSI 

Ranked Double Faults 

OEP-PSF -FC-DG 1 B 

OEP-PSF-FC-DG lA 

Nonseismic 

Probabill ty 

SF 

SF 

SF 

SF 

SF 

SF 

I.OE-02 

2.6E-03 

2.1E-03 

Nonselsmlc 

Probabill ty 

6.7E-03 

4.5E-03 

SF indicates that the cut set events are seismic failures, and their HCLPFs will be 

provided by the fragility analysis team. 

3-26 
Table 3-8. PORVs screening overview table. 

Component Screening Logic Event1 Va1ue2 

E-2 Out- Not required for Group A (Feed & Bleed) 

PR-M-16 Out-High HCLPF PP S-MOV-XX-PRM16 >0.3g 

PR-M-17 Out-High HCLPF PPS-MOV-XX-PRM17 >0.3g 

PR- S-11 Out-Not required for feed and bleed 

PR- S-12 Out-Not required for feed and bleed 

PR- S-13 Out-Not required for feed and bleed 

PR- S-14 Out-High HCLPF PP S- SOV-XX-PRS14 >0.3g 

P R- S-15 Out-High HCLPF PP S- SOV-XX-PRS15 >0.3g 

System In-Common cause failure for PORVs and PP S-CCF -FC-PORVs S.OE-4 w 

& block valves N 

......., 

1 ~ 2 - See Table 3-4 for standard footnotes for the screening overview tables. 
Table 3-9. POR V (no LOCA) cut sets. 

Ranked Single Faults 

CC W -HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BK W -EQ-CIRC 

OEP-TNK-EQ-TK62X 

ACP-TFM-EQ-57X68 

PCC-XHE-FO-ISOL 

OEP-PSF-FC-DG lA 

OEP-PSF -FC-DG 1 B 

PPS-XHE-FO-FDBLD 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

PPS-CCF-FC-POR VS 

No Double Faults 

Probability 

SF 

SF 

SF 

SF 

SF 

LOE-01 

6.7E-02 

6.7E-02 

5.0E-02 

l.OE-02 

2.6E-03 

8.0E-04 

SF indicates that the cut set events are seismic failures, and their HCLPFs will be 

provided by the fragility analysis team. 

3-28 
PCC-XHE-FO-ISOL 

OEP-PSF -FC-DG 1 A 

Table 3-10. POR V (small LOCA) cut sets. 

Ranked Single Faults 

CC W -HTX -EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK62X 

ACP-TF M-EQ-57X68 

PPS-XHE-FO-FDBLD 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

PPS-CCF -FC-POR VS 

Ranked Double Faults 

OEP-PSF -FC-DG 1 B 

OEP-PSF -FC-DG lB 

Nonseismic 

Probability 

SF 

SF 

SF 

SF 

SF 

5.0E-02 

l.OE-02 

2 .. 6E-03 

8.0E-04 

Nonseismic 

Probability 

6.7E-03 

4.5E-03 

SF indicates that the cut set events are seismic failures, and their HCLPFs will be 

provided by the fragility analysis team. 

3-29 
w 

I 

w 

0 

Table 3-11. PCC screening overview table. 

Component Screening Logic Event1 Value2 

AC-18 In-Unknown--H-cepr-Of--cnlTTer- --u-- - ------- ----ccw=ACX-tQ:.-cHJ[L-- --~---NA-W-~891 

E-3A Out-High HCLPF 

E-4A Out- Standby, recovery on 1 y 

E-4B In-Unknown HCLPF 

E-25 Out-High HCLPF 

E-54-1 thru 6 Out-High HCLPF 

E-82A Out-High HCLPF 

E-918 Out-High HCLPF 

E-928 Out-High HCLPF 

P-9A Out-High HCLPF 

P-98 Out-Hi.gh HCLPF 

P-7 Cooler Out-High HCLPF 

P-12A Cooler Out-High HCLPF 

P-14A Cooler Out-High HCLPF 

P-145 Coo 1 er Out-High HCLPF 

P-61A Cooler Out-Integral to pump 

P-61 S Coo 1 er Out-Integral to pump 

Penetration Coolers Out-High HCLPF 

PCC-A-53 Out-Return line isolation only 

PCC-A-216 Out-High HCLPF 

PCC-A-238 Out-Return line isolation only 

PCC-A-268 Out-High HCLPF 

PCC-A-270 Out-Replaced in procedure by PCC-A-268 

PCC-M-43 Out-Return line isolation only 

PCC-M-90 Out-High HCLPF 

PCC-M-150 Out-High HCLPF 

PCC-M-219 Out-High HCLPF 

PCC-HTX-XX-E3A 

CCW-HTX-EQ-4B5A 

PCC-HTX-XX-E25 

PCC-HTX-XX-E54X 

PCC-HTX-XX-E82A 

PCC-HTX-XX-E91B 

PCC-HTX-XX-E928 

PCC-MOP-XX-PTRNA 

PCC-MDP-XX-PTRNB 

PCC-HTX-XX-P7 

PCC-HTX-XX-P12A 

PCC-HTX-XX-Pl4A 

PCC-HTX-XX-P14 S 

PCC-HTX-XX-PEN 

PCC-AOV-XX-AV216 

PCC-AOV-XX-AV268 

PCC-MOV-XX-MV90 

PCC-MOV-XX-MV150 

PCC-MOV-XX-MV219 

>0.3g 

- NA (0.3lg) 

>0.3g 

>0.3g 

>0.3g 

>0.5g 

>0.3g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0 .. 5g 

>0.5g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 
Table 3-11 (Cont'd) 

Component Screening Logic Event1 Val ue2 

PCC-T-19 Out-High HCLPF PCC-TCV-XX-TCV19 >0.3g 

PCC-T-20 Out High HCLPF PCC-TCV-XX-TCV20 >0.3g 

TK-5 Out-High HCLPF PCC-TNK-XX-SRGTK >0.5g 

System Out-Common cause failure of pumps PCC-CCF-FC-PCCW 1.4E-4 

below cutoff (0.001) 

NA indicates that a HCLPF was not available at the time of screening, and the component remained in the analysis. The ~ value in parentheses is the HCLPF that was available later. LV 

~ 

1' 2 See Table 3-4 for standard footnotes on the screening overview tables. 
w 

I 

w 

N 

Table 3-12. sec screening overview table. 

Component Screening Logic Event 1 Val ue2 

AC-lA --fn-Unkno-wn-HCLP-FOTChilTer- CCW-ACX-EQ:T-HIIL -NAt0.389} 

AC-2 In-Unknown HCLPF of chiller CCW-ACX-EQ-CHILL NA (0.38g) 

E-3B 

E-5A 

E-5B 

E-828 

E-91A 

E-92A 

P-lOA 

P-108 

P-128 Coo·l er 

P-148 Cooler 

P-618 Cooler 

Penetration Coolers 

SCC-A-460 

SCC-A-461 

SCC-T-23 

SCC-T-24 

TK-59 

VE 21-3, 4 

System 

Out-High HCLPF 

In-Unknown HCLPF 

Out-Standby, recovery only 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-Integral to pump 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-Although failure impacts SCC line to 

penetration coolers, wall has hiqh HCLPF 

Out-Common cause failure of pumps below 

cutoff {0.001) 

SCC-HTX-XX-E3B 

CCW-HTX-EQ-485A 

SCC-HTX-XX-£828 

SCC-HTX-XX-E91A 

SCC-HTX-XX-E92A 

SCC-HTX-XX-PTRNA 

SCC-HTX-XX-PTRN8 

SCC-HTX-XX-Pl28 

SCC-HTX-XX-P148 

SCC-HTX-XX-PEN 

SCC-AOV-XX-AV460 

SCC-AOV-XX-AV461 

SCC-TCV-XX-TCV23 

SCC-TCV-XX-TCV24 

SCC-TNK-XX-SRGTK 

SCC-CCF -FC- SCCW 

>0.3g 

NA (0.3lg) 

-- >0.3g 

>0.5g 

>0.3g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

--

>O.Sg 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.5g 

>0. 5g 

1. 4 E-4 

NA indicates that a HCLPF was not available al-lfie-Eime of screening, and the component remained-in the-analysTs. The 

value in parentheses is the HCLPF that was available later. 

1, 2 - See Table 3-4 for standard footnotes on the screening overview tables. 
w 

I 

w 

w 

Component 

Circ. Water Pump House 

E-4A 

E-4B 

E-5A 

E-5B 

P-29A 

P-29B 

P-29C 

P-290 

System 

Table 3-13. SWS screening overview table. 

Screening logic 

In-low HClPF 

Out-standby, recovery only 

Out-Seismic failure considered for CCW 

Out-Seismic failure considered for CCW 

Out- Standby, recovery only 

Out-High HClPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HClPF 

Out-Common cause failure of pumps below 

cutoff (0.001) 

Event 1 

SWS-BKH-EQ-C I RC 

SWS-HTX-XX-E4B 

SWS-HTX-XX-E5A 

SWS-MDP-XX-PTRNA 

SWS-MDP-XX-PTRNB 

SW S-MD P -.X X-P T RNC 

SWS-MDP-XX-PTRNO 

SWS-CCF-FC-PUMPS 

1, 2 - See Table 3-4 for standard footnotes on screening overview tables. 

Value2 

0.30g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

1.4E-4 
w 

I 

w 

Component 

120 VAC Bus 1 thru 4 

120 VAC Bus lA thru 4A 

125 VDC Bus 1 thru 4 

125 VDC Cab. DC/CE-1 

125 VDC Cab. DC/CE-2 

125 VDC Panel DP/P 

125 VDC Panel DP/BU 

480 VAC MCC 7A,7B,7Bl 

480 VAC MCC 8A,8B,8B1 

480 VAC Bus 7,8 

4160 VAC Bus 5,6 

BATT-1 thru 4 

+::- BC-1 thru 4 

INVR-1 thru 4 

Transformer X-507 

Transformer X-608 

Bus Cross-Tie Breakers 

Main Control Board 

Electrical Control Board 

Aux. Logic Cabinets 

ESF Aux. Panels 

Air Cond. Control Panel 

Table 3-14. Electric power screening overview table. 

Screening Logic 

Out-Included with MCB 

Out-.Hi gh HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-Assumed to be upgraded 

low common cause failure 

Out-High HCLPF 

Out-High HCLPF 

In-Low HCLPF 

In-Low HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Event1 

OCP-BDC-XX-BUSl (etc.) 

ACP-PSF-LP-MCC7A{B) 

ACP-PSF-LP-MCCBA(B) 

ACP-PSF-LP-BllS7(8) 

ACP-PSF-LP-BUS5{6) 

DCP-RA1-XX-BA11{etc.) 

DCP-CCF-LP-BA11 

DCP-BA1-XX-BCH1(etc.) 

ACP-1FM-EQ-57X68 

ACP-1FM-EQ-57X68 

ACP-BKR-CC-315 

ACP-BKR-CC-416 

MCR-AC1-XX-CNTRL 

Value2 

>0.3g 

>0.3g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.3g 

l.OE-4 

>0.3g 

0.34g 

0.3g 

0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 
w 

I 

w 

U1 

Component 

SPDS Cabinets 

Instrument Racks 3 

Cable Trays3 

Impulse L i nes3 

OG-lA 

DG-18 

DG Output Breakers 

DG-1A~B Engine 

Control Panels 

DG-1A,B Distribution 

Panels 

DG-lA,B Control Panels 

DG-lA,B Air Inlet and 

Exhaust Dampers 

FN-20A 

FN-20B 

FN-31 

FN-32 

Screening Logic 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Table 3-14 (Cont•d) 

Out-Based on train and channel separation 

In-Random and common cause failures, 

High HCLPF 

In-Random and common cause failures, 

High HCLPF 

In-Combined with DG failure 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-High HCLPF 

Out-Switchgear load sufficiently decreased, 

fan cooling no longer required 

Out-Switchgear load sufficiently decreased, 

fan· cooling no longer required 

Event 1 

OEP-PSF-FC-DG1A 

OEP-CCF-FC-DGN 

OEP-PSF-FC-DGlB 

OEP-CCF-FC-DGN 

OEP-BKR-00-ACBlA(B} 

OEP-ACT-XX-1A{B)CTL 

OEP-ACT-XX-lA(B)CTL 

OEP-ACT-XX-lA(B) CTL 

OEP-SOD-XX-AIRA{B) 

OEP-MOD-XX-EXHA(B) 

OEP-FAN-XX-FN20A 

OEP-FAN-XX-FN20B 

ACP-FAN-XX-FN31 

ACP-FAN-XX-FN32 

Value2 

>0.5g 

>0.5g 

>0.5g 

6.7E-2 

2.6E-3 

6.7E-2 

2.6E-3 

>0.3g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 

>0.5g 
w 

I 

w 

0'1 

Table 3-14 (Cont'd) 

Component Sc. reen i ng Logic Event1 Value2 

-- P-33A Out-High HCLPF OEP-MDP-XX-P33A >0.5g 

P-338 Out-High HCLPF OEP-MDP-XX-P33B >0.5g 

PCC-A-493 Out-High HCLPF OEP-AOV-XX-AV493 >0.3g 

SCC-T-305 Out-Chained open OEP-AOV-XX-AV305 

SB 39-1 Out-Seismic failure impacts Group Not-A --- 0.35g 

components (FN-7A,B) 

SB 39-2 Out-Seismic failure impacts components which 

were screened out (FN-32 exhaust duct) 

SB El .39 floor Out-High HCLPF --- 0.35g 

TK-28A Out-High HCLPF OEP-TNK-XX-TK28A >0.5g 

TK-28B Out-High HCLPF OEP-TNK-XX-TK28B >0.5g 

TK-62A In-Unknown HCLPF OEP-TNK-EQ-TK62X NA ( 0.43g) 

TK-628 In-Unknown HCLPF OEP-TNK-EQ-TK62X NA (0.43g) 

TK-76A-l thru 6 Out-High HCLPF OEP-TNK-XX-76A-l(etc.) >0.5g 

TK-76B-l thru 6 Out-High HCLPF OEP-TNK-XX-768-l(etc.) >0.5g 

NA indicates that a HCLPF was not available at the time of screening, and the component remained 1n the analysis. The 

value in parentheses is the HCLPF that was available later. 

1, 2, 3 - See Table 3-4 for standard footnotes on screening overview tables. 
Table 3-15. Nonseismic event failure probability calculations. 

Mean Unavailability 

Event Beta or Failure Rate T1me(hr) Probability 

AFW-CCF-FC-AFW -- 1.0E-4 (steam binding) -- l.OE-4 

HPI-CCF-FC-HPSI 0.08 (MOV) 3.8E-3 (MOV FTO) 

0.17 (MOP) 3.2E-3 (MOP FTS) 

5.3E-5/hr (MOP FTR) 12 2.1E-3 

PPS-CCF-FC-PORVs 0.08 (MOV) 3.8E-3 (MOV FTO) 

0.08 (SOV) 6.3E-3 (SOV FTO) -- 8.0E-4 

PCC-CCF-FC-PCCW 0.03 (MOP) 3.2E-3 (MOP FTS) 

5.3E-5/hr (MOP FTR) 24 1.4E-4 

w SCC-CCF-FC-SCCW 0.03 (MOP) 3.2E-3 (MOP FTS) • w 5.3E-5/hr (MOP FTR) 24 1.4E-4 -.I 

SWS-CCF-FC-PtJMPS 0.03 (MOP) 3.2E-3 (MOP FTS) 

5.3E-5/hr (MOP FTR) 24 1.4E-4 

OEP-CCF-FC-DGN 0.05 (OG) 2 .1E-2 (OG FTS) 

1.26E-3/hr (DG FTR) 24 2.6E-3 

OEP-CCF-FC-P33X 0.11 (RHR pumps) 3.2E-3 (MOP FTS) -- 3.5E-4 

AFW-TOP-LF-P25B -- 1. 6E -2 (TOP T&M) 

3.2E-2 (TOP FTS) -- 5.0E-2 

1,3E-4/hr (TOP FTR) 12 

OEP-PSF-FC-DGlA,B -- 2.1E-2 (OG FTS) 

l.26E-3/hr (DG FTR) 24 

1.6E-2 (OG T&M) -- 6.7E-2 

AFW-XHE-FO-TRBMC -- -- -- 0.05 
w 

I 

w 

co 

Event 

AFW-XHE-FO-TRBLO 

AFW-XHE-FO-EFWXX 

PPS-XHE-FO-FOBLO 

PCC-XHE-FO-ISOL 

OEP-XHE-FO-FUEL 

Beta 

Table 3-15 (Cont'd) 

Mean Unavailability 

or Failure Rate Time(hr) Probability 

0.18 

0.28 

0.05 

0.1 

0.01 
(....) 

I 

(....) 

t.D 

DEMINERAUZED 

WATER 

STORAGE TANK 

(DWST) 

11<-21 1..;: • • CONDENSATE UAKEUP 

~from PWST (TK-16) 

t t 

~ piQ ..... E UAIN 

• E r>a-;>c::] us-r-t 63 srEAtJ I 

us..:..A-=173 US-P-168 

Figure 3-1 Auxiliory Feedwoter 

_j AJR ACC., 1 11<-2s 

AAN-A-338 t AAN-A-tOI 

lA-*-~ 1--.1.'--- lA 

INSIDE 

CONTAINMENT 

To Steqm Generators 

E-1-3 

E-1-2 

E-1-1 
w 

I 

.+:a 

0 

s 

c 

A 

T 

TK-~ 

54 

REFUELING 

CAVIlY 

WATER 

STORAGE 

TANK 

(RWST) 

TK-4 

----.;;;~~ Seal Water to 

Uotn Coolant Pump• 

To Chorglng 

Header 

INSIDE 

CONTAINMENT 

E-96 1--t><l .,,.. HJ [)( ~To Loop F"tll 

Header 

.. 9 ..- HSI-U-54 

HSI-U-55 

From 

Reaiduol 

Heat Exc;hangens 

LC 

CH-F-70 I 

CH-A-33 

HSI-U-40 

HSl-U-41 

HSI-U-42 

HSI-U-4J 

CH-A-J2 

Figure 3-2 High Pressure Safety Injection 

HSI-U-12 

HSI-U-1 1 

HSI-U-22 

To 

Loop 1 

To 

Loop 2 

I ..... 'li-"1~ 

HSI-tJ-21 1 

I 

I 

I 

I 

I HSi-U-J2 I 

I 

I 

I 

I 

To 

Loop J 

I.~ 

HSI-U-J1 

... ~~ 
w 

I 

.::. 

1--' 

RCS 

PUMP 

PR-S-13 PR-S-12 

PRESSURIZER 

E-2 

REACTOR <: > I 

(SURGE UNE) 

TO PRESSURIZER 

PR-S-11 I i I I > QUENCH TANK 

TK-8 

PR-S-14 PR-S-15 

PR-M-16 

PORVs 

PR-M-17 

Figure 3-3 Primary Pressure Relief 
w 

I 

..;::.. 

N 

PRIMARY 

COMPONENT 

COOLING 

SURGE 

TANK 

TK-5 

COOLING WATER RETURN 

COOLING 

i .. Ill I )I WATER 

SUPPLY 

P-9A 

Figure 3-4 Primary Component Cooling Water 
w 

I 

.;::.. 

w 

COOLING WATER SUPPLY 

PCC-U-4J 

E-82A 

P-14.\ 

P-145 

E-92B 

COOLING WATER RETURN 

E-72A,B 

E-81A,B,C 

C-3A. 

C-38 

E-94 

PCC-tJ-90 

E...:29 

Et5 

JlO 

I 

E-46 

E-9JA 

·r 

E-71A 

E-71B 

Figure 3-4 

P-11 

P-19 

i

-20 

-21 

-2.2 

P-65 

I 

JE-54-t 

E-54-2 

I E-54-3 

&E-54-4 

I

E-54-5 

E-54-& 

• I 

P-68A. I • 

P-66B l 

I 

I 

I 

I 

l 

I 

E..:25 I 

I 

I 

I 

J 

I 

I 

I 

I 

I 

I 

I 

I 

I 

I 

·-

PCC-U-219 

~r E-J9A 

E-398 

PCC-A-299 

PCC-A-2J8 

(cont.) 

_____ CONTAINUENT ......... __ ,_.,. 

E-53-1 

E-53-2 

E-53-3 

PENETRATION 

COOlERS 

{15) 

f 

P-1-1 

P-1-2 

P-1-3 
w 

I 

+=-

+=-

COOLING 

SECONDARY 

COMPONENT 

COOLING 

SURGE TANK 

TK-59 

~ 

WATER 

RETURN 

~ 

scc-A-461 

From 

non-isolated 

sec loads 

AIR 

t• I ACCUM ~ lA 

TK-110 

scc-T-23 

Figure 3-5 Secondary Component Cooling Water 

To non-isolated 

sec loads 

r 

COOLING 

WATER 

SUPPLY 
w 

I 

~ 

(J'1 

SCC SUPPLY 

SCC RETURN 

Figure 3-5 (cont.) 
FLOW 

FLOW 

TRASH RACK 

TRASH RACK 

TRAVELUNG 

SCREENS 

Figure 3-6 Service Water System 

3-46 

NO 

TO 

SD\L 

PIT 
::::! HSI-M-54 

lO FN-44A 

c 

..., HSI-M-50 en 

(...J 

I 

-....J 

)> 

(") 

-o HSI-M-55 0 

:E FN-448 en 

., HSI-M-51 

DG-1A 

DIST. 

PANEL 

DG-1B 

SUPPLY 

IJJ 

n 

I 

_. 

m 

n 

I 

N 

DIST. CAB 

OJ 

() 

I 

""" 

Lv-£ 

FN-20A 

P-33A 

PCC-M-90 

PCC-M-21 9 

HSI-M-41 - -

H~-M-12 

HSHA-4J ~~ 0 

H~-M-22 ~ 

HSI-M-32 

~ 

0 

0 

-....I 

OJ 

~ 

0 

0 

()) 

OJ 

__.. 

PR-S-14 --"' 

PR-M-16 

PR-M-17 

PR-5-15 -1m t 

FN-32 

P-33B 

rN-20B .........,. I ext 

HSI-M-40 

HSI-M-42 

HSI-M-21 

HSI-M-31 

Cl--

HSI-M-11 

CJ 

m 

(') 

I 

WI 

.Jlo. 

ext 

~ 

[I] 

c 

c.n 

P-29A 

P-29C 

P-298 

P-290 

P-145 

P-14A 

P-9A 

P-98 

tD 

c 

Vl 

(.,.1 

I.Jlo. 
INVERTER 

120 VAC 

VITAL BUS 1,1A 

INVERTER 

J 

120 VAC 

VITAL BUS 3,JA 

INVERTER 

2 

120 VAC 

VITAL BUS 2.2A 

INVERTER 

4 

120 VAC 

VITAL BUS 4, 4A 

MCC BA 

MCC SA 

f 

a:: 

C!l 

3 

VI t-1 - > V1W lt10 r--> nco nc a~ :J- ::I«'' uu lll'<t' CD'<t' ac 

::J IX: 

l:1 Ill ;:!!: Col ........ U1 

I-I ,_a.. Vli.IJ V10 > 100 a~ iS..o 1'11.0 CJ(.j CJ 0 :::1.- oc au m..,. 

125 VDC DIST CAB 2 

...: 

"tia.. 

-=.a 

uo 

au 

125 VDC DIST CAB 4 

Figure 3-8 DC Power 

3-48 

~ _J 

N 0 J-0. 

VJO - <I-- <n:: ==~ .... ,_ iS..c ~~ IZ CJO 6~ C!IO au a Ill OVl au 

IX: 

l:1 _J 

~ ... N 0 

m> m,_ ms- m.x: 

......... "j.x: ... a::: 1'10 IZ :::IIXJ C!l~ 6~ 00 m-.:- OV1 Q(J'J au 
To 

Auxiliary 

Boiler 

FL-F-13 

(FC) 

P-33A 

UNDERGROUND 

FUEL STORAGE 

TANK 

TK-28A 

UNDERGROUND 

FUEL STORAGE 

TANK 

TK-288 

P-338 

TK-76A1 

TK-76A2 

TK-76A3 

TK-76A4 

TK-76A5 

TK-76A6 

FUEL OIL 

DAY TANK 

TK-62A 

PCC SUPPLY 

FIRE 

PROTECTION-----!.,..~;.,.~ 

SYSTEM 

FUEL OIL 

DAY TANK 

,......-.::c---..JTK -628 

FL-F-27 

(FC) TK-7686 

TK-7685 

To TK-7684 

Auxiliary 

Boiler TK-7683 

TK-7682 

TK-7681 

PCC-A-483 PCC-A-493 

PCC-492 

SCC-304 

SCC-T-303 

Figure 3-9 Diesel Generators 

3-49 

BUS 

3 

BUS 

5 

BUS 

6 

BUS 

4 

3T5 

4T6 

4. ACCIDENT SEQUENCE ASSESSMENT AND RESULTS 

In order to obtain the plant level Boolean equation which represents the cut sets for the 

accident sequences, two event trees were developed. These event trees, one which 

includes a small LOCA and one which does not, are described in Section 4.1. The core 

damage sequences that are derived from the event trees are discussed in Section 4.2. 

The Boolean equations obtained from the analysis of each event tree are described ln 

Section 4.3. Section 4.4 presents the plant-level Boolean equation. 

4.1 E•1ent Trees 

Because it was not feasible to enter the reactor containment, the review team could not 

inspect and verify the seismic capacity of the LOCA sensitive piping. Therefore, the 

plant analysis was segregated into two cases: a seismic event and loss of offsite power 

with no LOCA, and a seismic event and loss of offsite power with a small LOCA. For 

this project a small LOCA is defined as one which requires the use of HPSI for core 

makeup. Based on data from other PRAs this would be a break in the range of roughly 

three-eighths inch to two-inch diameter. However, no plant-specific thermodynamic 

analyses were performed under this review to determine the J.ctual size range.. The first 

event tree considers a seismic event without a concurrent LOCA, the second tree 

considers a seismic event with a small LOCA. The minimal cut sets from the resulting 

accident sequences are then combined to determine the plant Boolean equation. 

4.1.1 No LOCA Case 

The event tree for the first case is presented in Figure 4-1. This event tree depicts the 

success or failure of the various front-line mitigating systems in response to a seismic 

event with accompanying loss of offsite power. The systems are displayed across the top 

of the figure. At each node in the event tree, success of the system ls represented by 

the upward branch, and failure by the downward branch. Each path through the event 

tree represents a sequence. Those sequences that end with a status of "OK" have 

successfully mitigated ·the seismic event, and do not result in core damage. Those 

sequences that end with a status of "CD" have failed to provide all necessary safety 

functions, and result in core damage. They are identified by the abbreviation for the 

initiating event (in this case, event T 1

) and the abbreviations for any front-line system 

failures (e.g., event L or P 1). These core damage accident sequences are then evaluated 

using the system fault trees to determine the minimal cut sets and Boolean equations. 

The initiating event for the first event tree is a seismic event which causes a loss of 

offsite power, designated as T 1

• Following this event the reactor must be made 

subcritical (event C). Potential accident sequences due to failure of the reactor 

subcriticality system were not evaluated further because the fragility analysis of the 

reactor internals and control rods found the seismic capacity to be greater than the 

review level earthquake. Nonseismic failures of the subcriticality system are also 

insignificant (approximately J.OE-5 per demand). 

Upon successful reactor subcriticality, the AF\V system is used for core decay heat 

removal (event L). Successful operation of this system (at least one of three pumps 

feeding one of three steam generators) will provide core cooling (early) and prevent core 

damage. Core inventory makeup is not required because the reactor coolant system 

(RCS) is intact for this event tree. Failures of the RCS integrity (such as a safety relief 

valve LOCA or RCP seal LOCA) are analyzed in the next section. 

4-L 
Should failure of the AFW system occur, the operator should begin feed and bleed (event 

P 1 ). In this case with no concurrent LOCA, both POR Vs (and the associated block valves) 

must be opened within approximately 30 minutes for system success. Failure to open 

both POR V lines (i.e., one or both remain closed) will result in inadequate core cooling 

and eventual core damage (sequence T 1 LP 1

). 

Following the successful initiation of feed and bleed, the HPSI system (event D) must be 

used to provide makeup to the RCS. This requires at least one of two HPSI pumps (the 

third pump is evaluated for recovery actions only) to provide injection to one of three 

RCS loops. Also included in HPSI success criteria is the long-term availability of the 

recirculation system. As discussed in Chapter 3, this included the availability of the 

containment spray pump area fans. Failure of the HPSI system is part of core damage 

sequence T 1 LD. 

4.1.2 Small LOCA Case 

The event tree for the second analysis is shown in Figure 4-2. For this case the initiating 

event is an earthquake that causes loss of offsite power, concurrent with a small LOCA 

(event S ). Larger LOCAs (beyond the makeup capacity of the HPSI system) were not 

evaluate~ based on the guidance in NUREG/CR-4334 and -4482 (Budnitz et a!., 1985, and 

Prassinos et al., 1986). Ths small LOCA could be caused by nonseismic or seismic 

failures. The nonseismic failures considered were a stuck-open POR V or safety valve, a 

significant reactor coolant pump (RCP) seal LOCA (greater than about 20 gallons per 

minute per RCP), and a coincident small LOCA. Based on generic ASEP data, the 

probability of a stuck open POR V that is not isolated by the block valve, or the 

probability of a stuck-open safety valve is equal to or less than the screening value of 

l.OE-3. The probability of a significant RCP seal LOCA with the RCP seal design at 

Maine Yankee is judged to be similarly low, based on discussions with Yankee Atomic. 

Also, the probability of a nonseismic small LOCA coincident with the earthquake event 

and recovery time is very small (less than l.OE-3). Therefore, nonseismic small LOCAs 

were screened out of the analysis. 

The small LOCAs induced by seismic failures that were considered were a stuck-open 

PORV or safety valve, and small breaks in LOCA sensitive piping. Based on the fragility 

analysis team review, stuck-open POR Vs or safety valves were judged to have HCLPFs 

greater than 0.3g, and were therefore screened out. However, because the review team 

was not able to inspect and verify the seismic capacity of small piping and impulse lines 

inside containment, the HCLPF of these lines is not known. Therefore, small LOCAs 

could not be screened out. The event tree for the small LOCA case is based on a seismic 

event that causes loss of offsite power and a small LOCA, and requires core inventory 

makeup for mitigation. 

Again the reactor must first be made subcritical (event C) and as described in Section 

4.1.1, the subcriticality system has high seismic capacity. Therefore, failure of the 

subcriticality function were not considered further. 

Following successful reactor subcri ticality, the AFW system is used for core decay heat 

removal (event L), with the same success criteria as described above. With the small 

LOCA, the HPSI system must be used to provide makeup to the RCS (event 0). Failure 

of HPSI at this point (no pumps feeding any RC loop) will lead to core damage sequence 

S2o. 

4-2 
Failure of the AFW system to provide decay heat removal will require the use of feed 

and bleed. Due to the small LOCA the operator need only open one POR V (with its 

associated block valve) within .30 minutes to successfully initiate feed and bleed (event 

P 2). Failure to open at least one POR V (i.e., both remain closed) within this time will 

result in sequence S2LP 2. With successful PORV operation, HPSI is again required to 

provide RCS makeup (event D), with the same success criteria described previously. 

Failure of AFW and HPSI result in core damage sequence s2LD. 

4.2 Core Damage Sequences 

To determine the cut sets for sequence T 1 LP 1 (seismic event with loss of offsite power 

followed by AFW failure, followed by POR V failure for feed and bleed) the pruned and 

merged fault trees for the AFW system and the PORVs (no LOCA case) developed in 

Section 3.2 were combined. By combining these two trees with an "AND" logic gate, 

failure of both systems is required to result in the sequence (T 1LP 1 ). A description of 

the basic events that make up the accident sequences is found in Table 4-1. 

The minimal cut sets for this sequence are listed in Table 4-2. As with the system cut 

sets, the seismic failures were arbitrarily assigned a probability of 1.0 for ranking 

purposes. There are no single faults which lead to this accident sequence. Of the double 

faults, some contain at least one seismic event, and others contain only non seismic 

events (e.g., random, common cause, human error). The event ACP-TFM-EQ-57X68 

(seismic failure of the station transformers) results in loss of power to both POR Vs and 

both motor-driven AFW pumps. Seismic failure of the DWST (AFW-TNK-EQ-DWST) 

leads to failure of the turbine-driven AFW pump, as does random failure of the TDP 

(AFW-TDP-LF -P25B) and failure to place the pump in service from the MCR (AFW -XHEFO-TRBMC). Seismic failure of the SCC and PCC coolers, the air conditioner chillers or 

the circulating water pump house (CCW-HTX-EQ-4B5A, CCW-ACX-EQ-CHILL, SWSBKW-EQ-CIRC) leads to failure of both the PCC and SCC, which in turn fails both 

diesels. Seismic failure of the diesel day tanks (OEP-TNK~EQ-TK62X) also results in loss 

of both DGs. The loss of both diesels has the same effect as seismic failure of the 

transformers. Generally, two separate faults are required to fail both the AFW motordriven and turbine-driven pumps. An exception is the common cause failure of AFW. 

4.2.2 T l LD 

The pruned and merged fault trees for the AFW system and the HPSI system were 

merged with an "AND" gate to determine the cut sets for this sequence (seismic event 

with loss of offsite power followed by loss of AFW, success of POR Vs, and loss of HPSI). 

The minimal cut sets for sequence T 1 LD are found in Table 4-3. Again there are no 

single faults. Most of the double faults contain at least one seismic event, the rest 

consist only of random, common cause or human failures. In this sequence, seismic 

failure of the transformers, SCC/PCC coolers, air conditioner chillers, circulating water 

pump house or DG day tanks leads to loss of power to both HPSI pumps, all HPSI motoroperated valves, and both motor-driven AFW pumps. The AFW faults are the same as 

those described for sequence TJ LP 1• Seismic failure of the R WST (HPI-TNK-EQ-R WST) 

or common cause failure (HPI-CCF-FC-HPSI) results in HPSI failure. 

It is interesting to note that PCC-XHE-FO-ISOL, which leads to PCC failure, does not 

appear here with the double faults, as it does with sequence T 1 LP 1

• This is because both 

PCC and SCC must fail to fail HPSI, while failure of only one was necessary to fail one 

of the two required POR Vs. 

4-3 
The fault tree for this sequence consists only of the HPSI pruned and merged system 

fault tree. Given the initiating event (seismic event with loss of offsite power and a 

small LOCA) followed by initiation of AFW, failure of HPSI wi11 result in core damage. 

Thus the minimal cut sets listed in Table 4--4- for sequence S2D are the same as those 

described for the HPSI system in Section 4.2 .. 2. 

Sequence S2LP 2 is similar to sequence T 1 LP 1' but the POR V success criteria is 

different. For s2LP2 only one PORV must be opened (i.e., both PORVs remain closed for 

failure). The system fault trees combined to determine the sequence cut sets were those 

for AFW and POR Vs (small LOCA), developed in Section 3.2. The minimal cut sets for 

this sequence are listed in Table 4-5. There are no single faults that lead to core 

damage. 

The major difference between the second order cut sets for sequence S_2~P 2 and for 

sequence T 1 LP 1 is that the former do not contain any events which fail PCC only. This 

is as expected, since both POR Vs must fail (and thus both DGs) in addition to AFW 

failure to give sequence S2LP 2• As PCC fails DG-1A, another event is needed to fall 

DG-1B. 

4.2.5 S2LD 

The final sequence consists of the same events (and thus the same system fault trees) as 

sequence T 1 LD, only the initiating events differ. Therefore, the cut sets for sequence 

s2LD listed in Table 4-6 are identical to those in Table 4--3. 

4.3 Boolean Equations for No LOCA and LOCA Cases 

The Boolean equations for the core damage sequences with no LOCA and core· damage 

sequences with a small LOCA are given in Table 4-7. The numbers correspond to the 

basic events listed in Table 4-1. Only the first and second order faults which contain at 

least one seismic failure event are included in the Boolean equations. Although it is 

possible for core damage to occur through nonseismic failures only, these were not 

evaluated further for this seismic margins project. Also, the final HCLPF calculations 

determined that some of the seismic failure events had final HCLPFs greater than 0.3g, 

so they were pruned out. 

The events included in the Boolean equations affect the systems as follows: 

Event 4: 

Event 7: 

Event &: 

Seismic failure of the station transformers (ACP-TFM-EQ57X68) also results in loss of power to all HPSI components, 

both POR Vs, and both AFW motor-driven pumps. 

Seismic failure of the R WST (HPI-TNK-EQ-R WST) results in 

loss of HPSI. 

Seismic failure of the DWST (AFW-TNK-EQ-DWST) results in 

loss of the turbine-driven AFW pump (the motor driven pumps 

have the PW ST as a backup). 

4-4 
Event 14: 

Event 15: 

Event 16: 

Event 17: 

Event 20: 

Event 22: 

Nonseismic common cause failure of the DGs (OEP-CCF-FCDGN) results in loss of power to all HPSI components, both 

POR Vs and both AFW motor-driven pumps. 

Nonseismic common cause failure of the AFW system (AF\VCCF-FC-AFW) results in total loss of AFW. 

Operator failure to refill the DG day and/or integral tanks 

(OEP-XHE-FO-FUEL) leads to failure of both DGs, resulting 

in loss of power to all HPSI components, both POR V s and both 

AFW motor-driven pumps. 

Operator failure to place the AFW turbine-driven pump in 

service from the MCR (AFW-XHE-FO-TRBMC) results in loss 

of the TDP. 

Collapse of the circulating water pump house (SWS-BK W-EQCIRC) will fail all SW pumps, which in turn falls PCC and 

SCC due to lack of SW flow through the coolers. The effects 

of PCC and SCC failure are loss of both DGs due to lack of 

cooling, which results in a loss of power for all HPSI 

components, both PORVs, and both AFW motor-driven pumps. 

Random failure of the AFW turbine-driven pump (AFW-TDPLF-P25B) results in loss of the TDP. 

Those numbered events in Table 4-1 that are not in this final list were either pruned 

because their final HCLPF was greater than 0.3g, subsumed into the above events 

because they were not in minimal cut sets, or were only in nonselsmic cut sets. 

4.3.1 No LOCA Case 

To determine the reduced Boolean equation for this event tree, sequences T 1 LP 1 and 

T 1 LD were combined with an "OR" gate, because occurrence of either of these sequences 

wdllead to core damage. 

The combinations of events described above that make up the reduced Boolean cut sets 

lead to core damage as follows: 

• Those cut sets that combine events 4 or 20 with events 8, 15, 17, or 

22 result in both sequence T 1 LP 1 and T1 LD. A fault from the first 

group (4 or 20) results in loss of all HPSI components, both PORVs 

and both AFW motor-driven pumps. A fault from the second group (8, 

15, 17, 22) results in loss of the turbine-driven AFW pump. 

• Events 14 or 16 also lead to loss of HPSI, both PORVs, and both AFW 

motor-driven pumps. However, only the cut sets where one of these 

events is combined with event 8 are included in the Boolean. These 

lead to both T 1 LD and T 1 LP 1. Coupling of 14 or 16 with 15, 17, or 22 

result in cut sets of two nonseismic events, which were screened out. 

• Finally event 7 ls combined with 15 to give sequence T 1 LD. Event 7 

leads to HPSI failure, and 15 to AFW failure. 

4-5 
4.:3.2 Small LOCA Case 

To determine the Boolean equation for the second event tree, sequences S2D, S2LP2 and 

S2LD were combined with an "OR" gate because occurrence of any one of these 

sequences will cause core melt. The reduced Boolean for this event tree consists only of 

single faults. This is because each of these faults will cause failure of HPSI, because 

this corresponds to sequence s2o. No other system failures are required for core damage 

to occur. Essentially, the cut sets that would lead to the failures of AFW and one POR V, 

or AFW and HPSI, all fall out as nonminimal cut sets or consist only of nonseismic events 

and are screened out. The events which make up the small LOCA Boolean are 4, 7, and 

20. 

4.4 Plant-Level Boolean Eguation 

Because the small LOCA sensitive piping could not be inspected and its seismic capacity 

verified, the Boolean equations for small LOCA and no LOCA cases were combined in 

two different ways. Sensitivity analysis could then be used to estimate the plant 

HCLPF. The sensitivity studies were performed by the fragility analysis team. 

The first method used split fractions to express the conditional probability of a seismic 

induced small LOCA given the seismic event. The two Boolean equations in Table 4-7 

were combined using these split fractions, and sensitivity studies can be performed by 

varying the split fractions. 

The second method used an additional term for the small LOCA Boolean equation that 

represented the HCLPF of the seismic induced small LOCA .. Sensitivity studies can then 

be performed by varying the small LOCA HCLPF valve. 

4-6 
Table 4-1. Basic event descriptions. 

Nonselsmic 

Number Event Description Probability 

1 CC W -HT X-EQ-4B5A Seismic failure of PCC/SCC coolers SF 

2 AF\V-TNK-EQ-TK25 Seismic failure of air tank for AFW SF 

AOVs 

3 AFW-XHE-FO-EFWXX Failure to align PWST to AFW MDPs 2.8E-O 1 

4 ACP-TFM-EQ-57X68 Seismic failure of station transformers SF 

5 HPI-CCF-FC-HPSI Nonseismic common cause HPSI failure 2.1E-03 

6 PPS-XHE-FO-FDBLD Failure to open PORV s for feed and 5.0E-02 

bleed 

7 HPI-T~K-EQ-R WST Seismic failure of the R WST SF 

8 AFW-TNK-EQ-DWST Seismic failure of the DWST SF 

9 PPS-CCF-FC-PORVS Nonseismic common cause PORV failure 8.0E-04 

10 PCC-XHE-FO-ISOL Failure to close PCC isolation valves LOE-0 1 

11 OEP-PSF -FC-DG 1 B Random failure of diesel generator 1B 6.7E-02 

12 OEP-PSF -FC-DG 1 A Random failure of diesel generator lA 6.7E-02 

13 AFW -XHE-FO-TRBLO Failure to start AFW TOP locally 1.8E-O 1 

14 OEP-CCF -FC-DGN Nonseismic common cause DG failure 2.6E-03 

15 AFW-CCF-FC-AFW Nonseismic common cause Al-W failure 2.0E-04 

16 OEP-XHE-FO-FUEL Failure to refill DG fuel tanks l.OE-2 

17 AFW -XHE-FO-TRBMC Failure to start AFW TOP from MCR 5.0E-02 

18 OEP-TNK-EQ-TK62X Seismic failure of DG day tanks SF 

19 CCW-ACX-EQ-CHILL Seismic failure of air cond. chillers SF 

20 SWS-BKW-EQ-CIRC Seismic failure of eire. water pump SF 

house 

21 AFW-TNK-EQ-PWST Seismic failure of the PWST SF 

22 AFW-TDP-LF-P25B Random failure of the AFW TOP 5.0E-02 

SF indicates that the event is a seismic related failure, and a HCLPF will be determined 

by the fragility analysis team. 

4-7 
Table 4-2. Sequence T 1 LP 1 cut sets. 

No Single Faults 

Ranked Double Faults 

ACP-TFM-EQ-57X68 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK 62X 

ACP-TFM-EQ-57X68 

AFW-XHE-FO-TRBMC 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BK W-EQ-CIRC 

OEP-TNK-EQ-TK 62X 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK 62X 

OEP-XHE-FO-FUEL 

OEP-CCF -FC-DGN 

OEP-XHE-FO-FUEL 

AFW -XHE-FO-TRBMC 

CCW-HTX-EQ-4B5A 

CCW -ACX-EQ-CHILL 

SWS-BK W-EQ-CIRC 

OEP-TNK-EQ-TK 62X 

ACP-TFM-EQ-57X68 

OEP-CCF-FC-DGN 

AFW-XHE-FO-TRBMC 

PCC-XHE-FO-ISOL 

OEP-PSF -FC-DG 1 B 

OEP-PSF-FC-DG lA 

PPS-XHE-FO-FDBLD 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

PPS-CCF -FC-POR VS 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

ACP-TFM-EQ-57X68 

AFW-TDP-LF-P25B 

AFW-TDP-LF -P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF -P25B 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

OEP-XHE-FO-FUEL 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-TDP-LF-P25B 

OEP-CCF-FC-DGN 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

Nonseismic 

Probabilitv 

SF 

SF 

SF 

SF 

SF 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

l.OE-02 

2.6E-03 

5.0E-04 

5.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

1.3E-04 

l.JE-04 

2.0E-05 

1.3E-05 

l.JE-05 

l.OE-05 

2.0E-06 

5.2E-07 

1.6E-07 

SF indicates that the cut set consists entirely of seismic related failures for which 

HCLPFs will be determined. 

4-8 
Table 4-3. Sequence T 1 LD cut sets. 

No Single Faults 

Ranked Double Faults 

ACP-TFM-EQ-57X68 

CC W -HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BK W -EQ-CIRC 

OEP-TNK-EQ-TK62X 

ACP-TFM-EQ-57X68 

AFW-XHE-FO-TRBMC 

CC W -HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BK W -EQ-CIRC 

OEP-TNK-EQ-TK 62X 

CC W -HTX -EQ-48 5A 

CCW-ACX-EQ-CHILL 

SWS-BK W -EQ-CIRC 

OEP-TNK-EQ-TK 62X 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

OEP-XHE-FO-FUEL 

AFW-XHE-FO-TRBMC 

ACP-TFM-EQ-57X68 

CC W -HTX -EQ-48 5A 

CCW-ACX-EQ-CHILL 

CCW-BK W-EQ-CIRC 

OEP-TNK-EQ-TK62X 

HPI-TNK-EQ-R W ST 

OEP-CCF-FC-DGN 

OEP-CCF-FC-DGN 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

HPI-CCF-FC-HPSI 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

ACP-TFM-EQ-57X68 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

OEP-XHE-FO-FUEL 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW -CCF -FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-TDP-LF -P25B 

AFW-XHE-FO-TRBMC 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

AFW-CCF-FC-AFW 

Nonseismic 

Probability 

SF 

SF 

SF 

SF 

SF 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

l.OE-02 

2.6E-03 

5.0E-04 

5.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

l.~E-04 

1.3E-04 

2.0E-06 

5.2E-07 

4.2E-07 

SF indicates that the cut set consists entirely of seismic related failures for which 

HCLPFs will be determined. 

4-9 
Table 4-4. Sequence S2 D cut sets.

Ranked Single Faults

ACP-TFM-EQ-57X68

HPI-TNK-EQ-RWST

CCW-HTX-EQ-4B5A

CCW-ACX-EQ-CHILL

SWS-BKW-EQ-CIRC

OEP-TNK-EQ-TK 62X

OEP-XHE-FO-FUEL

OEP-CCF-FC-DGN

HPI-CCF-FC-HPSI

Nonseismic

Probability

SF

SF

SF

SF

SF

SF

1.OE-02

2.6E-03

2.IE-03

Nonseismic

Ranked Double Faults Probability

PCC-XHE-FO-ISOL OEP-PSF-FC-DGIB 6.7E-03

OEP-PSF-FC-DGIB OEP-PSF-FC-DGIA 4.5E-0 3

SF indicates that the cut set consists entirely of seismic related failures for which

HCLPFs will be determined.

4-10
Table 4-5. Sequence s2LP 2 cut sets. 

No Single Faults 

Ranked Double Faults 

ACP-TFM-EQ-57X68 

CCW-HTX-EQ-4B5A 

CC\V-ACX-EQ-CHILL 

SWS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK62X 

ACP-TFM-EQ-57X68 

AFW-XHE-FO-TRBMC 

CCW -HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK62X 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

SWS-BKW-EQ-CIRC 

OEP-TNK-EQ-TK62X 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

OEP-XHE-FO-FUEL 

AFW-XHE-FO-TRBMC 

CCW-HTX-EQ-4B5A 

CCW-ACX-EQ-CHILL 

CCW-BKW-EQ-CIRC 

OEP-TNK-EQ-TK62X 

ACP-TFM-EQ-57X6& 

OEP-CCF-FC-DGN 

AFW-XHE-FO-TRBMC 

PPS-XHE-FO-FDBLD 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

PPS-CCF-FC-POR VS 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

ACP-TFM-EQ-57X6& 

AFW-TDP-LF-P25B 

AFW-TDP-LF -P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

OEP-XHE-FO-FUEL 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW -CCF -FC-AFW 

AFW-CCF-FC-AFW 

AF\V-TDP-LF -P25B 

OEP-CCF -FC-DGN 

AFW -CCF -FC-AF\V 

AFW-CCF-FC-AF\V 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

Nonseismic 

Probabili tv 

SF 

SF 

SF 

SF 

SF 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

l.OE-02 

2.6E-03 

5.0E-04 

5.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

2.0E-04 

1.3E-04 

1.3E-04 

l.OE-05 

2.0E-06 

5.2E-07 

l.GE-07 

SF indicates that the cut set consists entirely of seismic related failures for which 

HCLPFs will be determined. 

4-11 
Table 4--6. Sequence s2LD cut sets. 

No Single Faults 

Ranked Double Faults 

ACP-TFM-EQ-57X68 

CCW-HTX-EQ-lf.B5A 

CCW-ACX-EQ-CHILL 

SWS-BK W-EQ-CIRC 

OEP-TNK-EQ-TK62X 

ACP-TF M-EQ-57X68 

AFW-XHE-FO-TRBMC 

CCW-HTX-EQ-lf.B5A 

CCW-ACX-EQ-CHILL 

SWS-BK W -EQ-CIRC 

OEP-TNK-EQ-TK 62X 

CCW-HTX-EQ-lf.B5A 

CC\V-ACX-EQ-CHILL 

SWS-BK W-EQ-CIRC 

OEP-TNK-EQ-TK 62X 

OEP-XHE-FO-FUEL 

OEP-CCF-FC-DGN 

OEP-XHE-FO-FUEL 

AFW-XHE-FO-TRBMC 

ACP-TF M-EQ-57X68 

CCW-HTX-EQ-lf.B5A 

CC\V-ACX-EQ-CHILL 

CCW-BK \V-EQ-CIRC 

OEP-TNK-EQ-TI<62X 

HPI-TNK-EQ-R WST 

OEP-CCF-FC-DGN 

OEP-CCF-FC-DGN 

AF\V -:-CCF-FC-AFW 

AFW-CCF-FC-AFW 

HPI-CCF-FC-HPSI 

AFW-TNK-EQ-D\VST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

ACP-TFM-EQ-57X68 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-TDP-LF-P25B 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-XHE-FO-TRBMC 

AFW-TNK-EQ-DWST 

AFW-TNK-EQ-DWST 

AFW-TDP-LF-P25B 

OEP-XHE-FO-FUEL 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-CCF-FC-AFW 

AFW-TDP-LF-P25B 

AFW-XHE-FO-TRBMC 

OEP-XHE-FO-FUEL 

OEP-CCF-FO-DGN 

AFW-CCF-FC-AFW 

Nonseismic 

Probabilitv 

SF 

SF 

SF 

SF 

SF 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

5.0E-02 

l.OE-02 

2.6E-03 

5.0E-Olf. 

5.0E-Olf. 

2.0E-Olf. 

2.0E-04 

2.0E-04 

2.0E-Olf. 

2.0E-Olf. 

2.0E-Olf. 

1.3E-04 

L3E-04-

2.0E-06 

5.2E-07 

4.2E-07 

SF indicates that the cut set consists entirely of seismic related failures for which 

HCLPFs will be determined. 

4-12 
't • <· .. ~ .. ·~. ' ' • J 

. ~ . ' ' ; 

Table 4-7. Boolean equations for no LOCA and small LOCA 

accident sequences. 

No LOCA Case 

Core Damage= (4+20) * (8+15 + 17+22) + 8 * (14+16} + 7 * 15 

Small LOCA Case 

Core Damage = 4 + 7 + 20 

4-13 
.;:. 

I 

~ 

.;:. 

Seismic 

LOOP 

T1 

Reactor 

Subcri ti col 

c 

t Success 

t Foilure 

AFW/EFW 

L 

Figure 4-1 

Feed & Bleed 

Actions 

p1 

HPSI/R 

0 Status Sequence 

1 OK 

2 OK 

3 CD T1LD 

4 CD T1 LP1 

5 Not Evaluated Further 

Seismic Event, LOOP Event Tree 
~ 

I 

.... 

(J"1 

Seismic~ .LOOP 

LOCA Sz 

t Success 

t Failure 

Reactor 

Sub critical 

c 

f 

I 

AFW/EF'N 

L 

I 

Feed & Bleed 

Actions 

p2 

I 

HPSI/R 

D 

Status Sequence 

1 OK 

2 co sp 

3 OK 

4 CD sz-o 

5 CD sp2 

6 Not Evaluated Further 

Figure 4-2 Seismic Event. LOOP Concurrent with Small LOCA Event Tree 

' , I~ 

: .• !,,'· ' 

," 

"·.' 

5. ENGINEERING AND METHODOLOGY INSIGHTS 

Based on the trial plant review, several insights into seismic margin studies have been 

gained. Section 5.1 discusses engineering and operational insights, and Section 5.2 

discusses insights on the methodology and execution. 

5.1 Engineering and Operational Insights 

One of the most valuable results of a seismic margins review is finding and strengthening 

components and structures that may contribute to lower seismic capacity. Most of the 

items discussed below were noted during the two walkdowns or during the systems 

analysis, although a few, like the installation of new station batteries, were already 

planned. These changes have either already been performed or will be accomplished 

during the next refueling outage in March 1987. 

1. The station service transformers, X-507 and X-608, transform power 

from the 4160-V emergency buses to the 480-V emergency buses. As 

discussed in Chapter 4, failure of these transformers could lead to 

core damage in the case of a small LOCA, and contribute to core 

damage in the non-LOCA case. Because of their low seismic 

capacity, they are being upgraded during the outage. 

2. Station batteries 1 and 3 are being replaced with new lead calcium 

batteries that have high seismic capacity. If the older lead-antimony 

batteries had failed due to a seismic event, core damage could have 

resulted. 

3. Anchorages on the containment spray pump area fans FN-44A and B 

will be strengthened. Failure of these fans could have led to long 

term heat-up and failure of the containment spray pumps, with 

subsequent failure of high pressure safety recirculation if recovery 

actions were not effective. 

4. Block wall VE 21-1 will be strengthened to prevent its potential 

collapse from failing the containment spray area fans FN-44A and B 

discussed above. 

5. Block wall VE 21-3 will be strengthened to prevent its potential 

collapse from rupturing a SCC pipe, and thereby failing SCC. This 

failure by itself would not have caused core damage. 

6. The anchorages of the chillers for the computer room air conditioners 

AC-IA and B and the lab air conditioner AC-2 will be strengthened. 

Failure of the heat exchangers on these chillers could have failed the 

pressure boundary integrity of the SCC and PCC, and resulted in core 

damage upon loss of component cooling water. 

7. A procedure was developed to isolate nonessential PCC lines and heat 

exchangers inside containment following a large earthquake if the 

PCC surge tank low level alarm annunciates. Although the PCC was 

designed to seismic standards, the project team could not enter 

5-1 
containment to verify the capacity of the components. Isolating the 

PCC lines provides assurance that any potential failure of the PCC 

pressure boundary integrity inside containment will not fall the entire 

PCC system. After examining the effect of the isolation procedure 

on the system fault trees and Boolean equation, the minimal cut sets 

showed that a single failure still existed due to selection of an 

isolation valve powered by the opposite redundant train. Failure of 

that one power train would have failed both the SCC and the Isolation 

of PCC. The isolation procedure was revised accordingly to remove 

this potential single failure. This is one example of the ability of the 

fault tree logic modeling process to find small details that could 

affect seismic capacity, and to assist in formula tlng emergency 

procedures .. 

8. An unanchored monitor in the main control room panel was 

anchored. Its impact on other components and the seismic capacity 

of the plant therefore did not have to be evaluated. 

9. The emergency lights were strapped and anchored. 

10. A missing bolt on the anchorage of some level transmitters for the 

R \VST was replaced. 

11. Loose pressurized gas cylinders, a welding machine, and some heavy 

parts near the containment spray pump area fans were moved or tied 

securely. 

12. Anchorages for the DG day tanks were strengthened 

As can be seen, some of these changes had a major impact on the plant seismic capacity, 

while others probably did not. However, the changes brought about by the formal 

exercise of performing walkdowns and developing and solving the system fault trees 

demonstrates the power of the techniques, and the usefulness of a seismic margins 

review. 

5.2 Insights on the Methodology and Execution 

5.2.1 System Classification and Screening Guidelines 

Overall, the guidance given in NUREG/CR-4334 and -4482 was very helpful in classifying 

the systems according to "Group A" and "Not-A." However, there are three areas where 

the trial plant review has provided more insight. 

The first is with respect to the safety function of reactor subcriticality. This function 

can be performed by either the insertion of the control rods, or by boron injection using 

the boric acid transfer (BAT) system. Although the control rod drive mechanisms are 

screened out in Table 5-l of NUREG/CR-4334 for review earthquake levels of 0.3 g, the 

reactor internals are not screened out based on insufficient information. This meant that 

both the reactor internals and the BAT system components were included in the 

information gathering process before and during the first walkdown. The BAT system is 

fairly complex, with numerous components that are not initially screened out. 

Appreciable resources were expended in gathering information concerning the BAT 

system. This was unnecessary after the seismic capacity review of the reactor 

internals. It would probably be more efficient in future seismic margin reviews if initial 

5-2 
,., 

effort is placed in verifying that the reactor internals have high capacity, and only look 

at alternate means of subcriticality, such as the BAT system, if this is not the case. If 

necessary, this examination of alternate systems could be accomplished during the 

second walkdown. 

The other two insights concern the safety function of emergency core cooling (early). 

Guidance in NUREG/CR-4334 and -4482, based on evaluation of previous PRAs for PWR 

plants, states that the emergency core cooling (early) function is included in Group A, 

while the emergency core cooling (late) function is in Group Not-A, and therefore is 

screened out of the analysis. There is the caveat that this screening is conditional on not 

finding any extremely gross plant-specific differences. 

The systems analysis team therefore included the initial switchover phase from 

emergency core cooling injection (early) to emergency core cooling recirculation (late) as 

a screening verification step in the first plant walkdown. While the guidelines are 

ambiguous, this screening verification included long-term area cooling for the 

recirculation systems. It was determined that the containment spray pump area cooling 

fans FN-44A and B, and a block wall near the fans, VE-21, could not be screened out 

based on the first walkdown. Based on the plant Boolean equation for small LOCA, both 

of these items were single failures resulting in core damage in the long term if not 

recovered. HCLPFs for these items had to be calculated in order to determine plant 

seismic capacity. As noted above, the utility will !Tiake changes to these items to 

increase their capacity so that they do not impact over' all plant capacity. Based on these 

findings, guidance in NUREG/CR-4334 and -4482 should be revised to insure that 

potential failures such as these are explicitly evaluated during a seismic margins review. 

For the transient case (no LOCA), the emergency core cooling (early) function is defined 

in NUREG/CR-4334 and 4482 as achievement of residual heat removal.. The AFW or 

EFW system at Maine Yankee or other PWR plants will achieve this balance within the 

first hour. For most PWR plants, irrecoverable failure of the emergency ac power 

system (station blackout) will not prevent the turbine-driven AFW train from performing 

early residual heat removal, and therefore satisfy the emergency core cooling (early) 

function. However, in the longer term without ac power, the station batteries would 

deplete, resulting in loss of instrumentation and AFW control power. Core damage could 

occur if de power is not restored, and manual control of the turbine-driven AF\V train or 

other feedwater source fails.. Based on the guidelines, this long-term failure of AFW was 

screened out of the analysis. If it were assumed that battery depletion and loss of 

instrumentation and control power results in loss of AFW and other feedwater sources, 

then the Boolean expression for the no LOCA case would be dominated by the seismic 

failure singletons that result in station blackout: 

• 

• 

• 

• 

• 

Failure of the SCC and PCC heat exchangers E-5A and 4B 

Failure of the station service transformers X-507 and 608 

Failure ·of the DG day tanks TK-62A and B 

Rupture of SCC and PCC because of chiller heat exchanger failure 

for the air conditioners AC-lA, lB, and 2 

Structural failure of the circulating water pump house falling the SWS 

Explicit guidance on the treatment of these long-term battery depletion sequences would 

be helpful. 

5-3 
5.2.2 Preparation for Walkdowns and Documentation 

Based on the trial review, there are two areas where the project team experience could 

be helpful to future efforts. The first involves the identification of components to be 

evaluated during the first walkdown. The identification of the front-line system 

components was relatively straightforward because of the detailed nature of the 

available information, and the small number of components. However, identification of 

support system components was more difficult. This is because these systems are 

generally more complex, have many more components and branches, and are not 

generally documented as well. In addition, the references concerning interfaces between 

the front-line systems and support systems are often ambiguous. Finally, the actual 

physical nature and location of some items, such as distribution cabinets and panels, is 

not shown on plant drawings or documentation. Based on this experience, there are two 

recommendations. First, when reviewing the plant information, emphasize the interfaces 

with support systems such as ac and de power, cooling water systems, HVAC systems, 

and instrument air systems. Document the ambiguities for later clarification. Second, 

plan to spend considerable effort tracing down these support system components during 

the first walkdown, and be prepared to make substantial revisions to the component list. 

The second insight concerns documentation and information transfer between the 

systems analysis team and the fragilities team. Many of the components identified by 

the systems analysis team for HCLPF screening or evaluation were selected because of 

the potential for component rupture to cause flow diversion and consequent system 

failure. The component itself was not needed to fulfill a safety function, but the 

integrity of the component pressure boundary had to be assured for overall system 

success. The common example was heat exchangers for nonessential equipment whose 

seismic rupture would fail a necessary cooling water system. Since it can make a 

difference to the HCLPF assessment, the systems team must make a clear 

differentiation between components that are required to function for system success, and 

components that are required only to maintain pressure boundary integrity. 

5.2.3 Systems Analysis and Pruning Process 

There are a number of insights concerning the systems analysis (fault tree) and pruning 

process which may be helpful to future seismic margin reviews. 

1. The procedure developed to isolate the PCC lines and components 

inside containment, and the automatic rupture isolation system on the 

sec greatly reduced the number of components that had to be 

considered for HCLPF evaluation. Both the systems analysis and 

fragilities analysis efforts would have been larger, and eventually a 

containment walkdown might have been necessary. 

2. As discussed above, verifying the seismic capacity of the reactor 

internals to allow subcri ticality decreased the effort which would 

have been required for the BAT system evaluation. 

3. Early evaluation and screening out of potential recovery actions and 

alternate systems, such as the small positive displacement pump for 

core cooling injection, reduced the number of components that 

required systems and fragility evaluations. 

4. Being able to define all the components on one skid as one 

supercomponent, such as the DGs, reduces the systems analysis 

effort, but the evaluation for the fragility team may not be reduced. 

5-4 
..... "; 

5.. As the fault trees are developed, it is useful to keep a list of the 

failure modes which should be considered for each component. For 

example, the pump failure modes include fail to start, fail to run, and 

test or maintenance outage, as well as seismic failure, but the pump 

appears only once on the initial fault tree. This information is needed 

later for the quantification process. 

6. In the initial trees it is necessary to include all the components that 

require support systems, including those components such as motoroperated or air-operated valves that will likely be screened out later 

because of their high HCLPF and low nonseismic unavailability .. 

Otherwise, if they are pruned from the initial trees, their dependency 

on support systems may be overlooked in the rest of the analysis. 

Also, since physical interactions between the component and 

structures, such as block walls or restraints, must be checked, it is 

better to include the component in the initial fault trees. 

7. Although a few seismic interactions, such as the possible impacts of 

potential seismically induced fires, were not evaluated, the potential 

for firewater piping ruptures to damage equipment was reviewed. 

The DG control panels and distribution panels were located under 

firewater piping that had considerable lateral sway. Upon 

investigation, however, it was determined that the piping was dry, and 

two signals would be required to fill the piping. The probability of 

inadvertent actuation was therefore negligible. The PCC and SCC 

pumps were also located under sprinkler nozzles, but the motor 

housings for these pumps were designed to prevent water from 

entering. Therefore, ruptures of firewater piping were not evaluated 

to impact seismic capacity. 

5.2.4 Minimal Cut Set Evaluation Process 

Minimal cut sets were developed at four stages in this project. 

1. Front-line system level cut sets using partially pruned fault trees, 

including their support systems, were developed just before the 

second walkdown to provide some guidance to the fragility team. 

These pointed out some potentially important system minimal cut 

sets. Although plant or sequence level cut sets could have been of 

additional assistance, because the fault trees were still fairly large 

the number of minimal cut sets would have been large as well. The 

additional effort to develop plant level cut sets before the second 

walkdown is not judged to be an effective allocation of resources, but 

the effort to develop system cut sets is effective. 

2. Immediately following the second walkdown, the fault trees were 

pruned based on the information gathered. Sequence and plant level 

minimal cut sets were then developed and transferred to the fragility 

team for use in calculating the preliminary plant HCLPF o Because 

some component HCLPFs were not yet calculated, these cut sets 

were still considered preliminary o 

5-5 
3. The plant Boolean equations containing the minimal cut sets were 

revised by hand just prior to the draft report to the Peer Review 

Group to include additional fragility information. 

4.. Some final plant Boolean equations were developed by computer for 

this report. 

Although this is more effort than originally planned, it is probably similar to that 

required in future seismic margin reviews. 

5.2.5 Schedule and Resources 

The schedule for the systems analysis tasks was judged to be adequate. The resources 

expended for each task are presented in Table 5-l. these resources were adequate for an 

experienced systems analysis team. About 10 percent of these manhours are support and 

clerical resources. 

5-6 
Table 5-l. Systems analysis resource expenditure. 

TASK MANHOURS 

1 COLLECT INFORMATION 

1.1 First Round Information 20 

1.2 Additional Specific Information 34 

1.3 Visit AE 8 

2 REVIEW PLANT INFORMATION 

2.1 Review EQ Level 0 

2.2 Initial Systems Review 60 

2.3 Indentify Components for Group A 36 

2.4 Initial Screening of Components 50 

2.5 Design Analysis/Seismic Reports 0 

3 PLANT W ALKDO\VNS 

3.1 Target Areas for First Walkdown 50 

3.2 Perform First Walkdown 84 

3.3 Simplified Analysis 36 

3.4 Document First \Valkdown 34 

3.5 Perform Second Walkdown 26 

4 SYSTEMS MODELING 

4.1 Develop Event Trees and Fault Trees 539 

4.2 Derive Accident Sequences 70 

4.3 Boolean Equation/Minimal Cut Sets 50 

5 SEISMIC MARGIN EVALUATION 

5.1 HCLPF - Components 

5.1.1 HCLPF- CDFM 0 

5.1.2 HCLPF- FAM 0 

5.2 HCLPF - Plant Capacity 10 

6 REPORTING 

6.1 Internal Review of Report 22 

6.2 Letter and Draft Final Report 253 

7 MEETINGS 

7.1 Project T earn Meetings 104 

7.2 Peer Review Group Meetings 22 

7.3 NRC/ AC RS/Expert Panel Meetings 24 

TOTAL 1532 

5-7 



:;. 

REFERENCES 

Battle, R.E., E mergencv Diesel Generator Operating Experience. 1931-1983, 

NUREG/CR-4347, Oak Ridge National Laboratory, Oak Ridge, Tennessee, July, 1985. 

Budnitz, R.J., A mica, P .J, Cornell, C. A., Hall, W .J ., Kennedy, R.P ., Reed, J. W ., and 

I 

Shinozuka, M., An Approach to the Quantification of Seismic Margins in Nuclear Power 

Plants, NUREG/CR-4334, UCID-20444, Lawrence Livermore National Laboratory, 

Livermore, California, July, 1985. 

Carlson, D.D., Interim Reliabilitv Evaluation Program Procedures Guide, NUREG/CR2728, SAND82-1100, Sandia National Laboratories, Albuquerque, New Mexico, January, 

1983. 

Fleming, K.N. et al., Classification and Analvsis of Reactor Operating Experience 

Involving Dependent Events, EPRI-NP-3967, Electric Power Research Institute, Palo 

Alto, California, June, 1985. 

Fletcher, D.C., Accident Mitigation Following a Small Break with Coincident Failure of 

Charging and High Pressure Injection for the Westinghouse Zion PWR, EGG-CAAD-5428, 

EG&:G Idaho Inc., Idaho Falls, Idaho, April 1981. 

Kolb, G.J., Interim ReliabilitY Evaluation Program: Analysis of the Arkansas Nuclear 

One - Unit 1 Nuclear Power Plant, NUREG/CR-2787, Sandia National Laboratories, 

Albuquerque, New Mexico, June, 198 2. 

Nuclear Regulatory Commission, PRA Procedures Guide, NUREG/CR-2300, Washington, 

D.C., January, 1983. 

Prassinos, P.G., Ravlndra, M.K., and Savy, J.B., Recommendations to the Nuclear 

Regulatory Commission on Trial Guidelines for Seismic Margin Reviews of Nuclear 

Power Plants, NUREG/CR-4482, UCID-20579, Lawrence Livermore National Laboratory, 

Livermore, California, March, 1986. 

R-1 

APPENDIX A 

IDENTIFIERS AND SYMBOLS 

0 

Q 

0 

;· . . . ·~ .... '~. 

Table A-1. Fault tree symbols. 

AND CA TE (AC) 

OR GATE (OC) 

BASIC EVENT (BE) 

UNDEVELOPED !VENT (UE) 

DEVELOPED EVENT (DE) 

DESCRIPTION 

TRANSFER IN 

TRANSFER OUT 

A-1 

- Output fault occurs if all of the input 

faults occur. 

- Output fault occurs if at least one of 

the input faults occur. 

.. An initiating fault requiring no further 

development. 

- An event which is not developed 

further, because it is of insufficient 

consequence or information is 

unavailable. 

- An event that could be further 

developed or is developed elsewhere but 

is treated here as a primary event. 

- Contains the description of an event. 

Indicates that the tree is developed 

further at the occurrence of the 

corresponding TRANSFER OUT. 

- Indicates that this portion of the tree 

must be attached at the corresponding 

TRANSFER IN. 
', ....... / 

Table A-2. Fault tree event identifiers. 

All basic, developed, undeveloped events and tab-or gates are to be coded with the 

following format. 

XXX-YYY-ZZ-AAAAA 

where 

XXX = System Identifier 

YYY = Event &: Component Type Identifier 

ZZ = Failure Mode Code 

AAAAA = Selected by analyst, try to indicate what the failure involves 

(e.g., rather ~han PSlO, use PTRNA) 

Gates should be labeled with an alphanumeric 10, e.g., AFW 1, HPI.5, etc. 

A-2 
Table A-2 (Cont'd) 

System Identifiers 

ACP AC Power System 

AFW Auxiliary Feedwater System 

CIS Containment Isolation System 

CSS Containment Spray System 

DCP DC Power System 

HPI High Pressure Safety Injection System 

MCR Main Control Room 

OEP Onsite Electric Power System 

PCC Primary Component Cooling Water System 

PPS Primary Pressure Relief System 

RAS Recirculation Actuation System 

RPS Reactor Protection System 

SCC Secondary Component Cooling Water System 

SIS Safety Injection Actuation System 

S\VS Service Water System 

I ~ , ; \ 

I ~ • ' 

A-3 
Table A-2 (Cont'd) 

Event and Component Type Identifiers 

Air Cooling Heat Exchanger ACX 

Sensor/Transmitter Units: 

Flow FST 

Level LST 

Physical Position ZST 

Pressure PST 

Radiation RST 

Temperature TST 

Flux NST 

Circuit Breaker BKR 

Calculational Unit CAL 

Electrical Cable CBL 

Signal Conditioner CND 

Control Rods: 

Hydraulically-Driven CRH 

Motor-Driven CRM 

Ducting OCT 

Motor-Driven Compressor MDC 

Motor-Driven Fan FAN 

Fuse FUS 

Diesel Generator DGN 

Hydrogen Recombiner Unit HRU 

Heat Exchanger HTX 

Inverter INV 

Electrical Isolation Device ISO 

Air Cleaning Unit ACU ,. 

Load/Relay Unit RLY ·,-

Logic Unit LOG 

Local Power Supply LPS 

A-4 
Table A-2 (Cont'd) 

Event and Component Type Identifiers 

Motor-Generator Unit MGN 

Motor-Operated Damper MOD 

Solenoid-Operator Dam per SOD 

Pumps: 

Engine-Driven EDP 

Motor-Driven MDP 

Turbine-Driven TDP 

Manual Control Switch xsw 

Rectifier REC 

Transfer Switch TSW 

Transformer TFM 

Tank TNK 

Bistable Trip Unit TXX 

Air Heating Unit AHU 

Electrical Bus - DC BDC 

Electrical Bus - AC BAC 

Manual Damper XDM 

Pneumatic/Hydraulic Damper PND 

Battery BAT 

Valves: 

Check Valve CKV 

Hydraulic Valve HDV 

Safety/Relief Valve SRV 

Solenoid-Operated Valve SOY 

Motor-Operated Valve MOV ' ,• 

... Manual Valve XVM 

Air-Operated Valve AOV 

Testable Check Valve TCV 

Explosive Valve EPV 

A-5 
Table A-2 (Cont'd) 

Event and Component Type Identifiers 

Filter FLT 

Instrumentation and Control Circuit ICC 

Strainer STR 

Heater Element HTR 

Pipe Segment PSF 

Pipe Train PTF 

Actuation Segment ACS 

Actuation Train ACT 

AC Electrical Train TAC 

DC Electrical Train TDC 

Block Wall BKW 

Operator Action XHE 

Common Cause Event CCF 

Miscellaneous Aggregation of Events VFC 

Phenomenological Events PHN 

',· ,. 

A-6 
·,' 

Valves, Contacts, Dampers 

Fail to Transfer 

Normally Open, Fail Open 

Table A-2 (Cont'd) 

Failure Mode Codes* 

Normally Open, Fail Closed (Position) 

Normally Closed, Fail Closed 

Normally Closed, Fail Open 

Valves, Filters, Orifices, Nozzles 

Plugged 

Pumps, Motors, Diesels, Turbines, Fans, Compressors 

Fail to Start 

Fail to Continue Running 

Sensors, Signal Conditioners, Bistable 

Fail High 

Fail Low 

No Output 

Segments, Trains, and Miscellaneous Agglomerations 

Loss of Flow, No Flow 

Loss of Function 

Actuation Fails 

No Power, Loss of Power 

Failure (for miscellaneous fault agglomerations 

not based on segments or trains) 

Hardware 

FT 

00 

oc 

cc 

co 

PG 

FS 

FR 

HI 

LO 

NO 

LF 

FC 

FA 

LP 

VF 

HW 

* Grouping of failure modes by events or components are only suggestions. The failure 

modes listed may be used for any applicable event or component type. 

A-7 
··"'< ·r'.1. 

Battery, Bus, Transformer 

No Power, Loss of Power 

Short 

Open 

Tank, Pipes, Seals, Tubes, Walls 

Leak 

Rupture 

Seismic Failure 

Human Errors 

Fail to Operate 

MiscaHbrate 

Table A-2 (Cont'd) 

Failure Mode Codes* 

Fall to Restore from Test or Maintenance 

Normal Operations (unavailable due to planned activity) 

Maintenance 

Test 

Test and Maintenance 

Non-Specified Failure 

LP 

ST 

OP 

LK 

RP 

EQ 

FO 

MC 

RE 

MA 

TE 

TM 

XX 

* Grouping of failure modes by events or components are only suggestions. The failure 

modes listed may be used for any applicable event or component type. 

A-8 
APPENDIX 8 

AUXD..IARY FEEDWATER SYSTEM 

Table B-1. Auxiliary feedwater {AFW). 

safety Function: Supply water to the steam generators to remove reactor 

decay heat when main feedwater is not available. 

System Components: 

Tanks: TK-21 

TK-16 

Pumps: P-25A 

P-258 

P-25C 

Turbines: T-1 

Valves: Refer to Valve Table 

Support Systems : 

AC Power: P-25A 

P-25C 

DC Power: P-25A 

P-25C 

Air: TK-111 

TK-123 

TK-25 

HVAC: Pump Room 

Pump Cooling: E-86A 

E-86B 

E-86C 

Demineralized Water Storage Tank 

Primary Water Storage Tank 

Emergency Feed Pump 

Auxiliary Feed Pump 

Emergency Feed Pump 

Turbine for P-258 {Powered from Main 

Steam) 

4160V Emergency Bus 6 

4160V Emergency Bus 5 

125V DC Distribution Cabinet 3 

125V DC Distribution Cabinet 1 

Control AOVs 

{AFW-A-101,201,301) 

Isolation AOV's 

{AFW-A-338,339,340) 

Turbine Steam 

Control Valves 

{MS-P-168, MS-T~163) 

P-25A Discharge Rec i rcul ati on 

P-25B Discharge Recirculation 

P-25C Discharge Recirculation 

Actuation: Steam Generator Low Leve 1 

Instrumentation: Feedwater Control System {SG Level, Pressure) 

B-1 
Table B-2. AFW valve table. 

Operating 

Power Normal Position Fa i 1 

Valve Description ( SOV} Position (Actuation) Position 

AF\~-A-101 Flow control to SG E-1-1 120VAC 1A 0 0 0 

(1201A1) 

AFLl-A-201 Flow control to SG E-1-2 120VAC 1A 0 0 0 

{1201B1) 

AF~I-A-301 Flow control to SG E-1-3 120VAC 1A 0 0 0 

(1201C1) 

AFW-A-338 Flow control isolation valve (AFW-A-101) 120VAC 3A 0 0 0 

( 1205A) 

co AFW-A-339 Flow control isolation valve (AFW-A-201) 120VAC 3A 0 0 I 0 N (1205B) 

AFW-A-340 Flow control isolation valve (AFW-A-301) 120VAC 3A 0 0 0 

(1205C) 

MS-A-173 AFW pump B turbine trip and throttle valve mechanical 0 latched open c 

MS-P-168 Turbine steam supply pressure control 120 VAC Bus 4 0 0 0 

(1106) 

MS-T -163 Turbine steam supply control 125 VDC Batt 3 0 C( SIAS/CI S) C* 

(1102) 

* Fails open on loss of solenoid power. 
·.- ,.'. ·' 

Table B-3. AFW cooling requirements. 

P-25A • Oil cooler E-86A 

• Recirculation to DWST 

P-25B • Oil cooler E-86B 

• Recirculation to DWST 

• Designed to operate under elevated temperature conditions 

P-25C • Oil cooler E-86C 

• Recirculation to DWST 

T-1 • Does not have a high temperature interlock 

Pump Room • Open door, portable fan 

B-3 
~m FLOW FROM 

PS-1 

! 

I 

!COMMON ~AUSE 

l FAILURE JF AFW I 

I I 

AFW-CCF-FC-AFW 

(.-... .... 

I ! I .I 

- I ; NO. FLOW FF.DM 

l'PS-2 

I 1 

!FAILURE TO 

I 

"E!:~o~ '11~ 1 ! E" 

I 

r\ .... 1 .~ .. v 

IN ANY ~G 

W/ANY PU~P 

AFW1A ! 

(-... .. 

u I 

i FAILURE QF AF(4 ! 

! 

i 

i 

j 

! 

! 

! 

l 

AFW! 

1---__,.--....J ~LJ~~ AFW2 I AFW3 I 

,; 

/\PAGE 2 

L..l 

I 

FAILURE OF 

AFW-A-201 TO 

ALLOW FLOt4 

! ! 

! ) NO FLO~ FROM FAILURE QF I j FAILURE OF 

' !PS-7 AFW-A-339 TO I lAF;J-A-301 TO 

I i. ALLOW FLOW ,, I All ;;w I=LQW 

I

! J 

. I f 

I I ! II ..... 

t...---...,..----.1.1 l ~~........,..~=-_; ,._' ____ --..~ 

AFW-AOV-¥.X-2(l1 AFW5 I AFW-AOV-f.X-339 AFW-AOY-n-301 

(--...., ·\ PAG::' 2 (-..i t-.,, U L· ~ U 0 

!FAILURE OF 

!MAIN CQNTRCL I , RC"'!c.l Un • ;:>"~IEl::; 1111 ... ., 

I 

I 

MCR-ACT-~X-CNTRL (-... , I i I .. _. •' 

I NO FLOW FROM i 

IPS-3 I 

AFW4 I 

(\ 

kr·l I 

I

• ~m FLow FRot1 

PS-7 

I 

I 

I AFO-A-340 TO 

I I ALLOW FLDW 

II H'-.F-WS--~,----~ A'-F-W--A-0-Y-~k-X--Z-A_O--.~. ll ,. .. ,\ 

i \ PAEE 2 ! J LJ. ~ ........... 

Figure 8-1 Auxiliary Feedwater System Fault Tree. 

B-4 
. ' -

1 .. ' l1 POl= n.. .\F~' j

1 

,ttL ..... n ... :Jr r: " 

!PUMP P-258 

l 

' i 

I FAILURE TG I 

l rLACE P-25B IN I 

I ~·I' I'RE. !"iF rH L.Ut ._, 

AFW-A-101 TO 

ALLOW FLGW 

AFW-AOV-~X-101 (-•.. 

~ .. _. ) 

PS-9 

AFW6 

FAILURE OF 

DWST lTK -21 l 

\ c'"-vtr.. i r· .c II A~F:':""W6-A-~-__.: A'-F-W--T-tl-K --~-X --l:-·l4\,;-:T_. 

(--·~ 

\ ___ l 

Figure B-1 (cont.) 

B-5 

I ~JO FLOW FRGN 

lot:- ,, ..... I 

I 

I 

I NO FLOW FROl1 

!PS-7 

I 

AFW5 I 

/-'\ 

u ! 

rlO ~LOW FROM 

PS-8 

I 

!FAILURE OF 

IAFW-A-338 TO 

fALLOW FLOW 

I 

I . PS-10 

I

' I l I I 

AFW9 ~' --~~--~' AFW12 '~~~--~ f\ nA"".. A 1 \I Ut. r 

1-..::. 

!TURBINE DRIVE 

IT-l I . 

I 

AFW7A ,.,_ 

f \ , .. .,.. ... : , t 'f 

i 

IFAILUF:E QF 

jMS-?-168 Tu 

i ALLOW FL\JI~ 

j 

I 

AF~~-Pcv-n:-P t68 ,..-.. f i 

I I i I 

·-· 

/\PAGE 5 

u 

!FAILURE OF 

!MS-T-t63 TO 

jALLO:.l FLOW ) 

i 
I 

!FAILURE iD •

1 

PLACE P-258 iN 

SERVICE 

IREJ10TELY 

i 

AFW6B 

NO PONER FRO" 

TO START P-25B I 12S YDC BUS 4 

FROH CONTROL I I 

ROOM II 

AFW-Y.HE-FO-TRBHC D~C--P-~BD--C-~~-~oU.-94~ 

(1 ¢ 

FAILURE OF 

HS-T-163 AIR 

ACCUI1ULATOR, 

TK-25 

Figure B-1 (cont.) 

B-6 

FAILURE iO I 

PLACE ?-255 HI 1 

!SERVICE I 

I I 

AFW6A I 

.. -.. 

u I 

FAILURE iO I 

PLACE P-258 IN I 

SERVICE 

LOCALLY 

AFi'i6C 

JoPERATOR FAIL I OPERATOR 

l

·iD START P-258 FAILURE TO 

FROH CONTROL SiART P-25& 

R0011 LOCALLY 

AFij·XHE·FO-TRBiiC Afi~-XHE-=:'Q-TRBLO (j () 
AF!49 

NO ?OWER rRut1 

125 IJDC sus 1 

l@ 3iEAN GEN I I I LCW LEVEL ACT. I 

I NO FLJW TO 

lpl"\-::l I :l " 

f 1sr6NAL ! 

I I I I I 

! 

DCP·SDC-~P-3US1 RPS-4CT-FA-SGLEV 

0 <) AF~lO l 

0 

Figure B-1 (cont.) 

B-7 

I 

I FAILURE OF I I NO FLCW FROM 

IDWST lTK-21) I !PS-15 

I 1 I 

I I 

.I I 1..___..--------: 

AFW-TNK-kX-DWST AFWll 

(=) 

lFAILURE OF 

IPWST !TK-i61 

I I 

AFW-Tt~K-h -PWST 

r-··~ '.._,/ 

lu-P~PATr~ I 

''"" IJn I FA I LURE TO l'lt 

I LWE UP PWST 

l I ! . 

AFW-XHE-FD-EFWXI r-·.\ 

i.· ) .. _ 
AF~12 

Figure B-1 (cont.) 

B-8 

FAILURE QF 

ID

IJ""T nlJ. 'TU ·"·t' \ ,.,-.; I 

I 

NO FLO\i iu 

PS-1() 

AFW13 

UO FLO~ FROH 

PS-16 

I I 

I I '-- 1 --r--' 

AFW-TNK-MX-D~ST AFW14 ,. .......... 1 

I \ l .. ..._ 

FAILURE OF 

PWST \TK-16} 

I u. t1 """ 

!FAI~URE TO 

!LINE UP PWS7 

l 

i 

AFW-XHE-FO-EFWY. ,.,.--4 

I I t .... _ ... } 
FAILURE TO 

I PLACE P-258 IN I 

I SERVICE FROM I 

MCR I 

I I I I 

AFW6a I 

(''\ 

. I ~.,.,.., 

SE!S!HC 

FAILURE OF 

MS-T-l63 AIR 

ACCUMULATOR 

AFW-TNK-E!2-TK25 

n .......... 

I SEISMIC 

FAILURE OF THE 

1 DWST nK-21 l 

l 

~ • - J v -t0-:'1 t: 

I FA!LURE

1 

TO I I I PI "-!1 ·'CJ:' "- P-?C.B ' ~... IU I~ I 

I SERV! CE FROM I 

IMCR I 

AFt4-XHE :10-TRBHC 

( · .. 

l I I ' ·-..... ,i 

I 

NO FLOW FROM 

PS-9 

AFW6 I 

(\ 

'-r 

FAILURE TO 

PLACE P-2SB IN 

SERVICE 

I 

AFWGA I 

;'""\ 

( . ' 

FAILURE TO 

PLACE P-25B IN I 

I SERVICE FROM I 

IHCR I 

AFW-XHE-FO-TRBMC r-·, 

I J .......... 

NO FLOW FROM 

PS-7 

AFW5 

0 

I 

t!O FLOW FROM 

PS-8 

FAILURE TO i 

KEEP NIN LEVEL I 

IN ANY SS BY 

ANY pp I 

AFWl i 

r~ .. . I\ 

kr-·1 

I 

I 

I NDr~-SE!SHIC 

I COi1!10N CAUSE I 

I FAILURE OF AFW ! 

I I 

AFW-CCF-FC-AFW c--) 

.._ 

NO FLOW FROM 

PS-1B 

AFW9 

/\ 

I~ PASE 2 /\ PASE 3 

w w 

I 

J P.I\NDOM 

FAHURI.; 0~ 

11'·258 

I I 

ffr\.J·1'DP, u=·PZ.5'3 

0 

I FAILURE, TO I PLACE P-258 IN 

SERVICE 

LOCALLY 

AFWGL I 

i·· 

I \ ~"T"l I 

I FAILURE' iO I I pr ACE P-?lirt p1 I I SERVICE ...... " II' 

!LOCALLY 

I I AFW-XHE-~D-TRBLO (-'\ 

I ! I ........, ' 

Figure B-2 Auxiliary Feedwater System Fault Tree, Pruned and Merged. 

8-9 
t 

SEISMIC 

FAILURE )P 

TRANSFORMER 

(X-507, X-608) I 

I I I I 

ACP-TFM-EQ-57X68 

{'\ 

t._) 

SE!SiHC 

FAILURE OF THE 

DIESEL DAY 

TANK TK62A/B 

OEP-TNK-EQ-TK62I 

0 

POWER tiOT 

AVAILABLE FROt1 

DIESEL 

GENERATOR lA 

OEP-DGN- P-DGIA 

i"*'\, 

( I 

~ I 

I 

NON-SEISMIC 

COHMON CAUSE 

C•G FAtLURE 

I FAULTS IN 

'I DiESEL 

GENERATOR lA 

l 

I 

OEP-PSF-FC-DG1A ( .... _\ 

I ,I 

·-I 

SEISMIC 

FAILURE OF AC 

CHILLERS 

AC-iA/18/2 

.... ~ 

! tm FLOW FROM 

I PS-8 

l 

I ,.., 

r ·, 

_i '< 

~..,....j 

I 

POWER NOT !sEISMIC I 

AVAILABLE fROM I !FAILURE OF I 

416ll VAC BUS 51 II TRANSFORMER I 

i 

I (Y.-507, X-6~8) 

_ I I . I A~C--P---B--AC---r-lP-----BU--55-J ACP-TFH-tQ-57X68 

(\l (--,) 

-..._.; 

SEISMIC I 

FAILURE OF THE, 

INa FLow· ro 

I PS-8 . 

I 

AFWl~ ,.~·-.\ 

I. l 

I I 

SE!SM!C 

FAILURE OF. 

TRANSFORMER 

I (X-507' X-608) ! 

IDWST (TK-21) ,, 

I I ~.---1 _____,..-----J 

AFW-TNK-EQ-DWST AFWll 

NO POWER FROM 

DIESEL 

GENERATOR lA 

D61Al 

r···~ ( _I ........ 

SEISMIC 

FAILURE OF THE 

I PWST lTK-t6l I 

I I 

I FAILURE' F:EFILL I ! NO PCC CDOL!N6 ! 

I FUEL TANKS l l TO NECESSAF:Y I 

I (P33, VLV} ! 'SAFETY SYSTEM I II - I l t I ! I COMP ! ' I 

OEP-XHE-FG-FUEL PCCl (-.) 

. ...,., ~ 

! 

SEl5lHC : FP.ILUF:E TO I 

FAILURE OF !CLOSE PCC I I 

PCCi SCC COOLE.q !ISOLATION VLVS i E-~B/E-5A I I 

PC!:-XHE-Fu-iSQL (-.. ~ 

l ! • ... ._.111'· 

OPERATOR I 

FAILURE TO 

I OPEN PWST I SOli 

VALVES • 

I I AFW-XHE-~0-EFWXX ( ..... \ 

1. ! I . 

·--

i 

SEIS~[C 

FAILURE OF 

CIRC WATER 

PUMP HOUSE 

(SWSi ~ 

Figure B 2 (cont.) 

f3-l0 
I SEISMIC I 

FAILURE THE I TR ·•Js .. .,R)1J:'I:, 1 . Ha ru ... M .. 

I !X-557. X-608) I . I I 

' SE!S~IC 

FAILURE OF THE 

DIESEL DAY 

TANK iK6l~/B 

.... ,, f""""' .... , ..... 

POWER NOT 

!AVAILABLE FROM 

I

DtESEL I 

6ENERATOR 1B I 

I I 

OEP-DGN-LP-DS!B ...... r ., .I \ 

~--r--1 ! 

i, POWEF: NOT l 

l AVAILABLE FROM l 

!4169 VAC BUS 61 

I I 

ACP-BAC-LP-BUS6 

/"· 

j' \ 

~r-i 

40 FLOW FROM 

pc:-10 I "" .. 

! 

I 

I 

AFW12 

!CETt::Mtr 

!FAiLURE THE 

I T~J\~'SFQnMJ:'R (.-.H.~ l\na...a• I 

I !X-597, X-608) ! 

I . I 

ACP-TFN-EQ-57X68 

r-·~ I I -

INO FLOW TO 

j F'S-10 

I 

~FW13 

'

SEISMIC SElSMIC INO FLOW FROM 

FAILURE THE FAILURE OF THE, PS-16 

! TRANSFORMER I DWST (TK-2U II I I (X -597. X -6B8) I 

I . I I I !'---__,______. ACP-TCM-~D-~7X68 AFW-TNK-EQ-DWST AFW14 "'t~) wo' (-.,\ 

' .......... · '-._) 

I

NON-SE!SMIC !NO POWER FROH I 

COMMON CAUSE DIESEL 

SEISMIC I 

FAILURE OF THE I 

P!~Si (H:-16) I 

1

0PERATOR 

FAILURE TO 1 ! DS FAILURE ISENERHTOR 1B OPEN PWST !SOL ' 

OEP-CCF-FC-DSN ~~------~! DStBl ~! --~--~ {' .... '\ 

I I 1.._! 

I 

I FAILURE

1 

I FAULTS IN I REFILL I l nr!:'C:;:"l I !FUEL TANKS I U koW ........ 

I GENERATOR 1 B I I IP33, VLV) I I I 

I I I I I I I 

I 

SE!S!1IC 

FAILURE OF AC 

CHILLERS 

AC-lA/18/2 

l""HI9' 

I tm sec 

TO ~JECE 

SAFETY 

COMP 

SEISMiC 

I 

I 

COOLING I 

SSARY 

SYSiEl1 l 

I 

i 

I 

! 

FAiLURE OF 

PCC/SCC COOLEE 

E-48/E-SA 

,.. "' 

!VALVES I 

: 

SEISiHC 

· FAILURE OF 

CIRC WAT ..... 

L't 

PUMP HOU SE 

(C::'·'O\ ... rc ... 

.. 1"'\ I I -

Figure B-2 (cont.) 

8-~1 
j 
APPENDIX C 

HIGH PRESSURE SAFETY INJECTION SYSTEM 

Table C-1. High pressure safety injection (HPSI) and containment 

spray pump area cooling {CSPPCL). 

safety Function: Inject borated water into the reactor vessel immediately 

after a LOCA. Also for feed and bleed, post-accident 

core cooling and additional shutdown capability during rapid cooldown of RCS. Spray pump area cooling to · ensure long-term availability of containment spray 

pumps. 

System Components: 

Tanks: 

Pumps: 

Fans: 

Valves: 

Support Systems: AC Power: 

DC Power: 

Pump Cooling: 

Actuation: 

TK-4 Refueling Cavity Water Storage Tank 

P-14A (N.O.) Charging (HPSI) Pump 

P-148 (S) Charging (HPSI) Pump 

P-14S (Spare) Charging (HPSI) Pump 

FN-44A 

FN-448 

Spray Pump Area Fan 

Spray Pump Area Fan 

Refer to Valve Table 

P-14A 

P-148 

P-14S 

FN-44A 

FN-448 

P-14A 

P-148 

P-14S 

P-14A 

P-148 

P-14S 

Motors 

P-14A 

P-148 

P-14S 

FN-44A 

FN-448 

4160V Emergency Bus 5 

4160V Emergency Bus 6 

4160V Emergency Bus 5/6 

480V Emergency MCC 78 

480V Emergency MCC 88 

125V DC Distribution Cabinet 1 

125V DC Distribution Cabinet 3 

125V DC Distribution Cabinet 1, 3 

PCC 

sec 

PCC 

Air Cooled 

SIAS A 

SIAS B 

SIAS A/8 

SIAS A 

SIAS 8 

C-1 
E .. .34 

E-67 

E-96 

FL-.34B 

TK-.54 

Table C-1 (Cont'd) 

Additional Components Whose Failure May 

Lead to HPSI Failure 

Seal Water Heat Exchanger 

Reactor Coolant Regenerative Heat Exchanger 

Seal Water Heater 

Seal Water Supply Fll ter 

Spray Chemical Addition Tank 

C-2 

Isolated by 

CH-A-32 and 

CH-A-33. 

Catastrophic failure may 

cause failure of R WST 

and/or interconnecting line. 
Table C-2. HPSI/CSPPCL valve table. 

Operating 

Power Normal Position Fail 

Valve Description ( SOV) Position (Actuation Position 

CH-A-32 HPSI pump B discharge to charging header 125VDC DC/CE-1 0 C( SIAS-A) c 

CH-A-33 HPSI pump A discharge to charging header 125VOC 11C/CE-1 0 C( SIAS-A) c 

CH-F-38 Inlet to charging header 125VDC OC/CE-2 0 C( SIAS-B) c 

HSI-M-11 HPSI train B to Loop 1 injection MCC BA c O(SIAS-B) AI 

HSI -M-12 HPSI train A to Loop 1 injection MCC 7A c 0( SIAS-A} AI 

n HSI-M-21 HPSI train B to Loop 2 injection MCC BA c 0( SIAS-B) AI I 

w HSI-M-22 HPSI train A to Loop 2 injection MCC 7A c 0( SIAS-A) AI 

HSI-M-31 HPSI train B to Loop 3 injection MCC BA c 0( SIAS-B) AI 

HSI-M-32 HPSI train A to Loop 3 injection MCC 7A c 0( SIAS-A} AI 

HSI-M-40 HPSI train B discharge cross-connect to train A MCC BA c C* AI 

H SI -M-41 HPSI train A discharge MCC 7A c 0( SIAS-A} AI 

HSI-M-42 HPSI train B discharge MCC BA c 0( SIAS-B) AI 

HSI-M-43 HPSI train A discharge cross-connect to train B MCC 7A c C* AI 

HSI-M-50 RWST supply to HPSI pump A MCC 781 c 0( SIAS-A) AI 

C( RAS-A} 

* Open for recovery action. 
("'") 

I 

~ 

Valve 

HSI -M-51 

HSI-M-54 

HSI-M-55 

SL-P-3 

Table C-2 (Cont•d) 

Description 

RWST supply to HPSI pump B 

Recirculation supply to HPSI pump B 

Recirculation supply to HPSI pump B 

RCP seal water inlet 

Power Normal 

( SOV) Position 

MCC 881 c 

MCC 7Bl c 

MCC 8Bl c 

125VDC DC/CE-2 0 

{211) 

Operating 

Position Fa i 1 

{Actuation) Position 

0{ SIA 5-B) AI 

C( RAS) 

0( RAS.:.A) AI 

0{ RAS-B) AI 

C( SIAS-B) 0 
P-14-A 

P-14B 

P-14S 

Pump 

Cubicles 

Table C-3. HPSI cooling requirements. 

• Air-cooled motor 

• PCC cooled stuffing box (seals) 

• Shaft-mounted lube oil pump for bearings and gear (P-14A-2) 

• Electrical lube oil pump for standby cooling (P-14A-3) 

• Air-cooled motor 

• sec cooled seals 

• Shaft-mounted lube oil pump (P-14B-2) 

• Electrical lube oil pump (P-14B-3) 

• Air-cooled motor 

• PCC or SCC cooled seals (PCC preferred) 

• Shaft-mounted lube oil pump (P-14-C-2) 

• Electrical lube oil pump (P-14C-3) 

• Open area, natural circulation 

• Doors to fuel building, turbine building and the garage doors may be 

opened to increase flow. 

C-5 
FAILURE OF CS 

PUMP AREA 

I EXHAUSi FM, 

FN-44A 

I 

!LOSS OF HPSI l 

.. I 

I 

HP!l 

FiiiLURE ;JF 

EXHAUSi FAN 

FN-44A 

CSS2 

I ~~0 .~CTUA ii ON - .. ·M!:.t 1"'\1'1\:-o. l tt(I.Jil ::J!H::l 

IT ....... . t-;1111'1 H 

l 

iU"!~s nt 

I

I H;; T •' ;H· 'fl D~, l ...,.~ '""' ,,.,;:\ 

jiERn) OR . Rr'r! Ri' {I fHIG I ,.....,., • ,..., • \.:w•C 

17ER';t' • 1./ 

I ~ ''l' PRE !1'" 1 rH.I..~. ..r 

if1AIN CONTROL l R00!1 PANELS 

I 

j 

FAILURE uF 

FN-44A 

ACTUATION 

! i !:iSCFF!C:ENT 

l

r r;":l~jH•l•\j)oiT Wwf'• ••••• 

f S?E A~~· ?~~p 

iHREH CJQL!NG 

I 

t

, ,, FAILURE OF CS 

PIJMP AREA 

I •EXH~UST FAN, 

· F~~-o;4B 

I l ~--:o----..i ;......-~--....J: CSS4 I CSS-F~N-~X-FN44B 

0. (-\ 

I I 

I ! 

'-. .. _,.·, 

I 

joPERATOR 

!FAILURE TO 

jSiART CS PUMP 

I FH '·N F)'-•·4· J:i ,1 '+ H!i.i 

CSS· :<HE -Fu-FN44 

r--. 

I I 

\ J .. _., 

JFAILURE; OF 

'

EXHAUST FAN 

FN•44B 

I 

i 

I NO POWER FROM I 

l480 iJAC MCC 3B l I~ ae1 l 

I I ACP-3AC-LP-~CC8B 

i\ \:! 

t li"' ~c·u·· .. ""\' ! ;,u H i 111 w,1 

j F?.Oi1 31AS 

! iRA HI B 

! 

i 

! 

IFHi~uRE JF 

,FN-449 

i ACTJAiiON 

l 

I 

CS55 ,.-.., 

, ., 

i 4! 

I ; 

7 

CSS-~HE·F~-FN44X .-. 

{ \ I I 

\ ._,. ·' 

Figure C-1 High Pressure Safety Injection System Fault Tree. 

C-6 
-· •• ''ii .. r:· .. :-HtLU .. t. Jt' 

iSl·~-11 TO ~LLOW FLOW 

=-I-MGV-~X-~911 (--·. 

I I l ! .. _... 

LJ 

Hf'l32 I 

/'\ 

~l 

I NO FLOW iO ANY I 

I

OF 3 LOOPS 

FROM ANY OF Z 

1

PUMPS 

I 

HPl-?SF-FC-?UMPS / .. \ 

u I 

LOSS OF HPSI 

HP!! 

I I 

I

NO FL~W TO ! l NO FLOW TO UO FLCW TO 

INJECTION LOOP I INJECTION LOOP lNJEC~lON LOOP LJ 2 1.-3 __ __,1 

HPI2 I HPI'3 I HPI4 I 0 6 PAGE 1 /J. PAGE S 

!No FLOI,FROM 

!PS-4 

I 

HP!S I 

/\PAGE 7 

w 

NO ?C~ER FROM t~O ACTUATION HJO FLOW FROM 

!480 JAC MCC SA li I"FROH SlAS lPS-9 

I TRAIN 8 I 

I I I" I 

I l1 1..._.1...._~- AcP-sAc-LP-MccaA SIS-ACT-FA-TRNB HPI21 

.::[\ :.{1. \(1'" ·:.j /\PAGE 9 ~ 

Figure C-1 (cont.) 

C-7 

I 

I

I COMMON CAUSE 

FAILURE OF 

! HPSI 

I 
I 

FAILURE OF 

HSI-M-21 TO 

"ALLOW FlOW 

HPI33 I 

/\ 

J ) 

~J I l 

NO POWER FROH I INO ACTUATION 

480 VAC KCC 9A f FRO" SIAS I I TRAn~ e I 

'---~~' t I ACP-BAC-lP-HCCSA SlS-HCT-FA-TRNB 

.:~'·.A ..• ". .A , " ~L~ w ':::;' 

Figure C-1 (cont.) 

C-8 

NO FLOW TO 

HHECTION LOOP i 

2 I 

HP!3 

!\ 

'.I u 

1 

NO FLOW FROI1 

PS-5 

HP134 I 

l 

/\PAGE 4 ~ 

I 

NO FLOW FROM 

PS-9 

HPlZl I 

/\PAGE 9 LJ. 
!FAiLURE OF 

HSI-H-22 iO 

ALLOW FLOW 

I ··o _,.._., .. ~ ,.,.n.... I 

(~Qc~:r M~~"7A ~ 

AC?-3AC-LP-MCC7A 

,() 

v 

Finure C-1 (cont.) 

C-9 

t ~\o ,..., nu !"'l' ..... ,..,w i :~ i'l..io" tnun !PC:~::; 

I'"' "' 

I I 

I 

HP!34 t 

/\ 

~~ ! 

I

' NO AC7UAT!ON 

FR0;1 SIAS 

TRAIN A 

I 

Sl5-ACi-FA-iRNA ~~ ~--~? 'o/ 

I

, NO FLOW FROH 

PS-12 

LJ 

HP!i I 

fj_PASE 7 
J 

FAILURE OF 

HSI-H-31 TO 

ALLOW FLOW 

HP I -HOV·KX -MV31 

0 

NO POWER FROH 

480 VAC HCC BA 

1

NO FLOW FRGH 

PS-3 

HPI19 

I 

I 

NO ACTUATION 

FROH SIAS 

TRAIN B 

ACP-BAC-lP-MCCBA SIS·ACT-FA-TRNB l~ A 

<V 'I 

Figure C-1 (cont.) 

C-10 

iNU FLUW 70 

I !NJECTIOU LCDP :,. 

r 

I NO FLow' FROi1 

?S-6 

HPI35 I 

/\PAGE 6 

i.....:;. 

j I NO FLOW FROH I 

PS·9 I 

LJ 

HP!21 I 

/j_PAGE 9 
FAiLURE uF 

HS!w1w32 iO 

ALLOW FLOW 

HPI-:-!OV-~X-M\132 

(-.\ 

,_,. 

1 NO POWER FRG11 I 

1190 VAC MCC 7A I 

I I 

ACP-BAC-LP-MCC7A 

Figure C-1 (cont.) 

C-ll 

l"",1"\ l 

r:-:: 

HP!35 

I NO ACTUATION 

I F .. ·~·!.l ........ KUn :tf-i.:J 

TRAW A 

I 

SIS-ACT-FA-iRNA 

!.;.':.. 

·:~ ') ·~{i 

!NO FLOW FROM 

PS-12 

HPI7 

l\ :::l•O'" 7 l \ rH.~;:t. 

J-!. 
1 FH I LURE OF I rJG P~wE~ FRO!'! 1 

IH,.,.. i ·;..I_,_ "''" I " ,.0 1 I""· 'Tt:V 'J'r ., H .. ···r" n ... t. -· ; H 1i 

1

; 

I

I ALLOW FLOW 

I.__ _II i HP!-HOV-~X-HY12 ACP-3AC-LP-1CC7A (-~·,, _,;~-. 

<<. -~:~>· \ .' \Y~ ._...,. 'wl 

1 F' ILU·'"·E n ... I H M t.:t· 

I HSI-H-41 TO 

jALLOW FLOW 

I 

!FAILURE OF 

IHPSI ?UMP 

IP-14A 

l 

NO ?OWER FRO~ 

480 !JAC HCC 7A 

NO FLOW FROM 

PS-4 

HP!5 

INO ACTUATION ~-- !NO FLOW FROM 

j .... _A .N A 

1

FROM SIAS I'JPS-12 

SIS-ACT-FA-iRNA I'" l I HPI7 ~-'-------' l~, 

~~·l 

1

NG POwER FRQH I 125 VDC BUS 1 

' ! 

l 

!NO ACTUATION 

I FRCt1 SIAS 

IT::if:PI ,•, j"'n.rl n 

I 

l 

SlS-HCT-FA-TRNA 

FAILURE or 

HSI-r!-S':' TO 

ALLOW FLO~ 

NO Po~sq FROM 1 

480 VAC ~CC 7B ! 

~ 7Bl I 

P! -HOV-?:X ... :iV50 ~-- i I 

I ! \~,.._) 

Figure C-1 (cont.) 

C-12 

I 

! flO FLOW FRiJl1 ! PS-23 . 

HPI14 

NO ACilJA T1 ON 

FR'Ut1 S! AS 

TRAif4 A 

FLOW DIVERSION 

THROUGH PS-14 

HPI9 

/..,\PAGE 8 

.L_j, 

I

I NO PCC COOLING j 

I TO NECESSARY I 

I 5'--'"C"'/":1 ~-_,,...,.,...\.1 ! ht'L..; r ...tl~lt.n ! 

. !"0' p i i IJ M. I 

PCCi 

i 

f 

!FAILURE OF 

I El~Si 

i 

I 

! 

P.WSH 

f'H.:ll:. 
FAilURE OF 

CH-A-32 TO 

ISiOP FLOW I 

HPI -,40\J-~X -CHA32 

r-· .. 

~ I ........ 

Figure C-1 (cont.) 

r:-13 

I 'L!"'~' un r• 'P 'iUtG'H. .r. .. u. 

IPS-t4 

I 

HPI10 

/'\ 

l \ 

~r~ 

NO ACTUAi!ON 

FROi1 S!AS 

TRAm A 

SIS-ACT -Hi-iRMA 

i;~\ \l 

HPI9 ...... -...... 

u ! 

HP!ll 
riHLURE OF 

HSI-M-~2 TO 

ALLOW FLOW 

1PI-liOV-~X-itV42 (] 

i NG POWER FEOM . 

!480 YAC MCC SA l 

I I 

I f 

! 

jNO FLOW FRON 

•pc ... c I ,.; I 

I 

I 

HP121 

·'"· ( ., I ': 

~i·l ! 

I"O N H 'C"""" IUHI .... I'"M Lb 

I

. rr. '"'"'ON 1 ,., 

CI·'"'' M::l 

!TRAIN B 

I 

l FLOW DIVERSION ! 

I 'l'"'"'OU"'H DC- i ~ I ,nn l:l, ' .... •w . 

1 I 

I I I I I I 

HPl22 

!NO FLu~ FR0:1 

!PS-18 

I 

HPI23 I 

i\ PllGE iO r:\ U'" ~~ 

FAILURE OF l NO POWER FROM I NO POWER FROM I NO ACTUAiiON 

II HPSI PUMP I 114160 i)AC BUS 6 ,,' 1125 VDC BUS 3 I FROM SIAS 

P-148 I TRAIN B 

I I I I l I ..__l _ ____. 

HPI-HDP-ki-PTRNB ACP-BAC-LP-BUSb DCP-BDC-lP-BUS3 SIS-ACT-FA-TRNB 

('"''}, ~t~) .. i"\ i\ \.._, \j ·f~;,/ ··.;-v'/ 

HSl-H-51 TO 

!ALLOW FLOW 

I 

HPi -M01J-\X-MV51 

r~~, 

\ .._ .I 

I

' ~0. p~~ER FROM I 

't8li ·lAC MCC BB 1 

I& BBl I 

I I 

ACP-BAC-L?-HCCSB 

Figure C-1 (cont.) 

r:- 1. t1 

INo FLow FROM 

IPS-24 

I 

i 

I 

i 

HPI27 

/'\ 

~T-l I 

I t~D ACiUA i I ON 

1

1

FROM SIAS 

TRAIN B 

I 

i 

! 

SlS-ACi-FA-TRNB 

~l\ ··~f 

FAILURE DF THE l 

RWST (TK-4) I 

I I I I . I I t f ' 

l NO SCC COOLING 

ITO NECESSARY 

I r:o .. t:T" r:"'s-r-u 'W* ,t" ~ f f ':.I I 1 tr~ 

! CDHP I 

SCC1 

I 

1 ... ,, IL'lOC n::: l rH t..;," .. U1 

I R!~ST 

I 

RWSH 
~L-P-3 FAILS 

7G STOP FLOW 

:SEAL WATER ~SOL) 

I FLOW THROUGH I 

PS-15 

u ('•) 

I I 

FAILURE OF NO ACTUATION 

CH-A-33 iO FROM SIAS 

SiOP FLOW TRAIN A 

II ~------~ ~--------~ Hri-AOV~~X-CHA33 SIS-ACT-FA-TRNA 

{

r-...) fl~":. '& :'• ........, \~f 

I FLOW THROUSH I 

C_j 

HP128A I 

{'\ 

i \ 

~T4 

NO PO;iER FRC~ 

1

m vnc Bus 3 

I 

tiO ACTUATION 

FROM S1AS 

TRAiN B 

FLOW DIVERSION 

THROUGH PS-2b I 

I 

HPI28 I 

0 

I 

I ...... ''"E ..... I"Hll.. .. ·l'\ •.;r 

!SEAL ~AiER 

IHEHiER ~E-~6i 

I 

HPI-HTX-~X-SL~TR 

(--.~ 

I ' · ... _.: 

I FLOW ~!VERSION I 

I THROUGH PS-lS I 

LJ 

HP!22 I 

/'"'·· 

u I 

I FAILURE OF I 

[_J 

HP126B I 

/ .. \ 

I '~ 

~T-~ I 

!FAiLURE OF 

!sEAL ~ATER I ... L ........ \1"1 lt.l': 

j (FL-348} 

I 

l FliJW THROUGH II 

PS-26 uR PS-27 

I 

HPill 

0 l 

FAILURE OF RCS .,. 

jREGENEF.ATIVE r \H7i 

1 

HPI-HTX-~X-RESEN 

( ..... )! 

.. ,l .._. 

!FAlLCRE OF 

!CH-F-39 TO : ::j:ip ;1 nw ~ w w · ~w 

I 

HP1-FLT-¥X-SL~TR HPI-rlOV-~X-~~F3 .. -. ,.-.. i .\ ; 

\ .! : : ,.._.. "'....J*'. 

Figure C-1 (cont.) 

C-15 

1 FLOil DIVERstmt 1 

liHROUSH PS-27 

I 

I 

HP130 I 

0 

I FLOW iHROl.!GH 

IPS-27 I 

LJ 

HP!31 l 

/'"·· 

l-~\ ! 

i ~m ACTUATION 

I

I FRO~ SliiS I iRAHl B 

I 

I 
r 

FAULTS u• 

DIESa 

GENERATOR 1A 

OEP·PSF-FC-DS1A 

0 

·NO . POWER FROM 

DIE5EL 

GENERATOR lA 

DG1A1 I 

Q 

I 

0Pb.RA10R 

!=A1 L\..\Rt;. TO 

RE.f=l LL FUEL 

iAN I(~ 

OEP-Y.HE-FO-FUa 

0 

POWER NOT 

AVAtLABLE FROlt 

DIESEL 

SENERATCR 1A 

DEP-OGN-LP-OSlA 

9 

NON-SEISIHC 

COt1110N C~USE 

DS FAILURE 

OEP-CCF-FC-DSN c) 

I 

NO PCC CCOLINS i 

TO NECESSARY 

SAFETY SYSTEM 

COI'!P 

PCCl 

LOSS OF 

HPSI<SHORT 

iERl1~ OR 

REC!RC !LONG 

iEnH) 

HPI-CSl I 

Q 

I 

LOSS OF H?Si 

POWER NOT 

AVAILABLE FROM 

48B VAC KCC 7B 

OP. 781 

I 

INSuFFICIENT 

COHTAINMT 

SPRAY PUMP 

AREA COOLING 

NO POWE.~ FROK 

4SO VAC MCC BB 

& SBl 

ACP-BAC-l?-HCC7B ACP-BAC-L?-nCCSB 

SE!SltiC . 

FAILURE Of 

TRANSFOR"ER 

(X-507, 1·6BSJ 

ACP-iFH-EQ-57168 

0 

l 

s::!:~zc 

FAILURE OF Tr.£ 

iHEEE. DAY 

TANK TiC6ZA/B 

- -

6. PA~i; 3 

Figure C-2 High Pressure Safety Injection System System Fault Tree, Pruned and Merged. 

C-16 
I 

EE!S~IC 

Ft'tiL!!RE OF At 

CHII..l£~S 

AC·1A/1B/'2 

. 

I 

SE!El1IC 

FtiiLUR£ OF 

FCC/SCC COOLE.~ 

E-4BiE-:A 

.. . .. 

NO PCC CDOLINS I 

TC NECESSARY 

SAFETY SYSTEM 

CQl'IP 

PCCl I 

Q 

I 

FAILURE TO 

CLnSE PCC 

ISOLATION VLVS 

I 

Figure C-2 (cont.) 

C-17 

, 

SE!S~lC 

FAILURE OF 

C!RC WAt 

PUMP HO' Uc:: 

.... (SHS2 
J 

FAULTS IN 

DIESEL 

GENERATOR IB 

1 

POWER NOT 

AVAILABLE FROI1 

DIESEL 

GENERATOR lB 

POWER NOT 

AVAILABLE FRO:! 

480 vAc 11cc ae I 

I& S'lll I AC?-3AC-LP-~CC83 

Q 

I 

SE!51UC 

FAILURE OF 

TRANSFORMER 

(X-507, I-6BBl 

CEP-DSN-lP-OGlB AC?-TFM-EQ-57X68 

Q 0 I -' 

NO POWER FROH NON-SEISMIC 

DIESEL COM"ON CAUSE 

GENERATOR 18 DS FAlLURE 

I DGlBl DEP -CCF-FC -DGN 

Q 0 

I r 

O?'i:'\\ATnR 140 SCC COOLING SE!!:?IlC F~-\L~R·ii' 10 TO NECESSARY FAILURE OF THE R~F'I LL R.HiL SAFETY SYSTEM DIESEl. DAY 

IANK5 CCHP TANK nco2A/B 

OEl P-P C - - - -1: - I c. I OE?-iNK-Eil-iK62l .. F FC DStE OEP Y.HE • 0 FUEL ... cc1 o ·o 0 SE!S~[C 

FAILURE OF AC 

CHILLE.~S 

AC-lA/1812 

SE!:~IC 

FAILURE OF 

PCCiSCC COOLER 

E-48/E-SA 

ScZS~IC 

FAILURE OF 

CIRC WAiE.~ 

PUI'!P HOUSE 

tSWS) 

CC~-ACX-EQ-CHILL CCW-HTI-EQ-~B5A SWS-3KW-EQ-CIRC 0 0 0 

Figure C-2 (cont.) 

C-18 
I 

POWER rmT 

AVAILABLE FROM 

DIESEL 

GENERATOR 1Q 

OE?-DGN-lP-0619 

~,..,c,;;.~ 

I 

~JO FLOW TO 

INJEtiiON LOOP 

1 

I 

NO FLOW TO AUY ' 

OF 3 LOOPS 

FROM At•Y OF Z 

PUMPS 

! 

UQ FLOW TQ H~JECilON LOOP 

2 

I LOSS OF HPSI I 

u 

HPrl I 

1\ '4ri 

I 

NO FLOW TO 

lNJECTI ON LOOP 

3 

, T? HP ... I .. HPI~ l HPI4 I 

NO FLOW FROM 

Pt:-1 "" .. 

HPI32 

NO FLOW FROi1 

PS-4 

HP!S I 

/j_PAGE " 

/j_PASE 5" 

NO ?OWER FROM NO FLOW FROM 

480 lJAC MCC SA I PS-9 

5E!SM!C 

FAILURE J~ 

TRANSFGRt-1E~ 

(X-5G7, x-::as1l 

I 

I 

AC?-iFM-Eg-5iX68 

0 

I 

/\?AGE IO w 

Figure C-2 (cont.) 

C-19 

NON· .%15:-11 C. CUI"U~m~ CAUSE 

FliiLURE OF 

HPS! 
HPI33 

f . 

~m POWER FROM 1 

480 VAC MCC SA 

I·'· bJu" ::-1 I_!,.,, ~ "'1 u"' I 

HiJECTIOH LOOP I 

HP!3 I 

0 

I 

NO FLOW FROi1 

PS-5 

I 

~m FLOli FROM 

PS-9 

ACP-SAC-l?-HCCSA HPt2l 

~rAc.F::."'' 

I 

!J.rAGE rC> 

Figure C-2 (tont.) 

C-20 
NO FLOW FROM 

1

PS-5 

I 

HP!34 I 

/\ 

~~} I 

NO rO~ER FF.GM 

480 VAC MCC 7A 

NO FLOW F~OM 

ACP-oAC·LP-MCC7A HP!i I 

Q 

! 

POWER NQT SEISMIC 

AVAILABLE FRO~ FAILURE 0~ 

DIESEL TRANSFORMER 

GENERATOR !A <X-507, X-698} 

OEP-DSN-LP-OGlA ACP-TFM-EQ-57168 

DPAf.:s~j. 0 

Figure C-2 (cont.) 

C-21 

I

t\ ... 6.. 0 '• t'H t. I --.l 
I 

lCO FLOW FROM 

PS-3 

HPI19 I 

HO POWER FROH 

460 VAC HCC BA 

, ....... QlJ ·o ~ ,cu r:. " 1 1 

WJEi:iiON LCOP I 

3 ! 

I 

HPI4 I 

0 

I 

NiJ FLOW FROM 

PS-6 

HPI3S I 

/1PAGE S 

LJ 

HPI2l I 

/1PAGE 10 

Figura C-2 (cont.) 

C-22 
NO POWER FROM I NO FLOW FROI1 

4&0 YAC MCC 7A • PS-12 

AC?·BAC~~·MCC7A HP17 I 

D'AO\i (, ~PASE '1 

Figure C-2 (cont.) 

c ... 23 
NO FLOW FROM 

PS·4 

H?I'S I 

(\ 

~~ 

I 

NO POWE.i FROM 

480 VAC MCC iA 

.!.. 

Nu FLOW FROM 

PS-12 

- .. ACP-dAC-LP~~CC7A . HPI1 I 

~ ~AC:ae t~ 

tm rOWS FR011 

480 VAC HCC iA 

9 ~lO FLOW FROH 

PS-20 

ACP-5AC-lP-!iCC7A HPI16 I 

/1 v.u~ ~ ~~--~ ---:----

POWER NOT 

AVAILABLE FRO~ 

DIESEL 

EE.'lERATCR lA 

DE?-OSN-lP-061A 

6.'P'Af:IS :. 

NO Flu}IJ FRON 

PS-23 

HPI14 I 

y 

NO ?C)iE.~ FF.ul'! 

480 YAC ~CC 7S ~ iBl 

Figure C-2 (cont.) 

C-24 

NO PCC CCOLINB 

TO NECESSARY 

SAFEiY SYSiEM 

COMP 

PCCl 

FlilLURE OF THE 

RWSi <n:-4> 

HP1-!NK-~X-RW5i () 
~m ~LiJW FRON 

PS-9 

NO ?Cwt:R FROM NO FLui.i FROM 

480 VAC MCC SA · PS-iS 

ACP-3AC-lr-MCC8A HPl23 I 

!J.Ar!J'e<.l "" ~ 

Figure C-2 (cont.) 

C-25 

I 

POWER NOT 

AVAILABLE FRO~ 

DIE5a 

SENERATOR lB 

DEP-DSH-tP-DGlB 

I 

NO FLDW FF.Ot1 No sec coouus 1 

PS-24 TO NECESSARY 

SAFETY SYSiEM I 

COI1P 

I 

HPI27 I SCC1 

NO POWER FROM FAILURE OF iHE 

480 VAC MCC 88 RWSi <TK-4> 

& BB! 
j 
APPENDIX D 

PRIMARY PRESSURE RELIEF SYSTEM 

Table D-1. Primary pressure relief system (PPS). 

Safety Function: 

System Components: 

Valves: 

Support Systems: 

AC Power: 

Actuation: 

To provide feed and bleed capability. Also provides 

reactor coolant system overpressure protection. 

PR- S-14 

PR- S-15 

P.R-M-16 

PR-M-17 

PR- S-14 

PR-M ... 16 

PR- S-15 

PR-M-17 

Actuated by 

Power-Operated Relief Valve 

Power-Operated Relief Valve 

PORV Isolation Valve 

PORV Isolation Valve 

480V Emergency MCC 78 

480V Emergency MCC 78 

480V Emergency MCC 88 

480V Emergency MCC 88 

operator for feed and bleed 

D-1 
FAILURE OF 

IPR·-''-1'\ "'0 • :::> .... I 

lALLOW FLOW 

I 

PPS-SQV-~X-PRS15 ,.-.. .. ·, i j . I 11-_ .. ~ 

NO PmiER FROM • 

480 VAC 11CC SB I I~ 8B1 I 

I I 

I 

I 

!COMMGU CAUSE 

!FAILURE OF 

IPORViBLOCK 

i 

FHILURE TO 

SUPPDRT FEED ~"~ 

BLEED FUNCTION 

PPS! 

lPORVs FAIL TO I iip~~~ tSHALI 

!LOC~ HAS .. I rrc""·R""n· 1\lw Ul"(,t,..,} 

PPS-CCF-FC~ ?OR\{5 PPS2 I 

(1 0 

i 

LOSS OF 

!FUNCTION OF 

IPS-! 

PPS4 

I OPERATOR 

FAILURE TO 

I OPEN POR1r

1

S iO 

:BLEEU 

I 

I 

lLOSS OF 

!FUNCTION OF 

IPS-2 

I 

I 

PP55 

FAILURE OF 

PR..-M-17 TO 

!ALLOW FLOW 

I 

j 

1 

2 

Figure D-1 Power-Operated Relief Valve Fault Tree. 

0-2 

l ..... r.·n i"""j ..... 

I t'hu .. URt: 0!-

tMAIN C:JNT?.OL 

I ROOM PA~1ELS 

I 

nCR-ACl'-!:X-CNTRL 

(-.. ..• 

\ .! ~--
!FAILURE QF 

jPR-S-14 TO I ALLOW tlOW 

I 

Fiqure D-1 (cnnt.) 

D-3 

• ILQS3 Or 

! FutlCTION OF 

IPS-2 

I 

i 

I 

PPS5 

IC?ERAiOF: 

iF!iiLURE TO 

!OPEN POP.VS TO jBLE~D 

I 

PPS-XHE-FD-FDBLD 

r--·, 

'-} 

FAllVRE OF I 

PR-M-16 TO· II 

!ALLOW FLOW 

I I PPS-HOV-KX-~RHlb ,..-., 

f' I 

\.) 
!LOSS OF 

!FUNCTION OF 

I PS·l 

I 

FAILURE OF AT I 

LEAST ONE PGRV 1 

I !UNABLE TO 

OPEtH 

I 

PPS2 

LOSS OF FUf~CT I ON OF 

ps-z 

PPS5 

{jPAEE 3 

!FAILURE TO 

SUPPORT FEED & 

BLEED 1NO 

LOCAi 

I 

Figure 0-2 Power-Operated Relief Valve Fault Tree (No LOCA), Pruned and Merged. 

D-4 
I 

I 

I 

FAULTS IN 

DIESEL 

I 

POWER NOT 

AVAILABLE FRO" 

DIESEL 

BENERATOR 18 

1 

I 

LOSS OF 

FUNCTION OF ?S~l 

PPS4 I 

('\ 

~ri i 

POWER NOT ·! 

AVAILABLE FROM I 

480 VAC 11CC SB I 

& S!>l I 

I I I ! 

ACP-BAC-LP-MCC8B 

1\ 

.~ I 

SEISMIC 

FAILURE OF 

TRANSFORMER 

(X-507. X-608) 

- "1-t - - OEP DaN .j:' - ., .P DGlB ACP TFH ~Q s,XbB (\ r···\ ~·I '....) 

J,..------;, 

NO POWER FROM I !NOM-SEISMIC 

DIESEL I COM" ON CAUSE 

GENERATOR lB I IDS FAILURE 

I l ~------~ ~~~~~ D61B1 I OEP-CCF-FC-DGr4 

(\ (--., ~ 1,._) 

. , . r· I r I O?~~ATOR 

HIL.UR\i' 1) 

1 1 ND sec cocuNGI 

TO NECESSARY I 

SE!S!UC 

FAILURE OF THE 

GENERATOR 18 p~£f=t:.L. PJ'I... I 1"ANKS 

I I SAFETY SYSTEH I 

CCMP 

DIESEL DAY 

TANK TK62A/B 

. I I I I 

I 

OEP-PSF-FC-9818 OEP·XHE-FO-FUEL SCCl I - . ... 

,.,......,) r-\ 

( I I ......,; 1..._) 

SEISHIC 

FAILURE OF AC 

CHILLER!: 

AC-lA/fb/2 

..... 

/'\ 

{ \ ~i I 

SE!S~IC 

FAILURE OF 

PCC/SCC COOLER 

E-;B/E-SA 

.. .. 

FiaurP n-?. (cont.} 

o .. s 

I 

SE!S~lC 

FAILURE 

CIRC WAi 

PUMP HOU 

tsws: .... . -

OF 

~:;) .t:.:. 

SE 

I 

FAILURE TO 

OPERATE PORVS 

FDR FEED & 

BlEED 
FAULTS IN 

DIESEL 

GENERATOR 1A 

I 

SE!S~IC 

FAILURE OF AC 

CHiLLERS 

HC-iA/1812 

" ..... 

NO POWER FROM 

DIESEL 

GENERATOR 1A 

DGlAl I 

/\ 

I \ h---i 

I 

Q?~RA1~R 

;:,;tLl.(~; m RE.~ILL ';L\~\. I TAN~!':. I 

OEP-XHE-~0-F!JEL 

(""'i 

I .. _, ·' 

I 

SE!SMIC 

FAILURE OF 

PCC/SCC COOLER 

E-48/E-SA 

. ... '"' 

POWER NOT 

AVAILABLE FROI1 

DIESEL 

GENERATOR lA 

NON-SEISMIC 

COI1110N CAUSE 

D6 FAILURE 

OEP-CCF-FC-DGN {-..., 

~.) 

NO PCC COOLING l 

TO NECESSARY I 

SAFETY SYSTEM I 

COMP ! 

I 

I i 

PCCl 

(\ 

~..,...I 

i 

I 

!FAILURE TO I 

I CLOSE F'CC I 

!ISOLATION VLVS I 

I I 

I I I ' 

Figure D-2 (cont.) 

D-6 

!LOSS OF 

IFIJNCiiON OF 

1 ps~z 

I 

I 

I 

PPSS I 

i'·. J 't ~..,....J 

I 

I POWER NOT 

AVA rLABLE FR. OM i 

480 VAC l'!CC 7B ,

1

· 

OR 7Bl 

! 

ACP-BAC-~P-MCC7B 

"" l '\ 

t t 

SEISI1IC 

FAILURE Of= TRANSFOR~ER I 

I !X-507' X-6B8l I 

I ! 

ACP-iFK-EQ-57X68 r·. I I 

1,__) 

SEZSMIC I 

FH "ILU"'E r. t'i'" ... r n·: r. .. 

niESEL DAY 

TANK TK62Ai3 

OE?-TNK-EQ-TK62l 

0 

I 

SE!SMIC 

FAILURE 

CIRC WAT 

PUNP HOU 

(SWSl 

OF 

-~ t. .. 

SE 

SiiS-3KW-EQ-CIRC 

0 

FAILURE TO 

OPERATE PORVS I FOR FEED ~ ,BLEED 

I 

PPS-XHE-FD-FDBLD _,.--..., 

I' \ t ...... ./ 
LOSS OF 

FUNCTION OF 

PORVl 

FAILURE OF 

BOTH PORVS 

(UNABLE TO 

OPEN> 

PPS2 

LOSS OF 

FW•CTION OF 

PORV2 

PPS4 PPS5 

D_PASE 2 D_PAGE 3 

PPSl 

I 

NON-SEISMIC 

COMMDt• CAUSE 

FAILURE OF 

PORV/ISOL 

PPS-CCF-FC-PORVS 

r--· 

... _) 

Figure D-3 Power-Operated Relief Valve Fault Tree (Small LOCA), Pruned and MerQed. 

D-7 
! 

FAULTS IN 

DIESEL 

~GENERATOR 18 

I 

I 

NO POWER FROM 

DIESEL 

GENERATOR lB 

DS1Bl I 

I 

...... 

.r \ ~I 

I 

FAILURE REFILL 

FUEL TANKS 

<P33,1JLIJ) 

I 

SE!S!HC 

FAILURE OF AC CHILLEF~S 

AC-lA/lB/2 

I 

POWER NOT 

AVAILABLE FROM 

DIESEL I 

lsENERATDR tB I 

I I 

OEP-DGN-LP-D61B 

/\ l-;l I 

I 

NON-SEISMIC 

COMMOt~ CAUSE 

DS FAILURE 

J:' - - - O t ... P CCF FC DG.~ 

r--·, 

1.__,) 

I 

NO SCC COOLING ! 

I 

TO NECESSARY I 

SAFETY SYSTEM 

COMP 

SEISMIC 

FAILURE OF 

PCC/ SCC COOLE.~ 

E-4B/E-5A 

I 

I 

I 

CCW-ACX-EQ-CHILL CCW-HTX-EQ-~S5A (J () 

Figure D-3 (cont.) 

D-8 

LOSS OF 

FUNCT!ON OF 

PORV! 

! POWER NOT I I AVAILABLE FRON ! 

1480 VAC l'ICC 88 I I I 

I I 

ACP-BAC-LP-MCC8B , ... 1"' .. , 

{-rr"i 

IX-5e7, Y.-6~81 

ACP-TFM-EQ-57X68 ,,.-..., 

I I 1.._, 

r 

SEISMIC 

FAILURE OF THE 

DIESEL DAY 

TANK TK62A/B 

1"'\ - II ~ 

SEISMIC 

FAILURE OF 

CIRC WATER 

PUMP HOUSE 

ISWSl 

.. 

SWS-BKW-EQ-CIRC 

0 

FAILURE TO 

OPERATE PORYS 

I FOR FEED ~ 

'BLEED 

PPS-Y.HE-!=0-FDBLO 

(''\ 

! . ·' ........., 
I 

I FAULTS IN I DIESEl GENERATOR lA 

I 

OEP-PSF-FC-DStA (-~ .. '\ 

I I l ,I . ......,. 

I 

SEISMIC 

FAILURE OF AC 

CHiLLERS 

AC-!A/18/2 

"" u r""IIO\ ~· . 

I 

I 

I 

NO POWER FROM 

DIESEL 

GENEF:ATOR 1A 

POWER NOT 

AVAILABLE FROI't I 

DIESEL I 

GENERATOR 1A I 

OEP-06N- P-D61A 

l'h.'-... 

lf 

NOt~-SEISM!C 

COMHON CAUSE 

I IDS FA [LURE 

I I 

DGlAl ~--~~!! DEP-CCF-FC-DG~ 

~ ------~ 

l 

I 

FAILURE REFILL! 

FUEL TANKS I 

I 

(P33,VLYl 

(-\ 

I I I ,c . .._,.. 

( 

I NO PCC COOLING! I TO tlECESSARY 

SAFETY SYSTEM I 

I !COMP I 

! I I f I I I 

I I I 

PCCl 

! 

SEIS~IC ! FAILURE TG I 

FAILURE OF I CLOSE PCC 

PCC/SCC COOLER I ISOLATION VLVS I 

E-~B/E-5A : l ! 

! 

i I ··~ ~ .. , ?CC-'tHE-FO-!SOL _ .. _,__ i ., 

I I l ! ,._,... 

Figure D-3 (cont.) 

D-9 

I LOSS OF 

I FUNCTION OF 

l PORV2 

I 

! 

I 

PPS5 

I 

POWER NOT 

AVAILABLE FROM 

483 VAC MCC 7B I OR 7B1 

ACP-BAC-LP-MCC7B 

~~ SEISMIC 

FAILURE THE 

!TRANSFORMER 

I (X-5071 X-608) I 

I . I ACP-iFM-EQ-57~68 

('") 

~ .......... 

f 

SEISMIC 

FAILURE OF THE 

DIESEL DAY 

TANK TK62A/B 

,...f"''\ - Ul """ 

; 

SE!S~rc 

FAILURE 

CIRC WAT 

PUMP HOU 

ISWSl 

OF 

ER 

SE 

.,..,, "' 

~" SwS-:,,W-E9-CIRC 

0 

!FAILURE TO 

!OPERATE PORYS 

!

FOR FEED ~ 

BLEED 

l 

APPENDIX E 

PRIMARY COMPONENT COOLING VI A TER SYSTEM 

Table E-1. Primary component cooling (PCC). 

Safety Funct 1 on: 

System COmponents: 

Tanks: 

Pumps: 

Heat Exchangers: 

Valves: 

Support Systems : 

AC Power: 

DC Power: 

HVAC: 

Cooling: 

Actuation: 

Provide cooling water required by plant equipment for 

normal operation and decay heat removal during cooldown 

or accidents. 

TK-5 PCC Surge Tank 

P-9A (N.O.) PCC Pump 

P-9B ( S) PCC Pump 

E-4A ( s.) PCC Cooler 

E-4B (N.O.) PCC Cooler 

Refer to Valve Table 

P-9A 

P-98 

P-9A 

P-98 

Turbine Building 

P-9A 

P-9B 

Motors 

E-4A 

E-48 

P-9B 

4160V Emergency Bus 5 

4160V Emergency Bus 5 

125V DC Distribution Cabinet 1 

12SV DC Distribution Cabinet 1 

Oil Cooled 

Oi 1 Cooled 

Air Cooled 

sws 

sws 

Low PCC Header Pressure 

E-1 
Table E-2. PCC valve table. 

Operating 

Power Normal Position Fail 

Valve Description (SOV) Position (Actuation) Position 

PCC-A-216 Return from penetration coolers 125 VDC DP/P 0 C( SI AS-A/CIS-A) C** 

(3413) 

PCC-A-238 Return frQR air recirc. coolers 125 VDC DP/BU 0 C(CSAS-8) C** 

(3412) 

PCC-A-268 Return from CEA air coolers 125 VDC BATT 2 0 C( SIAS-8/CI S-8) . C** 

(3416) 

PCC-A-493 DG-1A cooling water outlet 125VDC (1730A) 0 

PCC-M-43 PCCW outlet from RHR heat exchanger MCC 781 c 0( RAS-A) AI 

rr1 PCC-M-90 PCCW isolation to auxiliary building MCC 7A 0 C(RAS-A) AI I 

N 

PCC-M-150 PCCW isolation to letdown heat exchangers MCC 7A 0 C(RAS-A) AI 

PCC-M-219 PCCW isolation to containment MCC 7A 0 . C(CIS-A) .AI 

PCC-T-19 Cooler supply temperature control pneumatic 0 0 0 

PCC-T-20 Cooler bypass temperature control pneumatic· c 0 c 

** Fail open on loss of solenoid power 
P-9A 

P-9B 

Turbine 

Building 

Table E .. J. PCC cooling requirements. 

• Air-cooled. motor 

• Oil-lubed bearings 

• Air-cooled motor 

• Oil-lubed bearings 

• Natural circulation 

E-3 
*AC-18 

C-3A 

C-3B 

*E-3A 

*E-25 

E-29 

E-30 

E-31 

E-34 

E-35 

E-39A 

E-398 

E-44 

E-45 

E-46 

E-53-1 

E-53-2 

E-53-3 

*E-54-1 

*E-54-2 

*E-54-3 

*E-54-4 

*E-54-5 

*E-54-6 

E-70 

E-71A 

E-718 

E-72A 

E-728 

E-75 

E-77A 

E-778 

E-81A 

E-818 

E-81C 

*E-82A 

*E-918 

*E-928 

E-93A 

E-938 

E-94 

E-100 

Table E-4. PCC cooling loads. 

Loads Location 

Control Room Air Conditioner 

Waste Gas Compressor (E-88A) 

Waste Gas Compressor (E-888) 

Residual Heat Removal Exchanger 

Fuel Pool Heat Exchanger 

Recovery Evaporator Distillate Condenser 

Recovery Evaporator Distillate Cooler 

Recovery Evaporator Bottoms Cooler 

Vent & AC Equip. Rm, El • 39' 0" 

Cont. Spray Pump 

Area, El. 14'6 11 

Fue 1 Bldg. , El • 25 • 0" 

Reactor Coolant Pump Seal Water Heat Exchanger 

High Pressure Drain. Cooler 

Neutron Shield Tank Cooler 

Neutron Shield Tank Cooler 

Letdown Heat Exchanger 

Waste Evaporator Bottoms Cooler 

Waste Evaporator Distillate Cooler 

CEA Drive Mechanism Air Cooler 

CEA Drive Mechanism Air Cooler 

CEA Drive Mechanism Air Cooler 

Reactor Containment Air Recirculation 

Reactor Containment Air Recirculation 

Reactor Containment Air Recirculation 

Reactor Containment Air Recirculation 

Reactor Containment Air Recirculation 

Reactor Containment Air Recirculation 

Pressurizer Quench Tank Cooler 

Degasifier Vent Condenser 

Degasifier Vent Condenser 

Degasifier Effluent Cooler 

Degasifier Effluent Cooler 

Waste Evaporator Distillate Condenser 

Reactor Coolant Sample Heat Exchanger 

Reactor Cool ant Sample Heat Exchanger 

Secondary Sample Heat Exchanger 

Secondary Sample Heat Exchanger 

Secondary Sample Heat Exchanger 

DG-1A Cooler 

Cooler 

Cooler 

Cooler 

Cooler 

Cooler 

Cooler 

Safeguard (LPSI) Pumps ~al Leakage Cooler 

Charging Pump Seal Leakage Cooler 

Degasifier Vent Cooler 

Degasifier Vent Cooler 

Waste Gas Compressors Aftercooler 

Blowdown Tank Cooler 

E-4 

Reactor 

Reactor 

Reactor 

Reactor 

Reactor 

Reactor 

Containment 

Containment 

Containment 

Containment 

Containment 

Containment 

PA8, El. 13'6" 
P-1-1 

P-1-2 

P-1-3 

*P-7 

P-11 

*P-12A 

*P-14A 

*P-14S 

P-19 

P-20 

P-21 

P-22 

P-65 

P-66A 

P-668 

Table E-4 (Cont'd) 

Loads 

Reactor Coolant Pump 

Reactor Coolant Pump 

Reactor Coolant Pump 

Auxiliary Charging Pump 

Recovery Evaporator Reboiler Pump 

LP SI Pump 

Charging Pump 

Charging Pump 

Recovery Evaporator Distillate Pump 

Recovery Evaporator Bottoms Pump 

Waste Evaporator Reboiler Pump 

Waste Evaporator Distillate Pump 

Waste Evaporator Bottoms Pump 

Degasifier Pump 

Degas i fi er ·Pump 

*Containment Penetration Coolers 

(Penetrations 9, 29, 30, 31, 32, 45, 46, 

53, 54, 55, 62, 64, 65, 66) 

Location 

PAB' El • 11 I 011 

Cant. Spray Pump Area, El • 

14'611 

PAB, El. 21'0" 

PAB, El. 21'0" 

* = PCC Loads not isolated by PCC-M-90, -150 or -219, whose failure may lead to 

failure of the PCC. 

E-5 
!FAILURE JF 

jLGAD IH fS-~ I . 

I 

PCC16 I 

/w ·'\. %8.. 2 ... t:. 

FAiLURE uf PCC i 

LOAD E-B2A I 

:Dfi-iA CQOLERI I 

I CC-HTX-~X-E92A 

(i 

L.,._) 

PCCl 

I Nu r=u:r~~ ra I 1 FLO~ alvERSIJN 

IPs-~ 1 IFRDM ?S-1 

I . 11 

PCC2 ~'--------~· ! PCC3 ~'--------~ I 

/\ PAS"" 9 ;'\ PA:''" ~·· t:. L-l''~t 

I 

I Cul1i1CN :AUSE 

IFAILUEE OF PCC 1 s·rsTEH 

I 

PCC -CCF-FC- ?CCt~ ~-- j "t 

I I ' .... ....!"'. ( 

Figure E-1 Primary Component Cooling System Fault Tree. 

E-6 
I I 

1 FAILURE OF PCC I FAILURE OF PCC 

ILGAD AC-lB I LOAD (P-7i 

I (CONTROL ROOM ! PUI'I? COOLER 

AC) I I 

I i ~~----~ PCC-~T ;:-~X ·?7 

{ ...... , I ! 

1,._,/ 

FAILURE OF PCC I 

LOAD ?-12A 

(PUMP CODLERl 

PCC·HTX-KX-?12A 

(-.'1 

'·· ..• ) 

FAILuRE

1 

oF Pee 1 I FAILURE, oF Pee j 

LOAD ?-t4S I ILOAD - tPU~P caoLERi I PENETRATtuN 

!COOLERS 

PCC-HTX-!X-Pl4S ~----~' PCC-HTX-~1-PEN 

~~--~~ 

r-·l r-·., \ 

~~ I ~~ ~ 

Fi~ure E-1 (cont.) 

E-7 

'FAILuRE oF 1 I LOAD !N PS-! I 

I I 

PCC!bA I 

/\PAGE 3 ,___.. 
_I_ 

FA I LURE OF PCC I 

LOAD E-3A tRHR 

HEAT 

EXCHANGER! 

I 

lF~ILURE OF I L~AD IN PS-1 

I 

FAILURE OF PCC fFAlLURE aF PCC 

LOAD E-25 ILDAO. !F!JEL POOL E-91B,SAFES 

COOL En) PUHPS SEAL CLR I 

I 

! 

Figure E-1 (ccr.t.) 

E-8 

.1. I 

!FAILURE OF ?CCI !FAILURE OF PCC 

LOAD E·92B,CHS j LOAD 

jPP SEAL LEAk ~~-54-!·:·~" ICLR I 1AlR •Eclr.c I I I . I !CLR) I o I 

PCC-HTX-KX-E92B (] PCC-HTX -KX -E5H 

r-·. 

l \ ~_,c 
!,.., ·u .l;'E·-•"'\•1 ,• II"L~Ji't. J y ti::iLUi1 

I 

,~, ~HRI'!Ur.!H l.i o r. .. s -.::: .•. I . ! I 

I I 

I l I 1 

PCCll 

tm POWER FRDI't 

I 480 VAC MCC 7A I 

PCC-HOV-~X-HVl~O ~----'' i ACP-BAC-LP-MCC7A I 

r--·) ,~ I il \;. \_,., '\#' 

-· il" ~I 'I,..,."'. ""·'J • t'LUii iJ 'ttl":::HUh I 

FROM PS-1 ! 

j 

I 

i 

/FLOW DIVERSION i I THROUGH PS-4 I 

LJ 

!FLON DIVERSION i IFLD~ DIVERSION 

iiHRCUGH PS-5 THROUGH ?S-6 

i 

PCC12 I ?CC13 I PCC14 I PCC15 i 

6PAGE 5 fJ.PASE 6 /J.PASE 8 

OPERATOR 

FAILURE iD 

CLOSE PCC I SOL I 

VALVES I 

PCC-XHE-!=0-lSOL 

(") 

.. _, 

Figure E-1 (cont.) 

E-9 

1

,\ :)·s.. ...., 'I 1-\ t:. w 
'FiiiLURE OF 

PCC-H-90 TO 

SiO? FLOW 

PCC-'!Ov-~X-l'!V90 c) 

FLOl~ ii FJEF:Sl uN !' 

iHROUr.:H ;::,:: -3 

LJ 

PCC12 ! 

/"\ 

~l I 

!NO POWER FRC/1 

480 VAC MCC 7A 

OPERATOR 

FAILURE TO 

CLOSE PCC I SOL ~ 

VAL \/ES 

ACP-BAC-LP-~CC7A PCC-XHE-FQ-!SOL .A ,.-.., 

.:< » ll ) \;'(/· ~ 

Figure E-1 (cnnt.) 

E-10 
FAILURE DF I 

PCC-M-219 TO I I STUP FL0\1 ! 

PCC-MOY-~X-MV219 ....... 

CJ 

FL0l~ DIVERSION i 

ru::.ni'"'H 1nnUUl:l 1" ::;.c_ "' • I 

I 

hiQ p nut.~ ...... U,. w I , ,1 , ... n r. !"I'( u I 

!490 I 11' A" 1.. l MC" .... ~· /H 

I t I t 

ACP-BAC-LP-~CC7A 

\() 

'{-'/ 

Figure E-1 (cont.) 

E-ll 

OPERAiOR I 

FAILURE TO I 

CLOSE PCC ISOL ! 

VALVES 

PCC-XHE-FO-ISQL 

r) 

~ ......... 
\FLOW ui\'ErtSION 

} THROUGH PS-.~ 

! 

l l 

PCC15 

!FAILURE oF I jNo POWER FROM iNo ACTUATiON 

.. "'rt:r-.r~-?16 __ n ..... r,n _ ! 11'it:: tl"u·r "'Ut:: 1 )lF;:;r~·.,. ::rll~ 

1 

. ...,. li .., 0 .... ... ! """n ""'.a.n"""· 

l SlOP FLww , I I TRA Pl A I l I I j I I I I I 

PCC-AOV-~X-AY216 

~------~ DCP-BDC-LP-BUSl ~--------' 515-~CT-FA-TRNA ~ --------~ r--.. , //.:~.. .'\ J i {~ : ·) {~·.;:··"u~_ :~) \ .i ··;:~;;·· '··' ~~ v 

Figure E-1 (cont.) 

E-12 
jFAILURE OF 

jPCC-A-268 TO 

iSTDP FLOW 

I 

PCC-AOV-¥.X-AV268 (-..... 

\ J ~-

?CC14 

1 NO PO~E~ ri=,On I. "11:\ •.~rc ~·1u·c 2. !'"" '" .. J 

l 

DCP-BDC-LP-8USZ. .:-,. 

{~) 

I QPE?.ATGR 

!

FAILURE TO 

CLOSE ?CC ISOL 1 VALI/ES 

I 

l NO .~Ci. SiGNAL l 

I iO CLOSE ! 

j PCC-A-2Ea'8 f 

I I I 

?CC21 

NO ACTUATION 

FROM SIAS 

TEAIN A 

Figure E-1 (cont.) 

E-13 
I FAILURE OF PCC !' 

'PUMP P-9A I 

I . I PCC-HDP-~X-PTRNA ',.-~~. 

t._.! 

~lO ?CC FLOW 

I THROUG" CO"LER I B 

PCC-HTX-f.X-E4B 

(') 

... _. 

I 

FAILURE OF 

PCC-T-19 TO 

ALLOW FLOW 

I 

NO FLOW FROM 

PS-12 

PCC-TCV-MX-TCV19 PCC7 I 

('1 ur:_··i, ~ ........ } 

NO FLOW FROM NO FLOW FROM 

PS-13 PS-14 

'~-r--___.11 '---,.-------< PCCB I PCC9 I 

/'\ /\ PAGE 1 0 

1-'r-·1 w 

i 

[NO POWER FROM ! NO POWER FROM 

1'160 VAC BUS 5 I 125 voc sus I 

I AC?-BAC-LP-BUSS i~ I~' DCP-BOC-LP-BUSl ~---...1 ,.;:\ 

-:~.;v." __ ,~!,~ /( \~~ •I '.~l' 

I

' FAILURE

1 

OF PCC \ 

SURGE TANK I 

\TK-51 

I I 

I ! PCC-TNK-~X-SRGTK (--1, 

I I I ._..,.~ ; 

Fi m1re E-1 (cant.) 

E-14 

I rJO FLow m 

IPS-l 

I 

I 

PCC2 6 

I 

NO sw FLOW iO I 

Hv H ... '·n~E:;. ,..~,~.~ 1 ·' I tH .. ruUil I 

I

' ANY OF 4 S~ 

PUNPS I 

I 

1FAll U'"'·:: i>F I ..... r. .. 'J 

PCC-T -21

) TO 

SiOP FLOW 
FtitLURE OF PCC ·I NO POWER FROM I 

PUMP P-~9 !4166 YAC BUS 51 

PCC·MDP·\X~?TRN8 ACP-3AC-lP·8USS 

0 ¢ 

r~O rluW FRO:t 

Pe""l~ 

PCC? 

I 

NO POWER FROM 

125 VDC SUS 1 

-

Figure E-1 (cent.) 

E-15 

! 

I 

l 

.I i 

F~ILURE OF PCC i 'NO ACT. SISNAL I 

SURGE iANK TO ?·9B 

\TK·S) 

I .. PCC-sllK-U-SRSTK PCC10 

0 0 

FAILURE OF PCC 

STANDBY PUHP 

ACi. StGNAL 

I 

OPERATOR 

FAILURE iO 

PLACE P-98 IN 

SERVICE 

PCC~rST-FA-Ac;pp PCC·XHE-FD·P9B 

0 0 

I 

APPENDIX F 

SECONDARY COMPONENT COOLING WATER SYSTEM 

Table F-1. Secondary component cooling (SCC). 

Safety Function : Provide cooling water required by plant equipment for 

normal operation and decay heat removal during cooldown 

or accidents. 

System Components: 

Tanks: TK-59 

Pumps: P-lOA (N.O.) 

P-10B ( S) 

Heat Exchangers: E-5A (N.O.) 

E-5B ( S) 

SCC Surge Tank 

sec Pump 

sec Pump 

SCC Cooler 

sec Cooler 

Valves: Refer to Valve Table 

Support Systems : 

AC Power: 

DC Power: 

Air: 

HVAC: 

Cooling: 

Actuation: 

P ... lOA 

P-108 

P-10A 

P-lOB 

TK-110 

Turbine 

Building 

P.,lOA 

P-lOB 

Motors 

E-5A 

E-5B 

P-lOB 

SCC-A-460, 461 

4160V Emergency Bus 6 

4160V Emergency Bus 6 

125V DC Distribution Cabinet 3 

125V DC Distribution Cabinet 3 

Isolation valves 

( SCC-A-460, 461) 

Oil Cooled 

Oil Cooled 

Air-Cooled 

sws 

SWS 

Low SCC Header Pressure 

Low Suet ion Pressure 

F-1 
Table F-2. sec valve table. 

Operating 

Power Normal Position Fail 

Valve Description ( SOV) Position (Actuation) Position 

SCC-A-460 Non-seismic supply header stop 125 VDC 3 DP/BU 0 c 0 

(1725A1 & A2) 

SCC-A-461 Non-seismic supply header stop 125 VDC 3 DP/8U 0 c 0 

(172581 & 82) 

SCC-T-23 Cooler bypass temperature control pneumatic c 0 c I 

SCC-T-24 Cooler supply temperature control pneumatic 0 0 0 

SCC-T -305 OG-18 cooler inlet temperature control 125 voc 0 

, (17308) I 

N 
P-lOA 

P-IOB 

Turbine 

Building 

Table F -3. SCC cooling requirements. 

• Air-cooled motor 

• Oil-lubed bearings 

• Air-cooled motor 

• Oil-lubed bearings 

• Natural circulation 

F-3 
*AC-1A 

*AC-2 

AC-3 

C-1A 

C-18 

C-lC 

*E:-38 

E-6A 

E-68 

E-6C 

E-60 

E-7A 

E-76 

E-8 

E-19A 

E-198 

E-20 

E-21A 

E-218 

E-21C 

E-78A 

E-788 

E-78C 

E-780 

E-78E 

E-78F 

E-79A 

E-798 

E-80A 

E-808 

*E-828 

*E-91A 

*E-92A 

E-101A 

E-1018 

*P-12B 

*P-148 

+P-145 

P-27A 

Table F·4. SCC cooling loads. 

Loads Locat1on 

Computer Room Air Conditioner 

Lab Air Conditioner 

Office Area Air Conditioner 

Control Air Compressor 

Control Air Compressor Control Air Compressor 

Vent & AC Equip. ~' El.39'0 11 

Vent & AC Equip. Rm, El.39'0" 

Residual Heat Exchanger Cont. Spray Pump Area, El. 14'611 

Generator Hydrogen Cooler 

Generator Hydrogen Cooler 

Generator Hydrogen Cooler 

Generator Hydrogen Cooler 

Turbine Oi 1 Cooler 

Turbine Oil Cooler 

Exciter Air Cooler 

Generator Seal Oil Unit {Air Side) 

Generator Seal Oil Unit (Hydrogen Side) 

Generator Leads Cooler 

Control Air Compressor Aftercooler 

Control Air Compressor Aftercooler 

Control Air Compressor Aftercooler 

Sample Cooler 

Sample Cooler 

Sample Cooler 

Sample Cooler 

Sample Cooler 

Sample Cooler 

Steam Generator Feed Pump Lube Oil Cooler 

Steam Generator Feed Pump Lube Oil Cooler 

Electro Hydraulic Governor Oil Cooler 

Electro Hydraulic Governor Oil Cooler 

DG-18 Cooler 

Safeguards (LPSI) Pumps Seal Leakage Cooler 

Charging Pump Seal Leakage Cooler PA8, El. 16 1 0" 

Turbine Drive Main Feed Pump Lube Oi 1 Cooler 

LP SI Pump 

Charging Pump 

Charging Pump Condensate Pump 

F-4 

Cont. Spray Pump Area, El • 14 • 611 

PAB, El. 21 1011 
Table F-4 (Cont•d) · 

P-278 

P-27C 

P-62A 

P-62B 

Loads 

Condensate Pump 

Condensate Pump 

Heater Drain Pump 

Heater Drain Pump 

*Containment Penetrations Cooler 

Penetrations 9,29,30,31 

Penetrations 32,45,46,47,62 

Penetrations 53,54,55,64,65,66 

Locations 

Containment Spray Pump Area 

Primary Aux. Tunnel Area 

Main Steam · Va 1 ve Area 

* Loads not isolated by non-seismic stop valves, whose failure may fail the SC C system. 

+ PC C cooling perferred. 

F- S 
I""~ "'I • '~:-' ~-

i"'i-11~~~rtt. : ..n· 

LCAD :~i ?S-11 

PAGE 2 

Ff.ilL:JEE ::iF .3CC 

LOAD ~-92B 

IDG-lB CCOLERl 

F~lLURE OF SCC ' 

LOAD \P-14B/ I 

PUMP COiJLER 

FAILURE uF 

jSCC-H-460 & 

'461 

!ACCUMULATOR, 

1

n:-uo 

SCC-TNK-KX-AC110 .-... I ·, t_) 

NO 3CC CJOU:iG 

TO >iEC£SSARY 

SAFE7Y 3Y 3TE:·1 

CGMP 

SCC1 

!rm FLJ~J FRO~ IFLm~ }lVERSIDN! I 

iPS-Hi !THROUGH PS-1 I iCCMHON CAUSE 

1

1 

1 I jl I F~ILURE oF sec 

l""'lw•"*•• 

I 

1

1 ,i, Jc•Jt:TI:M 

SCC2 ~--~--~'I SCC3 I~'--~~--~ SCC-2CF-FC-SCCW 

/

.·\ !'lAG .. } l'· ;- t. 3 

u 

NO POWER FROM 

125 IJDC 8US 3 

/"\ {.-.. l 

!rr'-, ,......,. 

FAILURE uF 

SCC-A-460 TO 

jSTDP FLOit 

I , 5CC-AOY-KX-AV460 r-·i ' .. _,• 

NO ISOLATION 

SIGNAL 

\?S-li50A 1 LD 

HDR PRESS} 

SCC-?ST-FA-i750A A 

{..' '~ ' I v 

Figure F-1 Secondary Component Cooling System Fault Tree. 

F-6 
FAILURE OF SCC l 

LOAD AC-2 !LAB ! 

AC) I 

I I 

SCC-ACX-n-AC: (~'j 

v 

FAlUJRE OF 

!LOAD IN PS-ll 

I 

l 

i 

I 

SCClA 

I FAILURE. oF sec I 

! LOAD - I 

j ~EN~TR~TlON 

;1..00LER~ 

i 

1

F .... ''RE "'F """'"' I HlLU \.1 ~L.L 

jLOAD P-12B 

1 \LPSI PUnP I COOLER) 

I 

SCC-HH-~X-?12B () 

FA I LURE OF SCC 1 

LOAD E-3B \RHR I 

HEAT I 

EXCHANGER! 

Figure F-1 (cont.) 

F-7 

1

FAILURE OF 

LCAD IN PS-11 

SCClB 

'F•~; Hl U·"E• nF ~"" j .. ,, u ... l.l. I 

! LOAD I 

1

E-91A,SAFEG I I PUMPS SEAL CLR I 

5CC-HTX-1:X -E91A 

('"' 

I j ~ ... _. 

1 FAILURE GF sec I I LOAD E -:t2A, CHB i 

P:l ...... ,,L L'""!t I ' , JtH t:rh. j 

CLR I 

! 

SCC-HTX- :X-E92A 

("') 

.. _,~ 
NO FLOW FROH 

PS-10 

SCC2 

I tm s;t FLGif TO No sec FLow 1 

I

HX HEADER FROM THROUSH COOLER 

tm FLow FROM 1 

[' I I ,:~~p~F 4 SW I t_j 

SWS 1 I SCC-HTY. -U -E5A ... ~ (~\ 

{(/ \ ... ) ,1'\ 

! FAILURE uF SCC i 

!PUMP ?-lOB I 

I I ; I 

I ! SCC-~DP-~X-PTRNB ,.---.. 

ii J -* 

FAILURE OF 

SCC-T-24 TO 

ALLOW FLOW 

~ 

NO FLOW FROH 

!PS-6 I 

w 

0 

I 

sees scc9 

/\ PAGE 4 i_j 

I Ho POWER FROM !' ltm POWER FRo~ j4160 VAC BUS 6 125 YDC BUS 3 

I II I I I 

! II~--------~ ACP-BAC-LP-BUS6 DCF-2DC-lP-3US3 -·~ /,\ .f.'\\ ,/'\ '··\v{~ ···~.v''~i .., 

FiourP. F-1 (cont.) 

F ... a 

jFA!LURE CF 

·scc-T-2.3 iO 

STOP FLOW 

SCC-TCY-Y.X-TCV23 .-. t:j 

I L'Q -· -... -... ""'1 I;~ rV .. /1'1 t'!iUi 

jPS-9 

i 

i 

I 

SCC10 

(\ 

f:_.PA3E 

IF"··· ""'I"' .... "rr i H li.Wl'\:;. 'JI' '.J..,-. 

lSiANDSY PUHP 

) ACT. S!GtiAL 

l 

! 

IN" '·"'T s·"'·""' l• U HI, • ii:ti'II-IL 

ITO P-lOB 

I 

! 

SCCli 

I u··o~:;:.A~"!:;. 1 1 '-" IL,, 

.,. F•":i ·1 LH!;'I•:: ru ..,, "" ~..... ' 

!PLACE P-10?. lN f SEE:.; ICE 

I 

SCC- :•:HE -FU-F·t Ci5 /''""""\oo-,1 

I I \ ,I ""-·· 
I I 

FAILURE OF SCC ~m POWER FROM 

PUHP P-lOA 4160 VAC BUS 6 

NO FLOW FROH 

PS-7 

sees 1 

("\~ 

T 

I 

!NO POWER FRO" 

125 VDC BUS l 

I 

NO FLDW FROM 

PS-9 

SCC-nDF-~X-?iRNA ACP-BAC-Lf-BUSa DC?-BDC-LP-BU53 · SCClU 

A 

I 

0 ./1\, 

\\ J~· 

w 

'i J,\ ",'f c I 

I I 

FAILURE OF SCC FLOW DIVEP.SION 

SURBE TANK THROUGH PS-14 

ITK-59> 

SCC-TNK-U-SRBTK SCC12 I 

0 Q 

I ! I 

FAILURE OF NO I SOLATION NO POWER FROH 

SCC-A-46Ci « SISNAL 125 VDC BUS 3 

461 (PS-1750Bt LO 

ACCUMULAiOR, HDR PRESS) 

TK-110 

., - ::iCC-Ita.-~:X-HCllO SCC-PSi-FA-175oB DC?-BDC-LP-BUS3 

(-...., !\ /'\i· jl~ ~ J.l '\) "' -~1 " 

I 

FAILURE OF 

SCC-A-461 iO 

SiOP FLOW 

5CC-AOY-~I-AV461 

{] 

Figure F-1 (cont.) 

F-9 

APPENDIX G 

SERVICE WATER SYSTEM 

Table G-1. Service water system (SWS). 

Safety Function: Provide cooling for the PCC and SCC systems. 

System Components: 

Pumps: P-29A {N.O.} 

P-29B { S} 

P-29C ( S) 

P-290 (N.O.} 

Heat ~changers: E-4A(S) 

E-4B(N.O.) 

E-SA(N.O.) 

E-58( S) 

Support Systems : 

AC Power: P-29A 

P-29B 

P-29C 

P-29D 

DC Power: p;..29A 

HVAC: 

Pump Cooling: 

P-298 

P-29C 

P-29D 

Pump House 

Pump Discharge 

South SWS Pump {for SCC) 

South SWS Pump {for SCC) 

North SWS Pump (for PCC) 

North SWS Pump (for PCC} 

PCC Cooler 

PCC Cooler 

SCC Cooler 

SCC Cooler 

480V Emergency Bus 7 

480V Emergency Bus 8 

480V Emergency Bus 7 

480V Emergency Bus 8 

125V DC Distribution Cabinet 1, 

125V DC Distribution Cabinet 3, 

125V DC Distribution Cabinet 1, 

125V DC Distribution Cabinet 3, 

G-1 
Table G-2. SWS cooling requirements. 

P-29A • Pump discharge recirculation (safety class) 

• Raw water (preferred) 

P-29B • Pump discharge recirculation (safety class) 

• Raw water (preferred) 

P-29C • Pump discharge recirculation (safety class) 

• Raw water (preferred) 

P-29D • Pump discharge recirculation (safety class) 

• Raw water (preferred) 

Clrc. Water 

Pump House • Natural circulation 

G-2 
tr:·· .. "R,.. -........ I· HlL\.! t. \Jr :;::~ 

!PUMP ?-290 

I 

I 

NO SW FLOW 

THROU6H SCC 

rOLER E-5A • 

SWS-~TX-n-ESA 

{=) 

NO FLaw FROM 1 INa FLow FROM 

IPS-5 ' IPs-o 

swss '~--~--~~ ~~--~-----~ I 

!NO FOWER FRGI'I 

I ·ar; v·c ::.u"' n -. .J ~H " :;. o 

I 

I 

I 

I 

SW56 ... l'· PAG .. ' '•, t l.......l 

! tW POi~ER FROM 

112~ VDC BUS 3 

I t 

'l 

;. 

I

I rn"' \.rv,lll .. o~,· 1 rH·u~:: w wit... 

'FAILURE OF SW 

PUMPS 

I 

tiO FLOW THRU 

PS-Z 

5WS2 

NO FLOW FROM 

PS-9 

SWS4 I 

(-.' 

u 

! 

'NO sw FLOW iO I I H'i HE. "E~. '"'"·""1 I A j HU I'( it\lll . 

I ANY oF i s~ I !PUMPS. . ~.. I 

~ /\ 

I \ 

kr·~ 

!FAILURE; OF SW i 

S'iSTEM 

r-u 

I 

NO SW FLOW 

THROUGH PCC I COOLER E-4B 

I 

!NO FLOW FRO~ 

I ,PS-8 

L...------:----'1 1...__----=-----~ swse 5ijS7 i 

/\ PAGE 3 ;'\ P'·G'" I ', 1 I , I H ,t L....:. w 

Figure G-1 Service Water System Fault Tree. 

G-3 

4 

! 

NO FLOW THRU 

PS-3 

5WS3 I 

/'\ I 'I ,.r 

INo FLOW FROM 

jPS-9 

I 

I 

I 

SWS4 

/\ 

I ': L...:J. 

i 
pm i=UW FROM IPH 

I 

i 

I 

SHS6 

r- • .• L''R .. :"IF "''. ,, I NO OQ"EO c-·oliol I )li! Oifi,IE;o. :'::iQ"' 

I

' r H i. i.i It \, ;:: M I R 1\ I t( II I !1\i ! wll II I h I i 

IPUrP P-:qc j 14ao YAC sus 7 1 i25 voc sus 1 

I ILJl1 

I 

SaS-~DP-~X-PiRNC I ACP·BAC-LP-BUS7 DCP-;DC-LP-BUSl ~· ~~ 

t-.·1 .\)#\, ,A:~. 1\ } ''l ;;"' '\. •• )' 

- v 

Figure G-1 (cont.) 

G-4 

jOPE?.AiGR 

/ ~~! ;~R~ .. !. ~or·" 1 .. 1Hnl ~li;N l:IY 

I su C'i~p: I w It.: ..... 

SWS-XHE-FO-PUMPS r-· .. l 

\.._) 
FA I LURE ilF SW \ 

rU"P P-'299 : 

SWS-MDP-~ X·PTP.~JB (j 

NO ?OWER FROM 

460 VAC BUS 3 

ACP-SAC-LP-SUSS 

\; f~ 

Figure G-1 (cont.) 

G-5 

INu FLOW FROM 

jPS-i 

I 

I 

/\ 

i i 

kr-~ I 

.,. NO POWER FROM 

1"5 VDC ~uo 't t•• w,.,..., 

I 

! ... ,.,c,.,~ .... R )Ui" ... tti'IIU• 

1

FA1LURE TO l ,. .... ,,R .. '"·T·"'."' ... Y l;,ln.l :J 11Nl.'tl 

ISH PUMPS 

SWS-IHE-FG-?UHPS (-..,, 

t.) 
SWS-MDP-kX-PTRNA 

r-·'1 

1..._,; 

Figure G-1 (cont.) 

G-6 

!No FLOW FROM 

!PS-3 

I 

I 

SWS8 

I NO on:.I""R ;::;.r1.1 • 1 ,J'!'ftu • nUn 

480 VAC BUS 7 

AC?-BAC-LP-BUS7 

.:{~:) '·\vj 

1

1 HO POWER FROM i''JI:' V"'''' r,}''' 1 ;.4J .L'lt UU~ 

I 

I 

DCP-9DC-LP-9U51 

<t;.~ \'/ 
APPENDIX H 

ELECTRIC POWER SYSTEM 

Table H-1. AC power. 

Safety Function: To provide operating power to plant equipment and 120V AC 

power to plant instrumentation. 

System Components: 

Buses: Buses 5, 6 4160V Emergency Buses 

Buses 7, 8 480V Emergency Buses 

Buses 1-4, 1A-4A 120V Vital Buses 

Transfomers: X-507 4160V Bus 5 to 480V Bus 7 

X-608 4160V Bus 6 to 480V Bus 8 

Motor Control 

Centers: MCC 7A, 7B and 781 460V Emergency MCCs 

MCC SA, 88 and 881 480V Emergency MCCs 

Inverters: INVR-1~ 120VAC Inverter 

INVR-2 120VAC Inverter 

INVR-3 120VAC Inverter 

INVR-4 120VAC Inverter 

Support Systems : 

AC Power: DG-lA Diesel Generator 

DG-18 Diesel Generator 

DC Power: OC-1 125 VDC Bus 

DC-2 125 VDC Bus 

DC-3 125 VDC Bus 

DC-4 125 VDC Bus 

HVAC: FN-31 Switchgear Room Supply Fan 

FN-32 Switchgear Room Exhaust Fan 

H-1 
Table Ha2. DC power. 

safety Function: Provide DC power and control for switchgear, vital bus 

inverters, vital SOVs and instrumentation. 

System Components: 

Buses: DC-1 

OC-2 

DC-3 

DC-4 

Batteries: BATT-1 

Battery Chargers: 

Distribution: 

Support Systems: 

AC Power: 

HVAC: 

BATTa2 

BATT-3 

BATT-4 

BC-1 

BC-2 

BC-3 

BC-4 

DC/CE-1 

DC/CE-2 

DP/P 

DP/BU 

BC-1, 2 

BC ... 3, 4 

FN-31 

FN-32 

FN-31 

FN-32 

125 VDC Bus 

125 VDC Bus 

125 VDC Bus 

125 VDC Bus 

60 Cell Station Battery 

60 Ce 11 Station Battery 

60 Cell Station Battery 

60 Cell Station Battery 

129V - 250A Charger 

129V - 250A Charger 

120V - 250A Charger 120V - 250A Charger 

Distribution Cabinet 

Distribution Cabinet 

Distribution Panel 

Distribution Panel 

480V Emergency MCC 7 

480V Emergency MCC 8 

480V Emergency MCC ?A 

480V Emergency MCC BA 

Protected Switchgear Room Supply Fan 

Protected Switchgear Room Exhaust Fan 

H-2 
Table H-3. Diesel generators (DG). 

Safety Function: Provide electric power to the plant emergency bus_es when 

normal power is not available. 

System Components: 

Generators: DG-1A( S) 

DG-1B( S) 

Diesel Generator 

Diesel Generator 

Pumps: P-33A Auxiliary Fuel Oil Transfer Pump 

P-336 Auxiliary Fuel Oil Transfer Pump 

Heat Exchangers: E-82A 

E-828 

DG-1A Cooler 

D G -1 B Coo 1 e r 

Tanks: 

Support Systems: 

AC Power: 

DC Power: 

Air: 

HVAC: 

DG Cooling: 

Actuation: 

TK-28A 

TK-28B 

TK-62A 

TK-62B 

Auxiliary Fuel Oil Supply Tank 

Auxiliary Fuel Oil Supply Tank 

Diesel Generator Day Tank 

Diesel Generator Day Tank 

DG-1A Distribution 

Panel 

DG-lA Engine Cont ro 1 Pane 1 

DG-1B Distribution 

Panel 

OG-18 Engine Control Panel 

P-33A 

P-338 

480 VAC MCC 7A 

480 VAC MCC 7 A 

480 VAC MCC SA 

480 VAC MCC 8A 

480 VAC MCC 7 A 

480 VAC MCC 8A 

DG-1A Start 1 & 2 125 VDC Battery 1 

Circuits and Control 

Power 

OG-18 Start 1 & 2 125 VDC Battery 3 

Circuits and Control 

Power 

TK-76-A1,A2,A3 DG-1A Air Receiver 

TK-76-A4,A5,A6 OG-1A Air Receiver 

TK-76-B1,B2,B3 DG-18 Air Receiver 

TK-76-B4,BS,B6 DG-lB Air Receiver 

FN-20A OG-1A Room Exhaust 

Bank 

Bank 

Bank 

Bank 

Fan 

FN-20B DG-lB Room Exhaust Fan 

Air Intake and Exhaust 

E-82A PCC 

E-828 sec 

DG Start and Load Shed/ Sequencer . 

H-1 

Dampers 
Table H-4. DG cooling requirements. 

DG-lA • The engine and turbocharger aftercoolers are water cooled by the PCC (or 

Fire Protection System), via E-82A and two cooling water pumps. 

• The generator is air cooled by a blower driven by a gear off the engine 

camshaft. 

DG-lB • The engine and turbocharger aftercoolers are water cooled by the SCC (or 

Fire Protection System), via E-828 and two cooling water pumps. 

• The generator is air cooled by a blower driven by a gear off the engine 

camshaft. 

Diesel • Combustion air via normally closed intake and exhaust dampers. 

Rooms • Exhaust FN-20A and FN-208. 

H-4 
! FAULTS lH 

•DG-tA LOAD 

SEGUENCER 

STAT1GN 

BP.TTERY 

FAULTS 

1

POWER NOT 

!AVAIL FRON 

!DIESEL 3EN iA 

I 

OEP-DGN-lP-DG1A A 

·:~~> 

I rr.i:·L.,.• c:; 

1

7.· \.: __• • ,..,,~: .. rr11 .. t ,..., .... L.~ 

l uur r· n •

1 c:; C,ii -o., I • ... w "'' • .. nL' 

j1 

i 

I 

OC?-5DC-~:x -BUS i ~--

{ \ I .I .... _ .... 

I POWER NOT I AVAILABLE FP.Ol11 

1125 VDC BUS 1 I 

DCP-aDC-lP-BUSl l'·, I I 

~-.j 

I 

I CJMHGN CAUSE 

!FAILURE JF 

I STATION 

~ f"'tl', .,._,...11'"' .• ,. .... I ~HIIt.:Ut.::l 

I 

POWE?. NOT 

AVAIU~BLE FRCM : 

141' .. uAC "I'S •: v I 

. ACP-3AC-LP-BUSS 

(\ 

,..,....~ 

IF'' .... ,., .... ,.. HlLUi1t. ,Jr 

l

i BAiiERV 

t1 ,.,f"".N.f'W'""' .Cl'1Httt::K 

! 

I 

FAULTS IN 4161) I 

liJAC BUS 5 I 

I I I 1 

ACP-?SF-LP-BUS5 (-.) 

\..._. 

!FAILURE OF 

i,...,,,..,..,.."''"'n'!"",'\.,..· 

l~l'lllL-MtJt:ni'\ 

!ROOM SUPPLY 

IFANt FN-3i 

AC?-FAN-~X-FN3i (---..• 

\ .... _, I 

AC?-BAC-LP-9US7 ;,..· .. ,. j 

I I, 

i-T-~ i 

I FAULTS HI 480 

VAC BUS 7 

l 

I 

ACP-PSF-LP-PUS7 t·---~, 

\,,) 

INSUFFICIENT 

SWITCHGEAR 

i ROOM COOL me 

I 

ACP-PSF-FC-3BRCL 

(~··i 

L....l 

! 

'1 OP. FAIL TO 

PI Acr: ~ ';l! nj I .... \.. . Ht~ .;1 

!SERVICE FOR 

'p--rovE-... 1 .t..... tn ; 

r 

ACP-XHE-FG-PRTFN 

(.-.... , 

I ! \ .. _ ... · 

Figure H-1 AC Power (MCC 7A) Fault Tree. 

H-5 

PG~EF; ~lOT 

Atf'A ILHBLE FROM ; 

460 !JAC MCC 7 A i 

AC?-9AC-LP-MCC7A 

/ ... \\ 

~T-) ; 

!FAILURE OF 1 

1

4 i 60-480\J XFHR '· .IJ Y-C::i'] ..,.,J I 

I ! ! 

j j 

I l 

ACP- TF11-~X -:t5(l7 ,.,.-.. ,. 

\ J .....,.. 

IFtilLURE OF 

tSHI!CHSEAR 

jROOn RETURN 

i r: A··~ F~·l- '" I' ·.1 l I ""' 

! 

i 

ACP-:= AN-~: X -FN32 1 ---·, ( .... __ , l 
·FAULIS 1N 

D6-1B LOAD 

SEQUENCER 

~EP-DGN-UF-3LDSG . :\ 

/ ' ~~" \ 

\..,;/ 

STATim~ 

SATiERY 3 

FAULTS 

POWER NOi 

AVAIL FROH 

jDIESEL 

I 

SEN tB 

I 

I fAULTS IN 125 

!VDC OIST. CAB. 

!3 

I 

ilCP-BDC-~:X -5US3 .. -. l ) \ I ... _.l 

I POWER NOT l 

I il\l' il , .... L,. i="'QM 1 f nrHL ... Htl t , 11 II, 

lt25 VDC BUS 3 I i I 

J I 

l I 

DCP-BDC-LP-BUS3 

('\ 

kr·l i 

lr-l!t.'l<lfi:•' "'·!'.::'"' ! ~ow•liiUr1 LnU ... t 

I F P, ! lURE GF I STAilDN 

l;i·"T'l=RIE·:: J ....-i1 t""". w 

I 

ilCP-CCF-lP-3AiT .... -.... 

II ) 

l"'--'' 

!POWER NOT J 

I

I AYA i LABLE FHot1 I 

14160 VAC BUS "I 

ACP-BAC-LP-BUS6 

·'""' I .\ 

I I 

~T'"t 

I FAULTS 

1

1N 4160 I 

r/AC BUS • 

1 

I I . ACP-PSF-LP-8US6 

r--· .. I l ~ .. _) 

!FAILURE OF !FAILURE OF '"'~-- .... I"':·.J ,,.,it'!'~I"'Lun.rr.,..,._ I t{i-1 i i tH ~ I t.V!! I Lli\:!1:1-l!i 

!

CHARGER 3 !ROOM SUPPLY 

I j:"'·ll .. l.l ..,, l I' Hn! r r~-.:; l 

DCP-8AT-MX-BCH3 t ! ACP-FAN-KX-FN31 ~' ------~ t-·., r--'t 

1

1 ) I I t.._,..· :......._./' 

PiJWER ;-~OT 

AVAIUBLE :=ROM 

480 VAC BUS 8 

HC?-BAC -LP- ~:US8 .... · .... r ., 

I l h-~ 

i 

FAULiS IN 480 

VAC BUS B 

AC?-~AC-L?-MCC8A 

AC?-?S~-L~-MCCSA r·-... ·:. I I i .: 

·---

!FAIL~RE OF I ! .... 'P."'' .. F"'"· I I 'fi6U-'t~iJY X f.i1 i 

I IX-608 . I 

I 1 I 

I i i ~---------- ACP-PSF-LP-3US8 

(-... ~~ 

I i l . ·' .,.__,.. 

ACP-PSF-tc-SSRCL _ ... ....,._. 

(.) 

y 

/DP. FAlL :u 

I

. PI t-..r·::: ;:H• \J '1 ~j 11...~1\J'- I I. h 

1 c;::;;.n-1

·"'J:" ;:;'iD 

1

-_._,,1 L. .... U1; 

!RECOVERY 

I 

l 

FAILURE OF 

SWITCHGEAR 

RDOt1 RETURN 

FAN! FH-32 

ACP-F HN-~: ;x -FN32 (-.,, 

; I \ . 

~-· 

Figure H-2 AC Power (MCC SA) Fault Tree. 

H-6 
FAULTS IN 

DIESEL 

3ENERAiOR lA 

1 DG- i .~ DU TPUi 

BREAKER FAlLS 

TO CLOSE 

OE?-BKR-00-ACBlA 

(-...', 

\ .. _) 

NO FUEL OIL 

SUPPLY TO 

1

11 1,..""E' '"ut::•\'E L} t::: L l:nul1~ 

i 

IBUS 3 TO 5 TIE 

i r,:-,,..."!.11f"" ..... - ... 1't ""'w 

11.'\Kt:.Hf .• t.:<; t'!-I!L~ '-to 110'")j I ;.;tt;,, I 

I 

I 

tiD BUS ~ LCSS 

OF VOLTAGE 

SIGNAL 

DG1A2 I OEP-ACT-F~-3US5 l\ PA6E 3 /\\ 

w \./ 

NO ?DWER FROM 

GE~JERH TDR 1 A 

US!Hl 

'!FAILURE OF D6 

r·io ::·A ........ T"G , H.,, ,;; I 111 .N 

~~~;~~~ A1R l t'd-"'t'!1::J 

I 

! 

!FAILURE GF 1

1 

I !)6-lA CONTROL i 

I ~ ". ,.. ... ~. • ~" • • Qll · I •:! u l J I i"( l tiU I l " 

~PANEL 

DEP-ACT-kX-lACTL (-., 

I t. 

\ .. _.) 

I uo CDMB~STIDN I AIR SUPPLY TQ 

1 ErmiNE 

I 

DGiA8 D81A11 

i\ PAG.. 4 0'' t:. 

I FAILURE OF i 

l DG1A AIR INLET I 

! DAMPER iO I 

IALLG\:J FLOW 

i I 

QEF-50D-~X-AlRA 

r-·'~ t J 

~-

FAILURE OF 

DS-tA RO'OM 

I

EXHAUSi FAN, 

FN-20A 

! 

OEP-FAN-KX-FN20A 

(..,_' 

I I I ! .... _ ... 

I ~'OMMI"~J "'AUC::: 

1 FAILURE .. oF "oGs 

I 

OEP-CCF-FC-DGN (.-.. ~, 

i i \'\.._.i 

!

INSUFFICIENT 

COOLING OF 

! I n·Ec~· 1 ..... L t '"t~ s·~·t ld .. 

I 

DG1At2 

/\ p .. ..,... ... I ·1 Htlt . I. 

u 

I

FAILL!R£ OF 

DS-1A EXHAUST 

OAHPE?. TO 

IALLDW FLOW 

I 

OEP-MOO-Y.X-EXHA 

(--'r 

l.l 

Figure H-3 Diesel Generator lA Fault Tree. 

H-7 

NO PiJ~~ER ?F.~~ 

125 VDC BUS ~ 

C?-BDC-LP-3US'1 
Figure H-3 {cont.) 

H-8 

NO PCC COOLING ! 

iO NECESSARY l 

'""'·Fr:"'y , .... 11S""::ll .. H .. 1 .:lT t .... I 

COMP 

. l I . 

I INSUFFICIENT 

iGOOLING QF 

I

I ii"t· ~l:~l ;N'·'t:;Hl't"' ;..! --\II-- .... 'wtl.li* 

! 

DG1A12 ... ---.. 

u 

!FAILURE OF 

IPCC-A-493 TO 

IALLUW FLO> 

! 

OEP-A01-~X-AV493 (-.. , 

1\._) 
lNO FUEL Oil 

I

SUPPL'i TO 

DIESEL EHSINE 

l 

I 

i 

~F~ILURE OF !OPERATOR DS~lA DAY FUEL !FAILURE TO 

TANK, TK-o2A . REFILL FUEL ' I I TANKS . 

NO rUEL iO i 

EMPTY DAY TANK I 

tTK-o2Al 

OEP ... iNK·U .. iK62A OEP·:<HE-FO·FUEL D61A7 I 

0 0 Q 

' I 

FAILURE OF . FAILURE OF OPERATOR 

DG·l·A DB·lA AUX FUEL FliiLURE TO 

UNDERGROUND OIL PUHP, REFILl FUEL 

FUEL TANK, P·llA TANKS 

TK·2BA 

C{;MMCN CAUSE I 

FAilURE OF AUY. ! 

FUE'L. PIJMP I - - OEP·iUK-~l-iK29A OEF-MDPO. X-P33A OEr·I.HE-FO-rUEL OEP-CCF-FC-P33X 

() (~) l(-} 

_, -

Figure H-3 (cont.) 

H-9 
.l 

FAILURE OF 

DS-lA SiART!N6 

AIR iANK 1 

1DS-1A 

STARHIN6 A!R BAt~K 1 

061ABA I 

~\ y 

FAlLURE OF 

DG·lA AIR 

STARTING TANK 

2 

FAILURE OF 

06-lA AIR 

STARTING TANK 

3 

1

FA1LURE OF JG 

AIR STARTING 

S'fSiEi'1 AIR 

RECVRS 

I 

FAILURE OF 

D6-1A STARTING 

AIR TAt~K 4 

lFA!LURE1 OF i l 06-iA STARTING ! 

lA .ltl .... -·~UV., t:lf·hm ._ I 

I I 

DSlASB I 

., 

I ' I ·~ 

~T~ 

.I 

FAILURE OF I 

D6-iA AIR 

S7ARTING TANK I 

5 I 

I I 

I 

FAILURE OF 

DG-lA AIR I 

STARTING TANK I 

l 

6 I . OEP-TNK-n-76A-1 OEP-iNK-U-76A-2 OE?-nu:-U-76A-3 DEP-1NK-n-76A-4 OEP-tNI<-U-7bA-5 - OEP-iNK-n-76A-c 

u ~ ~ rl~., n r) 0 u - v ~ J 

Figure H-3 (cont.) 

H-10 
APPENDIX I 

ACTUAnON SYSTEMS 

Table I-1. Actuation systems. 

Safety Function: To initiate the engineered safegu·ard systems. 

System Components: 

CIS Containment Isolation System 

CSAS Containment Spray Actuation System RAS Recirculation Actuation System SIAS* Safety Injection Actuation System RPS Reactor Protection System - High Pressurizer Pressure (PSR) - Low Steam Generator Level (SG LEV) 

Instruments: LIC-303 

AK,BK,CK 

LIC ... 304 

AK,BK,CK 

LT-1213 

A,B,C,O 

LT-1223 

A,B,C,D 

LT-1223 

A,B,C,D 

PIA-102 

A,B,C,D 

PS-2003 

A,B,C,D,E,F 

PS-2009 

A,B,C,O,E,F 

PS-2010 

A,B,C,D 

RWST Level (RAS-A) 

RWST Level (RAS-B) 

Support Systems: AC Power: Channel A 

Channel B 

Channel C 

Channel 0 

DC Power: CIS-A 

CIS-B 

CSAS-A 

CSAS-B 

RAS-A 

RAS-B 

SIAS-A 

SIAS-8 

Steam Generator 1 Level (SG LEV) 

Steam Generator 2 Level (SG LEV) 

Steam Generator 3 Level (SG LEV) 

Pressurizer Pressure 

Containment Pressure 

Containment Pressure 

Containment Pressure 

120 VAC Vital Bus 1 

120 VAC Vital Bus 2 

120 VAC Vital Bus 3 

120 VAC Vital Bus 4 

125 VDC Batt 1 

125 VDC Batt 3 

125 VDC Batt 1 

125 VDC Batt 3 

125 VDC Batt 1 

125 VOC Batt 3 

125 VOC Batt 1 

125 VOC Batt 3 

(SIAS, PSR) 

(CIS) 

(CSAS) 

(SIAS) 

* SIAS is the only actuation signal included in the system FT. 

1-1 
OPERATOR 

FAILURE TO 

UNBLOCK I 

'ACTUATE SIAS 

I 

lNO SIAS-ri 

IAUTOHATIC 

jACTUAi!ON 

I 

SISAl 

('\,, 

~-r~ 

--------- !~=~t ... ._u,_,l\ rT•: J1 :'~u-r I 

1

1· a,,_,aU,tt; DI:W'!J'"'i -wr\J t' .. Un 

!SIAS-A 

I 

SISA2 

0 I 

BIAS BLOCK 

I FAILED iO 

DISENABLE 

I 

I 

I 

SIS-ACT-FA-BLOCK lli·\ ..... 

\/ 

.. 

1

rW S!AS I A·urrMH·rrr I Ut II ... , 

i ·~~·r1 ~ ... ·r,,t f i;~ 1. H 11 L··' 

l 

I 

NO ACTUATION 

FRDi1 S!AS 

THAIN A 

DPERATOR 

FHILURE TO 

UNBLOCK J 

ACTUATE SIAS 

Figure I-1 Safety Injection Actuation System (Train A) Fault Tree. 

I 
NRC FORM 335 

12·841 

NRCM 1102, 

3201, 3202 

U.S. NUCLEAR REGULATORY COMMISSION 

BIBLIOGRAPHIC DATA SHEET 

SEE INSTRUCTIONS ON THE REVERSE. 

2. TITLE AND SUBTITLE 

Seismic Margin Review of the Maine Yankee Atomic 

Power Station 

Volume 2. Systems Analysis 

5. AUTHOR lSI 

D.L. Moore, D.M. Jones, M.D. Quilici and J. Young 

1. A !!PORT NUMSE R !Assigntd I>V TID C. lldd Vol No, if any/ 

NUREG/CR-4826 

UCID-20948 

Vol. 2 J. LEAVE BLANK 

4. DATE REPORT COMPLETED 

MONTH YEAR 

February l 987 Ei. DATE REPORT ISSUED 

MONTH YEAR 

March 1987 7. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS flncludeZ,p Code/ 8. PROJECT/TASK/WORK UNIT NUMBER 

Energy Incorporated 

Kent, WA 98031 9. FIN oR GRANT NuMBER 

Under Contract to: Lawrence L ivermo.re National Laboratory 

Livermore, CA 94550 A0461 

10. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Codttl 11a. TYPE OF REPORT 

Division of Engineering Safety 

Office of Nuclear Regulatory Research Technical 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 

b. PERIOD COVER En lfn~fusive dates/ June 1936 

thru March, 1987 

12. SUPPLEMENTARY NOTES 

13. ABSTRACT (200 words or 18ss} 

This Systems Analysis is the second of three volumes for the Seismic Margin Review of the Maine Yankee 'Atomi·~ Power Station. Volume 1 is the Summary Report 

of the first trial seismic margin review. Volume 3, Fragility Analysis, documents the results of the fragility screening for the review. The three 

volumes demonstrate how the seismic margins review guidance (NUREG/CR-4482) of 

the NRC Seismic Design Margins Program can be applied. 

The overall objectives of the trial review are to assess the s~ismic margins 

of a particular pressurized water reactor, and to test the adequacy of this 

review approach, quantification techniques, and guidelines for performing the 

review. Results from the trial review will be used to revise the seismic margin methodology and guidelines so that the NRC and industry can readily apply them 

to assess the inherent quantitative seimsic capacity of nuclear power plants. 

14. DOCUMENT ANALYSIS- a. KEYWOAOS/OESCRIPTORS 

Seismic f~argins 

Seismic Analysis 

b. IDENTIFIERS/OPEN·ENDED TERMS 

-~>"U.S.GOVE:RNMENT PRINTING QF"F'ICE:t1987-181-682:60J29 

15, AVAILABILITY 

STATEMENT 

Unlimited 

16, SECURITY CLASS IF !CATION 

(This p•gB) 

Unclassified 

(This rBport/ 

Unclassified 17. NUMBER OF PAGES 

18. PRICE 




UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300 

SPECIAL FOURTH·CLA.SS RATE 

POSTAGE & FEES PAID 

USNRC 

WASH. D.C. 

PERMIT No. G-87 
NUREG/CR-4826 

UCID-20948 

Vol. 3 

Seismic Margin Review of the 

Maine Yankee Atomic Power Station 

Fragility Analysis 

Prepared by M. K. Ravindra, G. S. Hardy, P. S. Hashimoto, M. J. Griffin 

EQE, Inc. 

Lawrence Livermore National Laboratory 

Prepared for 

U.S. Nuclear Regulatory 

Commission 
NOTICE 

This report was prepared as an account of work sponsored by an agency of the United States 

Government. Neither the United States Government nor any agency thereof, or any of their 

employees, makes any warranty, expressed or implied, or assumes any legal liability of re· 

sponsibility for any third party's use, or the results of such use, of any information, apparatus. 

product or process disclosed in this report, or represents that its use by such third party would 

not infringe privately owned rights. 

NOTICE 

Availability of Reference Materials Cited in NRC Publications 

Most documents cited in NRC publications will be available from one of the following sources: 

1. The NRC Public Document Room, 1717 H Street, N.W. 

Washington, DC 20555 

2. The Superintendent of Documents, U.S. Government Printing Office, Post Offici:! Box 37082, 

Washington, DC 20013·7082 

3. The National Tech meal Information Service, Springfield, VA 22161 

Although the listing that follows represents the majority of documents cited in NRC publications. 

it is not intended to be exhaustive. 

Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection 

and Enforcement bulletins, circulars, information notices, inspection and investigation notices; 

Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and 

licensee documents and correspondence. 

The following documents in the NUREG series are available for purchase from the GPO Sales 

Program: formal NRC staff and contractor reports, N RC·sponsored conference proceedings, and 

NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of 

Federal Regulations, and Nuclear Regulatory Commission Issuances. 

Documents available from the National Technical Information Service include NUREG series 

reports and technical reports prepared by other federal agencies and reports prepared by the Atomic 

Energy Commission, forerunner agency to the Nuclear Regulatory Commission. 

Documents available from public and special technical libraries include all open literature items, 

such as books, journal and periodical articles, and transactions. Federal Register notices, federal and 

state legislation, and congressional reports can usually be obtained from these libraries. 

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference 

proceedings are available for purchase from the organization sponsoring the publication cited. 

Single copies of NRC draft reports are available free, to the extent of supply, upon written request 

to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com· 

mission, Washington, DC 20555. 

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process 

are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available 

there for reference use by the public. Codes and standards are usually copyrighted and may be 

purchased from the originating organization or, if they are American National Standards, from the 

American National Standards Institute, 1430 Broadway, New York, NY 10018. 
~ ., ,. 

\ 

NUREG/CR-4826 

UCID-20948 

Vol. 3 

Seismic Margin Review of the 

Maine Yankee Atomic Power Station 

Fragility Analysis 

Manuscript Completed: February 1987 

Date Published: March 1987 

Prepared by 

M. K. Ravindra, G. S. Hardy, P. S. Hashimoto, M. J. Griffin 

EOE, Inc. 

Newport Beach, CA 92660 

Under Contract to: 

Lawrence Livermore National Laboratory 

7000 East Avenue 

Livermore, CA 94550 

Prepared for 

Division of Engineering Safety 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20566 

NRC FIN A0481 

ABSTRACT 

This Fragility Analysis is the third of three volumes for the Seismic Margin 

Review of the Maine Yankee Atomic Power Statjon. Volume 1 is the Summary 

Report of the first trial seismic margin review. Volume 2, Systems Analysis, 

documents the results of the systems screening for the review. The three volumes 

are part of the Seismic Margins Program initiated in 1984 · by the Nuclear 

Regulatory Commission (NRC) to quantify seismic margins at nuclear power plants. 

The overall objectives of the trial review are to assess the seismic margins of a 

particular pressurized water reactor, and to test the adequacy of this review 

approach, quantification techniques, and guidelines for performing the review. 

Results from the trial review will be used to revise the seismic margin methodology 

and guidelines so that the NRC and industry can readily apply them to assess the 

inherent quantitative seismic capacity of nuclear power plants. 

iii 
CONTENTS 

ABSTRACT .................................................................................................................................. iii 

LIST OF TABLES....................................................................................................................... vii 

LIST OF FIGURES ................................................................................................................... viii 

1. INTRODUCTION ........................................................................................................... . 

1.1 Background .................................................................................................. .. 

1.2 Objective of the Study ................... ; ........................................................ . 

1.3 Scope of the Study .................................................................................... . 

1-1 

1-1 

1-2 

1-2 

1.4 Organization of the Report ............................ ~...................................... 1·3 

2. REVIEW EARTHQUAKE LEVEL .............................. :.......................................... 2-1 

2.1 Screening of Components ......................................................................... 2-1 

2.2 Estimation of Seismic Capacities ........................................................ 2-1 

3. DESCRIPTION OF PLANT.STRUCTURES, SYSTEMS, AND 

COMPONENTS ............................................................................................................ 3-1 

3.1 Maine Yankee Plant/Structures and Systems ............................... 3-1 

3.1.1 Structures ........................................................................................... 3-1 

3.1.2 Systems ................................................................. ;............................... 3·1 

3.2 Seismic Design Criteria ........................................................................... 3-2 

3.3 Availability of Plant Design Data ..................................................... 3-2 

4. REVIEW OF PLANT INFORMATION AND WALKDOWN .................... 4-1 

4.1 Initial Screening of Components ......................................................... 4-1 

4.2 Review of Design-Analysis and Seismic 

Reevaluation Reports ............................................................................... 4-4 

4.3 Plant Walkdown ........................................................................................... 4·6 

4.3.1 Identification of Target Areas for First 

Walkdown ............................................................................................ 4-6 

4.3.2 Walkdown Procedures .................................................................. 4-22 

4.3.2.1 Walkdown Team .............................................................. 4-22 

iv 
CONTENTS (Continued) 

4.3.2.2 Procedures for Structures and 

Equipment ........................................................................... 4~23 

4.3.3 Walkdown Documentation ..................................... -.................... 4-29 

4.3.4 Walkdown Results .......................................................................... 4-30 

4.3.4.1 Walkdown Findings ....................................... ~.~.............. 4-30 

4.3.4.2 Plant Unique Features ................................................. 4-56 

4.4 Maine Yankee Component Modifications ....... ~............................... 4~59 

5. EVALUATION OF SEISMIC CAPACITIES OF COMPONENTS AND 

PLANT ............................................................................................................................. 5-1 

5.1 Review of Structural Models ............................................ ; ......... ~......... 5"1 

5.2 Simplified Analysis and Use of Screening Tools ....................... 5-2 

5.3 Second Walkdown ............................................................................... .'....... 5-4 

5.4 HCLPF Capacity of Components ..... ~.................................................. 5-4 

5.4.1 Fragility Analysis Method ......................................................... 5-4 

5.4.1.1 Methodology ....................................................................... 5-4 

5.4.1.2 Steel Structures ................................................................ 5·23 

5.4.1.3 Block Walls ......................................................................... 5-30 

5.4.1.4 Flat Bottom Storage Tanks ......................................... 5-32 

5.4.1.5 Inverter ................................................................................ 5-40 

5.4.1.6 Diesel Fuel Oil Day Tank ......... ~ ..... :.......................... 5-46 

5.4.1.7 Containment Spray Fans .~........................................... 5-52 

5.4.2 CDFM Method .................................................................................. 5-53 

5.4.2.1 Circulating Water Pumphouse .................................. 5-54 

5.4.2.2 Block Wall SB 35-3 ......................................................... 5-54 

5.4.2.3 Refueling Water Storage Tank ................................. 5-55 

\ ~.'. ~. 5.5 HCLPF Capacity of Plant ............................................................................ 5·56 

5.5.1 Accident Sequences ............................................................. :.............. 5-56 

5.5.2 Probabilistic Method ......................................................................... 5-59 

5.5.3 Deterministic Method ....................................................................... 5-67 

5.5.4 Sensitivity Studies .............................................................................. 5-68 

v 
CONTENTS (Continued) 

· 6. COMM:ENTS AND RECOMMENDATIONS ON SEISMIC MARGIN 

REVIEW METHODOLOGY .................................................................................. 6-1 

6.1 Selection of Review Earthquake Level .................................................. 6-1 

6.2 Use of Screening Guidelines ......................................................................... 6-2 

6.2.1 Extent of Review Needed ............................................................... 6-2 

6.2.2 Additional Screening Guidelines ................................................. 6-2 

6.2.3 Difficulties of Reviewing Certain 

Components ............................................................................................... 6-3 

6.2.3.1 Reactor Internals and CEDM ................................... 6-3 

6.2.3.2 Equipment Within Containment .... ~......................... 6-3 

6.3. Availability of Qualification Data ....... ~ ........ t......................................... 6-4 

6.4 Walkdown Procedures ...................................................................................... 6-5 

6.5 Guidance on CDFM Capacity Calculation Procedures ................... 6-6 

6.6 Staffing Requirements and Schedule ...................................................... 6-7 

6.6.1 Staffing Requirements ................................................................ 6-7 

6.6.2 Schedule .............................................................................................. 6M8 

6. 7 Applicability of Methodology for Other Plants ................................ 6-8 

7. RESULTS AND CONCLUSIONS ............... ~............................................................ 7-1 

7.1 Screened Out Systems and Components ................................................. 7-1 

7.2 HCLPF Capacities of Screened .. in Components ................................ ~ 7-2 

7.3 HCLPF Capacity of Plant ............................................................................ 7-32 

7.4 Identification of Low Capacity Components ..................................... 7M32 

7.5 Conclusions .......................................................................................................... 7-32 

REFERENCES .......................................................................................................................... R-1 

APPENDIX 

A. Maine Yankee Atomic Power Station Arrangement 

Drawings .................................................................................. ~.................................... A-1 

vi 
LIST OF TABLES 

3.2-1 Original Design Damping Values ........................................................ 3-5 

4.1·1 Initial Screening of Maine Yankee Components 

by Categories Based on Seismic Capacities ................................... 4-2 

4.3·1 Structure Separations ................................................................................. 4-8 

4.3-2 List of Block Walls Targeted for Walkdown ................................. 4-9 

4.3-3 Maine Yankee Atomic Power Plant First Walkdown 

List of Equipment Components ........................................................... 4-13 

4.3-4 Example Pump Walkdown Data Sheet ............................................. 4-31 

4.3-5 Example HV AC Components Walkdown Data Sheet ................ 4·33 

4.3-6 Example Block Wall Data Sheet .......................................................... 4-35 

5.3-1 Second Walkdown Component Review List ................................... 5-5 

5.4-1 Diesel Fuel Oil Tank Critically Stressed Areas .......................... 5-51 

5.5-1 Component Seismic Fragility Parameters ....................................... 5-57 

5.5-2 Probabilities for Nonseismic Failures .............................................. 5-58 

' 5.5-3 Summary of Plant Level HCLPF Capacities ................................ 5-64 

7.1-1 Summary of Group A Structures ......................................................... 7-2 

7.1-2 Summary of Group A Block Walls ...................................................... 7-3 

7.1-3 Summary of Maine Yankee Equipment Screening 

and HCLPF Capa.cities ............................................................................. 7 ·6 

vii 
LIST OF FIGURES 

2.1 0.3g pga Review Earthquake Ground Response 

Spectra, 5o/o and I OOJo Damping ............................................................. 2-2 

3.2~1 0.05g Design Earthquake Ground Response Spectra ........ ~........ 3-3 

3.2-2 O.IOg Hypothetical Earthquake Ground Response 

Spectra .............................................................................................................. 3-4 

4.3-1 Block Wall Evaluation Procedure ....................................................... 4-38 

4.3-2 Example of a PORV at the El Centro Steam Plant .................. 4-44 

4.3-3 Maine Yankee PORV PR-S-14 and PR·S-15 ................................. 4-45 

4.3-4 Maine Yankee CS Pump Suction MOV's CS-M-91 

and CS-M-92. Valves with Long Operator Arms ....................... 4-47 

4.3-5 Example of Long Operator Valves Documented in 

the Data Base from the Coalinga Area ........................................... 4-48 

4.3-6 Steam Lines Spanning Between the Steam and Feed 

Water Valve Area and the Turbine Building ............................... 4-50 

4.3-7 Photographs of Typical Maine Yankee Cable Trays ................ 4-52 

4.3-8 Photographs of Typical Maine Yankee Instrument 

Racks and Component Attachments .................................................. 4-54 

4.3-9 Photographs of the Maine Yankee Control 

Room Ceiling ................................................................................................ 4-55 

4.4~1 Block Wall VE 21-1 .................................................................................... 4-57 

4.4-2 Conceptual Seismic Retrofit to Block Wall 

VE 21-1 ............................................................................................................ 4-58 

viii 
'j ~ I 

LIST OF FIGURES (CONTINUED) 

4.4~3 Photograph of the Alarm Message Display Mounted 

on the Main Control Board .................................................................... 4-60 

4.4·4 Photographs of the Maine Yankee Station Service 

Transformer Internal Core/Coil Assembly Before 

the Anchorage Modification ................................................................. 4-62 

4.4-5 Proposed Anchorage Modification for the 

Maine Yankee Station Service Transformers 

Internal Core/Coil Assemblies ............................................................. 4-63 

4.4-6 Maine Yankee Air Conditioners in the 

Unmodified Condition ............................................................................. 4-64 

4.4·7 Proposed Anchorage Modifications for the 

Maine Yankee Air Conditioners ......................................................... 4-65 

4.4-8 Maine Yankee Containment Spray Fans Before the 

Installation of the Anchorage Modifications ............................... 4-67 

4.4-9 Proposed Anchorage Modifications for the Maine 

4.4-10 

5.4-1 

Yankee Containment Spray Fans ........................................................ 4-68 

Photographs of the Unstrapped Emergency Lighting 

Units in the Control Room and Welding Cart Near 

the Containment Spray Fans ................................................................. 4-69 

Fragility Curves ........................................................................................... 5-9 

5.4-2 Median, 5o/o Nonexceedance, and 95o/o Nonexceedance 

Fragility Curves for a Component ..................................................... 5-11 

5.4-3 Family of Fragility Curves for a Component .............................. 5-13 

ix 
LIST OF FIGURES (CONTINUED) 

5.4-4 Circulating Water Pumphouse .............................................................. 5-24 

5.4-5 Circulating Water Pumphouse, Typical Braced Frame ................ 5·25 

5.4-6 Turbine/Service Building, EI. 35'-0"/El. 39'·011 ............................ 5·28 

Turbine/Service Building, El. 61 '-0" ................................................. 5-29 

5.4-8 Block Wall SB35-3 ....................................................................................... 5-31 

5.4-9 Refueling Water Storage Tank ............................................................ 5-34 

5.4-10 Model for Determination of Tank Resistance 

Against Seismic Base Moment .............................................................. 5-36 

5.4·11 Beam Model of Tank Bottom Plate ................................................... 5-37 

5.4-12 Yield Line Model of Anchor Bolt Chair Top Plate .................. 5-39 

5.4-13 Maine Yankee 12 kVA Inverter and Battery Charger, 

and a Close-up of the Anchorage Addition .................................. 5-41 

5.4-14 Diesel Fuel Oil Day Tank (TK-62B) ................................................. 5-47 

5.4-15 Diesel Day Tank Critical Areas .......................................................... 5-48 

5.4-16 Diesel Day Tank Critical Areas .......................................................... 5-49 

5.4-17 Retrofit Anchorage Design for the Day Tank ............................ 5-50 

5.5-1 Seismic Fragility Curves of a Component ..................................... 5-60 

Fragility Curves for Small LOCA Core Damage ........................ 5-62 

5.5-3 Seismic Fragility Curves for No LOCA Core Damage ............ 5-63 

X 
1.1 Background 

CHAPTER 1 

INTRODUCTION 

A seismic margin review methodology has been developed by the NRC Expert 

Panel as documented in [Budnitz et al., 1985] and [Prassinos, et al., 1985]. The 

objective of the methodology is to estimate a high confidence value of the seismic 

capacity of a nuclear power plant. The methodology has been derived using the 

results and insights gained in conducting over 20 published and unpublished 

seismic probabilistic risk assessments, the data on actual performance of structures 

and equipment of industrial facilities and power plants in major earthquakes, and 

the data on qualification of equipment in nuclear power plants. The methodology 

consists of selecting a review earthquake level, screening out categories of plant 

structures and equipment which have seismic capacities generically higher than the 

review earthquake level, screening out components that are not essential for certain 

plant safety functions, performing plant walkdown to confirm that the screening 

of components (i.e., structures and equipment) is acceptable, and estimating the 

seismic capacities of the components and of the overall plant. The seismic capacity 

of a component or plant in this context is the so·called High Confidence Low 

Probability of Failure (HCLPF) capacity; it is a conservative representation of 

capacity and in simple terms corresponds to the earthquake level at which it is 

extremely unlikely that failure will occur. It can be viewed as approximately 

equal to the earthquake level for which we have 95% confidence that the 

probability of failure is less than 5%. At this· time, seismic margin methodology 

has been developed for pressurized water reactors based on the review of system 

analysis results and insights of a number of seismic PRAs on these reactors; for 

boiling water reactors additional reviews of seismic PRAs are required to develop 

seismic margin review procedures. 

The seismic margin review methodology has been developed and some preliminary 

guidelines for performing the margin r~view have been indicated in the foregoing 

references. It was felt necessary to apply the methodology to a selected plant on a 

trial basis in order to check whether the methodology is implementable and 

whether the guidelines need further revision and amplification. For this purpose, 

the Maine Yankee Atomic Power Plant was selected. It is expected that the 

methodology and margin review guidelines will be revised subsequent to this trial 

plant review. 

The NRC selected the review level earthquake to be 0.3g peak ground acceleration 

with a NUREG/CR·0098 50th percentile spectral shape. Maine Yankee concured 

with this selection. · 

The trial plant review is conducted by the Lawerence Livermore National 

Laboratory (LLNL) and is being funded by the U.S. Nuclear Regulatory 

Commission. It consists of two major aspects: system analysis and fragility 

evaluation. The contracts to perform these two aspects of the trial plant review 

have been awarded to Energy International Incorporated and EQE Incorporated, 

respectively. Volume 3 of the report describes the fragility evaluation aspect of 

1-1 
this seismic margin review. Volume I summarizes the entire seismic margin review 

study on Maine Yankee; Volume II discusses the system analysis aspects of the 

study. 

1.2 Objective of the Study 

This fragility aspect of the trial plant seismic margin review is aimed at achieving 

the following objectives: 

o Apply the seismic margin review methodology developed by the 

NRC Expert Panel and identify areas where the margin review 

guidelines need modification and clarification, 

o Estimate the HCLPF seismic capacity of the Maine Yankee 

Atomic Power Plant and identify any seismic vulnerabilities in 

the plant . 

. 1.3 ScoPe of the Study 

The primary purpose of this study has been to evaluate the NRC Expert Panel 

seismic margin review methodology using Maine Yankee as the trial plant; the 

secondary purpose is to assess the seismic margin of Maine Yankee. The scope of 

the study is defined with these two purposes in view. 

o The methodology outlines two approaches for estimating the 

component and plant-level seismic capacities: probabilistic and 

deterministic. The guidelines for the deterministic evaluation of 

the capacities are not sufficiently developed. Therefore, only the 

probabilistic approach is used in estimating both the component 

and plant-level seismic capacities in this study. Although this 

does not check out all the features of the Expert Panel 

methodology, it does provide an estimate of the seismic margin of 

Maine Yankee. Much further work is needed to develop more 

definitive guidelines for deterministic evaluation of seismic 

margins in the Expert Panel methodology. 

o The seismic capacities of structures and equipment are estimated 

using the structural models and qualification analysis results 

provided by the Maine Yankee utility. The adequacy of the 

structural models and the reasonableness of the seismic responses 

of structures and equipment were confirmed by cursory review 

and based on judgment. The floor response spectra generated by 

Maine Yankee are judged to be adequate for this seismic margin 

review. However, this cannot be construed to be a detailed 

review duplicating the review done as part of the plant QA/QC 

program in its licensing. 

o The limitations of the Seismic Margin Review Methodology as 

outlined in the cited references by the Expert Panel also apply to 

1-2 
the study described in this report (i.e., relay issues, design and 

construction errors, operator errors under seismic stress, etc). 

o The screening criteria for components developed by the Expert 

Panel based on their generic high seismic capacities (i.e., 

components denoted by letter "C" in Table 5-l of NUREG/CR· 

4334) are assumed to be applicable using minimal review and 

engineering judgment. In general EQE is in agreement with the 

Panel's recommendations. 

1.4 Organization of the Report 

Chapter 2 discusses the review level earthquake specified for this trial plant 

seismic margin review. A general description of the Maine Yankee . plant 

structures, systems, and components is given in Chapter 3. The processes of plant 

design review, initial sc·reening of components, and the plant walkdown are 

detailed in Chapter 4. Review of structural models, simplified analysis and second 

walkdown for additional data, and HCLPF capacity calculation for components 

and plant are described in Chapter 5. Feedback on the methodology in the areas 

of selection of review earthquake level, screening guidelines, walkdown procedures, 

HCLPF capacity calculation, staffing requirements, and applicability to other 

plants are discussed in Chapter 6. The results of the· study in terms of HCLPF 

capacities of components and plant, and any seismic vulnerabilities in the plant, 

are highlighted in Chapter 7. Appendix A consists of a set of general arrangement 

drawings showing the structures and equipment in the Maine Yankee plant. 

1-3 

"'I ,i 

CHAPTER 2 

REVIEW EARTHQUAKE LEVEL 

For this trial application of the NRC Expert Panel seismic margin review 

methodology, the NRC staff has specified a review earthquake level of 0.30g pga 

anchored to the median NUREG/CR-0098 ground response spectrum for rock sites 

(Figure 2.1). In this chapter, the implications of this review earthquake level will 

be discussed from the viewpoint of screening of components and seismic capacity 

calculations. 

2.1 Screening of Components 

The guidelines developed by the Expert Panel for screening of components based 

on their generic seismic capacities are considered applicable for review 

earthquakes of magnitudes less than 6.5, with 3 to 5 strong motion cycles and a 

total duration of 10 to 15 seconds. The spectral content of this earthquake is 

characterized by a broad-band spectra in the structural frequency range of 1 to 7 

Hz. 

The selected review earthquake for Maine Yankee meets all of the above 

requirements. It is, therefore, concluded that the screening guidelines given in the 

Expert Panel reports NUREG/CR-4334 and 4482 are applicable to the present 

seismic margin review. 

2.2 Estimation of Seismic Capacities 

The concept of HCLPF capacity requires that it is associated with a defined 

response spectrum and a specified nonexceedance probability. 

The HCLPF capacity used for screening as well as the calculated HCLPF capacities 

. for particular components not initially screened out and the final plant level 

HCLPF capacity are considered to be valid provided ground motion from any 

earthquake does not exceed the review earthquake level spectrum for more than 

16% of the spectral frequencies within the range of interest. The review 

earthquake level spectrum is a spectral shape defined by the 50% exceedance 

spectrum specified in NUREG/CR-0098 and anchored at 0.3g pga for the initial 

screening. The seismic margin for the components and plant is referenced to this 

spectrum but anchored to the pga corresponding to the HCLPF capacity. 

This definition of spectra used to determine a HCLPF capacity does not in any 

way refer to the probability of occurrence of an earthquake. It is no more than an 

arbitrary spectrum used to define the HCLPF capacity that recognizes the 

dependency of a component capacity on the frequency content of the spectrum and 

not just the pga. 
-0 . 

N 

0 

.::1" 

0 

'liD 

0 . 

co 

0 . 

N 

.::1" . 

:::'liD 

., . 

•co c . 0 

--

a:: 

"" a.. 

N 

co 

()~~~~~~~~~--~~~~~~~,~~~~~~~ 

0 

... 

0 

'loD 

0 

co 

0 

00 0 -\D ... Sd I • J.ll :l01 31\ 

2-2 

0 

0 

., 

Q.. 

u 

,. 

>-

- u :: CIQW .g 

N 

0 

w 

a: 

La.. 
The results of the seismic margin study are interpreted as follows. The HCLPF 

capacity of the structures, equipment and plant are conditional on the actual sitespecific spectrum not exceeding the target spectrum; exceedance is defined as the 

event when 16 percent of the spectral ordinates exceed the target spectrum over 

the frequency range of interest. It is assumed that the. spectrum peak-to-peak and 

earthquake direction variabilities (Sec. 5.4.1.1) are removed from the hazard 

analysis leading to the selection of the review earthquake. The review earthquake 

is specified by the same spectrum (Figure 2.1) in two horizontal directions and 2/3 

of the horizontal spectrum in the vertical direction. It is also assumed that the 

review earthquake level is specified as the higher of the response spectra from the 

two orthogonal horizontal directions. 

2-3 

CHAPTER 3 

DESCRIPTION OF PLANT STRUCTURES, SYSTEMS, AND COMPONENTS 

3.1 Maine Yankee Plant/Structures and Systems 

Maine Yankee plant is located on the west shore of the Back River approximately 

3.9 miles south of the center of Wiscasset, Maine. The plot plan is shown in 

Appendix A. 

Maine Yankee is a one-unit, 3-loop PWR supplied by Combustion Engineering with 

a rated capacity of 825 MWe. It began commercial operation in December 1972. 

3.1.1 Structures 

The major structures on the site are the reactor containment, primary auxiliary 

building, fuel building, turbine building, service building, and circulating water 

pumphouse. They are founded on hard rock. 

The reactor containment is a steel-lined reinforced concrete cylinder with a 

hemispherical dome and an essentially flat reinforced concrete foundation mat. 

The turbine building houses the turbine generator and the two diesel generators. 

The service building consists of the main control room, switchgear room, shops, ,and 

employee facilities. Portions of the turbine/service building, control/switchgear 

building, and diesel generator enclosure have reinforced concrete walls and slabs. 

The service building above EI. 39 ft 0 in. and the remaining portions of the 

turbine/service building are of structural steel framing with diagonal bracing and 

reinforced concrete slab with metal roof deck. 

The primary auxiliary building houses pumps, and tanks used for purification and 

processing of water from the reactor coolant system. It is made of reinforced 

concrete walls and slabs. 

The circulating water pumphouse contains the circulating water pumps and service 

water pumps. Below El. 21 ft 0 in., the structure has reinforced concrete walls and 

slabs. Above this elevation, it is fabricated structural steel framing with diagonal 

bracing and reinforced concrete slab. 

Other structures of interest to this study are the ventilation equipment room, 

containment spray pumphouse, and the main steam valve house. All of these 

structures are constructed of reinforced concrete walls and slabs except that the 

interior structure of the main steam valve house is of structural steel framing with 

diagonal bracing and metal gratin g. 

3.1.2 Systems 

In this study, the main focus is on the systems (i.e., front line and supporting) that 

support the Group A System functions. These systems are described in Volume 2 

of this report. 

3-1 
3.2 Seismic Design Criteria 

The original design of the Class I structures and components was based on a 

"Housner Spectrum" anchored to 0.05g for the design earthquake and O.lOg for the 

hypothetical earthquake (SSE). (See Figures 3.2-1 and 3.2-2.) 

The damping values used for the design of different structures, piping and 

equipment are summarized in Table 3.2-1. 

The following structures and components are characterized as Class I: 

o Reactor containment, including penetrations 

o Reactor vessel and its internals 

o Reactor coolant system 

o Reactor containment crane 

o Chemical and volume control system 

o Residual heat removal system 

o Safety injection system 

o All components affecting the ability of the control rods to scram 

o Containment spray system 

o Spent fuel pool and racks 

o Component cooling system 

o Circulating water system intake structure 

o Service water system 

o Emergency generators 

o Refueling water storage tank 

o Control room 

o Emergency steam generator feed pumps and piping 

The structures and components (including piping, cable trays, and HV AC systems) 

were designed and qualified to the requirements of the applicable AISC, ACI, and 

ASME codes of 1960 to 1970 versions. 

-The plant structures and equip.ment have been reevaluated at various times for 

earthquakes larger than the original SSE and for more recent regulatory standards 

[Stevenson, 1983]; [Hashimoto et al., 1984]; [Whittier, 1986]. Certain upgrades to 

improve the seismic capacity of the plant have also been made, e.g., additional 

anchoring of electrical equipment, and strengthening of block walls. The present 

study has focused on evaluating the seismic capacity of the plant in its current 

state, including certain modifications installed during the course of the margin 

review, or to be installed during the March 1987 outage. 

3.3 Availability of Plant Design Data 

Maine Yankee is an older plant designed and constructed before the development 

of quality assurance programs and seismic qualification methods presently existing 

in the nuclear industry. Hence, the qualification reports on certain equipment 

items were not available. Because of the reevaluation efforts, additional 

information and new structural models have become available. However, the 

extent of information available is not comparable to that normally available for a 

modern nuclear power plant (e.g., near term operating license plant). 

3-2 
< 

~ 

,-

~ 

... 

~ 

' w ; r I -· ...__., w 

BfQUfNCY h••l 

Figure 3.2-1 0.05g Design Earthquake Ground Response Spectra. 
c:i 

'"" +"' 

(,) 

u 

Q. 

r.l'.l 

u. 

en 

= 0 

Q. 

en 

u 

~ 

"0 

= ::::;) 

0 

'"" 

- {.'j • • u u ...¥j -

~ cd 

::::;) 

0" 

"' .c 

5 +"' 

'"" cd ... ~ f -cd 

(,) 

..... 

+"' 

u 

..c:: 

+"' 

0 

Q. 

>. 

::c 

00 

0 

ci 

N I 

N 

~ 

0 

s... 

::::;) 

Of) 

i.L: 

VIU>CITY 

3-4 
·,· ,• .. 

1. 

2. 

3. 

4. 

5. 

6. 

7. 

Table 3.2-1 Original Design Damping Values 

Percent of 

Design 

Earthquake 

Reactor containment structure 2.0 

Reinforced concrete structure, other 

than containment structure, founded 

on rock or soil 2.0 

Reinforced concrete supporting structure 

(not founded on soil or rock) 2.0 

Steel-framed structures, including 

supporting structures and foundations 

Bolted or riveted 3.0 

Welded 1.0 

Reactor vessel, internals and control 

rod drives 

Welded assemblies 1.0 

Bolted assemblies 3.0 

Mechanical equipment,including pumps, 

fans and similar items 2.0 

Piping systems 1.0 

3-S 

Critical Damping 

Hypothetical 

Earthquake 

s.o 

5.0 

s.o 

s.o 

2.0 

1.0 

3.0 

2.0 

2.0 
The structural models were generally reviewed by EQE; the floor response spectra 

generated for the review earthquake spectrum were provided by Maine Yankee. 

The spectra were judged to be realistic and representative of the seismic response 

for the review earthquake. Seismic capacities of structures and equipment were 

estimated by comparing the original design-analysis (or reevaluation) spectral 

values with the new spectral values and taking into account the margins and 

variabilities due to differences in damping, methods of analysis and testing, and 

procedures for mode combination and directional components. Where the original 

design analysis or reevaluation information was not available or applicable, the 

structure or equipment was analyzed by EQE to establish its seismic capacity. 

3-6 
CHAPTER 4 

REVIEW OF PLANT INFORMATION AND WALKDOWN 

In this chapter, a discussion of the plant information gathered during a review of 

drawings, and FSAR and seismic analysis/qualification reports will be given. The 

plant design information for systems and components supporting Group A 

functions is used to perform the initial screening of components based on the 

Expert Panel recommendations. The review of plant information is also aimed at 

identifying target areas and developing a strategy for the first plant walkdown. 

This chapter describes the procedures used in the plant walkdown, documentation, 

and salient findings. 

4.1 Initial Screening of Componenu 

Energy Incorporated provided a list of systems and equipment at Maine Yankee 

that support the Group A functions. From this list the structures containing the 

equipment were identified. NUREG/CR-4334 gives guidance on screening out 

certain structures and equipment from further consideration based on their 

generically high seismic capacities. 

For the chosen review earthquake level of 0.3g pga, the Expert Panel report 

recommends that the margin review may generically screen out the component 

categories after satisfying some caveats as shown in Table 4.1·1. Since the scope of 

this study includes both an evaluation of the seismic margin review methodology 

and a trial application to the Maine Yankee Atomic Power Station, the initial 

screening could not eliminate a large number of generic categories of components 

from further consideration in the review, walkdown, and analysis. This is so 

because the Expert Panel's recommendations on screening had to be generally 

confirmed as appropriate to Maine Yankee. 

The following items were screened out at this stage: 

o Containment structure 

o NSSS supports 

o Soil liquefaction potential 

For the screened-in components, further evaluation consisted of two levels: 

1. For those components identified by the Expert Panel as having 

HCLPF capacities larger than 0.3g pga, a minimal evaluation was 

done during plant review and walkdown to confirm the 

applicability of the Panel's recommendations for the Maine 

Yankee component categories. 

This was done in the following manner: 

o Control rod drive mechanism: Review of CE design 

o Structures: All concrete structures housing Group A systems 

and equipment were reviewed 

o Valves: Sampling review 

4-1 
,. '~ ' 

Table 4.1"~1 Initial Screening of Maine Yankee 

Components by Categories Based on Seismic Capacities 

Component 

Expert 

Recommendation Remarks 

I. Containment c Maine Yankee has a reinforced 

concrete containment structure; 

previous study by Hashimoto, et 

al (1984) confirms that the 

HCLPF capacity is in excess of 1 g. 

2. NSSS Supports c Studies done by LLNL as part of 

DEGB evaluation included the 

Combustion Engineering (CE) 

NSSS Supports. The NSSS Supports have HCLPF capacities in 

excess of 0.3g. 

3. Reactor Internals X To be evaluated. 

4. Control Rod Drive ·C To be confirmed by a review of 

Mechanism the CE design. 

5. Concrete Structure c To be confirmed by a review and 

Failures (Shearwalls, walk down. 

diaphragms, impact) 

6. Steel Structures X To be evaluated. 

7. Block Walls X To be evaluated. 

8. Piping X Panel•s caveats to be addressed. 

9. Valves c To be confirmed during walkdown. 

10. Heat Exchangers X Support and anchorage to be 

evaluated. 

11. Tanks X To be evaluated. 

4-2 
Table 4.1~1 Initial Screening of Maine Yankee 

Components by Categories Based on Seismic Capacities (Continued) 

Component 

12. Batteries and Racks 

13. Active Electrical 

Equipment 

14. Diesel Generators 

15. Pumps 

16. Soil Liquefaction 

17. HVAC Systems 

Fans and Cooler 

Units 

Ducting 

18. Cable Trays and 

Cabling 

19. Control Room 

Ceilings 

20. Dams, Levees 

and Dikes 

Expert 

Recommendation 

X 

X 

c 

c 

c 

X 

c 

c 

X 

X 

C = HCLPF capacity is larger than 0.3g pga. 

Remarks 

To be evaluated during review 

and walkdown. 

Anchorage to be reviewed. 

To be confirmed during walkdown. 

To be confirmed during walkdown. 

Rock site. 

Units on vibration isolators to 

be reviewed 

To be confirmed during walkdown. 

To be confirmed during walkdown. 

To be evaluated. 

To be evaluated. 

X = HCLPF capacity needs to be established through review, 

walkdown, and/or calculations. 

4-3 
o Diesel generators and peripherals: Complete review 

o Pumps: All pumps were evaluated 

o HV AC ducting: Sampling review 

o Cable trays and cabling: Sampling review 

2. For those components identified by the Expert Panel as requiring 

a review and walkdown, a detailed evaluation was performed as 

described in this report; it includes steel structures housing Group 

A systems. 

4.2 Review of Design-Analysis and Seismic Reevaluation Reports 

Initial and subsequent data collection efforts concentrated on those structures and 

components identified by Energy Incorporated required for the reactor 

subcriticality and early emergency core cooling. The plant specific seismic 

qualification information was primarily made available to EQE by Maine Yankee. 

In a few instances, outside vendors were contacted for additional information that 

was either lacking or proprietary for certain components (e.g., reactor vessel 

internals and the control element drive mechanism). The following provides a list 

of the types of information collected for the Maine Yankee Atomic Power Station 

as part of the margin evaluation. 

The data collected for use in the margin evaluation can be organized into four 

main categories: 

o Drawings 

o Maine Yankee reports and calculations 

o Independent review and reports 

o External information 

Drawings 

-

Types of drawings collected from Maine Yankee include the following: 

o Maine Yankee structural, architectural, and excavation design 

drawings 

o Design sketches of block wall seismic retrofits 

o Maine Yankee plant general arrangement drawings including 

floor elevations showing equipment locations 

o Support and or anchorage drawing details for equipment 

o Equipment vendor drawings indicating component construction 

(e.g., configuration, size, and materials used) 

Maine Yankee Reports and Calculations 

Typical types of reports and calculations collected include the following: 

4-4 
o Sections from the Final Safety Analysis Report (FSAR) 

o Structural steel, roof deck, and block wall construction 

specifications 

o Tables summarizing Maine Yankee block wall information 

o Maine Yankee generated calculations for seismic evaluation of 

block wall and retrofit design 

o Maine Yankee generated calculations for equipment seismic 

qualification. 

o A review of the Maine Yankee FSAR Amendment No. 35, Volume 

II, which documents the seismic qualification of vital instrument 

and electrical equipment. The amendment consists primarily of 

certified letters from vendors regarding conformance of their 

components to the seismic requirements of the Maine Yankee 

component procurement specification; however, a few calculations 

were provided and used in the component evaluations. 

Independent Reviews and Reports 

Several independent reviews and reports conducted for selected structures and 

equipment components at Maine Yankee collected include: 

o Cygna report describing structure dynamic analysis models 

(Cygna Report BM·Y .. MY-80006·5, April 1982) 

o Cygna reports describing dynamic analyses performed for several 

Maine Yankee critical equipment components 

o Cygna computations for building floor spectra generation using 

the 0.18g pga NUREG/CR~0098 50th Percentile Ground Response 

Spectra 

o Report by J. D. Stevenson, "Seismic Review of the Maine Yankee 

Nuclear Power Plant, .. which analyzed several critical components 

at Maine Yankee 

o Report by Structural Mechanics Associates, Inc., "Conservative 

Seismic Capacities of the Maine Yankee Reactor Containment 

Including and Excluding Design Incident Pressure," [Hashimoto et 

al., 1984]. 

External information not available through Maine Yankee 

Several examples of information collected include: 

4-5 
o Outline drawings and seismic qualification data for the Maine 

Yankee reactor vessel internals and the control element drive 

mechanism obtained from Combustion Engineering. 

o Information regarding dimensional data and support 

configuration for the Maine Yankee station service transformer 

internal core/coil assembly collected from contacts with the · 

manufacturer, General Electric Medium Voltage Transformer 

Division. 

o Lateral load capacity of vibration isolators supporting the Maine 

Yankee computer room air conditioners and the laboratory air 

conditioner, collected from contacts with the manufacturer, 

Vibration Mountings and Controls Inc .. 

During the course of the margin evaluation, several requests were made for 

additional plant or component qualification data. The additional requests were 

required as a result of the following: 

o The equipment component list was being refined, adding and 

deleting systems and components 

o Low capa.city components from the first plant walkdown were 

identified requiring additional component specific data to 

complete the evaluation. 

4.3 Plant Walkdown 

4.3.1 Identification of Target Areas for First Walkdown 

Target areas for the first walkdown were developed from the initial equipment list 

provided by the system analysts. This list identified preliminary equipment 

components as critical to reactor subcriticality and early emergency core cooling. 

From the preliminary equipment component list the critical structures housing this 

equipment were identified. The following provides a list and brief discussion of 

the structures and equipment identified as target areas for the first walkdown. 

Structures 

Structures identified for the first walkdown were those identified as housing 

Group A components. 

o Containment structure 

o Primary auxiliary building 

o Circulating water pumphouse 

o Turbine/service building 

o Containment spray pumphouse 

o Main steam valve house 

o M.C.C. room 

o Aux feed pumphouse and purge air exhaust area 
o Fire water pumphouse 

o Fuel oil pumphouse 

o Appendix R diesel room 

The containment internal structure was not targeted for walkdown due to 

inaccessibility. The location of these structures are identified on the plant layout 

drawings in Appendix A. 

Structure Separations. Based upon a review of the structural drawings, a number 

of separations involving Group A structures were identified. These separations are 

listed in Table 4.3-1 along with their gap widths as indicated on the design 

drawings. 

Block Walls. Summary tables providing information on the Maine Yankee block 

walls were a vail able prior to the first walkdown. These tables describe the 

locations of all block walls in the plant, identify any safety-related equipment on 

or near the block walls, and categorize the block walls in terms of their .seismic 

safety status. 

These summary tables were used to develop a list of block walls to be inspected 

during the first walkdown. This list is shown in Table 4.3~2. Nearly all block 

walls in the plant are included in this list. Certain block walls were excluded 

because they are located in areas that obviously did not house Group A 

components, based upon a comparison of block wall and equipment locations. 

These walls are located in the administration building, front office, fuel building, 

gas house, LSA storage building, RCA building, office areas of the service 

building, and parts of the yard. Walls inside containment were not targeted since 

they were known to be inaccessible for the walkdown. Block walls affecting safety 

related equipment that were not walked down are assumed to have seismic 

capacities comparable to Group A walls. 

Dams. Levees. and Dikes. Review of the drawings indicated that water for the 

fire pond is enclosed by a dike adjacent to the fire water pumphouse. Because of 

the proximity of the fire water pond to the plant, this dike was targeted for 

walkdown to determine if it could fail during a seismic event and cause flooding. 

Eguioment Components 

The Maine Yankee structural drawings and general plant layout drawings were 

reviewed to locate the preliminary equipment items identified by the system 

analyst. A general walkdown sequence organized by structure was developed for 

maximum use of time during the first walkdown. Walkdown data sheets were 

developed for specific classes of equipment components to be used in recording 

manufacturer and dimensional information necessary for a fragility evaluation 

(reference Section 4.3.3 for a discussion and example of a typical walkdown sheet). 

In most cases the equipment walkdown data sheets were lengthy as sufficient 

vendor information had not been received from Maine Yankee prior to the first 

walkdown. The first walkdown list of equipment components reviewed at the 

Maine Yankee Atomic Power Station are listed in Table 4.3-3. 

4-7 
'," 

' ~ ~ ~ 

Table 4.3·1 Structure Separations 

Structures 

Containment, containment spray 

pumphouse 

Containment, ventilation equipment 

room 

Containment, main steam valve 

house 

Containment, M.C.C. room 

Containment, aux. feed pumphouse/ 

purge air exhaust room 

Containment, fuel building 

Containment, equipment hatch 

shield 

Containment spray pumphouse, 

ventilation equipment room 

Ventilation equipment room, main 

steam valve house 

Primary auxiliary building, 

service building 

4-8 

Floor 

Elevations 

El. 14'·6, El. 30'-0, 

El. 40'·0 

El. 21 '·0, El. 40'·0 

El. 21 '·0, El. 68'-0 

EI. 21 '·0, El. 46'-0, 

El. 68'-0 

El. 9'-0, El. 22'-0, 

El. 33' .. 0 

El. 44'·6 

Elt 20'-0, El. 54'·0 

El. 21'-0, El. 40'·0 

El. 21'·0, El. 40'-0 

El. 35'·0 

Separation 

Gap 

3" 

3" 

3" 

3" 

3" 

12'' 

3" 

3" 

3" 

3" 
: I' ~ -· 

Maine Yankee 

Wall ID No. 

ARD 20 

1-2 

cs ~4 

1 

FPH 20 

1-2 

PAB 11 

1 

PAB 11 

2 

PAB 11 

3 

PAB 11 

4 

PAB 21 

1-6 

PAB 21 

7 

PAB 21 

8 

PAB 21 

9 

Table 4.3-2 List of Block Walls Targeted for Walkdown 

Building 

Appendix R 

Diesel 

Containment 

Spray 

Fire Pump 

House 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Room 

Location 

El. 21' 

Sump El. 4'-0" 

El. 20'-0" 

Door Leading to 

Filter Cubicles 

El. 11 '-0" 

El. 11 '-0" 

El. 11 '-0" 

El. 11 '-0" 

Boric Acid 

Storage Area 

El. 21 '~0" 

Deaera tor Vent 

Condenser Cubicle 

El. 21 '~0" 

Waste Evaporator 

Cubicle 

El. 21'-0" 

Sampling Cubicle 

Behind Charging 

Pumps, El. 21 '-0" 

4-9 

Wall 

Location 

Dividing Wall 

RHR HX 3A Sump 

Two walls near P-4 

East Door Jam 

Shielding Blocks Around 

Letdown Lines 

Loose Blocks Near East 

Wall of the Primary Drain 

Tk Cubicles 

Loose Blocks on East 

Side of Seal Water Cooler 

Cubicle 

Between Lines 7 and 9 

and Columns F and H 

East Wall 

East Wall 

North Wall 
Maine Yankee 

Wall ID No. 

PAB 21 

10 

PAB 36 

1-2 

PAB 36 

3 

PAB 36 

4 

SB 21 

1-3 

SB 21 

4-7 

SB 21 

8-10 

SB 21 

11-13 

SB 21 

14-15 

SB 21 

17·19 

SB 35 

1-4 

Table 4.3M2 List of Block Walls Targeted for Walkdown (Continued) 

Building 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Primary 

Auxiliary 

Service 

Service 

Service 

Service 

Service 

Service 

Service 

Room 

Location 

Curbs Charging 

Pump Cubicles 

Degasifier Vent 

Condenser Area 

(Evaporator 

Cubicle) 

El. 36' .. 0" 

Waste Gas Surge 

Drum Area 

El. 36'~0'' 

El. 36'·0" 

Control Room 

El. 21 '-0'' 

Corridor Along 

C-Line 

El. 21 '-0" 

Corridor, Along 

C-Line 

El. 21 '·0" 

Aux. Boiler 

Room, El 21 '-0" 

Elevator 

Enclosure 

El. 21 '-0" 

Control Room 

El. 21 '-0" 

Cable Tray 

Room, El. 35'-0" 

4-10 

Wall 

Location 

South End 

South and East Wall 

Around PAB Non-Nuclear 

Safety Class Charcoal 

Filter 

Removable Shield Wall 

East Wall 

Primary Vent Hi-Range 

Monitor Shield Blocks 

Control Room Entrance 

Between Column 4 and 

Elevator 

Between Columns 1 & 4 

East, West, North 

Walls 

Elevator Enclosure 

Toilet, Southt West 

and East Wall 

Battery Room 3 and 4 

Area; South, North 

and West Wall 
Maine Yankee 

Wall ID No. 

SB 35 

5-6 

SB 35 

7 

SB 35 

8 

SB 39 

1 

SB 39 

2 

SB 39 

3 

SB 45 

1-3 

SB 45 

4-5 

SB 45 

6 

SB 45 

7 

Table 4.3-2 List of Block Walls Targeted for Walkdown (Continued) 

Building 

Service 

Service 

Service 

Service 

Service 

Service 

Service· 

Service 

Service 

Service 

Room 

Location 

Elevator 

Enclosure 

El. 35'-0" 

Elevator 

Enclosure 

El. 35'-0" 

Cable Tray Area 

El. 35'-0" 

Vent and Air 

Condition 

Equipment 

El. 39'-0" 

Cable Tray Area 

El. 35'-0" 

Cable Tray Area 

El. 35'-0" 

Switchgear 

Room, El. 45'-6" 

Elevator 

Enclosure 

El. 45'-6" 

Switchgear 

Area, El. 45'-6" 

Switchgear 

Area, El. 45'-6'' 

4-11 

Wall 

Location 

Enclosure Except 

North Wall 

Elevator Enclosure East 

Wall - North Side 

East Wall 

Wall Along 7-Line 

Between Columns E and F 

South Wall Along Line 7 

Between Columns D and E 

South Wall Along Line 7 

Between Columns C and D 

Battery No. 2 and 1 

Area; South and West 

Wall and SafetyRelated SWGR Room 

Enclosure Except East 

Wall 

East Wall 

North Wall 
Maine Yankee 

WalllD No. 

SB 61 

1 

SB 77 

1 

SB 61 

2 

SB 77 

2 

TB 21 

1 

TB 21 

2 

TB 21 

3-8 

TB 21 

9-11 

TB 21 

12 

TB 21 

13 

TB 21 

14 

VE 21 

1-4 

y 28 

Table 4.3-2 · List of Block Walls Targeted for Walkdown .(Continued) 

Building 

Service 

Service 

Turbine 

Turbine 

Turbine 

Turbine 

Turbine 

Turbine 

Turbine 

Vent 

Equipment 

Area 

Yard 

Room 

Location 

E.levator 

Enclosure 

El. 61 '-0" and 

El. 77'-4" 

Elevator 

Enclosure 

El. 61'-0" and · 

El. 77'-4" 

Lube Oil Room 

El. 21 '-0" 

Corridor Along 

C-Line Columns 7 

to 9, El. 21'-0" 

Corridor Along 

C-Line Columns 1 

to 7, El. 21'-0" 

El. 21 '~O" 

El. 21 '-0" 

El. 21 '·0" 

El. 21'·0" 

El. 21 '·0" 

RWST 

4-12 

Wall 

Location 

Enclosure Except 

South Wall 

Elevator Shaft and 

Equipment Room -

South Wall 

North Wall 

Adjacent to PCC Hx 

E-4A, E-4B 

Corridor at Service/ 

Turbine Building 

Line 

Walls Separating Feedwater Pumps P-2A and 

·P-2B 

Wall Between Turbine 

Pedestals North Side 

East Wall (Inside) 

Column 9 Doorway 

East Wall (Inside) 

Column 10 Above 

Doorway 

Walls Near Entrance 

to Containment 

Spray Building 

R WST Shielding Blocks 
Table 4.3-3 Maine Yankf;e Atomic Power Plant 

First Walkdown List of Equipment Components 

EQUIPMENT ITEM 

TANKS 

1. Boric Acid Storage Tank TK-2 

2. Boric Acid Mix Tank TK~3 

3. Refueling Cavity Water Storage Tank TK-4 

4. Primary Component Cooling Surge Tank TK-S 

5. Volume Control Tank TK-6 

6. Demineralized Water Storage Tank TK-21 

7. Auxiliary Fuel Oil Supply Tank (buried) TK-28A 

8. Auxiliary Fuel Oil Supply Tank (buried) TK-28B 

9. Chemical Spray Addition Tank TK-54 

10. Secondary Component Cooling Surge Tank TK·S9 

11. Emergency Diesel Day Tank TK-62A 

12. .. Emergency Diesel Day Tank TK-62B 

13. DG-1 A Compressed Air Tank TK-76A 1 

14. DG-lA Compressed Air Tank TK-76A2 

15. DG-IA Compressed Air Tank TK·76A3 

16. Diesel Starting Air Receiver lA TK-76A-4 

17. Diesel Starting Air Receiver lA TK-76A-S 

18. Diesel Starting Air Receiver lA TK-76A-6 

19. DG-IB Compressed Air Tank TK-76Bl 

20. DG-1 B Compressed Air Tank TK-76B2 

2L DG-IB Compressed Air Tank iK-76B3 

22. Diesel Starting Air Receiver IB TK-76B-4 

23. Diesel Starting Air Receiver IB TK~76B-S 

24. Diesel Starting Air Receiver IB TK-76B-6 

25. Chemical Feed Tank TK-89 

26. Sample Tank Tk-94 

27. Chemical Additive Tank (Pipe capped off) 

28. Fuel Tank DG-2 

PUMPS 

1. 

2. 

3. 

4. 

5. 

Fire Pump (Diesel) P-5 

Boric Acid Transfer Pumps P·6A 

Boric Acid Transfer Pumps P-6B 

Boric Acid Transfer Pumps P-6C 

Auxiliary Charging Pump P-7 

BUILDING 

AND 

SYSTEM 

BAT 

BAT 

HPSI 

PCC 

CH 

AFW 

FO 

FO 

cs 

sec 

FO 

FO 

DG 

DG 

DG 

DG 

DG 

DG 

DG 

DG 

DG. 

DG 

DG 

DG 

AFW 

PCC 

sec 

DG 

ASDA 

BAT 

BAT 

BAT 

CH 

ELEVATION 

PAB + 36' 

PAB + 11' 

Yd. + 20' 

SB + 61' 

PAB + 11' 

Yd.+ 20' 

APR+ 21' 

APR + 21' 

Yd.+ 21' 

SB + 70' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AB + 21' 

AF + 20' 

TB + 21' 

PAB + 21' 

Yd + 21' 

FP + 20' 

PAB + 21' 

PAD+ 21' 

PAB + 21' 

PAB + 11' 
,, 

,·, 

Table 4.3-3 Maine Yankee Atomic Power Plant 

First Walkdown List of Equipment Components (Continued) 

EQUIPMENT ITEM 

PUMPS (Continued) 

6. Primary Component Cooling Pump P-9A 

7. Primary Component Cooling Pump P-9B 

8. Secondary Component Cooling Pump P~lOA 

9. Secondary Component Cooling Pump P-lOB 

1 0. Charging Pump P-14A 

11. Charging Pump p .. l4B 

12. Charging Pump P-14S 

13. Emergency Feed Pump P-25A 

14. Auxiliary Feed Pump P-25B 

15. Emergency Feed Pump P-25C 

16. Service Water Pump P-29A 

17. Service Water Pump P-29B 

18. Service Water Pump P-29C 

19. Service Water Pump P-29D 

20. Auxiliary Fuel Oil Transfer Pump P-33A 

21. Auxiliary Fuel Oil Transfer Pump P-33B 

22. Service Water Sampling Pump P-38 

23. Containment Spray Pump P-61 A 

24. Containment Spray Pump P-61 B 

25. Containment Spray Pump P-61 S 

26. Boric Acid Mix Tank Pump P-81 

27. SWS Mussel Control Pump P-86 

28. Chemical Feed Pump P-115 

29. Service Water Seal Pit Sample Pump P-116 

30. Fuel Pumps ( 4 pumps total on DG skid) 

31. Lubrication Oil Pumps (8 pumps on DG skid) 

32. Water Pumps (4 pumps total on DG skid) 

33. DG-2 Fuel Pump 

HEAT EXCHANGERS 

I. 

2. 

3. 

4. 

Pressurizer E-2 

Residual Heat Removal Heat Exchanger E-3A 

Residual Heat Removal Heat Exchanger E-3B 

Primary Component Cooling Heat Exchanger E-4A 

4-14 

BUILDING 

AND 

SYSTEM 

PCC 

PCC 

sec 

sec 

HPSI 

HPSI 

HPSI 

AFW 

AFW 

AFW 

SW 

sw 

sw 

sw 

FO 

FO 

sw 

cs 

cs 

cs 

BAT 

sw 

AFW 

sw 

DG 

DG 

DG 

ASDA 

PCC 

PCC 

sec 

PCC 

ELEVATION 

TB + 21' 

TB + 21' 

TB + 21' 

TB + 21' 

PAB + 21' 

PAB + 21' 

PAB + 21' 

AF + 21' 

VA + 21' 

AF + 21' 

cw + 7' 

cw + 7' 

cw + 7' 

cw + 7' 

Yd + 21' · 

Yd + 21' 

TB + 21' 

cs + 14' 

cs + 14' 

cs + 14' 

PAB + II' 

cw + 7' 

AF + 20' 

on Diesel 

on Diesel 

on Diesel 

Yd + 21' 

RC + 20' 

cs + 14' 

cs + 14' 

TB + 21' 
Table 4.3-3 Maine Yankee Atomic Power Plant 

First Walkdown List of Equipment Components (Continued) 

EQUIPMENT ITEM 

HEAT EXCHANGERS (Continued) 

5. Primary Component Cooling Heat Exchanger E-4B 

6. Secondary Component Cooling Heat Exchanger E-SA 

7. Secondary Component Cooling Heat Exchanger E·SB 

8. AC Electric Compressor Fan Cooler E-20 

9. Seal Water Heat Exchanger E-34 

10. Diesel Generator Heat Exchanger E-82A 

11. Diesel Generator Heat Exchanger E-82B 

12. Oil Coolers E-86A 

13. Oil Coolers E-86B 

14. Oil Coolers E-86C 

15. Charging Pump Seal Leakage Cooler E-92A 

16. Charging Pump Seal Leakage Cooler E-92B 

MISCELLANEOUS COMPONENTS 

1. Control Air Compressor .C-lOA 

2. Control Air Compressor C-lOB 

3. DG-1 A Starting Air Compressor C-SIA 

4. DG-1 B Starting Air Compressor C·SlB 

5. Diesel Generator DG·IA 

6. Diesel Generator DG-IB 

7. Diesel Generator DG-2 

8. Primary Ejector EJ·2A 

9. Primary Ejector EJ-2B 

10. Atmospheric Steam Dump Valve Silencer S-1 

11. Traveling Screen SR-lA 

12. Traveling Screen SR-IB 

13. Traveling Screen SR-lC 

14. Traveling Screen SR-ID 

15. Turbine for P-2SB T-1 

16. Charging Pump Seal Leakage Cooler E-92B 

4-15 

BUILDING 

AND 

SYSTEM 

PCC 

sec 

sec 

Elec 

PCC 

PCC 

sec 

AFW 

AFW 

AFW 

sec 

PCC 

ASDA 

ASDA 

DG 

DG 

DG 

DG 

ASDA 

SPC 

SPC 

SPC 

sw 

sw 

sw 

sw 

AFW 

PCC 

ELEVATION 

TB + 21' 

TB + 21' 

TB + 21' 

TB + 39' 

PAB + 11' 

AB + 21' 

AB + 21' 

AF + 20' 

VA + 20' 

AF + 20' 

PAB + 11' 

PAB + II' 

VA + 21' 

VA + 21' 

AB + 21' 

AB + 21' 

AB + 22' 

AB + 22' 

AB + 21' 

TB + 39' 

TB + 39' 

VA + 40' 

cw + 21' 

cw + 21' 

cw + 21' 

cw + 21' 

VA +21' 

PAB + 11' 
l"' •• 

., 

Table 4.3-3 Maine Yankee Atomic Power Plant 

First Walkdown List of Equipment Components (Continued) 

BUILDING 

AND 

EQUIPMENT ITEM SYSTEM ELEVATION 

ELECTRICAL DISTRIBUTION SYSTEMS 

1. 4160V Emergency Bus 5 Elec SB + 46' 

2. 4160V Emergency Bus 6 Elec SB + 46' 

3. 480V Emergency Bus 7 Elec SB + 46' 

4. 480V Emergency Bus 8 Elec SB + 46' 

5. 480V Emergency Motor Control Center MCC·7A Elec SB + 46' 

6. 480V Emergency Motor Control Center MCC-7B Elec RMC + 21' 

7. 480V Emergency Motor Control Center MCC·7Bl Elec cs + 20' 

8. 480V Emergency Motor Control Center MCC·BA Elec SB + 46' 

9. 480V Emergency Motor Control Center MCC·BB Elec RMC + 21' 

10. 480V Emergency Motor Control Center MCC-8B1 Elec cs + 20' 

11. 120V AC Vital Bus 1 . Elec SB + 21' 

12. 120V AC Vital Bus lA Elec SB + 21' 

13. 120V AC Vital Bus 2 Elec SB + 21' 

14. 120V AC Vi tal Bus 2A Elec SB + 21' 

15. 120V AC Vital Bus 3 Elec SB + 21' 

16. 120V AC Vital Bus 3A Elec SB + 21' 

17. 120V AC Vital Bus 4 Elec SB + 21' 

18. 120V AC Vital Bus 4A Elec SB + 21' 

19. 125V DC Bus 1 Elec SB + 46' 

20. 125V DC Bus 2 Elec SB + 46' 

21: 125V DC Bus 3 Elec SB + 46' 

22. 125V DC Bus 4 Elec SB + 46' 

23. Station Battery No. 1 (Lead Antimony) Elec SB + 46' 

24. Station Battery No. 2 (Lead Antimony) Elec SB + 46' 

25. Station Battery No. 3 (Lead Antimony) E1ec SB + 35' . 

26. Station Battery No. 4 (Lead Antimony) Elec SB + 35' 

27. Battery Charger No. BC·1 E1ec SB + 46' 

28. Battery Charger No. BC-2 Elec SB + 46' 

29. Battery Charger No. BC-3 Elec SB + 46' 

30. Battery Charger No. BC·4 Elec SB + 46' 

31. Inverter No. INVR-1 Elec SB + 46' 

32. Inverter No. INVR·2 Elec SB + 46' 

33. Inverter No. INVR-3 Elec SB + 46' 

34. Inverter No. INVR-4 Elec SB + 46' 

35. Station Service Transformer X-507 Elec SB + 46' 

36. Station Service Transformer X-608 Elec SB + 46' 

4-16 
: !' 

'. 

Table 4.3-3 Maine Yankee Atomic Power Plant 

First Walkdown List of Equipment Components (Continued) 

BUILDING 

AND 

EQUIPMENT ITEM SYSTEM ELEVATION 

ELECTRICAL DISTRIBUTION SYSTEMS (Continued) 

37. 480V MCC-981 (normally off MCC-9B) Elec APR+ 20' 

38. 120V Vital Bus 7 Elec AF + 20' 

39. 125V Bus 6 Elec APR+ 20' 

40. Station Battery 6 Elec APR+ 20' 

41. Battery Charger No. 6 Elec APR+ 20' 

42. Inverter No. 7 Elec APR+ 20' 

43. Alternate Shutdown Panel Elec AF + 21' 

44. 480V MCC liB (mix-tank agitator) Elec FB + 21' 

45. 480V MCC-llD (for pump P-86) Elec cw + 21' 

46. 480V MCC 9B (mix-tank pump) E1ec PAB + 21' 

47. Diesel Generator Control Board DG·IA Elec AB + 27' 

48. Diesel Generator 480V Distribution Cab 1 A Elec AB + 22' 

49. Di·esel Generator Control Board DG·lB Elec AB + 22' 

50. Diesel Generator 480V Distribution Cab IB Elec AB + 22' 

52. 120V Distribution Cabinets (Diesel Backed) Elec SB + 46' 

53. Main Con trot Board Elec SB + 21' 

HVAC 

1. DG·1A Room Exhaust fan FN-20A HV AB + 31' 

2. DG-IB Room Exhaust Fan FN·20B HV AB + 31' 

3. Protected SWGR Room Supply Fan FN-31 HV SB + 39' 

4. Protected SWGR Room Exhaust Fan FN-32 HV SB + 55' 

5. AC Electric Compressor Fan Fan FN·33 ? TB + 39' 

6. Fan FN-61 CH PAB + 21' 

VALVES 

1. Aux Feedwater Regulating Valve AFW-A-101 AOV AFW AF + 23' 

2. Aux Fecdwater Regulating Valve AFW-A-201 AOV AFW AF + 23' 

3. Aux Feedwater Regulating Valve AFW·A·301 AOV AFW AF + 23' 

4. Block Valve for AFW-A-101 AFW-A-338. AOV AFW AF + 23' 

5. Block Valve for AFW-A-201 AFW-A-339 AOV AFW AF + 23' 

6. Block Valve for AFW·A-301 AFW-A-340 AOV AFW AF + 23' 

7. Boric Acid VCT Isolation valve BA-A·32 BAT 

8. Boric Acid VCT Isolation valve BA·A·80 BAT 

9. Boric Acid Flow Control valve BA-F-30 BAT 

10. Emergency Boration Isolation valve BA·M-36 BAT 

4-17 
Table 4.3-3 Maine Yankee Atomic Power Plant 

First Wa1kdown List of Equipment Components (Continued) 

BUILDING 

AND 

EQUIPMENT ITEM SYSTEM ELEVATION 

VALVES (Continued) 

II. Emergency Boration Isolation valve BA-M-37 

12. HPSI Pump B Discharge to charging header 

CH-A-32 

13. HPSI Pump A Discharge to charging header 

CH-A-33 

14. VCT Discharge to HPSI Pumps CH-M-1 MOV 

15. VCT Discharge to HPSI Pumps CH-M-87 MOV 

16. Containment Spray Header Isolation Valve CS-M-1 

17. Containment Spray Header Isolation Valve CS-M-2 

18. Spray Chern Tank Isolation Valve CS-M-66 MOV 

19. Spray Chern Tank Isolation Valve CS-M-71 MOV 

20. CS Pump Containment Suction CS-M-91 MOV 

21. CS Pump Containment Suction CS-M-92 MOV 

22. HPSI Discharge to Loop 1 HSI·M-11 MOV 

23. HPSI Discharge to Loop I HSI-M-12 MOV 

24. HPSI Discharge to Loop 2 HSI-M-21 MOV 

25. HPSI Discharge to Loop 2 HSI-M-22 MOV 

26. HPSI Discharge to Loop 3 HSI-M-31 MOV 

27. HPSI Discharge to Loop 3 HSI-M-32 MOV 

28. HPSI Discharge to SI Header HSI·M-40 MOV 

29. HPSI Pump Discharge HSI-M-41 MOV 

30. HPSI Pump Discharge HSI-M-42 MOV 

31. HPSI Discharge to SI Header HSI-M-43 MOV 

32. HPSI Suction from R WST HSI-M·50 MOV 

33. HPSI Suction from R WST HSI-M-51 MOV 

34. CS Discharge to HPSI Pump HSI-M-54 MOV 

35. CS Discharge to HPSI Pump HSI-M-55 MOV 

36. R WST Discharge to LPSI LSI-M-40 MOV 

37. RWST Discharge to LPSI LSI-M-41 MOV 

38. Decay Heat Release Valve MS-A-162 AOV 

39. AFW Pump B turbine throttle valve MS-A-173 

40. Main Steam Stop Check Valve MS-M-10 MOV 

41. Main Steam Stop Check Valve MS-M-20 MOV 

42. Main Steam Stop Check Valve MS-M-30 MOV 

43. Decay Heat Release MS-M-161 MOV 

44. Auxiliary Steam Supply Valve MS-M-255 MOV 

45. Turbine Steam supply pressure control MS-P-168 

46. Steam Generator Safety Valve MS-S-12 

47. Steam Generator Safety Valve MS-S-13 

4-18 

BAT 

HPSI 

HPSI 

CH 

CH 

cs 

cs 

cs 

cs 

cs 

cs 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

HPSI 

cs 

cs 

cs 

cs 

ASDA 

AFW 

MS 

MS 

MS 

SPC 

MS 

AFW 

SPC 

SPC 

PT + 13' 

PT + 13' 

PAB + 21' 

PAB + 24' 

cs + 19' 

cs + 19' 

Yd + 29' 

Yd + 29' 

cs .. 08' 

cs - 08' 

PAB + 23' 

PAB + 23' 

PAB + 23' 

PAB + 23' 

PAB + 23' 

PAB + 23 

PAB + 23' 

PAB + 23' 

PAB + 23' 

PAB + 23' 

Yd + 21' 

Yd + 2P 

cs + 19' 

cs + 19' 

Yd + 28' 

Yd + 28' 

VA + 43' 

VA +21' 

VA + 49' 

VA + 49' 

VA + 49' 

VA + 43' 

VA + 43' 

VA + 21' 

VA + 39' 

VA + 39' 
Table 4.3-3 Maine Yankee Atomic .Power Plant 

First Walkdown List of Equipment Components (Continued) 

EQUIPMENT ITEM 

VALVES (Continued) 

48. Steam Generator Safety Valve MS-S-14 

49. Steam Generator Safety Valve MS-S-15 

50. Steam Generator Safety Valve MS-S-16 

51. Steam Generator Safety Valve MS-S-17 

52. Steam Generator Safety Valve MS-S-22 

53. Steam Generator Safety Valve MS-S-23 

54. Steam Generator Safety Valve MS-S-24 

55. Steam Generator Safety ·valve MS-S.;25 

56. Steam Generator Safety Valve MS-S-26 

57. Steam Generator Safety Valve MS-S-27 

58. Steam Generator Safety Valve MS-S-32 

59. Steam Generator Safety Valve MS-S-33 

60. Steam Generator Safety Valve MS-S-34 

61. Steam Generator Safety Valve MS-S-35 

62. Steam Generator Safety Valve MS-S-36 

63. Steam Generator Safety Valve MS-S-37 

64. Return from Penetration Coolers PCC-A .. 216 

65. Return from Penetration Coolers PCC .. A-238 

66. Return from RCP Coolers PCC-A-252 

67. Return from RCP Coolers PCC-A-254 

68. Return from CEA Air Coolers PCC-A-268 

69. Return from CEA Air Coolers PCC-A-270 

70. Return from Drain Cooler PCC-A-299 

71. Return from Drain Cooler PCC-A-300 

72. Return from Drain Cooler PCC-A-302 

73. Diesel 1 A Cooling Water Outlet PCC-A-493 

74. PCCW Outlet from RHR Heat Exchanger PCC-M-43 

75. PCCW Isolation to BR & LW Coolers PCC-M-90 

76. PCCW Isolation to Letdown Heat Exchangers 

PCC-M-150 

77. PCCW Isolation to Containment PCC-M-219 · 

78. Pressurizer Safety Valve PR-S-11 

79. Pressurizer Safety Valve PR-S-12 

80. Pressurizer Safety Valve PR-S-13 

81. Power-Operated Relief Valve PR-S-14 

82. Power-Operated Relief Valve PR-S-15 

83. Power-Operated Block Valve MOV PR-M-16 

84. Power-Operated Block Valve MOV PR-M-17 

85. Nonseismic Return Header Stop SCC-A-460 

4-19 

BUILDING 

AND 

SYSTEM 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

SPC 

PCC 

PCC 

PCC 

PCC 

PCC 

PCC 

? 

PCC 

PCC 

PCC 

PCC 

PCC 

PCC 

PCC 

SRV 

SRV 

SRV 

PORV 

PORV 

PORV 

PORV 

sec 

ELEVATION 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' 

VA + 39' , 

VA + 39' 

PT + 12' 

PT + 12' 

RC + 01' 

PT + 12' 

RC + 01' 

RC + 01' 

RC + 01.' 

RC + 01' 

AB + 24' 

cs + 02 

PAB+11' 

PAB + 21' 

PAB+l1' 

RC + 65' 

RC + 65' 

RC + 65' 

RC + 66' 

RC + 66' 

RC + 64' 

RC + 64' 

TB + 43' 
· Table 4.3-3 ·Maine Yankee Atomic Power Plant 

First Walkdown List of Equipment Components (Continued) 

EQUIPMENT ITEM 

VAL YES (Continued) 

86. Nonseismic Return Header Stop SCC·A-461 

87. RCP 1 Seal Water Return MOV SL-M·29 

88. RCP 2 Seal Water Return MOV SL·M·40 

89. RCP 3 Seal Water Return MOV SL .. M·51 

PIPING 

CABLE TRAY AND CONDUIT 

INSTRUMENT RACKS 

CONTROL ROOM CEILING 

Legend 

Auxiliary Feedwatcr 

BUILDING 

AND 

SYSTEM 

sec 

SL 

SL 

SL 

SYSTEM 

AFW 

ASDA 

BAT 

CH 

cs 

DG 

FO 

HPSI 

HV 

PCC 

PORV 

PPC 

sec 

SL 

SPC 

SRV 

sw 

Alternate Shutdown Decay Heat Removal 

Boric Acid Transfer 

Charging· 

Containment Spray 

Diesel Generator Starting 

Fuel Oil 

High Pressure Safety Injection 

Area Heating and Ventilation 

Primar:y Component Cooling 

Power-Operated Relief Valve 

Primary' Pressure Control 

Secondary Component Cooling 

Seal Water 

Secondary Pressure Control 

Safety Relief Valves 

Service Water 

4-20 

ELEVATION 

TB + 42' 

RC + 2' 

RC + 2' 

RC + 2' 
Table 4.3·3 Maine Yankee Atomic Power Plant 

First Walkdown List of Equipment Components. (Continued) 

EQUIPMENT ITEM 

Legend (Continued 

BUILDING 

AB 

AF 

APR 

cs 

cw 

FP 

PAB 

PT 

PV 

RC 

RMC 

SB 

TB 

VA 

YD 

Turbine Building Auxiliary Bay 

Auxiliary Feed Pumphouse 

Appendix R Diesel 

Containment Spray Pumphouse 

Circulation Water Pumphouse 

Fire Pumphouse 

Primary Auxiliary Building 

Pipe Tunnel 

Purge Air Valve Room 

Reactor Coolant 

Reactor Motor Control Center Room 

Service Building 

Turbine Building 

Steam and Feed Water Valve Area 

Yard 

4-21 

BUILDING 

AND 

SYSTEM ELEVATION 
Piping and Valving. For margin review levels up to 0.3g pga the Panel's guidelines 

recommend a sample review of accessible piping systems to verify that no problems 

exist, such as inflexibility of piping runs between adjacent buildings. A sample 

piping system to be reviewed in detail was selected by mutual agreement between 

the system analysis and the fragility analysis teams prior to the first walkdown. 

Additionally, piping as encountered during the course of the equipment component 

walkdown was reviewed, verifying no anomalies exist. 

A sample review of valves from the preliminary valve list was determined to be 

the most effective way to confirm the Panel's guidelines regarding valve capacities. 

The selection of valves to be reviewed during the walkdown was based on a study 

performed by Maine Yankee [Henries et al., March 1986]. The study qualified the 

MYC critical valves using EQE's earthquake experience data for valves. Valves 

not enveloped by experience data were targeted for walkdown review. A more 

descriptive discussion on the targeted valves occurs under Section 4.3.2.2, Walkdown 

Procedures for Valves. 

Cg.ble Trays. Cable tray systems throughout the plant were targeted for general 

survey to determine if they could be screened out generically in conformance with 

the review guidelines. This general survey was planned to be supplemented by a 

detailed inspection of a representative cable tray run, with the selection of this run 

to be made during the walkdown. 

Instrument Racks. A sample of instrument racks, although not specifically 

required for review by the Expert Panel guidelines, were targeted for walkdown 

review. The instrument racks support critical system actuation instruments and 

components. Although many of the critical instrument racks are located inside 

containment, a sample review was conducted on similar instrument racks 

encountered during the course of the equipment walkdown. 

Control Room Ceiling. The control room suspended ceiling system was targeted for 

walkdown since inspection is required by the panel's guidelines. 

4.3.2 Walkdown Procedures 

4.3.2. t Walkdown Team 

The fragility analysis team that conducted the plant walkdown consisted of 

engineers experienced in structural and equipment fragility analysis, seismic 

analysis and design of nuclear power plants, assessment of actual earthquake 

experience, and seismic margin studies, and probabilistic risk assessments. The 

analysis team was divided into two groups as follows: 

Ravindra (EQE) 

Hashimoto (EQE) 

Prassinos (LLNL) 

Hardy (EQE) 

Quilici (EI) 

Murray (LLNL) 

Swan (EQE) 

Moore (EI) 

Griffin (EQE) 

As can be observed, each group consisted of fragility analysts and system analysts. 

The close interaction between the two aspects of the review was considered 

important. During the course of the walkdown, some of the members of the two 

groups were switched to confirm the findings of the other group and to ensure that 
certain items are not missed by either of the groups. At the end of each day, the 

groups met to compare notes and to identify the areas and items to cover in the 

next day's walkdown. 

4.3.2.2 Procedures for Structures and Equipment 

Structures 

Information necessary for seismic evaluation of civil structures is normally 

obtained from the design drawings rather than a walkdown. A complete set of 

drawings for the Maine Yankee structures was available prior to the first 

walkdown. These drawings were reviewed to obtain a general understanding of 

construction and configuration of the structures and to identify any specific data 

to be obtained during the walkdown. 

Walkdown of the targeted Group A civil structures was performed to determine the 

following: 

o Verify that the structures are in general conformance with the 

design drawings. 

o Identify any gross deficiencies that would imply a reduction in 

seismic capacity. 

o Confirm that structure separations indicated ori the drawings 

were provided. 

o Obtain structural details not available from the drawings. 

The first two items above were obtained in a general manner, rather than 

performing a rigorous walkdown of all seismic load resisting structural members. 

The latter two items were specifically identified for inspection. Targeted structure 

separation gaps were determined before the walkdown and tabulated. Specific 

structural data not contained in the design drawings were obtained in the 

walkdown. For example, as-built sketches of weld and bolt details for certain 

structural steel connections were developed. 

Block Walls. In preparation for the first walkdown, the following tasks were 

performed: 

o A target list of block walls in the plant was compiled using Maine 

Yankee summary tables. 

o Detailed walkdown data sheets were prepared to facilitate the 

compilation of block wall information. 

o Locations of block walls were highlighted on mechanicat layout 

or architectural drawings to permit wall identification during the 

walk down. 

4-23 
o Information, if already available, was entered into the walkdown 

data sheets in advance of the walkdown. For example, wall 

identification numbers and locations were recorded based upon 

the Main_e Yankee summary tables. Also included were any 

a vail able sketches of seismic retrofits. 

Block walls identified on the target list were inspected to the extent possible. The 

following information was typically obtained and recorded on the walkdown data 

sheets: 

o Location 

o Dimensions 

o Boundary conditions 

o Seismic retrofits, if any 

o Other physical conditions (any cracking, gaps at boundaries, 

openings or penetrations) 

o Identification of Group A components or lifelines directly 

attached or nearby 

These data were supplemented by photographs and sketches. 

During the walkdown, it was possible to identify several walls which obviously do 

not pose hazard to Group A components. Since the purpose of the walkdown was 

only to verify that their failure will not damage Group A components, the detailed 

data listed above were not recorded for them. 

Equipment 

In general, preparation for the first walkdown began with a review of the 

available equipment data. This prereview was used to accomplish two goals: 

1. To obtain as much familiarity with the equipment component as 

possible prior to the first walkdown. 

2. Identify areas where additional details were required to assess 

component capacity and any possible low capacity items requiring 

a detailed review during the walkdown. 

Typical data reviewed prior to the first walkdown included a review of: 

o Plant general layout drawings determining equipment location. 

o Structural drawings determining equipment support and 

anchorage details. 

o If available, equipment vendor drawings or data to determine 

configuration, size, and material properties. 

Equipment walkdown data sheets were developed for each class of equipment 

(reference Section 4.3.3 for examples of walkdown data sheets used). These data 

sheets were used as checklists to verify details identified during the data review 

4-24 
and to record walkdown inspection notes and details. To expedite and make 

efficient use of the time available during the walkdown~ the data sheets were 

filled out as completely as possible prior to the first walkdown. 

The majority of the equipment components identified by the system analysts were 

reviewed during the first walkdown. Components located in highly radioactive 

areas (containment was not accessible at Maine Yankee) were not reviewed. 

Generically reviewed components included piping, valving, ducting~ cable trays, 

and instrument racks. The following provides specific procedures used for the 

review of different classes of equipment inspected during the walkdown. 

Tanks. Design drawings for the tanks and their foundations and or supports were 

available prior to the first walkdown. These drawings were reviewed to obtain a 

general understanding of the tank configurations and anchorage details. 

Walkdown procedures for the tanks included the following: 

o Verification that the overall tank configuration and anchorage 

details conform with the design drawings. 

o Review of piping and other attachments to identify any potential 

sources of damage due to seismic anchor point motion. 

o Inspect any unique features, which are not common to tanks~ but 

were identified during review of the drawings. 

An example of the latter item is the concrete enclosure surrounding the 

demineralized water storage tank (DWST). The enclosure was inspected to confirm 

the separation gap from the tank itself and to determine if the enclosure access 

room was sealed to prevent loss of tank contents to the environment in case of 

tank failure. 

Pumps. Historical performance during past earthquakes of horizontal and vertical 

pumps have shown HCLPF capacities greater than 0.3g acceleration levels 

(reference NUREG/CR-4334). The Panelts recommendations for horizontal and 

vertical pumps are that for a margin review level of 0.3g, a high HCLPF capacity 

exists. The walkdown procedures concentrated on verifying pump and motor 

anchorage, type of anchorage, foundation configuration and integrity, as well as 

any interaction potential from attached or adjacent components. This review 

aimed at a confirmation of the Panel's recommendations based on our judgment as 

well as documentation of the pump configuration and anchorage via the walkdown 

data sheets and photographs. 

Heat Exchangers. Walkdown procedures for heat exchangers concentrated on 

reviewing the supports, support saddles, anchorage details and interaction potential 

from attached or adjacent components. This is consistent with the Panel's 

recommendations for establishing he.at exchanger capacity. Heat exchanger 

internals were not reviewed as past PRAs have established their capacity to be 

greater than that of their supports or anchorage. Data sheets were used to record 

configuration and dimensional data from the walkdown for heat exchanger 

support, anchorage, and attached or adjacent component interaction potential 

4-2S 
details that were not available from the plant data reviewed prior to the 

walk down. 

Diesel Generators. The Panel's guidelines for diesel generators recommends a 

review similar to that for pumps. Past performance of diesel generators 

demonstrates HCLPF capacity levels of at least O.Sg pga. The Maine Yankee diesel 

generators were reviewed for anchorage and support integrity, noting if any 

vibration isolators were present, and the review of the peripherals for positive 

anchorage. The two major peripherals reviewed during the walkdown were the 

engine control panel and the heat exchanger, both mounted on the diesel generator 

skid. Walkdown data sheets were developed and used during the walkdown to 

record any problem areas encountered. Photographs were also used to document 

items reviewed during the walkdown. 

Electrical Distribution Eauipment. The PanePs recommended walkdown procedures 

for a 0.3g pga review earthquake for electrical distribution equipment include 

reviewing that the cabinet or enclosures and internal instruments and components 

are positively anchored. Past performance of electrical distribution equipment 

during earthquakes suggests HCLPF capacities near 0.5g, providing the equipment 

and internals, instruments, breakers, contactors, etc., are positively anchored. This 

supports the Panel's recommendation regarding the walkdown evaluation of 

electrical distribution equipment. It should be noted that relays have been 

specifically excluded from the Maine Yankee seismic margin study. 

The walkdown procedures at Maine Yankee concentrated on reviewing and 

collecting cabinet or enclosure anchorage details on the larger equipment items for 

a subsequent analytical review. The smaller (wall-mounted distribution cabinets) 

equipment items were reviewed for positive anchorage, but typically very few 

details were recorded. These smaller items were judged to have a HCLPF capacity 

of greater than 0.3g pga during the walkdown review. Internal component 

anchorage was inspected for positive attachment to the cabinet framing or cabinet 

walls for all· electrical equipment components reviewed during the walkdown. 

Additionally, any interaction problems or seismic deficiencies observed were 

recorded. Walkdown data sheets and photographs were used to record and 

document the walkdown findings. 

HV AC. The Panel's recommended walkdown procedures for a margin evaluation 

of HV AC equipment include reviewing the component for positive anchorage for a 

0.3g pga review earthquake level acceleration margin level. Additionally, if a 

component is supported from vibration isolators, then an evaluation is required\ to 

establish lateral stability. The historical performance of HV AC equipment 

supports the Panel's recommendations. HVAC equipment positively anchored, as 

well as vibration isolator supported equipment having positive lateral restraints 

installed, have performed well during past earthquakes. 

The procedures for the walkdown review of the Maine Yankee HV AC equipment 

followed the Panel's recommended guidelines. Two procedures for reviewing the 

HVAC equipment were used. 

I. For HVAC equipment found mounted on vibration isolators, a 

detailed walkdown review was performed. 
2. For HV AC equipment found positively anchored to a supporting 

structure, an engineering judgmental evaluation was performed 

during the walkdown review. 

The review for HV AC equipment mounted on vibration isolators included 

recording the dimensional data and support configuration sufficient to perform an 

analytical evaluation after the walkdown. Comprehensive data sheets were 

developed to record and document details and sketches. Photographs were also 

used to record the walkdown findings. 

The review of components in the second case included several air intake and 

exhaust dampers, and exhaust fans. The walkdown review assessed anchorage and 

any seismic deficiencies present in order to judge that the component had a 

HCLPF capacity greater than 0.3g pga. The predominant form of documentation 

for these components was the use of photographs to record the walkdown findings. 

Little details or notes were recorded on the walkdown data sheets. 

HV AC Ducting. The Panel's guidelines state that for HVAC ducting, a HCLPF 

capacity exists for acceleration levels up to 0.3g pga. The walkdown procedures 

used to confirm the panel's guidelines consisted of two approaches: 

1. Inspect the ducting in close proximity to the .HV AC equipment 

components to be reviewed. 

2. Inspect a sample ducting system selected during the walkdown. 

Inspection items included reviewing the vertical- and lateral-load-resisting members 

of the ducting system and any possible anchor point displacements that could 

impart significant loads to connecting equipment. Documentation consisted of 

noting any anomalies and taking several photographs. 

Valves. For review earthquake levels up to 0.3g pga the Panel's recommended 

walkdown procedures require a review of valves encountered during the sample 

piping system walkdown. Areas of concern to be reviewed during the walkdown 

include observing for interaction potential between the valve operator and adjacent 

structure or component and reviewing possible anchor point displacements between 

piping and valve. Historical performance of valves during earthquakes at this 

acceleration level supports the Panel's walkdown review recommendations. 

In addition to reviewing valves during the piping walkdown, a sample list of 

valves targeted for review was developed. The method used to select a valve for 

review was determined after a prior review of a Maine Yankee (MYC) document 

[Henries et al.t March 1986], which evaluated the seismically critical valves using 

seismic experience data. The document recorded items such as valve description 

and function, height above grade, cast iron body or yoke, pipe diameter, type of 

operator, actuator weight, distance between the pipe center line and the top of the 

operator, and the evaluation conclusions. The margin study valve list was 

reviewed against the MYC document. The selected valves included those valves not 

previously shown to be within the bounds of the experience data or were not 

included in the document list. 

4-27 
The review of the targeted valves was conducted during the course of two 

walkdowns. The review procedures consisted of inspecting each valve on the list 

accessible to the walkdown teams. Seven valves targeted for review were located 

in containment and could not be reviewed during the walkdown. For these valves, 

vendor literature and any previous photographs taken by Maine Yankee were 

requested for review to confirm the Panel's recommendations. Walkdown 

procedures consisted predominantly of inspecting the valve and area around it, 

noting any potential interaction or anchor point displacement problems. 

Documentation consisted of taking several photographs and in some instances 

recording specific valve data on walkdown data sheets. This was limited to valves 

not included in the Maine Yankee document, because those valves listed had this 

information tabulated. 

Piping. Past seismic PRA studies and earthquake experience data have shown that 

welded steel piping systems have a very high resistance to seismic loads. 

NUREG/CR·4334 contains the Expert Panel recommendation that "piping systems 

in nuclear power plants have HCLPF capacities greater than 0.5g pga." The panel 

recommends that for a 0.3g pga review level earthquake, a walkdown of a sample 

piping system should be conducted and that piping between buildings should be 

inspected. 

Specific areas associated with the piping which were reviewed during the plant 

walkdown for Maine Yankee were: 

o Walkdown of the auxiliary feedwater system (AFW) 

o Assessment of piping which spans between two buildings 

Underground piping systems could obviously not be addressed on the plant 

walkdown, and were evaluated based on their design drawings. Two other piping 

failure modes that were addressed during the walkdown included the impacting 

failures of valve operators and the damage of piping caused by the failure of 

anchorage of attached equipment. The valve clearance issue and the equipment 

anchorage issue are addressed as a part of the margin evaluation for the specific 

equipment component and not as a part of the piping margin review. 

The sample system to be walked down for the seismic margins review was the AFW 

system. The AFW system was chosen based on its importance to the safety of the 

plant and because of the variety of piping sizes, supports, and components 

(branches, elbows, reducers, tee connections, etc.) in the system. The procedure for 

walking down the piping system included following the piping layout drawings to 

verify support locations, assessing system interaction potential to the piping, and 

taking detailed configuration information for piping details that are judged to be 

of paten tial concern. 

Cable Trays. In the first walkdown, inspection of the cable trays was performed at 

two levels: 

o General survey of cable tray systems in the plant 

o Detailed inspection of a representative cable tray run 

4·28 
The general survey was performed to obtain an overview of cable tray construction 

throughout the plant. This included a review of the variety of cable tray system 

layouts, support configurations, and construction details. The inspection also 

considered items identified hi' the review guidelines as being of potential concern, 

including failure of taut cables due to large relative displacement, severing of 

cables caused by sharp edges at the ends of cable trays, and weld failure. 

As a part of the walkdown of cable tray systems, a representative run was 

inspected in detail to obtain specific information on the construction. This run, 

located in the cable spreading room, was selected since it exhibited many of the 

features common to cable trays throughout the plant. Information was documented 

on walkdown data sheets. Information collected included run location and layout 

supplemented with sketches, support configuration details and spacings, cable tray 

configuration and loading, tray and support connection details, interfaces with the 

structure and other components, and any potential problems. 

Instrument Racks. Instrument racks were not specifically addressed by the Panel 

in establishing its margin review guidelines. Historical performance during 

earthquakes of a variety of instrument rack configurations documented in the 

experience data base suggests that a HCLPF capacity of greater than 0.3g pga 

exists. Walkdown procedures of the Maine Yankee instrument racks consisted of 

reviewing a sample of racks encountered during the c.ourse of the equipment 

walkdown. The reviewed racks were inspected for positive anchorage and 

similarity to documented data base racks. Attached instruments and components 

were also inspected for positive anchorage to the rack. Walkdown documentation 

consisted of taking several photographs. 

Several of the most critical instrument racks at Maine Yankee are located inside 

the containment building and could not be reviewed during either the first or the 

second walkdown. Yankee Atomic engineers have photographic records of these 

instrument rack installations which were taken on previous plant outages. These 

photographs were used as a basis for evaluating the seismic margin inherent in the 

instrument racks located in containment and the components (transmitters, 

transducers, sensors, etc.) supported on these racks. The impulse lines and 

electrical leads which enter and exit these instrument racks were assessed by the 

systems analysts on the basis of multiple trains being separated from one another. 

Control Room Ceiling, Sketches provided prior to the first walkdown illustrated 

design modifications incorporated in the suspended ceiling system over the control 

room. These modifications consisted of safety wiring the T·bar ceiling and light 

fixtures to the concrete slab overhead. The ceiling system was inspected during 

the ·walkdown to verify that the safety wiring was installed consistent with the 

sketches and appeared adequate. In accordance with the PanePs guidelines, other 

fixtures above the control room were inspected to identify any other potential 

hazards to personnel and equipment below. 

4.3.3 Walkdown Documentation 

Walkdown documentation for equipment and structures consisted of recording the 

findings using walkdown data sheets and photographs. The walkdown data sheets 

were developed for each particular class of component indicating specific 

4-29 
information required to confirm and verify the Panel's recommendations as well as 

to record details sufficient to perform a seismic fragility evaluation if necessary. 

Typical examples of equipment data sheets used during .the Maine Yankee 

walkdowns for equipment are presented in Tables 4.3-4; 4.3-5, and 4.3-6 for pumps, 

HV AC components and block walls, respectively. The d~ta sheets reflect the 

varying levels of information required between different classes of equipment 

depending on their seismic ruggedness (e.g. pumps require little review other than 

to verify anchorage and interaction potential whereas HVAC components supported 

by vibration isolators require a detailed review recording greater degrees of 

information necessary for a fragility evaluation). 

Photographs were also used to record details of the walkdown. Photographs 

provide a permanent record of what was reviewed and support any notes or details 

taken during the walkdown. System interaction concerns are typically documented 

with photographs. Additionally, photographs are used in the fragility evaluation 

to confirm details taken or to provide additional clarification. Photographs are a 

valuable part of the complete walkdown documentation. 

4.3.4 Walkdown Results 

4.3.4.1 Walkdown Findings 

Structures 

The original seismic design criteria for the Maine Yankee civil structures are 

described in Section 3.2. With the exception of the 10 CFR 50, Appendix R diesel 

room, and the turbine/service building, all of the Group A structures listed in 

Section 4.3.1 were categorized as Class 1 structures by the original design basis. 

However, the following areas within the turbine/service building were considered 

Class 1: control room, cable room, switchgear room, service building area housing 

the control room air conditioning, breathing air, and switchgear room ventilating 

equipment, diesel generator enclosure, and turbine building portion housing the 

component cooling heat exchangers, pumps and air compressor receivers. The 

Appendix R diesel system was subsequently deleted from the Group A systems by 

the systems analyst. 

Seismic load-resisting systems for the Group A civil structures are composed of 

reinforced concrete and/or structural steel. HCLPF capacities were not established 

by the Expert Panel for steel frame structures. The following steel frame 

structures are therefore screened in and require a seismic capacity evaluation: 

o Circulating water pumphouse, portion above El. 21'-0" 

o Turbine/service building, steel framed portions 

o Main steam valve house, interior steel structure 

During the first walkdown, one of the diagonal braces for the mainsteam valve 

house steel structure was found to be missing. This was considered in the 

evaluation of the structure HCLPF capacity. 

The Expert Panel concluded that concrete containments, concrete shear walls, 

diaphragms, and footings, special nonductile details, and impact between buildings 

4-30 
TABLE 4.3-4 

EXAMPLE PUMP WALKDOWN DATA SHEET 

ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS 1/ 

SHEET NO. ---

8227-09 LLNL Seismic Margin Review 

JOSNO ___ JOB------------------ SV .. ___ DATE---

LLNL Plant Wa 1 kdown Data Sheet CHK'D DAJE CLIENT SUBJECT ---

FACILITY: Maine Yankee Atomic Power Station 

COMPONENT: -----------------------------

LOCATION: Building = ~~~~~~~~~~~~~~~~~~­

Elevation = -------------------

1.0 C0~1PONENT DATA: 

Manufacturer Plant ID Number=------------------------ = ________ ,;....__.....;.... ______________ _ 

Model 

Function = ----..----------------------

Photograph (overall} Roll No. ____ Frame No.s -----

2.0 AREAS REQUIRING DETAILED REVIEW: 

2. 1 Anchorage: Number and size of anchor bo 1 ts = ------------ Type of anchor bolts ... ------------ Description of foundation = -------------

Photograph Roll No. ___ Frame No.s -----

Note: Provide a sketch of anchorage plan with dimensions and indicate 

any foundation deficiencies observed in space provided below. 

4-31 
"tABLE 4·.3-4 (CONTINUED) 

EXAMPLE PUMP ~ALKDOWN DATA SHEET 

ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS 2/ 

SHEETN0---

8227-09 LLNL Seismic Margin Review 

JOB NO. JOB------------------ BY------- DATE---

CLIENT---- llNl SUBJECT ______ Plant Wa 1 kdown .............,......._.,..::_.,...._ Data Sheet _ _......._____ c o HK" D"'JE ,..... ---

3.0 ADDITIONAL COMMENTS OR OBSERVATIONS: 

Note any system interactions. 
TABLE 4.3-5 

EXAMPLE HVAC COMPONENT WALKDOWN DATA SHEET 

ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS SHEET NO. _l_/ __ 

JOBNO ~8227 .. 09 JOB LLNL Seismic Mar9in· Review BY·--- DATE __ _ 

CLIENT LLNL SUBJECT Plant Wal kdown Data Sbeet CHK'D D .. ~TE ---

FACILITY: Majne Yankee Atomic Power Station 

COMPONENT: ---------------------------

LOCATION: Building Elevation • • ~~~~~~~~~~~~~~~~~~~ ___ ...,.._. _____ ,_...., ______ _,..... __ _ 

1.0 COMPONENT DATA: 

Plant 10 Number ~ ~~~-~~~~~~~~~~~~~~~~~~~-

Manufacturer • --------~---------------

Model Function • = ----~----------------------------------~----- _______ --.ot!" _____ ____,_~---------

Photograph (over~ll) Roll No.---- Frame No.s -----

2.0 AREAS REQUIRING DETAILED REVIEW: 

2.1 Housing: Overall dimensions = --....---------- Anchorage: Type of anchorage (vibration isolators?) • -------- Lateral restraints on isolators • ------------ Number and size of anchor bolt$ • ------------ Type of ancha.r bolts • ------------ Description of foundation/supt. 1\'1 ------------

Photograph Roll No.·--- Frame No.s -----

Note: Provide a sketch of anchorage plan with dimensions and indicate 

any foundation/support deficien~i~s observed in space provided 

below. If vibration isolat~d sketch lateral restraint. 

4-33 
TABLE 4.3-5 (CONTINUED) 

EXAMPLE HVAC COMPONENT WAL~DOWN DATA SHEET 

ENGINEERING. PlANNING AND MANAGEMENT CONSULTANTS 

SHEET NO. _2_/ --

) JOB NO. _8227-09 JOB __ ....;l::.:l.:.;.:NL=-=S;..:.e..:..:is:::.:.:m~i.::.c....:.M~a!..!..r!Lg1!..!..n!..-.!!Re~vwi.::.ew.:.:._ _____ BV ___ DATE _ _...,..._ 

CLIENT I I Nl suB~CT Plant Walkdown Data S~eet CHK'D DATE---

3.0 ADDITIONAL COMMENTS OR OBSERVATIONS: 

Note any system ; n,teract; ons. 

~: . 

4-34 
TABLE 4.3-6 

EXAMPLE BLOCK WALL DATA SHEET 

ENGINEERING. PLANNING AND MANAGEMENT CONSULTANTS SHEETNQ __ _ 

BY·--- DATE __ _ 

CHKD __ DATE __ _ 

BLOCK WALL DATA SHEET 

Wall 10 Number: 

Building: 

Floor: 

Location: 

Ref. Drawing Number: 

Film Roll Number: Frame Numbers: 

Wall Dimensions:· W X H X T 

Lateral Supports: 

Any Visible Cracking? 

Any Gaps At Boundaries? 

Any Openings and/or Penetrations? 

Group A Components Directly Attached: 

Group A Components Nearby: 

Supported or Adjacent Lifelines: 

Group A Components or Lifelines Likely to Be Damaged By Wall Failure: 

4-35 
JOB NO ---"'-"~---=-

TABLE 4.3-6 (CONTINUED) 

EXAMPLE BLOCK WALL DATA SHEET 

MANAGEMEN1 CONSULTANTS 

SHEET NO.---

---"'~~-~occ__-=-~~---------- BY---- DATE __ '--

cun\IT U. ... IJ{.... 3UBJECT-...~.C>...~:.:JJ~~~~.........:::~...s!!:_--------- CHK'O __ DATE __ _ 

BLOCK WALL DATA SHEET 

Wall ID Number: 

PLAN 

ELEVATION 

4-36 
should have HCLPF capacities greater than 0.3g. The Group A concrete structures 

were reviewed to confirm that this conclusion is appropriate for Maine Yankee. 

The containment structure consists of. a base mat, a cylindrical shell, and a 

hemispherical dome. The wall is reinforced in a two-way pattern. The 

containment structure is comparable to other reinforced concrete containments 

evaluated in past fragility evaluations. In addition, HCLPF capacities have 

already been developed in [Hashimoto et al., 1984] using approaches essentially the 

same as the fragility analysis and CDFM methods. Even with the concurrent 

effects of the design incident internal pressure, the HCLPF capacity was found to 

be equal to or greater than l.Og. 

The other reinforced concrete structures are typically composed of integral walls 

and slabs. A review of the design drawings was performed to verify that they are 

adequate to resist the review level earthquake. This review considered the ability 

of the shear walls to transmit the overall seismic loads into the foundation, the 

availability of local load paths to deliver inertial loads to the walls, and the 

presence of any nonductile detailing. The concrete structures were found to be 

comparable to other structures analyzed in past fragility evaluations. No inherent 

weaknesses in the seismic load-resisting capabilities were noted. The Group A 

concrete structures were concluded to have HCLPF capacities greater than 0.3g and 

were therefore screened out. 

A review of the structural drawings indicates that the buildings are typically 

separated by gaps of three inches or more. Readily accessible building separations 

were inspected during the first walkdown. The walkdown confirmed that these 

separations are present. Three-inch separation is greater than gaps provided for 

most nuclear plant structures. Since the Maine Yankee structures are typically of 

shear wall construction and founded on rock, lateral seismic displacements will be 

small. It is concluded that impact between buildings is very unlikely at the review 

earthquake level of 0.3g. 

Block Walls. The Maine Yankee block walls are typically unreinforced. A seismic 

evaluation of the block walls was conducted in response to I & E Bulletin 80-11. 

As a result of this evaluation, seismic retrofits were installed to increase the 

resistance of certain block walls supporting or adjacent to safety-related 

equipment. 

The review guidelines specify that a margin evaluation is required for these types 

of block walls. However, an evaluation should not be necessary for all walls in the 

plant. Walls that can be screened out include those that are not supporting or 

adjacent to Group A components, and those that can collapse but not cause damage 

to adjacent Group A components. 

To identify the subset of block walls requtnng a HCLPF calculation, the block 

wall evaluation procedure in Figure 4.3-1 was developed. Prior to the walkdown, 

block walls in the plant were identified and located. Based upon the walkdown, 

the walls were screened in or out depending on their potential hazard to Group A 

components. All walls supporting attached Group A components were screened in. 

For a wall adjacent to (i.e., within about one wall height) Group A components, an 

assessment was made of the likelihood of damage to the components should the 

4-37 
I 

r 

WALLS SUPPORTING I 

GROUP A COMPONENTS 

I SCREENED IN I 

REVIEW AVAILABLE INFORMATIONI 

I 

IDENTIFY ALL BLOCK WALLS USING 

MAINE YANKEE SUMMARY TABLE 

I 

LOCATE BLOCKWALLS ON EQUIPMENT 

LAYOUT DRAWINGS 

I 

WALLS WHOSE 

COLLAPSE 

COULD RESULT 

IN IMPACT 

ONTO GROUP 

A COMPONENTS 

WILL IMPACT 

CAUSE DAMAGE 

TO GROUP A 

COMPONENTS? 

I 

l WALKDOWN I 

l 

I 

WALLS SUPPORTING 

LIFELINES {I.E., 

CABLE TRAYS, 

.pIPING, ETC • ) 

ARE THESE 

LIFELINES 

PART OF GROUP 

A SYSTEMS? 

BY SYSTEMS 

ANALYST 

y r 

I 

WALLS WHOSE 

COLLAPSE 

COULD RESULT 

IN IMPACT 

ONTO LIFELINES 

(I.E., CABLE 

TRAYS, PIPING, 

ETC.) 

ARE THESE 

LIFELINES 

PART OF GROUP 

A SYSTEMS? 

BY SYSTEMS 

ANALYST 

. I 

OTHER 

WALLS 

I 

SCREENED 

OUT 

y I N .....--"'---.... I SCREENED IN I ....--..._-...,N 

ISCREENED OUTI WILL IMPACT 

CAUSE DAMAGE 

TO GROUP A 

LIFELINES? 

I SCREENED OUT I 

Y I I N 

SCREENED INI SCREENED OUT 

Y I I N 

ISCREENED IN SCREENED OUTI 

I 

I 

PERFORM FRAGILITY ANALYSIS 

TO CALCULATE HCLPF 

Figure 4.3-1 Block Wall Evaluation Procedure 

4-38 
wall collapse. Examples of walls screened out at this point include walls shielded 

from adjacent components by built·up steel framing and walls adjacent to 

components judged to be capable of withstanding the impact without damage. 

Block walls affecting lifelines such as piping and cable trays were screened in the 

same manner. If necessary, confirmation of the Group A status of these lifelines 

was provided by the systems analyst. All other block walls not supporting or 

adjacent to Group A components were screened out. Block walls affecting safety 

related equipment that were not walked down are assumed to have seismic 

capacities comparable to the Group A walls. Review of the block walls inside of 

containment was conducted on the basis of architectural and design modification 

drawings. 

Dams. Dikes. and Levees. Walkdown of the fire water pond indicated that grading 

around the pond would cause the water to flow away from the plant even if the 

surrounding dike should fail during an earthquake. The dike was therefore 

screened out since its failure would not have any impact on plant safety systems. 

Equipment 

Tanks. Except for the buried diesel fuel oil storage tanks, TK-28A and B, all the 

tanks identified on the Maine Yankee margin review equipment list required a 

walkdown review and capacity evaluation. The tanks reviewed during the 

walkdown were all found to be anchored (i.e., no unanchored tanks were observed). 

The majority of tanks observed were consistent with the design drawings and 

details reviewed prior to the walkdown. Few anomalies affecting tank capacity 

were found during the course of the walkdown. The results of specific tank 

findings are discussed in greater detail below. 

The auxiliary fuel oil supply tanks, TK-28A and B, were reviewed during the 

walkdown. The fuel oil supply tanks are buried with only the upper portion of the 

tank visible; consequently, only the visible connecting piping was reviewed for 

possible failure caused by large relative displacements of the ground surrounding 

the buried ·tanks. The tank fill lines were the only observable piping. The fill 

lines are attached to the top of the tank; consequently, if a piping failure occurred 

a loss of diesel oil would not result. 

The emergency diesel day tanks, TK-62A and B, were reviewed in detail during the 

walkdown. The review found the tank elevated and welded to a braced steel 

frame. The support frame was bolted to the concrete floor with two expansion 

anchor bolts per column. Under each column was observed to be a shim plate and 

approximately one inch of grout. This installation detail varied between the four 

columns due to the sloping floor with the distance between the top of concrete and 

bottom of column base plate varying from 1.75 in to 2.5 in. This observation 

brought up the concern of how much of the anchor bolt was embedded in the 

concrete floor. To verify bolt embedment, Maine Yankee plant personnel 

ultrasonic tested (UT) each of the anchor bolts. The UT inspection confirmed that 

the bolt embedment did not extend into the concrete floor. Maine Yankee 

subsequently modified the day tank anchorage by installing new anchor bolts 

which are fully embedded into the concrete floor slab (Section 4.4 describes the 

modification). 

4-39 
.•.: .. 

Pumps. The critical pumps identified on the margin equipment list were reviewed 

during the walkdown. For a p.3g pga review level earthquake the Panel requires a 

review of the pump anchorage and any potential interaction problems. Horizontal 

pumps at Maine Yankee were determined to be well anchored on the basis of the 

first walkdown. Piping anchor point ·displacement and interaction potential 

concerns were reviewed for each pump with no areas of concern observed. Maine 

Yankee horizontal pumps were subsequently screened out with a HCLPF capacity 

judged to be greater than 0.3g pga. 

The Expert Panel guidelines recommend a capacity evaluation for vertical pumps 

if the shaft length is cantilevered greater than 20 ft. There were three sets of 

vertical pumps identified on the Maine Yankee margin review equipment list: 

o Service Water Pumps 

o Containment Spray Pumps 

o Auxiliary Fuel Oil Transfer Pumps 

The Maine Yankee. service water pumps were identified from the initial drawing 

and vendor data review as vertical pumps with long cantilevered shafts greater 

than 20 ft in length. The service water pumps shaft base support could not be 

reviewed during the plant walkdown as required by the Panel's recommendations; 

however, the vendor drawings identified the pumps to have 26 ft shaft lengths 

with pin supports at the base. Consequently, with respect to this issue, the service 

water pumps were screened out per the Panel's guidelines. 

The two other sets of vertical pumps identified on the equipment list for review 

were observed to have shaft lengths less than 20 ft in length. These pumps were 

subsequently screened out per the Panel's guidelines. 

Two vertical pumps (service water and containment spray pumps) were observed 

during the walkdown to have a possible weak connection between the motor and 

pump interface due to a small number of bolts and small bolt circle dimension. 

Dimensional data and motor name plate data was recorded in order perform a 

capacity calculation for these pumps. 

Heat Exchangers. All heat exchangers identified on the margin review equipment 

list were reviewed during the walkdowns except for the reactor containment air 

recirculation coolers, the reactor coolant regenerative heat exchanger, the seal 

water heat exchanger and seal water heater due to inaccessibility (high radiation 

areas). · Per the Panel's guidelines the exchangers were reviewed for support and 

anchorage integrity as well as any potential interaction or seismic anchor point 

displacement problems. The supports and anchorage require an evaluation to show 

capacity greater then 0.3g pga. The following briefly describes the types of heat 

exchangers reviewed at Maine Yankee and the walkdown findings. 

The heat exchangers reviewed at Maine Yankee can be classified into four 

categories. 

1. Vertically oriented heat exchangers. 

4-40 
.. ', . ~' . 

2. Horizontally oriented heat exchangers supported from a concrete 

pier or steel support frame. 

3. Small heat exchangers mounted directly to a larger supporting 

component. 

4. Heat exchangers not accessible during the walkdown due to high 

local radiation levels. 

The only vertical heat exchangers reviewed were the residual heat removal (RHR) 

heat exchangers which were observed to be supported from anchor lugs at 

approximately the midpoint of the exchanger shell. Additionally, bottom lateral 

supports in both horizontal directions were observed. An analytical evaluation of 

the supports and anchorage was performed. 

The standard horizontal heat exchangers were all observed to have an adequate 

support system (concrete pier or braced steel frame). All were observed to be well 

anchored with no anomalies noticed. An analytical evaluation of the supports and 

anchorage was performed. 

The small skid-mounted heat exchangers included as a peripheral component of the 

component to which it was attached. These heat exchangers were observed to 

typically provide a bearing cooling function, and varied from small diameter 

(approximately 5 to 6 inches) cylindrical units to just helical coils attached to the 

side of the pump casing. Capacity for these exchangers was judged to be greater 

than 0.3g pga during the walkdown. 

The heat exchangers not accessible for the walkdown were evaluated via a drawing 

and vendor data review with the support from photographs previously taken by 

Maine Yankee engineers. The reactor containment air recirculation coolers were 

the only component that fell into this category requiring an evaluation. The 

reactor coolant regenerative heat exchanger, the seal water heat exchanger, and 

seal water heater did not require a review, because valves PCC-M-90 and PCC-M219 have a HCLPF capacity greater than 0.3g pga, which will isolate the PCC 

system if these components fail. 

Diesel Generators. The walkdown findings verified the Panel's recommendations 

that a high capacity exists for diesel generators at the 0.3g pga review level 

earthquake . The diesel generator skid assembly rests on a large grout pad which 

is used for leveling purposes. The anchors for the skid consist of large "J" bolts 

embedded well into the concrete foundation. These 1 1/4-inch-diameter "J" bolts 

are judged to have a very high seismic capacity and the effect of these anchors 

passing through the grout pad is felt to be minimal. The diesel generator 

peripheral components which were mounted on the skid were reviewed for 

structural integrity and observed to be well anchored. No potential interaction 

problems were observed from surrounding components. The diesel generators and 

peripheral components were subsequently screened out. 

Electrical Distribution Equipment. In general the electrical equipment reviewed 

for the margins program were observed consistent with the Panel's guidelines 

stating that active electrical equipment can survive ground accelerations up to 0.5g 

4-41 
pga, provided that the cabinets and instruments are anchored. The equipment 

reviewed at Maine Yankee was well anchored except for one anomaly that 

occurred with the station service transformers which is described below. Most of 

the critical electrical equipment at Maine Yankee had seismic anchorage upgrades 

installed during one of several previous seismic evaluations conducted by the 

Utility. Upgraded anchorage and supports were evident on the switchgear, motor 

control centers, inverters, distribution cabinets, main control board, and battery 

chargers. The instruments and internal components of most electrical cabinetry 

were all observed to be well anchored. However, one noncritical component in the 

main control board was observed unanchored. This component has been 

subsequently anchored by Maine Yankee (component reviewed during the second 

walkdown was observed positively anchored). The following discusses several 

walkdown findings with regard to particular types of electrical equipment 

reviewed at Maine Yankee. 

The station battery cells at Maine Yankee were observed to be of lead-antimony 

flat-plate construction. Data on the performance of lead-antimony batteries during 

earthquakes is lacking at the present time. However, extreme corrosion degraded 

electrical performance has been recorded for aged cells such that a HCLPF 

capacity cannot be confidently established. The Expert Panel (NUREG/CR-4334) 

suggests that battery cells are not vulnerable at a 0.3g earthquake and can be 

screened out. EQE feels that this recommendation has not been proven to be 

reasonable for all emergency batteries in nuclear power plants. Maine Yankee 

batteries proved to be a case where further examination/review was necessary. 

Section 4.4, Maine Yankee Component Modifications, discusses this finding fur.ther. 

The inverters were one of the components which had an upgraded anchorage from 

its originally installed condition. The anchorage modifications were somewhat 

unorthodox due to adjacent equipment constraining the placement of the anchorage 

addition. Details were recorded and an anchorage capacity evaluation was 

performed. Additionally, the back shear panel of the cabinet had ventilated 

cutouts which raised some concern about the lateral load resistance of the cabinet 

frame. After the front doors were opened it was observed that the transformer 

was bolted to the bottom base framing with few component attachments occurring 

above the b1;1.se framing level. 

The station service transformers enclosure framing was observed to be well 

anchored and also had an anchorage upgrade installed. However, when the 

internal core/coil assembly was reviewed it was found to be unanchored. T~e 

core/coil assembly was observed to be supported by rubber isolators in four pla.'c-es 

with no vertical uplift or lateral restraints installed. The transformer internal 

core/coil assembly has been subsequently identified as requiring additional 

anchorage to be installed during the next refueling outage (March 1987). Section 

4.4, Maine Yankee Component Modifications, describes the modifications which are 

scheduled to be installed. 

Numerous wall-mounted panels were reviewed during the course of the walkdown. 

All of these panels were observed to be light in weight and well anchored with a 

minimum of four bolts per panel, one at each corner. The HCLPF capacities for 

the wall mounted panels were determined to be greater than 0.3g pga review level 

4-42 
earthquake using the methods described under Section 5.2, Simplified Analysis and 

Use of Screening Tools. 

HV AC Comnonents. The walkdown review consisted of inspecting several fan 

units, air conditioner units and dampers. Except for the diesel generator exhaust 

fans, all the fan units and air conditioner units were observed to be supported 

from vibration isolators; consequently, per the Panel's guidelines the fans and air 

conditioners required an evaluation to establish lateral stability. An initial 

evaluation determined the lateral load capacity to be deficient for both items and 

Maine Yankee subsequently scheduled both items to be upgraded during the March 

1987 refueling outage. Section 4.4 Maine Yankee Component Modifications 

describes these modifications. 

The diesel generator exhaust fans and the diesel generator air intake and exhaust 

dampers were observed to be well anchored into the building walls. A capacity 

evaluation based on the walkdown inspection judged the fans and dampers to have 

a HCLPF capacity greater than 0.3g pga, thus were subsequently screened out. 

Overhead HVAC ducting in the Maine Yankee plant was observed to be well 

supported with lateral bracing in both horizontal directions. The vertical and 

lateral supports were found to be threaded rods anchored into threaded inserts in 

the concrete ceiling and adjacent walls. The ducting reviewed at each component 

was observed to have flexible joints at the interface connections. Consequently, 

lateral movement of the ducting will not impart significant loads to the connected 

component. Maine Yankee ducting was judged to have a HCLPF capacity greater 

than 0.3g pga, thus was subsequently screened out. 

Valv;s. The sample list of valves reviewed during the walkdown were all observed 

to be seismically adequate such that a high confidence exits that the valves 

addressed in the Maine Yankee margin study will survive a 0.3g pga margin 

earthquake. These findings agree with the PanePs guidelines on valve performance 

in a 0.3g pga margin earthquake. The walkdown findings of the sample list of 

valves is discussed below for specific types of valves reviewed. 

The power operated relief valves (PORV's) could not be reviewed during the 

walkdown due to inaccessibility (high local radiation). This included several other 

valves as well (Reference Table 7.1-3 in Chapter 7 for the capacity evaluation of 

the valves). However, vendor data and photographs were provided by Maine 

Yankee to aid in EQE's review of the valves. The vendor data illustrated the 

construction and operation of the POR V's. The function is essentially a solenoid 

operator cantilevered from the valve body, not unlike a standard MOV, actuating 

the valve upon the presence of an operation signal. The Maine Yankee POR V is 

quite similar to the PORV's located at the El Centro Steam Plant which was 

subjected to the 1979 Imperial Valley earthquake. Figures 4.3-2 and 4.3-3 show a 

comparison of the Maine Yankee PORV's and those from the El Centro Steam 

Plant. 

There were several valves reviewed which had long "reach rods 11 between the 

operator and the valve body. The configuration of these valves is such that 

·operators can operate the valves from a safe distance and not be exposed to 

dangerous levels of radiation. The valve operator was well anchored to a pedestal 

4-43 
Solenoid Enclosure Box 

Figure 4.3-2 Example of a PORV at the El Centro Steam Plant. 

4-44 
Figure 4.3-3 Maine Yankee PORV PR-S-14 and PR-S-15. 

4-45 
while the operator rod was connected to the operator with a universal joint. The 

operator rod was then routed through the floor supported at regular spacings 

before connecting to the valve body. At each support point and at the valve body 

was a universal joint. From the walkdown review it was judged that these valves 

have a HCLPF capacity greater than the 0.3g pga review level earthquake. 

Earthquake experience data supports our judgment on these "reach rod" valves. 

Valves similar to the long "reach rod" valves were found at several Coalinga area 

sites subjected to the 1983 earthquake. Figures 4.3-4 and 4.3-5 show a comparison 

of the Maine Yankee valves and a sample of those from the Coalinga sites. 

Several Maine Yankee valves fell out of the bounds of the earthquake experience 

data based on conservative estimates of operator weight and operator height. 

These valves were reviewed and found to be reasonably close to the experience 

data bounds. Seismic anchor point displacements and interaction potential were 

assessed for all valves addressed on the walkdown. No problems were observed and 

the valves subsequently judged to have a HCLPF capacity of greater than 0.3g pga. 

Table 7w3 in Chapter 7 identifies the Maine Yankee valves and also indicates 

which valves were judged to have a HCLPF capacity greater than 0.3g pga 

compared to those where experience data supported the evaluation. 

Piping. The auxiliary feed water system (AFW) was the sample p1p1ng system 

selected for a detailed walkdown review. The AFW extends from the Containment 

penetration to the demineralized water storage tank. 

Three Maine Yankee drawings were used (verification of support· types, support 

spacing, unsupported spans, etc.) during the walkdown to aid in the walkdown: 

o Drawing 12365.10-MKS-103RI-4 shows the piping from the 

containment penetration to anchor H-102 

o Drawing 12365.10-MKS-103Nl-4 shows the piping from anchor H102 to the discharge side of the emergency feedwater pump (P25A) 

o Drawing 12365.10-MKS-103Nl-4 shows the p1p1ng from the 

suction side of Pump 25A to the point where the piping goes 

underground to the DWST. 

The. walkdown of the AFW piping produced nothing which altered the Panel's 

assessment of a high confidence exits that the piping will survive a 0.3g pga review 

level earthquake. The piping was observed to be of all welded steel construction 

with standard fittings (tees, elbows, and branch connections). The supports were 

typically welded steel frames constructed of angles or boxbeams and attached to 

the walls/ceiling/ floors by welding to embedded or structural steel, and on 

occasion with expansion anchors. Interaction with surrounding components was 

assessed throughout the AFW piping system, but no possible damage scenarios were 

found. 

Piping systems which span between two structures were also found to have high 

capacities. Maine Yankee is a rock site, thus very little relative motion between 

4-46 
Figure 4.3-4 Maine Yankee CS Pump Suction MOV's CS-M-91 and CS-M-92. 

Valves with Long Operator Arms. 

4-47 
f' ' ,.,. ~ .... 

'"':: ~ 

Figure 4.3-5 Example of Long Operator Valves Documented in the Data 

Base From the Coalinga Area. 

4-48 
the foundations of different structures would be expected to occur in a 0.3g pga 

earthquake. The areas which are of a possible concern for a plant on a rock site 

are places where short spans of piping traverse two buildings at a location which 

will experience large relative displacements. This could occur high up in a 

relatively flexible structure. Maine Yankee has very few piping systems which 

traverse two structures at elevations above grade. The most critical location at 

Maine Yankee exists for the steam lines running between the steam and feedwater 

valve area and the turbine building (Figure 4.3-6). These interbuilding pipes were 

judged to have a high capacity for two reasons: 

o The length of piping is long relative to the diameter of the 

piping. The relative building displacements required to fail the 

piping would have to be very large given the distance between 

buildings and the enormous ductility which has been 

demonstrated for welded steel piping. 

o The steam and feed valve area structure is a very stiff concrete 

structure and will have a small displacement at the mid-height 

location of these steam pipes. Displacements of the turbine 

building on the other end of the pipe will be limited by the 

stiffer control building. 

The cast iron service water header from the service water pumps to the PCC and 

SCC Heat Exchangers was found to be predominately buried. The only exposed 

portion of the piping was from Pump Discharge to ground penetration in the 

Pumphouse. The exposed portion of the cast iron service water piping was 

retrofitted with large seismic supports which were anchored back to the pumphouse 

structure. Short span lengths and adequate bracing were observed during the 

walkdown, such that capacity was judged greater than the margin earthquake level 

of 0.3g. The buried portion of the service water header could not be inspected; 

however, Maine Yankee calculations were reviewed with the subsequent capacity 

of the buried service water header determined to be greater than 0.3g. 

Cable Trays and Cabling. Under the PanePs guidelines, cable trays and cabling are 

screened out for review earthquake levels less than 0.3g pga. The Expert Panel 

recommended that example cable trays be inspected to verify that they are 

adequately anchored and braced and to confirm that taut cables will not be 

affected by anticipated relative displacement or any sharp edges at the ends of the 

trays. 

Cable tray walkdown effort was focused on the following areas which contain 

much of the cable trays in the plant: 

o Turbine/service building 

Cable vault, El. 21'-0" 

Cable tray room, El. 3 5' -0" 

Cable tray area, El. 35'-0" 

Switchgear room, El. 45'-6" 

Turbine hall, El. 21 '-011 

o Circulating water pumphouse 

o Primary auxiliary building, El. 21 '-0" 
Figure 4.3-6 Steam Lines Spanning Between the Steam and Feed Water Valve 

Area (Left) and the Turbine Building. 

4-50 
o M.C.C. room 

M.C.C. rooms, El. 21 '-0" and El. 33'~4" 

Containment penetration area, El. 46'~0" 

o Ventilation equipment area 

Cable trays in other areas are generally similar to cable trays in the areas listed 

above. 

Typical Maine Yankee cable trays are shown in Figure 4.3·7. The following 

general description of the Maine Yankee cable trays is obtained from the 

walkdowns: 

o Cables are typically contained in 24-inch-wide aluminum, ladder 

type cable trays. 

o Maximum cable tray loading was found in the service building 

cable vault where inserts were installed within the trays to permit 

cable fill somewhat in excess of 100% relative to the tray itself. 

o Cable tray supports are typically unbraced rod hung trapezes, 

screwed into Phillips Redhead concrete inserts. Braced cantilever 

bracket floor to ceiling column supports are used in the cable 

vault. Wall-mounted cantilever brackets were found at a few 

locations in the plant. 

o Support spacings are six feet or less. 

o Supports carry a maximum of six tiers of cable trays. 

Cable tray systems similar to those found at Maine Yankee have been subjected to 

shake table testing and actual earthquakes. The test programs have shown that 

cable tray systems are capable of withstanding significant seismic input levels 

without gross damage that would compromise cable integrity. [EQE Inc., 1986] 

·presents a summary of the performance of cable tray and conduit systems in past 

earthquakes. Experience data shows that cable tray and conduit systems 

constructed according to normal industrial standards have a large capacity for the 

absorption of seismic inertial loads. There have only been a few instances of local 

damage to cable tray systems due to actual earthquakes. These occurrences did not 

compromise the structural integrity of the cable tray systems or the integrity or 

function of the supported cables. The single instance of cable tray collapse 

involved an anomalous support configuration not observed at Maine Yankee. 

To confirm that the Maine Yankee cable trays can be screened out at the 0.3g pga 

review earthquake level, they were compared to cable trays in the earthquake 

experience data base [EQE Inc., 1986]. This comparison was performed for a 

number of parameters important to cable tray seismic adequacy, including the 

following associated with seismic input, seismic response, and system integrity: 

o Input Parameters 

Peak ground acceleration 

Duration of strong ground motion 

4-51 
Figure 4.3-7 Photographs of Typical Maine Yankee Cable Trays. 

4-52 
; ' 

Frequency content of ground motion 

Soil type 

Building type and size 

Elevation in ·building 

o Response Parameters 

Extent and complexity of systesm 

System interfaces 

System interactions 

o System Integrity Parameters 

Support type and support members 

Connection details 

Cable tray type 

Cable tray loading 

Support span 

Number of tiers per support 

The Maine Yankee cable trays were found to be enveloped by the experience data 

base for these parameters, thereby demonstrating that they can be screened out for 

the 0.3g pga review earthquake. 

Potential concerns identified by the Expert Panel include faih:~re of taut cables due 

to large relative displacements, severing of cables caused by sharp edges at the 

ends of cable trays, and failure of welds. Review of the Maine Yankee cable trays 

did not identify any of these conditions. 

Based upon a comparison of the Maine Yankee cable trays with cable trays in the 

earthquake experience data base [EQE Inc., 1986], and walkdown of the systems for 

particular problem areas, it is concluded that the cable trays and cabling can be 

screened out for the review earthquake level. 

Instrument Racks. The sample of instrument racks reviewed during the course of 

the walkdown were all judged to have a HCLPF capacity of greater than the 0.3g 

pga review level earthquake. The racks reviewed were observed to be similar to 

rack configurations documented in the experience data base. The racks were 

constructed of structural angle with bracing near the base well anchored to the 

concrete ftoor or walls with expansion anchors. The rack member connections 

were of all welded construction. The instruments and attached components were 

all observed to be well anchored to the rack face plate. Typical Maine Yankee 

instrument racks are shown in Figure 4.3-8. 

Control Room Ceiling. Control room ceilings are screened out for review 

earthquake levels less than 0.3g pga. This is subject to verification. that the 

ceilings are adequately braced and that other overhead fixtures are properly 

anchored. 

4-53 
., 

• t ~ ' 

Figure 4.3-8 Photographs of Typical Maine Yankee Instrument Racks and 

Component Attachments. 

4-54 
Figure 4.3-9 Photographs of the Maine Yankee Control Room Ceiling. 

4-55 
The suspended ceiling system over the control room is the conventional T-bar type 

with fibrous acoustic panels typically 2'-0" square. The T-bars are safety wired to 

the concrete slab above. The safety wiring is judged to be sufficient to prevent 

collapse of the T-bar ceiling to the floor below. The acoustic panels are not 

positively attached to the T-bars~ Suspended ceiling panels have been dislodged in 

past earthquakes. However, this has not been a source of damage to control panels 

or injury to operators. Furthermore, attempts to push the Maine Yankee ceiling 

panels out met with significant resistance due to the presence of the safety wire 

loops at the T-bar intersections. Damage resulting from falling ceiling panels is 

considered unlikely. 

The control room and battery room light fixtures are made of sheet metal and are 

suspended from the slab above. They are safety wired to the slab similar to the T· 

bar ceiling (Figure 4.3-9). This should be sufficient to prevent the light fixtures 

from dropping due to the 0.3g review level earthquake. It may be possible for 

translucent panels and light tubes to become dislodged during an earthquake. 

However, as with the ceiling panels, the likelihood of damage to equipment and 

personnel below is small. 

Other ceiling fixtures above the control room include HV AC ducting and conduit. 

As with the rest of the plant, the ducting is typically rod hung and braced by rods 

attached to the ceiling by concrete inserts. Conduit above the control room is 

typically about two inches in diameter or less. It is supported by threaded rod 

hangers anchored to the ceiling by concrete inserts, similar to the cable trays. 

Ducting and conduit are judged to be adequate for the review level earthquake of 

0.3g. 

In conclusion, review of the control room ceiling and other overhead fixtures 

confirmed that they could be screened out. 

4.3.4.2 Plant Unique Features 

During the review and walkdown of the plant, particular attention was paid to 

identify any unique features. The unique features are defined as the ones that 

either have proved to be important contributors in the past seismic PRAs or have 

not been considered in the previous PRAs. The example of the first kind is the 

Jocasse dam at Oconee. At Maine Yankee, the earth dike enclosing the fire water 

pond was examined for potential safety significance to the plant. It was found 

that terrain around the plant is such that the water from the pond would flow 

away from the plant in case of a dike failure. 

Other features that were not addressed in the previous seismic PRAs and that 

should be examined in future seismic margin studies as a result of this study are: 

o Mussel pump 

o Steel structures 

o Lead-antimony batteries 

o Anchorage of the transformer core/coil assembly 
' ....... 

Figure 4.4-1 Block Wall VE 21-1. 
G. 18" 

~ I 

3/16"6 HOLE(TYP.j 

SOLID CONC. BLOCK WALL 12'-11 1/2" (REF.) 

L6x6x 1/2 

l6x6x 1/2 continuous 

along top of wall 

Bar 1/4" x 2" (5 total) from 

l6x6 to steel anchored to 

floor slab (not shown) 

e EV 

Figure 4.4-2 Conceptual Seismic Retrofit to Block Wall VE 21-1. 

4-58 

A b. • 

I> • '> I I) 

p, 

{) ,fl. 

. ~ . 
The report by Energy Incorporated (Vo1.2) discusses the significance of mussel 

pump and where similar situations could occur in other plants. The Panel report 

should include some discussions and guidance on reviewing steel structures and 

about their seismic capacities as discussed in Chapter 6. The topics of leadantimony batteries and transformers are covered in later sections. 

4.4 Maine Yankee Component Modifications 

Several Maine Yankee components were identified during the margin evaluation as 

potentially having a relatively low capacity. These components included critical 

equipment items identified as part of the margin review and noncritical items that 

posed a possible interaction hazard to critical equipment. During the course of the 

margin review Maine Yankee took upon itself to modify those components 

identified as having a potentially low capacity. The installation of the 

modifications occurred in two stages. Modifications which could be installed 

without affecting plant operation were undertaken. Modifications which would be 

disruptive and pose a risk to plant operation and maintenance personnel were 

scheduled for the March 1987 refueling outage. The modifications completed 

before the refueling outage were verified during the second walkdown; however, a 

third wa1kdown during or after the March 1987 outage is necessary to verify the 

modifications scheduled to be completed during that time period (reference Section 

6.4). The specific component modifications described below are reflected in the 

structures, equipment and plant capacities presented in Section 5. 

Structures 

Block Wall VE 21-1. Block Wall VE 21-1 is located in the ventilation equipment 

room. This wall is adjacent to the containment spray fans, Fans 44A and 44B, and 

their ducting and filter (Figure 4.4-1). It is freestanding and built integral with an 

intersecting wall at one side. Structural drawings show this wall to be 18 in. thick 

by I 0 ft 0 in. high, and constructed of solid concrete block. 

Failure of Wall VE 21-1 due to out-of -plane seismic response could result in 

damage to the adjacent ducting and filter, thus causing the fans to draw air from 

the ventilation equipment room itself rather than the containment spray 

pumphouse. This wall may be particularly vulnerable to seismic effects since it is 

freestanding. Conceptual sketches for seismic retrofits call for structural steel 

restraints to be added at the top and bottom. · Top supports will consist of steel 

angles spanning back to the concrete wall of the containment spray pumphouse 

(Figure 4.4-2). Calculations based upon the preliminary retrofit sketches provided 

indicate that the modifications should be sufficient to provide the wall with a 

HCLPF capacity greater than 0.3g. 

Equipment 

The Emergency Diesel Day Tanks TK-62A & B. The day tanks were observed 

during the walkdown to be elevated tanks supported from a braced steel frame. 

Each column of the frame was observed anchored to the concrete floor with two 

expansion anchor studs with provisions for the installation of two additional studs. 

Under each column was a leveling plate and grout pad which raised concern about 

4-59 
,.·,. 

Figure 4.4-3 Photograph of the Alarm Message Display Mounted on the Main 

Control Board. 

4-60 
' . 

the expansion anchor embedment into the concrete floor. To confirm bolt 

embedment each bolt was ultrasonic tested by Maine Yankee technicians and found 

to be only partially embedded into the concrete floor. The modification for the 

day tank anchorage consisted of installing two additional bolts through existing 

unused holes in the column base plates. Seven-inch Hilti Qwik Bolts were installed 

with a four-inch minimum embedment into the concrete floor. The modification 

was performed by Maine Yankee without affecting plant operation or safety. 

Main Control Board Alarm Message Display. The Alarm Message Display is located 

on the Main Control Board and is not critical to plant operation. However, during 

the first walkdown the component was observed to be unanchored (Figure 4.4-3) 

and posed a significant risk to adjacent components in the main control board (the 

component was mounted approximately six feet above the floor). This was a 

classic example of system interaction. There were anchor bolt holes in the support 

such that it was believed the component was removed for maintenance, replaced 

and not anchored. Maine Yankee subsequently reinstalled the bolts which were 

verified during the second walkdown. 

Station Batteries No. 1.2.3 & 4. The station batteries were observed to have leadantimony flat-plate battery cells during the first walkdown. Lead-antimony flatplate batteries have been subject to controversy regarding their seismic 

performance, particularly in the aged state. 

There is no positive data substantiating the lack of performance during 

earthquakes. However, there is sufficient data to suggest a reduced capacity over 

lead-calcium type batteries where a wealth of test data exists. This lack of data 

precludes the determination of a HCLPF capacity for lead-antimony battery cells. 

Maine Yankee had previously scheduled Station Battery No. 1 for change out 

during the March 1987 refueling outage based strictly on the battery age. 

Subsequent to the first wa1kdown, Maine Yankee personnel decided to replace both 

batteries No. 1 and No. 3 during the March 1987 refueling outage. Station 

Batteries No. 2 and 4 were not scheduled to be changed out at this time, but Maine 

Yankee anticipates replacing them at the next refueling outage. The component 

and the plant HCLPF capacities presented in Section 5 are based on the Station 

Batteries No. I and 3 being replaced with new C & D lead-calcium batteries, LC-25, 

supported from battery racks, Model No. RD-938G Type 2. The Station Batteries 

No. 1 and 3 battery cells and racks should be verified for correct installation 

during the third walkdown. 

Station Service Transformers X-507 & X-608. The station service transformer 

enclosures were observed to be well anchored during the walkdown. The review of 

the internal transformer core/coil assembly was observed to be not anchored, but 

supported by four rubber isolators with no vertical uplift or lateral restraints. 

Figure 4.4-4 shows several photographs of the transformer core/coil assembly in the 

unanchored state. The transformers in the unanchored state have a low seismic 

capacity. Maine Yankee engineers designed a modification to securely anchor the 

core/coil assembly by utilizing the existing shipping studs. The electrical 

implications of this modification were discussed with the manufacturer, General 

Electric. Figure 4.4-5 is a sketch of the proposed modification. Maine Yankee has 

4-61 
Figure 4.4-4 Photographs of the Maine Yankee Station Service Transformer 

Internal Core/Coil Assembly Before the Anchorage Modification. 

4-62 
EXISTING 3/4"6 SHIPPING STUD 

NEW 3/8" THICK BY 4"6 OR SQ 

PLATE WASHE.R 

EXISTING 4'x1' CROSS BAR 

NEW 1/2" THICK x 4" sq SHIM PLATE 

5/8"6 JACK STUD 

NEW NUT AND WASHER 

RUBBER ISOLATOR 

C 6.38" x 4.69" BASE CHANNEL 

ELEVATION VIEW OF TRANSFORMER CORE/COIL 

ASSEMBLE ANCHORAGE ( TYP 4 PLACES) 

TRANSFORMER CORE & 

COl L ASSEMBUZ 

Figure 4.4-5 Proposed Anchorage Modification for the Maine Yankee Station 

Service Transformers Internal Core/Coil Assemblies. 

263nb/mrgsch4 4-63 
Figure 4.4-6 Maine Yankee Air Conditioners in the Unmodified Condit!ion. 

4-64 
\ ... ~. ~ 

NEW L 3 1/2 LONG 

(REF, CUT TO SUIT) 

1/811 

\ 

SECTION A-A 

NEW fj_ 

NEW L 

SECTION B-8 

Figure 4.4-7 

SECTION C-C 

U8" (TYP. 

~ MAY BE SIMILAR 

TO SECTION C-C 

SECTION D-D 

Proposed Anchorage Modification for the Maine Yankee Air 

Conditioners. 

4-65 

MAX. 
scheduled the anchorage modification to be installed during the March 1987 

refueling outage. The transformer and plant HCLPF capacities represent the 

modified condition. The station service transformers should also be inspected 

during the third walkdown to verify the modification has been installed. 

HV AC Computer Room and Lab Air Conditioners. The air conditioners reviewed 

during the walkdown were found to be supported from vibration isolators. 

Conversations with the isolator manufacturer determined the isolators to have no 

vertical uplift capacity and marginal lateral load resisting capacity. Maine Yankee 

developed a modification that would provide vertical and lateral load capacity to 

be installed during the March 1987 refueling outage. Sketches of the modification 

were provided to EQE for the capacity evaluation of these components. Figures 

4.4-6 and 4.4-7 show photographs of one air conditioner and a sketch of the 

proposed modification provided to EQE, respectively. The HCLPF capacities of 

air conditioners represent this modified condition. The air conditioners should 

also be inspected during the third walkdown to verify the modification has been 

installed in accordance with the proposed design provided to EQE. 

HVAC Containment Spray Fans. The fans were observed during the walkdown to 

be supported from vibration isolators. The configuration of the isolators was such 

that vertical uplift was provided. Lateral stability of the isolator configuration 

was identified as marginal in resisting the seismic loadings. The existing 

containment spray fan isolator assembly was scrutinized by both the fragility 

analysts and by the peer review group members. The earthquake experience data 

base was not useful as a resource for this particular isolator configuration due to a 

lack of similarity with the data base isolator systems. A rigorous fragility analysis 

of the cold formed steel attachment strip and the attachment bolt was deemed to 

be nearly impossible to complete due to low cycle fatigue, lack of ductility, and 

stress concentration issues associated with the attachment strip. Maine Yankee 

engineers agreed to design an anchorage upgrade based on the good engineering 

judgment that a modification was necessary in order to ensure that the fan could 

withstand (with high confidence) the 0.3g review level earthquake. Maine Yankee 

provided EQE with a sketch of the proposed modification for the margin 

evaluation. Figures 4.4-8 and 4.4-9 show photographs of the fans and a sketch of 

the proposed modification provided to EQE, respectively. The fans and plant 

HCLPF capacities represent this modified condition. The fans should also be 

inspected during the third walkdown to verify that the modification has been 

installed. 

Miscellaneous Components. Miscellaneous components include noncritical 

components which were identified during the walkdown to posses a potential risk 

to plant safety by failing and impacting a critical equipment component. All of 

these components fall into the general category of system interaction which the 

Expert Panel identifies in its review guidelines. These components included the 

alarm message display discussed above, emergency lighting units throughout the 

plant, compressed gas boilers, and a welding cart on wheels near the containment 

spray fans. Figure 4.4-10 shows photographs of the emergency lighting units and 

the welding cart. All components that were identified during the first or second 

walk down have been subsequently modified, secured, or removed by Maine 

Yankee. These modified components were evaluated during the second walkdown 

(or based on verbal communication by Maine Yankee engineers in the case of the 

welding cart) and judged to have a HCLPF capacity greater than the 0.3g pga. 

4-66 
Figure 4.4-8 Maine Yankee Containment Spray Fans Before the Installation of 

the Anchorage Modification. 

4-67 
(TYP.) 

1/2" (TYP.) 

TYPICAL PLAN VIEW 

(MIRROR IMAGE OPPOSIT END) 

(MAX) 

SECTION A-A 

SKETCH FOR FN 44A/B SHT. 1 of 2 

Figure 4.4-9 

--#- 1 /8" ( TYP. ) 

SECTION B-B 

Proposed Anchorage Modification for the Maine Yankee 

Containment Spray Fans. 

4-68 
Figure 4.4-10 Photographs of the Unstrapped Emergency Lighting Units in the 

Control Room and Welding Cart Near the Containment Spray Fans. 

4-69 
-~: ... 
. ·.:",' 

,; : 

CHAPTER 5 

EVALUATION OF SEISMIC CAPACITIES OF COMPONENTS AND PLANT 

5.1 Review of Structural Models 

ln~structure response spectra for Maine Yankee were generated by Cygna 

specifically for use in the trial plant application. Input consisted of acceleration 

time-histories matching the median NUREG/CR~0098 ground response spectrum. 

The dynamic analyses were performed with structural models recently developed 

for use in other Maine Yankee analyses. These models were reviewed to verify 

that they are adequate to predict responses for the review level earthquake. The 

floor response spectra generated by Cygna were judged to be adequate for this 

margin review. Since seismic analyses of structures and components developed for 

the original design were generally not used to calculate HCLPF capacities, the 

original design structure dynamic models were not obtained or reviewed. 

[Cygna, 1982] describes the dynamic models generated for the following structures: 

o Containment structure 

o Containment internal structure 

o Containment spray pumphouse 

o Main steam valve house 

o Primary auxiliary building 

o Turbine/service building 

This report contains the following information: 

o Computer programs used 

o General modeling assumptions 

o Description of individual structure modeling considerations 

o Sketches of seismic load resisting elements 

o Overall structure mass properties 

o Calculated frequencies, modal participation factors, and mode 

shapes 

This information was reviewed to verify the adequacy of the dynamic models in 

the fallowing manner: 

1. Review overall approach and assumptions. 

2. Review individual structure models . 

2.1 Review general model layout. 

2.2 Verify that the major load paths are included. 

2.3 Verify accuracy of the overall masses. 

2.4 Review reasonableness of the eigen solutions. 

The overall approach and assumptions were judged to be consistent with practices 

within the nuclear industry and generally adequate to predict overall seismic 

response. All structures were modeled as being linear elastic with fixed base 

boundary conditions. 

5-1 
The containment was analyzed by a three-dimensional finite-element model using 

shell elements with uniform mass densities. The effects of concrete cracking for 

the containment analysis were conservatively modeled by using upper and lower 

bounds on the modulus of elasticity. Ac9uracy of the containment model was 

confirmed by an independent analysis described in [Hashimoto, 1984]. 

The other structures were analyzed using dynamic models with masses lumped at 

major floor elevations. Stiffness matrices for the lumped mass models were 

generated from more refined finite-element representations of the load resisting 

elements. Uncracked stiffness properties were used for concrete shear walls. Floor 

slabs were modeled as being infinitely stiff in their own planes. 

With the possible exceptions noted below, the dynamic models of the structures 

were found to be consistent with practices used in the nuclear industry and 

generally adequate to predict overall seismic response. Model descriptions and 

sketches in [Cygna, 1982] were reviewed to verify that major load-resisting 

elements were included. Independent, approximate calculations were performed 

when possible to confirm accuracy of the lumped masses. The eigen solutions were 

reviewed using engineering judgment to assess whether they were reasonable given 

the structure configurations. 

Review of the turbine/service building model indicates that the diesel generator 

enclosure walls contribute significantly to the overall stiffness of the lowest story. 

However, the load path from the rest of the structure to the diesel generator 

enclosure is relatively flexible. Although decoupling the diesel generator enclosure 

from the remainder of the structure reduces the overall structure stiffness, the net 

effect on seismic response is lessened since exclusion of the diesel generator 

enclosure also reduces the overall structure mass. Rather than develop a new 

structure model, correction of the dynamic response was estimated in assessing the 

structural response factor for the fragility evaluations. 

Stiffnesses of the shear wall structures other than containment were based upon 

uncracked properties. On-going scale model testing being conducted for the NRC 

has indicated that this may result in an overestimation of the actual stiffness. The 

effect of this issue on the plant HCLPF capacity was assessed by sensitivity studies 

described in Section 5.5.4. 

Group A structures that were not dynamically analyzed by Cygna are the 

circulating water pumphouse and the fuel oil pumphouse. Review of the 

circulating water pumphouse indicates that the concrete portions at and below the 

operating floor can be considered essentially rigid since the structure is supported 

by several heavy concrete walls and the exterior north, south, and west walls were 

poured against excavated rock. Dynamic analysis of the steel superstructure 

supporting the roof slab was independently generated for the fragility evaluation. 

Dynamic response of the fuel oil pumphouse was not required since no Group A 

components are housed in or directly attached to it. 

5.2 Simplified Analysis and Use of Screening Tools 

NUREG/CR-43-34 developed a set of initial screening criteria for seismic margin 

studies as outlined in Section 4.1 of this report. These initial screening criteria are 

5-2 
based on the results of past probabilistic risk assessments and on actual earthquake 

experience data. These initial screens identified classes of equipment and 

structures which have consistently demonstrated high seismic capacities. Within 

the remaining classes of equipment and structures which have not been screened 

out there typically exists a range of seismic capacities. Simplified analysis 

techniques can be utilized to separate out components within a category which 

have very high seismic capacities and do not warrant a detailed fragility analysis. 

Simplified analyses for the Maine Yankee margin study were conducted using 

either of two methods: 

o Fragility derivation using conservative response and capacity 

parameters and an estimate for the variability (/jR + t3u = 0.7) 

o Deterministic evaluation using conservative values similar to the 

CDFM approach outlined in NUREG/CR-4482 

The 120-V ac vital bus 1 A through 4A is a good example of a Maine Yankee 

component whose seismic margin was evaluated on the basis of a simplified 

analysis. These 120-V buses are wall-mounted panels containing circuit breakers 

and are located in the control room (service building at +21 ft). NUREG/CR-4334 

does not explicitly screen out active electrical equipment and states that anchorage 

should be verified for the cabinet and for individual components in the cabinet. 

The walkdown of these 120-V buses showed them to be relatively lightweight 

components with overdesigned wall anchorage. A simplified conservative fragility 

analysis was conducted utilizing the following parameters: 

o Weight = 245 lb (specified on drawing) 

o Spectral acceleration = peak spectral value for 7% damping for 

the appropriate floor spectra (cabinet is judged to be rigid) 

o Center of gravity = 8 in. away from the wall (conservative since 

the cabinet is only 10 in. deep) 

o Anchor bolt capacity = industry specified allowable of onequarter of the ultimate strength 

o f3R + f3u = 0.1 

The simplified fragility analysis on these panels resulted in a HCLPF capacity 

greater than O.Sg. Components whose HCLPF capacities were calculated via 

simplified analysis methods to be greater than 0.3g were automatically screened 

out, and detailed fragility derivations were judged to be unnecessary. 

The conservatisms most commonly utilized in the simplified analyses are: 

o Natural frequency - conservative estimate on the frequency or 

use of the frequency corresponding to the peak spectral 

acceleration 
o Capacity - conservative estimate such as the Code allowable, 70% 

of the ultimate, or failure at the yield strength 

0 Combined randomness and 

conservatively estimated based 

components at Maine Yankee 

5.3 Second Walkdown 

uncertainty variability 

on calculations for similar 

The second walkdown was performed to accomplish the following tasks: 

o Obtain additional detailed information on equipment and 

structures inspected during the first walkdown. 

o Survey components added to the Group A equipment list after the 

first walkdown was performed. 

HCLPF capacities were calculated for certain components immediately following 

the first walkdown. In determining these capacities, additional information needs 

were identified. The first task listed above was performed to obtain this 

information for finalization of component HCLPF capacities. A minor amount of 

effort was also devoted towards additional documentation of items reviewed in the 

first walkdown. A limited number of components were added to the Group A 

equipment list after the first walkdown was completed based upon more detailed 

system analyses. The second task was performed to obtain information on these 

added components for screening and/or HCLPF capacity determination. Structures 

and equipment reviewed in the second walkdown are listed in Table 5.3-1. 

5.4 HCLPF Capacity of Components 

In the following, the fragility analysis methodology is described with some 

illustrative examples. 

5.4.1 Fragility Analysis Method 

5.4.1.1 Methodology 

In this method, the component HCLPF capacity is calculated using the fragility of 

the component. The seismic fragility of a structure or equipment is defined as the 

conditional probability of its failure at a given value of peak ground acceleration. 

The methodology for evaluating seismic fragilities of structures and equipment is 

documented in [Ravindra and Kennedy, 1983], [PRA Procedures Guide, 1983], and 

[Kennedy and Ra vindra, 1984] and has been developed and applied in over 20 

seismic PRAs. 

5-4 
Table 5.3-1 Second Walkdown Component Review List 

TURBINE BUILDING 

COMPONENT FLOOR ELEVATION 

1. Structural details 

2. Block wall TB 21-2 

3. Primary Component Cooling Heat Exchanger E-4A 

4. Primary Component Cooling Heat Exchanger E-4B 

5. Secondary Component Cooling Heat Exchanger E-5A 

21' - 0" 

21'- 0" 

21' • 0" 

21' .. 0" 

2P .. 0" 

37' - 0" 

37' - 0" 

6. Secondary Component Cooling Heat Exchanger E-SB 

7. Cooler Supply Temperature Control Valve PCC-T-19 

8. Cooler Bypass Temperature Control Valve PCC-T-20 

SERVICE BUILDING 

COMPONENT 

1. . Structural details 

2. Block Wall SB 21-7 

3. Block Wall SB 21-17 

4. Block Walls SB 35-1 to 4 

5. Block Walls SB 39-1 and 2 

6. Block Wall SB 45-6 

7. Block Wall SB 61-2 

8. Block Wall SB 77-2 

9. General inspection of the Switchgear Room 

IO.Protected SWGR. Room Supply Fan FN-31 

ll.Computer Room Air Conditioner AC-1A 

12.Computer Room Air Conditioner AC-1B 

13.La b Air Conditioner AC-2 

14.General inspection of the Control Room 

15.Cable Trays 

16.Con trot Room Ceiling 

TURBINE BUILDING AUXILIARY BAY 

FLOOR ELEVATION 

21' - 0" 

21' .. 0" 

35' - 0" 

39' - 0" 

45' .. 0" 

61' .. 0" 

77' - 0" 

45' - 6" 

39' .. 0" 

39' .. 0" 

39' - 0" 

39' - 0" 

21' .. 0" 

21' .. 0" 

COMPONENT FLOOR ELEVATION 

1. Auxiliary Fuel Oil Supply Tanks Tk-28 A & B 

2. Diesel Fuel Oil Day Tank Tk-62 A & B 

3. DG Compressed Air Tanks Tk-76A & Bl-6 

4. DG Room Exhaust Fans FN-20A & B 

5. Diesel Generator Air Intake & Exhaust Dampers 

6. DG-1 B Cooler Inlet Temperature Control Valve 

SCC-T-305 

5-5 

21' - 0" 

21'- 0'' 

21' - 0" 

31' - 0" 

31' .. 0" 

23' - 0" 
Table 5.3-1 Second Walkdown Component Review List (Continued) 

PRIMARY AUXILIARY BUILDING 

COMPONENT FLOOR ELEVATION 

1. Charging Pump Seal Leakage Coolers E-92A & B 11' - 0" 

2. Seal Water Heater E·96 

3. Auxiliary Charging Pump P-7 Lube Oil Cooler 11' - 0" 

4. Return from Penetration Coolers PCC-A-238 12' - 011 

5. Seal Water Supply Filter FL-34B 

6. PCCW Isolation from the RHR Heat Exchanger 11' - 0" 

PCC-M-90 

7. PCCW Isolation to Containment PCC-M-219 11' - 0" 

8. Cable Trays 

STEAM AND FEED WATER VALVE AREA 

COMPONENT FLOOR ELEVATION 

1. Structural details 

2. AFW Pump B Turbine Throttle Valve MS-A-173 21' - 0" 

3. Turbine Steam Supply Pressure Control MS-P-168 21'- 0" 

CONTAINMENT SPRAY PUMP HOUSE 

COMPONENT FLOOR ELEVATION 

1. Residual Heat Removal Heat Exchangers E-3A & B 14' - 0" 

2. Reactor Coolant Regen era ti ve Heat Exchanger 

E-67 

3. Safeguards Pumps Seal Leakage Cooler E-9IA & B 00 - 0" 

4. LPSI Pump Coolers 21' - 0" 

5. Containment Spray Pump P-6IA, B & S Coolers 14' - 0" 

6. Containment Penetration Cooling Lines 

7. Block Wall VE 21-1 21'- 0" 

,! I FUEL BUILDING 

COMPONENT FLOOR ELEVATION 

I. Block Walls 

2. Fuel Pool Heat Exchanger E-25 21' - 0" 
Table 5.3-1 Second Walkdown Component Review List (Continued) 

MCC ROOM 

COMPONENT 

1. Cable Trays 

MISCELLANEOUS COMPONENTS 

COMPONENT 

1. Instrument Racks 

FLOOR ELEVATION 

FLOOR ELEVATION 

various 

5-7 
The objective of fragility evaluation is to estimate the ground acceleration 

capacity of a given component. This capacity is defined as the peak ground 

acceleration value at which the seismic response of a given component located at a 

specified point in the structure exceeds the component's resistance, resulting in its 

failure. The ground acceleration capacity of the component is estimated using 

information on plant design bases, responses calculated at the design-analysis stage, 

as-built dimensions, and material properties. The ground acceleration capacity is a 

random variable which can be described completely by its probability distribution. 

However, there is uncertainty in the estimation of the parameters of this 

distribution, the exact shape of this distribution, and in the appropriate failure 

model for the component. For any postulated failure model and set of parameter 

values and shape of the probability distribution, a fragility curve depicting the 

conditional probability of failure as a function of ground acceleration can be 

obtained. Hence, for different models and parameter assumptions, one could 

obtain different fragility curves. A satisfactory way to consider these 

uncertainties is to represent the component fragility by means of a family of 

fragility curves obtained as above; a subjective probability value is assigned to 

each curve to reflect the analyst's degree of belief in the model that yielded the 

particular fragility curve. When represented in this fashion, the fragility curves 

need not appear to be smooth S-shaped curves, approximately parallel to each 

other; they could intersect each other and they may not even be nondecreasing 

functions of peak ground acceleration. The only requirement is that fragility 

being a probability should be between 0 and I (see Figure 5.4-1). 

At any acceleration value, the component fragility (i.e., conditional probability of 

failure) varies from 0 to 1; this variation is represented by a subjective probabillty 

distribution. On this distribution we can find a fragility value (say, 0.0 I) that 

corresponds to the cumulative subjective probability of 5%. We have 5% 

cumulative subjective probability (confidence) that the fragility is less than 0.01. 

Similarly, we can find a fragility value for which we have a confidence of 95%. 

Note that these statements can be made without reference to any probability model. 

Using this procedure, the median and high (95%) and low (5%) confidence fragility 

curves can be drawn. On the high confidence curve, we can locate the fragility 

value of 5%; the acceleration corresponding to this fragility on the high confidence 

curve is the so called HCLPF capacity of the component. By characterizing the 

component fragility through a family of fragility curves, the analyst has expressed 

all his knowledge about the seismic capacity of the component along with the 

uncertainties. Given the same information, two analysts with similar experience 

and expertise would produce approximately the same fragility curves. 

Development of the family of fragility curves using different failure models and 

parameters for a large number of components in a seismic margin review or seismic 

PRA is impractical if it is done as described above. Hence, a simple model for the 

fragility was proposed as described in the above cited references. In the following 

this fragility model is described. 

Fragility Model 

The entire fragility family for an element corresponding to a particular failure 

mode can be expressed in terms of the best estimate of the median ground 

acceleration capacity, Am, and two random variables. Thus, the ground 

acceleration capacity, A, is given by 

5-8 
1. 0 

LJ.J 

0::: 

::::::> 

_J 

......... 

<( 

L.L.. 

L&.. 

0 

>-

1-

-

_J 

-

co 

<( 

ca 

0 

0::: 

0... 

_J 

<( 

z 

0 

1- -......... 

0 

z 

0 

u 

0 

PEAK GROUND ACCELERATION 

\' 

Figure 5.4-l Fragility Curves. 
(5.4-1) 

in which ~R and ~U are random variables with unit medians, representing, 

respectively, the inherent randomness about the median and the uncertainty in the 

median value. In this model, we assume that both fR and €u are lognormally 

distributed with logarithmic standard deviations, f3R and f3u, respectively. The 

formulation for fragility given by Eq. (5.4-1) and the assumption of lognormal 

distribution allow easy development of the family of fragility curves which 

appropriately represent fragility uncertainty. For the quantification of fault trees 

in the plant system and accident sequence analyses, the uncertainty in fragility 

needs to be expressed in a range of conditional failure probabilities for a given 

ground acceleration. This is achieved as explained below: 

With perfect knowledge (i.e., only accounting for the random variability,€R), the 

conditional probability of failure, f 0 , for a given peak ground acceleration level, a, 

is given by 

(5.4-2) 

where c;l> (') is the standard Gaussian cumulative distribution function. The 

.relationship between f 0 and a is the median fragility curve plotted in Figure 5.4-2 

for a component with a median ground acceleration capacity Am = 0.90g and IJR = 

0.30. For the median conditional probability of failure range of 5% to 95%, the 

ground acceleration capacity would range from 0.55g to 1.48g. 

When the modeling uncertainty €u is included, the fragility becomes a random 

variable (uncertain). At each acceleration value, the fragility f can be represented 

by a subjective probability density function. The subjective probability, Q (also 

known as "confidence") not exceeding a fragility f' is related to f' by 

when .. 

(5.4-3) 

Q = P[f < f' I a] i.e., the subjective probability (confidence) that 

the conditional probability of failure, f, is less than f for 

a peak ground acceleration a 

c;l> -1(·) = the inverse of the standard Gaussian cumulative distribution 

function. 

For example, the conditional probability of failure f' at acceleration 0.4g that has 

a 95% nonexceedance subjective probability (confidence) is obtained from Eq. (5.4-

3) as 0.22. The 5% to 95% probability (confidence) interval on the failure at 0.4g is· 

0 to 0.22. Subsequent computations are made easier by discretizing the random 
' ·.,_: 

95% 

Nonexceed. 

ProbabilityConfidence 

-o Curve Median 

- Fragility 

Curve := 0.6 

.D 

c:a 

.D 

0 

._ 

a.. 0.4 

c:o 

c 

0 ;a 0.2 

c 

0 

(.) 

Figure 5.4-2 

- 5% 

Nonexceed 

Probability 

(Confidence) 

Curve 

t 0.4 0.8 1.2 1.6 2.0 SSE Peak Ground Acceleration. g 

Median, 5% Nonexceedence, and 95% Nonexceedence Fragility 

Curves for a Component. 

5-11 
variable probability of failure f into different intervals and deriving probability 

qi for each interval (Figure 5.4-3). Note that the sum of qj associated with all the 

intervals is unity. The process· develops a family of fragility curves, each with an 

associated probability qi. 

The median ground acceleration capacity Am, and its variability estimates·~R and 

~U are evaluated by taking into account the safety margins inherent in capacity 

predictions, response analysis, and equipment qualification, as explained below. 

Failure Modes 

The first step in generating fragility curves such as those in Figure 5.4-3 is to 

develop a clear definition of what constitutes failure for each of the critical 

elements in the plant. This definition of failure must be agreeable to both the 

structural analyst generating the fragility curves and the systems analyst who must 

judge the consequences of component failure. Several modes of failure (each with 

a different consequence) may have to be considered and fragility curves may have 

to be generated for each of these modes. For example, a motor-actuated valve may 

fail in any of the following ways: 

I. Failure of power or controls to the valve (generally related to the 

seismic capacity of the cable trays, control room, and emergency 

power). Since they are not related to the specific item of 

equipment (i.e., motor actuated valve) and are common to all 

active equipment, such failure modes are most easily handled as 

failures of separate systems linked in a series to the equipment. 

2. Failure of the motor. 

3. Binding of the valve due to distortion and, thus, failure to 

operate. 

4. Rupture of the pressure boundary. 

It may be possible to identify the failure mode most likely to be caused by the 

seismic event by reviewing the equipment design and considering only that mode. 

Otherwise, fragility curves are developed based on the premise that the component 

could fail in any one of all potential failure modes. 

Identification of the credible modes of failure is largely based on the analyst's 

experience and judgment. Review of plant design criteria, calculated stress levels 

in relation to the allowable limits, qualification test results, seismic fragility 

evaluation studies done on other plants, and reported failures (in past earthquakes, 

in licensee event reports and fragility tests) are useful in this task. 

Structures are considered to have failed functionally when they cannot perform 

their designated functions. In general, structures have failed functionally when 

inelastic deformations under seismic load are estimated to be sufficient to 

potentially interfere with the operability of safety-related equipment attached to 

the structure, or fractured sufficiently so that equipment attachments fail. These 

failure modes represent a conservative lower bound of seismic capacity since a 

5-12 
(1) 1. 0 ..... 

:::J 

co 

u.. 

-0 

>-

-

.D 

('lj 

.c 

0 

..... 

a.. 

('lj 

c 

0 

'0 -

c 

0 

(.) sbE o.4 o.a 1.2 1.s 2.0 

Peak Ground Acceleration, g 

'. 

·: 

Figure 5.4-3 Family of Fragility Curves for a Component. 

5-13 
, .. 

larger margin of safety against collapse exists for nuclear structures. Also, a 

structural failure has been generally assumed to result in a common cause failure 

of multiple safety systems, if these are h'oused in the same structure. For example, 

the service water pumps in Zion were assumed to fail when the crib house pump 

enclosure roof collapses. 

For piping, failure of the support system and plastic collapse of the pressure 

bounda.ry are considered dominant failure modes. Failure modes of equipment 

examined may include structural failure modes (e.g., bending, buckling of supports, 

anchor bolt pullout, etc.), functional failures (binding of valve, excessive 

deflection), and relay trip or chatter. 

Consideration should also be given to the potential for soil failure modes (e.g., 

liquefaction, toe bearing pressure failure, base slab uplift, and slope failures). For 

buried equipment (i.e., piping and tanks), failure due to lateral soil pressures may 

be an important mode. Seismically indu.ced failures of structures or equipment 

under impact of another structure or equipment (e.g., a crane) may also be a 

consideration. Seismically induced failures of dams, if present, resulting in either 

flooding or loss-of -cooling-source, should also be investigated. 

Estimation of Fragility Parameters 

In estimating fragility parameters, it is convenient to work in terms of an 

intermediate random variable called the factor of safety. The factor of safety, F, 

on ground acceleration capacity above the safe shutdown earthquake level 

specified for design, ASSE' is defined as follows: 

A= F ASSE 

F!O!C A~UH\1 ~"i~mi~ ~sUla~it~ Qf ~l~m"nt 

Actual response due to SSE 

= Actual capacit~ 

Calculated capacity 

X Calcylat~d capacity 

Design response due to SSE 

X Design re~oonse due to SSE 

Actual response due to SSE 

5-14 
F is further simplified as: 

F= Actual capacity 

Design response due to SSE 

X Design response due to SSE 

Actual response due to SSE 

(5.4-4) 

Note F can also be defined with reference to a different earthquake such as the 

review earthquake level in this margin study. 

The median factor of safety, F m' can be directly related to the median ground 

acceleration capacity, Am, as: 

(5.4-5) 

The logarithmic standard deviations of F, representing inherent randomness and 

uncertainty, are identical to those for the ground acceleration capacity A. 

For structures, the factor of safety can be modeled as the product of three random 

variables: 

(5.4-6) 

The strength factor, Fs, represents the ratio of ultimate strength (or strength at 

loss-of-function) to the stress calculated for AssE· In calculating the value of Fs, 

the nonseismic portion of the total load acting on the structure is subtracted from 

the strength as follows: 

(5.4-7) 

where S is the strength of the structural element for the specific failure mode, PN 

is the normal operating load (i.e., dead load, operating temperature load, etc.) and 

PT is the total load on the structure (i.e., sum of the seismic load for AssE and the 

normal operating load). For higher earthquake levels, other transients (e.g., SR V 

discharge, and turbine trip) may have a high probability of occurring 

simultaneously with the earthquake; the definition of PN in such cases should be 

extended to include the loads from these transients. 

The inelastic energy absorption factor (ductility), F p., accounts for the fact that an 

earthquake represents a limited energy source and many structures or equipment 

items are capable of absorbing substantial amounts of energy beyond yield without 

loss-of -function. A suggested method to determine the deamplif ica tion effect 

5-15 
resulting from inelastic energy dissipation involves the use of ductility modified 

response spectra [Newmark, 1977]. The deamplification factor is primarily a 

function of the ductility ratio defined as the ratio of maximum displacement to 

displacement at yield. More recent analyses [Riddell and Newmark, 1979] have 

shown the deamplification factor to be a function of system damping. One might 

estimate a median value of for low-rise concrete shear walls (typical of auxiliary 

building walls) of 4.0. The corresponding median FIJ value would be 2.4. The 

variabilities in the inelastic energy absorption factor, FIJ., are both estimated 

as (3_R = 0.21 and f3u = 0.21, taking into account the uncertainty in the predicted 

relationship between Fp., IJ, and system damping. 

The structure response factor, F~s, recognizes that in the design analyses 

structural response was computed ustng specific (often conservative) deterministic 

response parameters for the structure. Because many of these parameters are 

random (often with wide variability) the actual response may differ substantially 

from the design a,nalyses calculated response for a given peak ground acceleration. 

The structure response factor, FRS' is modeled as a product of factors influencing 

the response variability: . 

where 

F.p 

(5.4-8) 

spectral shape factor representing variability in ground 

motion and associ a ted ground response spectra 

= direction factor representing the variability in the two 

earthquake direction response spectral values about the mean 

value 

= damping factor representing variability in response due to 

difference between actual damping and design damping 

= modeling factor accounting for uncertainty in response due to 

modeling assumptions 

mode combination factor accounting for variability in 

response due to the method used in combining dynamic modes 

of response 

earthquake component combination factor accounting for 

variability in response due to the method used in combining 

earthquake components 

Fso = factor to reflect the reduction with depth of seismic input 

Fss = factor to account for effect of soil-structure interaction. 

The median and logarithmic standard deviations of F are expressed as: 

5-16 
:I'' 

(5.4-9) 

and 

(5.4-10) 

The logarithmic standard deviation ~F is further divided into random variability, 

f1R, and uncertainty, Pu· To obtain the median ground acceleration capacity A 

the median factor of safety, Fm, is multiplied by the safe shutdown earthqu~e peak ground acceleration. · 

For equipment and other components, the factor of safety is composed of a 

capacity factor, Fe; a structure response factor, FRS; and an equipment response 

(relative to the structure) factor, FRE· Thus, 

(5.4-11) 

The capacity factor Fe for the equipment is the ratio of the acceleration level at 

which the equipment ceases to perform its intended function to the seismic design 

level. This acceleration level could correspond to a breaker tripping in a 

switchgear, excessive deflection of the control rod drive tubes, or failure of a 

steam generator support. The capacity factor for the equipment may be calculated 

as the product of F_g and Fll . The strength factor, Fs, is calculated using Eq. (5.4· 

7). The strength, S, of equipment is a function of the failure mode. Equipment 

failures can be classified into three categories: 

1. Elastic functional failures 

2. Brittle failures 

3. Ductile failures. 

Elastic functional failures involve the loss of intended function while the 

component is stressed below its yield point. Examples of this type of failure 

include the following: 

o Elastic buckling in tank walls and component supports 

o Excessive blade deflection in fans 

o Shaft seizure in pumps 

The load level at which functional failure occurs is considered the strength of the 

component. 

Brittle failure modes are those which have little or no system inelastic energy 

absorption capability. Examples include the following: 

o Anchor bolt failures 

o Component support weld failures 

o Shear pin failures 

Each of these failure modes has the ability to absorb some inelastic energy on the 

component level, but the plastic zone is very localized and the system ductility for 

an anchor bolt or a support weld is very small. The strength of the component 

5-17 
failing in a brittle mode is therefore calculated using the ultimate strength of the 

material. 

Ductile failure modes are those in which the structural system can absorb a 

significant amount of energy through inelastic deformation. Examples include the 

following: 

o Pressure boundary failure of piping 

o Structural failure of cable trays and ducting 

o Polar crane failure 

The strength of the component failing in a ductile mode is calculated using the 

yield strength of the material for tensile loading. For flexural loading, the 

strength is defined as the limit load or load to develop a plastic hinge. 

The inelastic energy absorption factor, Fl', for a piece of equipment is a function 

of the ductility ratio,p. The median value Fl' is considered close to 1.0 for brittle 

and functional failure modes. For ductile failure modes of equipment that respond 

in the amplified acceleration region of the design spectrum (i.e., 2 to 8 Hz): 

(5.4-12) 

where € is a random variable reflecting the error in Eq. ( 4-16) and has a median 

value of 1.0 and a logarithmic standard deviation, ~U' ranging from 0.02 to 0.10 

(increasing with the ductility ratio). For rigid equipment, Fp is given by 

f • f 0.13 (5.4-13) p, p, 

Again, f is a random variable of median equal to 1.0 and logarithmic standard 

deviation ranging from 0.02 to 0.10. 

The median and logarithmic standard deviation of ductility ratios for different 

equipment are calculated considering recommendations of [Newmark, 1977]. This 

reference gives a range of ductility ratios to be used for design. The upper end of 

this range might be considered to represent approximately the median value, while 

the lower end of the range might be estimated at about two logarithmic standard 

deviations below the median. 

The equipment response factor F.RE' is the ratio of equipment response calculated 

in the design to the realistic equipment response; both responses are calculated for 

design floor spectra. FRE is the factor of safety inherent in the computation of 

equipment response. It depends upon the response characteristics of the equipment 

and is influenced by some of the variables listed under Eq. (5.4-8). These variables 

differ according to the seismic qualification procedure. For equipment qualified 

by dynamic analysis, the important variables that influence response and 

variability are as follows: 

o Qualification method (QM) 

o Spectral shape (SA) - including the effects of peak broadening 

and smoothing, and artificial time history generation 

o Modeling (affects mode shape and frequency results) (M) 

5-18 
' ~ I 

o Damping (6) 

o Combination of modal responses (for response spectrum method) 

(MC) 

o Combination of earthquake components (EC) 

For rigid equipment qualified by static analysis, all variables, except the 

qualification method, are not significant. The equipment response factor is the 

ratio of the specified static coefficient divided by the zero period acceleration of 

the floor level where the equipment is mounted. If the equipment is flexible and 

was designed via the static coefficient method, the dynamic characteristics of the 

equipment must be considered. This requires estimating the fundamental 

frequency and damping, if the equipment responds predominantly in one mode. 

The equipment response factor is the ratio of the static coefficient to the spectral 

acceleration at the equipment fundamental frequency. 

Where testing is conducted for seismic qualification, the response factor must take 

into account the following: 

o Qualification method (QM) 

o Spectral shape (SA) 

o Boundary conditions in the test versus installation (BC) 

o Damping (6) 

o Spectral test method (sine beat, sine sweep, complex waveform, 

etc.) (STM) 

o Multi-directional effects (MDE) 

The overall equipment response factor is the product of these factors of safety 

corresponding to each of the variables identified above. The median and 

logarithmic standard deviations for randomness and uncertainty are estimated 

fallowing Eqs. (5.4-9) and (5.4-1 0). 

The structural response factor, FRS' is based on the response characteristics of the 

structure at the location of component (equipment) support. The variables 

pertinent to the structural response analyses used to generate floor spectra for 

equipment design are the only variables of interest to equipment fragility. Timehistory analyses using the same structural models used to conduct structural 

response analysis for structural design are typically used to generate floor spectra. 

The applicable variables are as follows: 

0 

0 

0 

0 

Spectral shape 

Damping 

Modeling 

Soil-structure interaction 

For equipment with a seismic capacity level that has been reached while the 

structure is still within the elastic range, the structural response factors should be 

calculated using damping values corresponding to less than yield conditions (e.g., 

about 5% median damping for reinforced concrete). The combination of 

earthquake components is not included in the structural response since the variable 

is to be addressed for specific equipment orientation in the treatment of equipment 

response. 

5-19 
Median Fm and variability ~R and~ u estimates are made for each of the 

parameters affecting capacity and response factors of safety. These median and 

variability estimates are then combined using the properties of lognormal 

distribution in accordance with Eqs. (5.4-6), (5.4-8), and (5.4-11) to obtain the 

overall median factor of safety Fm and variability ~R and ~U estimates required 

to define the fragility curves for the structure or equipment. For each variable 

affecting the factor of safety, the random ~) and uncertainty <8rr) variabilities 

must be separately estimated. The differentiation is somewhat J"lldgmental, but it 

can be based on general guidelines. Essentially,~ represents variability due to 

the randomness of the earthquake characteristics for the same acceleration and to 

the structural response parameters which relate to these characteristics. The 

dispersion represented by f3u is due to factors such as the following: 

o Our lack of understanding of structural material properties such 

as strength, inelastic energy absorption, and damping. 

o Errors in calculated response due to use of approximate modeling 

of the structure and inaccuracies in mass and stiffness 

representations. 

o Usage of engineering judgment in lieu of complete plant-specific 

data on fragility levels of equipment capacities, and responses. 

Information Sources 

For structures such as concrete shear walls, prestressed concrete containment, steel 

frames, masonry walls, field-erected tanks, and buried structures, the fragility 

parameters are generally estimated using plant-specific information. For major 

passive equipment (e.g., reactor pressure vessel, steam generator, reactor coolant 

pump, recirculation pump, major vessels, heat exchangers, and major piping), it is 

preferable to develop plant-specific fragilities using original design analyses. 

Because of the large quantities of other types of passive equipment (e.g., piping 

and supports, cable trays and supports, HV AC ducting and supports, conduit, and 

miscellaneous vessels and heat exchangers), it is generally necessary to use generic 

fragilities. For active equipment, use of a combination of generic and plantspecific information is needed to develop fragilities. 

Several sources of information are utilized in developing plant-specific and generic 

fragilities for equipment. These sources include the following: 

I. Seismic qualification design reports 

2. Seismic qualification test reports 

3. Plant safety analysis reports 

4. Seismic qualification review team (SQRT) submittals 

5. Seismic qualification report summaries 

6. Past earthquake experience and expert opinion 

7. United States Corps of Engineers shock test reports 

8. Specifications for the seismic design of equipment 

Sources I to 5 are plant-specific; sources 6 to 8 are generic data collected for 

similar types of equipment. 

5-20 
Equipment fragility development is accomplished by grouping equipment into a 

number of. categories. 

In seismic margin studies, a relevant quantity is the HCLPF acceleration c.apa9ity 

of the component. This quantity considers both the uncertainty and randomness 

variabilities and is the acceleration value for which we have 95% confidence that 

the conditional probability of failure is less than 5%. That is, it is an acceleration 

value for the component for which we are highly confident there is only a small 

chance of failure given this ground acceleration level. 

HCLPF capacity = Am exp{ .. 1.65 <(3u +(jR)} · 

Spectral Shape Factor • FsA 

(5.4-14) 

The review earthquake level is specified in this study as the NUREG/CR-0098 

median ground response spectrum (rock) anchored to 0.3g pga. It is interpreted as 

the 84% nonexceedance spectrum for Maine Yankee. For calculating the median 

capacity and variability, we need to estimate the median spectrum for Maine 

Yankee and the variability in the spectral ordinates. It is interpreted that the 

review earthquake level (target spectrum) is a 84% nonexceedance level, in that a 

future earthquake will exceed the target spectrum only if 16% of the spectral 

ordinates exceed the target over the frequency range of interest. 

In calculating the spectral shape factor FsA we need to consider the peak-to·peak 

response spectrum variability which reflects the observation that response spectra 

in each horizontal direction for real earthquakes have hills and valleys relative to 

a mean spectrum. In other words, the spectral ordinates at a given frequency will 

occur randomly either above or below the mean spectrum. 

The peak-to-peak variability for multidegree of freedom structure is different 

from a single·degree of freedom structure. In general, the logarithmic standard 

deviation due to peak-to-peak variability is less for a multidegree of freedom 

structure. 

In fragility analysis, the response spectrum logarithmic standard deviation, 

assuming independent dynamic modes, is given by the following equation [Reed et 

al., 1986]: 

where 

= 

~ 

~Rim 4] ~;-) ~pp xm 

(5.4-15) 

= logarithmic standard deviation for building response in the xdirection (similat: for y-direction) 

5-21 
R = xm 

median building response for the ith mode for an earthquake 

in the x-direction 

median building response combined for all modes for an 

earthquake in the x-direction 

When a single mode is dominant, 

The value of {3pp has been estimated to be 0.20 [Reed, et al., 1986]. 

In this present study, {Jpp is conservatively assumed to be 0.20. Since the review 

level spectrum is assumed to be at a 84% nonexceedance value, the median 

spectrum is obtained by multiplying the target spectrum by exp(- {Jpp). 

Therefore, the median spectral shape factor for structural response factor 

calculation is given by 

FsAm = exp(0.20) = 1.22 · 

An additional factor to be considered is the horizontal earthquake direction 

variability. This occurs since at a given frequency the response spectrum value in 

one horizontal direction is different from the direction 90 degrees away. If the 

response of the critical structural element depends equally on the two horizontal 

earthquake components, then the directional variability has little effect on the 

response~ In contrast, if one direction dominates, then direction variability may 

significantly effect response variability. 

When co-linear responses for two horizontal earthquake directions are combined, 

the logarithmic standard deviation for total response due to peak-to-peak and 

direction variability, becomes for the SRSS peak response rule: 

where 

l 

2 ? 

2 

(:~~y 2 I R2 }''" :>.l 

fl ; · ., -·rn · '· 

p~J X ·-r 

\ -~2 ) y (5.4-16) 

logarithmic standard deviation for direction variability from a 

statistical analysis of the ratios of spectral accelerations for a 

fixed direction (e.g., north/south) to the corresponding 

geometric mean values. Since we are assuming that the 

maximum of the two horizontal components is used in 

specifying the review earthquake, the resulting distribution of 

the ratio will be approximately lognormal with a median 

larger than 1.0 .. 

5-22 
' ~ .. ', ' ' 

In the absence of such a detailed studyt~due to directional effect is judged to be 

between 0.10 and 0.20 and is taken as 0.15 in this margin study. The total 

variability due to peak-to-peak variability and directional effects is taken as 

It is assumed that these variabilities (peak-to-peak and direction) are removed from 

any hazard analysis leading to selection of margin earthquake selected for review. 

It should be noted that the issue of exceedance of the target response spectrum 

(e.g., 16% assumed in the Maine Yankee specific spectrum) is not the same as the 

confidence levels associated with the target spectrum as determined from the 

hazard curves. In the case of the hazard curves, the confidence is associated with 

the uncertainty of the underlying geophysical parameters, while the exceedance of 

the target response spectrum is a randomness consideration associated with the 

variabilities of earthquake time history "signatures" [Reed et al., 1986]. 

5.4.1.2 Steel Structures 

HCLPF capacities were determined for the following steel framed structures: 

o Circulating water pumphouse steel superstructure 

o Turbine/service building, steel framed portions 

o Main steam valve house steel interior structure 

The circulating water pumphouse (Figure 5.4-4) houses the service water pumps 

and other components of the service water system. This structure is constructed of 

both reinforced concrete and structural steel framing. In the vicinity of the 

structure, grade is located at El. 20'-0" and the rock line is at approximately El. 0'-

0". The portion of the structure at and below El. 21 '-0" consists of thick reinforced 

concrete walls and slabs. The exterior north, south, and west walls were poured 

against rock and the exterior east wall faces the Back River. As discussed in 

Section 4.3.4.1, the concrete portion of the pumphouse is screened out for the 0.3g 

review level earthquake. 

The 12-inch-thick concrete roof slab at El. 37'-0" is supported by structural steel 

beams and columns. The columns are typically anchored to the concrete floor at 

El. 21 '-0". The steel superstructure is enclosed by metal siding. Resistance to 

lateral seismic loads is provided primarily by diagonal bracing (Figure 5.4-5). The 

diagonal braces are built up from two channel sections arranged to form a T shape. 

A single diagonal brace is provided at each side of the building. 

Review guidelines are not established for steel framed structures. Consequently, 

the circulating water pumphouse steel superstructure was analyzed to determine its 

fragility. There are no Group A components attached directly to the steel framing. 

Damage to the Group A service water system components can only occur if the 

roof slab collapses. 

5-23 
c;::::;::;;:=:::::~~. r-.J~;: ===f~"~_,_~~ -(l_,,._) -\j'J'. ..;;,..,,._,_.;-.....::...--w+.---· =:.-_-_,_,.~,::.;~':'-: ...... :::::::::::;:::::-::=-::::1-L-~ -··~. 

. 

i 

~ :lC.REEH 

.:::~. A0ll£Y bM1 1 I:~:~~.: 1 

I 

9-s ........ 

~~~~~f.-.-.4--'-' ........ 1'!1 

I : t wt..n.A ....... ll, II jl, 

PUMP 

I I· I I II -.....4 

1

1

1 1):, ......... 

w ..........., ~. 

,, .._. 

..._, 

:I 

l 

.._. h·IS'·"'~ 

.'" ··' 

IO'·i:l· 

Figure 5.4-4 Circulating Water Pumphouse. 

5-24 
C9x13.4 

SECTION A-A 

Figure 5.4-S Circulatina Water Pumphouse, Typical Braced Frame. 

S-2S 
Seismic analysis of the steel superstructure above El. 21 '~0" was performed using a 

single-degree-of -freedom dynamic model. The concrete portion of the pump house 

was considered to be essentially rigid since it is built into rock and the shear walls 

are very stiff. Mass lumped at the roof level included contributions from the slab, 

steel framing, siding, and attached equipment. The roof slab was treated as a rigid 

diaphragm. Horizontal stiffnesses for the dynamic model were based on the 

stiffnesses associated with the braced frames. Seismic responses in the two 

horizontal directions were analyzed independent of each other since little coupling 

is expected. 

Fundamental frequencies in the two horizontal directions were both estimated to 

be about 3.7 Hz. Seismic input to the steel superstructure consisted of the specified 

review level earthquake ground response spectrum since the concrete structure is 

essentially rigid. Median damping was estimated to be approximately 15% based 

upon the recommendations of NUREG/CR-0098 for bolted steel structures at or 

near the yield point. Resulting overall seismic inertial loads were distributed to 

the braced frames in proportion to their relative rigidities. Accidental torsional 

moments were based upon a minimum 5% eccentricity. 

A number of potential failure modes were considered in the evaluation of the 

circulating water pumphouse, including the following: 

o Diagonal brace buckling 

o Column buckling 

o Diagonal brace connection failure 

o Column base connection failure 

o Roof diaphragm shear failure 

Buckling of the diagonal braces was found to be the initial failure mode. The 

results of testing on carbon steel compression members are compiled in [Hall, 1981 ]. 

This study derives empirical equations to determine average buckling stresses and 

associated variabilities. These data were used to estimate the buckling capacity of 

the diagonal braces and its logarithmic standard deviation. Inelastic energy 

absorption due to ductile response was not included since the structure is 

essentially elastic until buckling occurs. A HCLPF capacity of O.I9g was 

calculated for initial diagonal brace buckling. 

As previously noted, damage to the service water system components can occur 

only if the roof slab collapses. Known damage to steel framed structures in past 

earthquakes with estimated peak ground accelerations in the range of 0.5g is 

uncommon. There have only been a few instances of gross collapse of steel framed 

structures in actual earthquakes of magnitudes larger than expected for the Maine 

Yankee site. The overall evidence on the performance of steel framed structures 

in past earthquakes indicates that the collapse capacity should be at least 0.3g. 

Furthermore, the results of static and dynamic testing demonstrate that additional 

reserve capacity past buckling is provided by the following two sources: 

o Even in the buckled state, a steel member can still resist axial 

compression by developing a plastic hinge at midlength. 

Although the capacity degrades with increased axial 

displacement, significant displacements in excess of that at 

buckling are still attainable. 
''• 

o The reversing nature of the dynamic seismic input causes 

displacements to cycle rather than monotonically increase to a 

total loss of stability. 

Based upon past earthquake performance of steel structures and test data, the 

HCLPF capacity for collapse of the pumphouse steel structure is judged to be at 

least 0.30g; corresponding estimated fragility parameters are shown in Table 5.5-1. 

The turbine/service building (Figures 5.4-6 and 5.4-7) was built as a single integral 

structure. Most of the Group A components housed in this building are located in 

the control/switchgear building and the diesel generator enclosure. These portions 

of the structure are constructed of heavy reinforced concrete shear walls and slabs 

which have been screened out for the review level earthquake of 0.3g. 

Only the steel framed portions of the turbine/service building that could fail with 

resulting damage to Group A components were subjected to detailed evaluation. 

These areas included the following: 

o Service building floor at El. 39'-0", bounded by column lines 1/2, 

7, C, and F. 

o Service building roof at El. 61 '·0" 

o Turbine building floor at El. 35'-0" over the PCC/SCC system 

Approximate comparisons of initial load distributions and load carrying capacities 

indicated that the turbine building diagonal braces may buckle. However, even if 

this occurs, overall structural integrity of the turbine building will be maintained 

since lateral displacements will be limited by the massive concrete turbine 

pedestals once the one inch separation gap with the operating floor is closed. 

The portion of the service building floor at El. 39'-0" bounded by column lines 1/2, 

7, C, and F was evaluated since its failure may result in damage to the PCC lines 

running to the chillers. Resistance to seismic load in the N -S direction is provided 

by diagonal bracing, W36 building columns on Column Line C, diesel generator 

enclosure shear walls, and control/switchgear building shear walls. The capacity 

of the load path to the diesel generator enclosure is limited by the shear capacity 

of the adjacent roof deck. Because the El. 39'-0" floor is discontinuous at Column 

Line 7, the capacity of the load path to the control/switchgear building is limited 

by the weak axis bending capacity of the columns on Line 7 which must transmit 

E-W seismic loads to the floors at El. 35'-0"' and El. 45'-0". Accounting for the 

available load paths and their ductilities, a HCLPF capacity of 0.38g was 

conservatively estimated. 

The steel building roof at El. 61 '·0" over the chillers and the turbine building steel 

framing over the PCC/SCC components at grade both have relatively low masses. 

They were evaluated using conservative methods. Steel framing in these two areas 

were found to be capable of maintaining their integrity at peak ground 

acceleration levels well in excess of 0.3g. 

The steel framed structure within the main steam valve house provides support for 

the main steam lines before they enter containment and associated components. 

The structure consists primarily of platforms with metal grating supported by steel 

beams and columns. It is essentially independent from the exterior concrete 

structure with the exception of three horizontal struts spanning between the steel 

framing and the east exterior concrete wall. These struts were included as pipe 

supports only. 

5-27 
DIESEL 

GENERATOR 

ENCLOSURE 

NORTH c) 

SERVICE BLDG. ROOF EL 39' 

SERVICE .BLDG. FLOOR EL 39' 

TURBINE BLDG. MAZZANINE 

FLOOR EL 39' 

0 ® 

SERVICE BLDG FLOOR 0 

EL 35', EL 45'6" ......,._.....,_......0 

CONTROL 

BLDG. 

EL 35' 

EL 45' .... 6" 

tURBINE BLDG. 

MEZZANINE FLOOR 

EL 35' 

TURBINE PEDESTAL 

Figure 5.4-6 Turbine/Service Buildina, El. 35'·0" /EI. 39'-0". 

S-28 
NORTH 

c) 

CONTROL 

SERVICE BLDG ROOF BLDG. 

ROOF 

' ,. '1.', ·' 

TURBINE BLDG. OPERATING FLOOR 

I TURBINE I 

.. ' ' 

Figure .. 5.4-7 Turbine/Service Building, El. 61 '-0". 

5-29 

® 0 

0 
i "'' 

r r'. 

Resistance to lateral seismic load is provided primarily by diagonal bracing at each 

side of the structure. This bracing consists of either double angles or composite 

channel sections welded together to form .a T shape. During the first walkdown, it 

was noted that the diagonal brace for the north side of the structure at the lowest 

story was missing. This member was apparently removed due to interference with 

the access door. 

Reevaluation of the steel structure in the as-built condition was performed by 

Cygna. Results consisted of overall seismic loads from the dynamic model and 

localized load distributions to individual structural members from more detailed 

static models. The following potential failure modes were evaluated using 

conservative methods: 

o Column failure under combined axial load and bending 

o Diagonal brace buckling 

o Brace connection failure 

o Column base connection failure 

o Beam failure due to biaxial bending 

o Pullout of pipe support anchorage 

Particular attention was given to members in the vicinity of the m1ss1ng brace 

since their loads may be increased to accommodate the load path discontinuity. 

The HCLPF capacity for this structure was found to be greater than 0.3g based 

upon conservative methods. 

5.4.1.3 Block Walls 

A list of block walls supporting or adjacent to Group A components is presented in 

Chapter 7. These walls are separated in to two categories: 

o Walls that were screened out on the basis that their failure will 

not damage Group A components. 

o Walls for which HCLPF capacities were determined. 

Evaluation of the Group A walls did not identify any with HCLPF capacities less 

than 0.3g. This can be attributed to the seismic retrofits already installed or to be 

added during the next outage. Verification that the block wall capacities are 

greater than 0.3g was performed either by determining the actual fragilities or by 

utilizing simplified, conservative analyses. As an example, the fragility evaluation 

of Wall SB 35-3 is described in the following. 

Wall SB 35-3 (Figure 5.4~8) is located at the west side of the service building 

battery room at El. 35'-0". Out-of -plane collapse of this wall will result in damage 

to Battery Group 3 mounted on the floor next to the wall. It is built of 12-inchthick, hollow block units which are assumed to be normal weight. The wall is 7'-0" 

high by 19'-8" wide. It is mortared up into the overhead concrete beam. 

Original block wall construction specifications for Maine Yankee were available. 

Concrete block units were specified to meet ASTM C90 requirements. The mortar 

was specified to meet C270, Type M or N requirements. Testing on a limited 

number of unit and mortar samples taken from the battery room at El. 45'-6" 

determined the following average compressive strengths: 

o Block unit, f m= 3800 psi on the net area 
a. Looking from the Battery Room 

. ~ ' \~ .... ·. f; 

b. Looking from the Cable Spreading Room 

Figure 5.4-8 Block Wall SB35-3. 

5-31 
o Mortar, m0 = 2700 psi 

The mortar strength would imply the use of Type M mortar at the El. 45'·6" battery 

room. Because use of Type M mortar throughout the plant could not be confirmed, 

Type N was assumed for other walls. An average Type N mortar strength of 825 

psi was estimated by scaling the results from the· test samples. 

This wall was analyzed to determine its resistance against out·of ~plane seismic 

response. Because of its low height-to-width ratio, it was analyzed as a one-way 

member spanning vertically. The top and bottom boundary conditions were 

represented as simple supports. The fundamental out-of -plane response frequency 

was calculated using uncracked stiffness properties and found to be into the rigid 

response range of the floor spectra. · The maximum stress due to out-of-plane 

response was based upon inertial loads due to the floor ZPA and the elastic section 

modulus. This stress was combined with those due to in-plane and vertical seismic 

inertial loads by SRSS. Additional in-plane loads may be developed by relative 

structure displacements between the top and bottom of the wall. However, these 

loads are stiffness dependent and will tend to dissipate once the wall cracks. 

Relative structure displacements are small and are considered insufficient to cause 

wall collapse. 

Initial cracking was evaluated by comparing the calculated elastic stress against 

the median mortar modulus of rupture, including additional resistance provided by 

dead load compression. Conservative allowable stresses for use in the design of 

masonry structures are presented in ACI 531. The commentary to ACI 531 notes 

that factors of safety in masonry are generally held at a minimum of three. 

Review of available test data indicated that the average factor of safety is greater 

than this value. The median cracking stress was estimated to be three times the 

ACI 531 allowable value to account for potential reduction in capacity due to 

differences between laboratory and field workmanship. A relatively large 

coefficient of variation of 0.3 was assigned to the cracking stress to account for 

scatter in the test data, workmanship effects, etc. Inelastic energy absorption 

effects were not considered since behavior prior to cracking is essentially linear. 

A HCLPF capacity of 0.60g was estimated for initial cracking of Wall SB 35-3. 

Even if the wall cracks through at· midheight, stability can be maintained by 

arching action which has been exhibited in testing of confined walls. The wall 

will tend to behave as two rigid bodies pivoting about the midheight crack. This 

rotation causes the wall to push outward against the top and bottom supports, thus 

developing comp_ression stresses at the points of contact which oppose the 

transverse loading. The resistance developed by arching action may be limited by 

compression or shear in the wall, or by bending of the floor beams above and 

below. An approximate analytical model was developed to account for arching 

action. While this model was not correlated against test results, it demonstrated 

that, even using very conservative assumptions, this wall could potentially develop 

a HCLPF capacity well in excess of 0.3g through arching action. 

5.4.1.4 Flat Bottom Storage Tanks 

The following flat bottom storage tanks were evaluated: 

o Refueling water storage tank, TK-4 (R WST) 

o Demineralized water storage tank, TK-21 (DWST) 

5-32 
.. ~ J... ' ' 

~. ' ) 

o Primary water storage tank, TK-16 (PWST) 

o Spray chemical addition tank, TK-54 (SCAT) 

While the SCAT is not actually part of the Group A systems, its failure could 

result in the failure of interconnecting piping to the R WST. Fragility parameters 

for the RWST, DWST, and PWST are shown in Section 5.5-1. Calculated HCLPF 

capacities for these tanks are listed in Chapter 7. These capacities were 

determined using essentially the same methodology for all tanks. A detailed 

description of the evaluation of the R WST follows. 

An overall view of the R WST and a typical anchor bolt detail are shown in Figure 

5.4·9. The R WST has an inside diameter of 40'-0". The shell has a total height of 

38'-0" to the springline and varies in thickness from 3/16 in. to 3/8 in. The 

spherical shaped head has a total rise of 6'-10" and is 5/16 in. thick. The bottom 

plate is 1/4 in. thick. The shell, head, and bottom plate are fabricated from A240 

Type 304 stainless steel. The R WST is anchored to its concrete ring wall 

foundation by a total of 8 anchor bolts spaced at 45 degrees. These bolts are 2 

inches in diameter and are fabricated from A307 steel. They are attached to the 

tank shell by chairs built up from stainless steel plate (Figure 5.4-9). 

Seismic response of the RWST was calculated following the recommended approach 

described in NUREG/CR-1161. The following response modes were included: 

o Horizontal impulsive 

o Horizontal convective (sloshing) 

o Vertical fluid 

o Vertical tank 

Effective masses and mass centroids to determine overall tank seismic loads due to 

the horizontal impulsive and convective modes were calculated from equations in 

NUREG/CR-1161. The impulsive mode fundamental frequency of 5.6 Hz was 

based upon coefficients developed by [Haroun, 1981]. The frequency of the 

convective mode was found using equation for rigid tanks. The median impulsive 

mode damping of 7% was based upon the recommendations of NUREG/CR-0098 

for welded steel structures at or near the yield point. Damping of 0.5% was 

assigned to the convective mode. Overall tank seismic loads due to the impulsive 

and convective modes were found by factoring the effective masses, mass centroid 

heights, and spectral accelerations corresponding to the modal frequencies. 

Contributions from the two horizontal modes were then combined by SRSS. Tank 

response due to vertical excitations accounted for amplified fluid response 

associated with tank shell flexibility in the breathing mode and vertical response 

of the tank itself . 

Failure of the RWST is controlled by buckling of the shell due to overall seismic 

moment at the tank base. Conventional design practice for anchored tanks utilizes 

a stress distribution at the tank base derived from elementary beam theory. In this 

stress distribution, plane sections are assumed to remain plane with the neutral axis 

at the tank centerline under pure bending. This distribution results in unrealistic 

seismic capacities and is considered to be overly conservative. When the vertical 

5·33 
Vt 

I 

w 

+;.. 

a. Overall View b. Anchor Bolt Chair 

Figure 5.4-9 Refuelin£ Water Storage Tank. 
force provided by the tank weight is overcome by the base moment, the tank will 

uplift and mobilize anchor bolt resistance. Because the anchor bolts are not as 

stiff as the shell, the neutral axis will shift toward the compression side to 

maintain equilibrium between the applied and resisting forces. Also, if the anchor 

bolts or their chairs can deform in a ductile manner, additional load redistribution 

will occur as uplift increases and anchor bolts around the tank perimeter are 

progressively stressed past yield. This behavior is acceptable since ductile 

deformation of the anchor bolts or their chairs will not directly cause a loss of 

tank contents, unless local deformations become excessive. 

To more accurately determine a realistic tank capacity against overall seismic base 

moment, the analytical model shown in Figure 5.4-10 was developed. Tank 

deformations at the base and at the elevation through the top of the anchor bolt 

chairs are assumed to be linearly distributed. This is appropriate since the 

foundation and the portion of tank above the top of the chairs are stiff compared 

to the anchor bolts. The distribution of forces at the base of the tank resulting 

from this assumption is shown in . Figure 5.4-10. Compressive strains in the shell 

are found by distributing the displacements over the height of shell between the 

tank base and the top of the anchor bolt chairs. Shell compressive stresses are 

found by factoring the strains by the elastic modulus. So long as the anchor bolts 

or their chairs are elastic, anchor bolt strains are determined by assuming that the 

total uplift displacement results in a uniform strain over the length of the bolt 

from the top of the chair to the anchor plate embedded in concrete. 

Uplift of the tank base also mobilizes some additional resistance associated with 

the contained fluid bearing on the tank bottom plate. The fluid hold-down force 

provided by the tank bottom plate· is accounted for in the design of unanchored 

tanks by modeling the plate as a beam subjected to a vertical displacement at one 

end and uniform pressure along its length by the fluid weight. A plastic 

mechanism is assumed to develop in this beam. While this representation is 

considered to be conservative for unanchored tanks, it may be unconservative for 

an anchored tank whose uplift displacements may not be sufficient to develop this 

plastic mechanism. Accordingly, the fluid holdown force was based on the same 

beam representation for unanchored tanks described in [Wozniak, 1978], but limited 

to elastic behavior only (Figure 5.4-11). The magnitude of the hold-down force 

varies around the tank perimeter according to the distribution of uplift 

displacements. The overall tank base moment capacity is found by limiting the 

maximum shell compressive stress to the value causing buckling and solving for 

force equilibrium between the applied loads and resisting forces. 

Design buckling stresses for cylindrical shells subjected to bending moment were 

developed in [NASA, 1966]. The NASA design approach was adapted to predict the 

median RWST buckling stress. The design buckling stress coefficients in [NASA, 

1966] are values having a 90% confidence of exceedance. Based upon buckling 

stress data reported in [Weingarten, 1965] and [Gerard, 1957], the design 

coefficients for buckling due to bending are estimated to have a median factor of 

safety of 1.4. Increase in the tank buckling stress due to internal pressurization by 

the contained fluid was included. Stainless steel does not have a sharply defined 

yield plateau, and instead exhibits a rounded stress-strain curve. To account for 

potential inelastic buckling, the plasticity correction factor was derived from a 

Ramberg-Osgood-Hill representation of the material stress-strain curve. 

The tensile capacity of the anchor bolts may be limited by any of the following: 

5-35 
~ Extcrna l Loads 

- I I· 

. ~r--- --- . ~--··=-en--·.. . -··-·--· .. -·-·-···-- --~ ---·- . --- ' . - : ~ 

j 

' I 

; Neutral Axis 

a. Assumed Displacement Disttibution 

b. Distribution of Shell Compressive 

~~ Str~sses 

J ! p :S Anchor Bolt/Bolt Chair Yield Force 

c. Distribution of Anchor Bolt Forces 

d. Distribution of Fluid Holddown Force on Tank Bottom Plate 

Figure 5.4·10 Model for Determination of Tank Resistance Against 

Seismic Base Moment. 

5·36 
WEIGHT OF WATER 

l 1 1 I l l l 

r REACTlON ON TANK SHELL 

TANK SHELL 

BOTTOM PLATE 

UPLIFT DISPLACEMENT POINT OF CONTACT WITH GROUND 

~---------~~ ~ 

c. BOTTOM R.. DEFORMED SHAPE 

b. BOTTOM R. MOMENT DIAGRAM 

.'··: 

Figure 5.4-ll Beam Model of Tank Bottom Plate. 

5-37 
o Anchor bolt yield 

o Anchor bolt fracture at the threads 

o Anchor bolt pullout from the concrete foundation 

o Yield of the tank shell locally at the anchor bolt chair 

o Bending of the anchor bolt chair top plate 

o Failure of anchor bolt chair welds 

o Stability of the anchor bolt chair stiffener plates 

Based upon a review of these items, capacity of the anchor bolts was found to be 

limited by bending of the anchor bolt chair top plate. Under applied loading 

imposed by the anchor bolt, the top plate is subjected to bending as it spans 

between its transverse supports consisting of the vertical stiffener plates and the 

tank shell. The available resistance was based upon yield line theory with yield 

lines developing as shown in Figure 5.4~12. At the supports, hinge moment 

capacities are governed by the bending capacities of the vertical stiffener plates 

and the tank shell since they are thinner than the top plate itself. 

A240 Type 304 stainless steel has a minimum specified yield strength at 0.2% offset 

of 30 ksi and a minimum specified tensile strength of 75 ksi. The minimum 

elongation is specified to be 40%. Based upon values reported by [Smith, ASME] 

median yield and tensile strengths are estimated to be 37 ksi and 84 ksi, 

respectively. Because stainless steel does not exhibit a sharply defined yield 

plateau characteristic of carbon steel and because of the large strain ductility 

a vail able, derivation of plastic moment capacities from the yield strength would be 

overly conservative. Instead, an effective yield stress was estimated. The momentcurvature relationship was derived for a flat plate using a Ramberg-Osgood-Hill 

fit to the median strength properties. From a review of the resulting curve, an 

effective yield stress of 69 ksi was judged to provide a reasonable approximation 

to the moments at the relatively high curvature demands to which the top plate is 

likely to be subjected. The median anchor bolt tensile capacity as limited by 

bending of the chair top plate was estimated using this effective yield stress in 

conjunction with the yield line representation. 

Following this methodology, the RWST was found to have a HCLPF capacity of 

0.21 g against shell buckling. This capacity included a median system ductility of 

l. i to account for some limited benefit of inelastic energy absorption due to 

nonlinear response of the tank as the anchor bolt chair top plates yield around the 

perimeter. Because the seismic response and capacity of the tank is the same in 

any horizontal direction, the prescribed ground spectrum which is the larger of the 

two horizontal components is the one that will lead to tank failure. Consequently, 

values for Fso and~ lfJ of 1.0 and 0, respectively, were used. Other potential tank 

failure modes considered and found to have greater capacity than shell buckling 

included sliding at the base and yield of the shell due to hoop stress induced by 

hydrostatic and hydrodynamic pressure. 

The other flat bottom storage tanks were evaluated using essentially the same 

methodology. Although the DWST is smaller than the R WST and has the same type 

of anchorage, its HCLPF capacity is lower because it is fabricated from B209~5052· 

0 aluminum. The PWST is similar to the DWST, but it is anchored by twenty 

anchor bolts, one inch in diameter. 

5-38 
.·· I\ 

/ 

/ 

I 

I 

) 

YIELD LINES 

a. YIELD LINE PATTERN 

--- - -- - --- - - --- --- - - _/, 

y 

b. DEFORMED SHAPE 

I I 

I I 

I I 

I I 

: I 

1 I 

I I 

I I 

--At--

Figure 5.4-12 Yield Line Model of Anchor Bolt Chair Top Plate. 

5-39 
5.4.1.5 Inverter 

The Maine Yankee inverters are located in the protective switchgear room of the 

turbine/service building at elevation 45'~6" (25 ft. above grade). The inverters are 

12 8 kV A static inverters manufactured by Solid State Controls Inc. (SCI). They are 

constructed of an all welded steel tubular frame with bolted shear panels on three 

sides and bolted doors on the front. The back panels have ventilated slits to 

provide cooling for the inverter internals. SCI provided the initial qualification of 

the inverters by comparing the Maine Yankee 12~kVA units to a larger 15·kVA 

unit shake table tested as part of the qualification program for the Three Mile 

Island Nuclear Station (reference Maine Yankee FSAR Amendment No. 35, Volume 

II). The tested 15-kVA inverter was similar in configuration and size to the Maine 

Yankee inverter, but approximately 1000 pounds heavier. Specific test data was 

not provided such that a fragility evaluation could be extracted from the tests. 

The results however, did confirm that the critical failure mode of the inverter is 

its anchorage. The test results further confirmed the high capacity of the internals 

and attached components. The inverters were also compared to the GERS, October 

1986 draft which supported the capacity of the internals and attached components. 

The Maine Yankee inverters are one electrical component which has unpergone an 

anchorage upgrade since its original installation. The original installation (which 

is still present) bolted the inverter to four base channels which were in turn was 

expansion bolted to the supporting floor. In 1983 Maine Yankee installed an 

anchorage upgrade to the inverters increasing the capacity. Figure 5.4~13 shows 

two photographs of the inverter and anchorage addition. A conservative static 

analysis was performed by Maine Yankee in evaluating the anchorage addition. To 

arrive at a more realistic capacity, a fragility evaluation was performed in the 

present study using available data and as·built details collected during the 

walk downs. 

The anchorage of the inverter to the concrete floor was identified as the critical 

failure mode. The failure mechanism was defined as the existing anchorage in 

combination with the anchorage upgrade. 

Capacity Factor. The failure of the existing anchorage to the concrete floor was 

determined to be the most critical failure mode. A flexibility study was performed 

to demonstrate the existing anchorage was much stiffer than the anchorage 

upgrade. Calculations determined the existing anchorage was 99% stiffer than the 

anchorage upgrade. The primary area of flexibility found in the anchorage 

upgrade was the long extension of the angle required to clear the base channels in 

order to bolt the entire assembly to the floor. 

The method used to arrive at the inverter capacity consisted of computing two 

bounding cases for anchorage failure: 

o Lower bound: The existing 1/2 in. diameter Red Head self· 

drilling anchors resist the entire seismic load. 

5-40 
,"•. 

Figure 5.4-13 Maine Yankee 12 kVA Inverter and Battery Charger, and a 

Close-up of the Anchorage Addition. 

5-41 
o Upper bound: The existing Red Head anchors combined with the 

anchorage upgrade fail at the same time. 

The actual failure mechanism will lie between these two bounding cases. The 

theorized failure mechanism will be the existing anchors loading up first since this 

load path is more stiff. ·However, some slippage in the existing anchors will occur 

as ultimate load is reached. This condition will begin to transfer more load into 

the anchorage upgrade as the displacement increases until failure occurs. It is 

difficult to ascertain how much slippage (thus load transfer) will occur; 

consequently, the second bounding condition was assumed. The median failure 

capacity was assumed to lie more closely to the lower bound, since this is the 

stiffer load path. Consequently, median capacity was assumed to be defined by the 

following: 

0 

0 

Aupper represents a +3~ on Am 

A lower represents a -1~ on Am 

(5.4-17) 

(5.4-18) 

Strength Factor. The inverter was modeled as a single-degree-of -freedom system 

where the strength factor was computed using static analysis techniques. The 

turbine/service building floor spectra at 7% damping generated from the 0.18g 

NUREG/CR-0098 50th percentile ground response spectra was used to obtain the 

spectral accelerations at the inverters fundamental frequency. 

A conservative horizontal fundamental frequency was calculated using only the 

welded frame and neglecting the bolted shear panels. This yields a lower bound on 

the median horizontal fundamental frequency. A horizontal fundamental 

frequency of approximately 13 Hz was computed. This was used for both 

horizontal directions as the side to side direction was not believed to be 

significantly stiffer. The rear shear panels were observed to have ventilating slits 

and the front doors had cut-outs for instruments. 

The vertical fundamental frequency of the inverters was judged to be greater than 

30 Hz. Vertical response spectra was not developed for points in the floor slab. In 

or~der to account for the amplification of the floor spectra at points in the slab, the 

amplification factors developed from a study by [Structural Mechanics Associates, 

1983] were used. The amplification factors are based on the the floor vertical 

fundamental frequency. A conservative floor vertical fundamental frequency of 

18 Hz was computed for the slab in the protective switchgear room which yielded 

an amplification factor of 1.5. The factor was applied to the ground response 

spectra at the inverters vertical fundamental frequency to arrive at the spectral 

acceleration. Since the inverters vertical fundamental frequency was judged 

greater than 30 Hz, the 1.5 factor was applied to the 0.18 ZPA of the ground 

response spectra. 

The following presents the strength factor computed for the inverter HCLPF 

capacity determination based on the the combined failure mode of the existing 

anchorage and the anchorage upgrade. The strength factor discussion below 

presents the strength factor determination for the existing anchorage, the 

anchorage upgrade, and the combined strength factor resulting from both. 

5-42 
\: 

Existing anchorage strength factor. The existing anchorage consists of the 1/2 in. 

Red Head self ~drilling· anchors bolted through the inverter base channels into the 

concrete floor. The lower Qound strength factor determination assumed this 

failure mode to transfer the entire seismic loading. The seismic loads were applied 

to the C.G. of the inverter assumed located at the geometric center of the cabinet. 

This is slightly conservative because the majority of the inverter weight is bolted 

to the bottom base framing (the heaviest component, transformer, was observed to 

be positively bolted to the inverter base framing). The seismic loadings were 

applied in the three orthogonal directions considering 100-40-40 as median centered 

to result in the most severe loadings. One hundred percent of the east-west seismic 

accelerations yielded the highest bolt loads. The loads were resolved using 

standard static techniques to obtain bolt shear and tension loads. 

Criteria for determining the strength factor for expansion anchors was that 

recommended by [URS Corporation/John A. Blume & Associates, Draft 1986]. 

Shear-tension interaction was based on the quadratic formulation as the inverter 

bolt loadings were in the area of high tension and low shear (V /V rna <0.4). In 

this range of high tension and low shear the quadratic and the linear formulation 

produce negligible differences in the results. The median anchor bolt values 

tabulated in the Blume report were considered median. These values were used 

only if the Blume criteria was satisfied regarding edge distance requirements and 

anchor spacing. The uncertainty was computed based on a factor of 2 on the 

median values as representing a 95% 'confidence level. 

Anchorage Upgrade Strength Factor. The critical failure mode identified for the 

anchorage upgrade was shear in the fillet weld attaching the angle to the inverter 

base channels. The fillet weld is heavily stressed due to the configuration of the 

anchorage upgrade. The angle attaching to the inverter base channels extends out 

away from the inverter, thus creating a long lever arm which induces shear in the 

weld. 

The strength factor for the weld was computed based on considering median shear 

strength as 70% of the weld materials ultimate tensile capacity. Median ultimate 

tensile capacity for material strength was judged to be 20% higher than the code 

value. 

Combined Strength Factor. The combined strength factor considered both of the 

anchorage systems as resisting the seismic loadings. As discussed previously, the 

existing anchorage is stiffer than the added anchorage upgrade; consequently, will 

begin to transfer the loading initially. After slippage occurs in the existing 

expansion anchors, the anchorage addition will begin to share in the load transfer 

until idealistically both anchorage systems will fail simultaneously. This represents 

an upper bound on the actual failure mechanism. The strength factor for this 

failure mode was computed by adding the individual strength factors computed 

previously. 

The realistic failure mode lies between the lower bound, the existing anchor bolts, 

and the upper bound where the combined anchorage systems fail together. The 

median strength factor was computed from Eqs. (5.4-17) and (5.4-18). 

5-43 
Ductility. Inelastic energy absorption was considered as part of the strength 

factor. Median ultimate material strength properties were used in computing the 

strength factor. The inverter overall capacity factor, random variability and 

uncertainty computed are: 

Fe = 12.26 

fJi = 0.00 

~u = 0.46 

Eauipment Resoonse Factors 

Qualification Factor. The inverters respond in a single-degree-of-freedom manner. 

The static analysis used to compute the strength factor was considered median 

centered with no uncertainties or random variabilities. 

Spectral Shape Factor. The unsmoothened and broadened floor response spectra 

were not provided for the margin study. There is typically conservatism, 

particularly in the amplified regions of the spectra, in the smoothing and 

broadening. No attempt was made to recover this conservatism; however, 

conservatism in the development of the synthetic time history for input to the 

dynamic structural models for the computation of floor response spectra did occur. 

Cygna performed the structural modeling and subsequent floor response spectra 

generation. A synthetic time history was developed such that the response spectra 

enveloped the 0.18g NUREG/CR-0098 50th percentile ground response spectra. 

Cygna computed the statistical results of the conservatism in the enveloped spectra 

over the NUREG spectra. This conservatism was assumed to be included in the 

floor response spectra. The spectral shape factor used and associated random 

variabilities and uncertainties derived from the statistical results are: 

Fss = LIS 

~R = 0.066 

l3u = 0.00 

Damping Factor. Seven percent damping was considered median centered for 

welded and bolted structures. The inverter analysis used 7% damping. Five 

percent damping was considered as representing a minus one beta uncertainty on 

the median. The random variability was considered to be two tenths of the 

uncertainty. The resulting values used are: 

F0 = 1.0 

~R 0.01 

fJu = o.os 

Modeling Factor. A median horizontal fundamental frequency was computed in 

determining the spectral accelerations for the inverter analysis. To establish the 

uncertainty on the computed median frequency a range of frequencies was 

considered. A range of 10 to 20 Hz was considered in order to compute the 

uncertainty. The highest spectral acceleration in the frequency range was 

considered to represent a 95% confidence on the median spectral acceleration. The 

5-44 
random variability was considered to be equal to zero. The modeling factor and 

the associated random variability, and uncertainty are: 

1.0 

= 0.00 

== 0.17 

Mode Combination. The inverter was modeled as a single-degree-of -freedom 

system. There is some uncertainty as to the contribution of higher modes to the 

overall response of the inverter not considered in the simplified static analysis. An 

uncertainty of 0.10 was assumed to reasonably represent the contribution of higher 

modes. The mode combination factor, random variability, and uncertainty are: 

FMC 1.0 

f3R = 0.00 

l3u = 0.10 

Earthquake Component Combination Factor. 100-40-40 rule was used to compute 

the strength factor. This was considered median centered. Random variability 

based on past PRA studies, considering median coupling and both horizontal 

directions contributing to failure equal to 0.10 was used. The factor, random 

variability, and uncertainty are: 

FECC = l.O 

{jR = 0.10 

(ju = o.oo 

The combined equipment response factors for the inverter are: 

FER 1.15 

f3R 0.12 

f3u = 0.20 

Structural Response Factors 

The structural response factors for the turbine/service building included: 

0 

0 

0 

0 

0 

Soil-structure interaction 

Modeling 

Damping 

Spectral shape 

Directional effects affecting equipment response 

Refer to Section 5.4.1.1 for a discussion of the structural response factors. The 

combined structural response factors used for the inverters are: 

5-45 
FRS = 1.35 

f3R = 0.25 

~u 0.24 

Inverter Ground Acceleration Capacity. Combining the capacity factor, equipment 

response factor, and the structural response factor the resulting median ground 

acceleration capacity and HCLPF capacity for the inverter was computed as: 

Am 

PR 

Pu 

= 3.43g 

= 0.28 

= 0.59 

HCLPF Capacity = 0.82g 

5.4.1.6 Diesel Fuel Oil Day Tank 

The Maine Yankee diesel fuel oil day tank is a horizontal cylindrical tank 

supported about 5 feet off the floor by a braced frame support structure. The day 

tank is located at grade in the diesel genera tor building. Figure 5.4-14 shows 

photographs of the tank taken during the first walkdown. The critical failure 

mode evaluation for the tank was based on a previous finite element analysis 

[lrnpell, 1985]. The results of that analysis identified the seven most critical areas 

of the tank/support system and compared the resulting seismic stresses to ASME 

Code allowables. Table 5.4-1 specifies the factor of safety (allowable stress divided 

by the seismic stress) for the seven critical areas assessed. Figures 5.4-15 and 5.4-16 

are outline drawings of the day tank and show locations for these seven critical 

areas which were analyzed. The most critical day tank failure mode is shown from 

the factors of safety to be the anchor bolts, while the second most critical failure 

mode is the one-half-inch bolts on the 2" x 2" x 1/4" bracing. The demonstrated 

factor of safety on the bracing bolts is over three times the factor of safety on the 

anchor bolts. Thus, the anchor bolts are judged to be the primary failure mode 

and the contribution to risk of all other failure modes is judged to be negligible 

-due to their much higher capacity. It should be noted that in addition to the areas 

analyzed in the lmpell analysis, the fuel piping (threaded) exiting the day tank 

bottom was evaluated for displacement induced loads and found to have a 

relatively high seismic capacity. 

Capacity Factor 

The capacity factor for the day tank anchor bolts is determined by the ratio of the 

bolt ultimate failure threshold to the response at the bolt due to the review level 

earthquake. The capacity factor and its associated variability are described in 

detail in Section 5.4.1.1. The ultimate failure threshold is based on the anchor bolt 

test data presented in [URS, 1986]. Close inspection of the Maine Yankee day tank 

anchor bolts revealed that the expansion anchors utilized for the original plant 

design were not fully embedded into the floor slab (see Section 4.4 for complete 

description). Maine Yankee engineers devised a retrofit to ensure that the full 

strength of these anchor bolts would be developed by installing eight new Hilti 

K wik bolts (two per leg) as shown in Figure 5.4-17 (On one tank only six bolts 

could be installed; The calculated capacity is still greater than 0.3g). The day 

tank fragility is based on this retrofitted condition. 

5-46 
Day Tank Support Legs 

and Original Anchor Bolts 

Day Tank -with Attached Piping 

and Splash Shield 

Figure 5.4-14 Diesel Fuel Oil Day Tank (TK-62B). 

5-47 
.·.·,, 

0 1/2" BOLTS 

(FOUR PLACES) 

TWO 1/2". 

(;\ANCHORS PER 

\.!._) LEG 

~ 

1 /4" FILLET® 

1/4"· ANGLE 

I 0 1/4 FILLET WELD 

x - dir 

c) 

FOUR 3/8" 

BOLTS PER LEG 

2 1/2" NOMINAL PIPE 

* FAILURE MODES CIRCLED CORRESPOND TO THOSE ON TABLE 5.4-1 

Figure 5.4-15 Diesel.Day Tank Critical Areas. 
' ... ~ ( •' \ 

(.;\ TWO 1 /2" ANCHORS 

\.!_) PER LEG 

C) 1/4" FILLET WELD 

0 FULL PENETRATION WELD 

~ 15\. FOUR 3/8"/J BOLTS 

\:_) PER LEG 

2 1/2" NOMINAL PIPE 

*FAILURE MODES CIRCLED CORRESPOND TO 

THOSE ON TABLE 5.4-1 

Figure S.4·16 Diesel Day Tank Critical Areas. 

S-49 
FUEL OIL DAY TANK (TK 62A,B) 

NEW 1/2"~ HILT! IN EXISTING HOLE (TYP.) 

NOTE 3 

EXISTING 1 /2"~ STUD & NUT 

~ SUPPORT LEG GROUT(DEPTH VARIES) 

Q SUPPORT LEG 

1/2"Fl 

(BOLT HOLES AT 90°, AZMITHAL ORIENTATION ABOUT 

SUPPORT LEG VARIES) 

ob ~ 

·~·, I _L MIN EMBEU 

I 

Figure 5.4-17 Retrofit Anchorage Design for the Day Tank. 

5-50 
Table 5.4·1 Diesel Fuel Oil Day Tank Critically Stressed Area 

Critical Area 

1. Welding on 1 1 /2" Pipe Braces 

2. 1 /2" Diameter Bolts at 2"x2"x 1 /4" 

Longitudinal Bracing 

3. Fillet Weld Attaching 2"x2"xl/4" Bracing 

to the Center Stiffener Plate 

4. Weld of Tank to Saddle Plates 

5. Bolting of Vertical 2 1/2" Pipe Posts 

to the Saddle Plates 

6. Welds at top of Tank Saddle 

7. Anchor Bolts 

* Conservatively based on a 1/4" fillet weld instead of 

the actual full penetration weld. 

5-51 

Factor of Safety 

5.95 

3.28 

High 

High 

2.44 

7.14* 

1.03 
' ·, 

I ~ ' 

The individual median capacities of the Hilti Kwik bolts are reported in [URS 

1986] to be 5.97 kips in tension and 8.80 kips in shear. These values are reportedly 

based on the 2.25-inch minimum embedment for these bolts. Although the Maine 

Yankee day tank anchors have a specified minimum embedment of 4 inches, the 

2.25-inch embedment strengths are conservatively judged to be applicable for this 

situation. The reason for the use of the 2.25-inch embedment data is that the 

relatively close proximity between anchors (4.5 inches between centers) on each day 

tank leg will lead to some degradation of the rated strength as embedment goes 

past this 2.25-inch embedment depth. Although the derated strength of a four inch 

embedded bolt in this situation is greater than the 2.25-inch embedment strength, 

the 2.25-inch embedment strength has conservatively been assumed for this 

fragility analysis. 

The uncertainty on the median tensile and shear strength values stipulated above is 

quantified by estimating the 95% probability of exceedance level to be at 50% of 

the median values. This factor of one-half was judged appropriate as a method· of 

quantifying the various uncertainties associated with expansion anchor capacities 

(i.e., concrete pours, installation procedures, drill tolerances, dynamic loads and 

material properties). The shear and tensile loads have been combined using· the 

method recommended in [URS, 1986]. Section 5.4.1.5 describes this interaction 

equation for the Inverter fragility description. 

Response Factors 

The day tank natural frequencies were rederived from those presented in the 

[lmpell, 1985] analysis because the weight of the tank contents was neglected in the 

frequency calculation. This error is significant in calculating the tank frequency 

and subsequently in deriving the seismic accelerations since the weight of the 

contents is much greater than the weight of the tank itself. The actual tank 

frequencies shifted out of the ZPA region and into the amplified section of the 

spectra (14 Hz and 17 Hz). Spectral accelerations for all three earthquake 

components were based on the most recent floor spectra (0.18g NUREG/CR-0098) 

scaled to the 0.3g review earthquake level. Seven percent median damping was 

used with 5o/o as the 95% probability of exceedance value. Earthquake components 

were combined using the median centered 100%, 40%, 40% technique. The 

remaining equipment and structural response factors were calculated using the 

methodology presented in Section 5.4.1.1. 

Ground Acceleration Capacity 

The median ground acceleration capacity of the day tank was calculated to be 

1.4lg based on the failure of the anchor bolts. The HCLPF capacity is calculated 

to be 0.43g. Since this HCLPF capacity is greater than the 0.3g review level 

earthquake for Maine Yankee, the conservative assumption of 2.25-inch embedded 

anchors is considered acceptable. 

5.4.1.7 Containment Spray Fans 

The Maine Yankee containment spray fans are located in the ventilation equipment 

room at elevation 21'-0" (grade). The fans are Westinghouse Silent Vane, size 8030, 

5-52 
13500 cfm units. The. fan units are supported from vibration isolators with seismic 

restraints installed in all directions. The seismic restraints are to be installed 

during the next Maine Yankee refueling outage (see Section 4.4). Figure 4.4-8 and 

Figure 4.4-9 show photographs of the fans in the unmodified condition and a 

sketch of the proposed seismic restraints, respectively. . The fragility evaluation 

was based on this modified condition. The entire fan assembly is anchored to a 

reinforced raised concrete pad. 

The anchorage of the fans was identified as the critical failure mode to investigate 

and will be. presented below. However, another possible failure mode was 

identified during the walkdown. The fan units have cantilevered shafts. It was 

postulated that the cantilevered fan blade on the end of the shaft could fail or 

cause premature bearing failure as a result of seismic excitations. The concern for 

the cantilever shaft on the containment spray fans was alleviated based on 

numerous similar configurations which had survived large 111agnitude earthquakes. 

Several vendors also verified the fact that this cantilever variety fan is very 

common and represents approximately 50% of the fans used in power plant 

applications. These cantilever fans are designed to withstand the high operating 

loads, and, the vendor felt the safety factor on the operating loads was more than 

adequate to offset the seismic load. 

The containment spray fans anchorage HCLPF capacity was computed using a 

conservative simplified analysis as outlined in Section 5.2. Several conservative 

assumptions were used in the analysis to show a high HCLPF capacity. These 

included: 

o The existing anchorage was conservatively assumed to carry no 

loads. 

o Peak spectral acceleration values were used since the vibration 

isolation system frequency was not known. 

o Conservative dimensional data for the anchorage modification 

was used. Only a conceptual sketch was provided to the fragility 

team by Maine Yankee for the anchorage modification. Figure 

4.4-9 shows the sketch received from Maine Yankee for the 

anchorage modification. This modification will be verified 

during the third walkdown. 

o Overturning was assumed to be resisted by the outer anchor bolts 

alone. 

Even with these conservative assumptions the fragility evaluation for the 

containment spray fans sh.owed a HCLPF capacity greater then 0.5g pga which is 

greater than the margin earthquake for Maine Yankee. 

5.4.2 CDFM Method 

For the purpose of comparison, HCLPF capacities were determined using the 

CDFM method for the following selected components: 

5-53 
o Circulating water pumphouse 

o Block wall SB 35~3 

o Refueling water storage tank 

Application of the CDFM method was based on the recommendations in 

NUREG/CR-4482. HCLPF capacities found by the CDFM and fragility analysis 

methods are compared in the following discussion. One major difference between 

the two methods common to all of the examples is the use of the median 

NUREG/CR-0098 ground response spectrum to determine structure loads and instructure response spectra for the CDFM evaluation. For the fragility analysis, 

conservatism in this spectrum was accounted for in the structure response factor 

described in Section 5.4.1.1. The structure damping value of 7% that was included 

in the generation of in-structure response spectra is judged to be conservative for 

the structures considered. Other major differences between the two approaches in 

their application to the selected examples are discussed below. 

5.4.2.1 Circulating Water Pumphouse 

Differences between the CDFM and fragility analysis methods for the circulating 

water pumphouse included the following: 

o Damping 

o Diagonal brace buckling capacity 

o Effective inelastic energy absorption factor 

Seismic response of the steel superstructure was determined using I 0% damping. 

This value is recommended by NUREG/CR-0098 as being a conservative value for 

bolted steel structures at or near the yield point. The diagonal brace buckling 

capacity was determined using two different formulations: 

o Factored AISC allowable 

o Conservatively biased value based on test data 

The AISC allowable buckling capacity for working stress design was factored by 

1.7 as permitted by Section 2 of the AISC specifications for plastic design. 

Alternatively, the test data in [Hall, 1981] used to establish the median capacity 

was reviewed to establish the 84% exceedance value. These two formulations were 

found to provide essentially the same conservative capacity. 

The HCLPF capacity against initial brace buckling was calculated to be 0.22g. 

This value is higher than the corresponding capacity of 0.19g determined by the 

fragility analysis method. 

As noted in Section 5.4.1.2, the HCLPF capacity for collapse of the pumphouse steel 

structure is judged to be greater than 0.3g. This corresponds to an increase factor 

accounting for the reserve capacity against collapse of 1.35. 

5.4.2.2 Block Wall SB 35-3 

Conservative seismic response of Wall SB 35-3 for the CDFM method was based 

upon in-structure response spectra generated for the median NUREG/CR-0098 

5~54 
ground spectrum. A conservative estimate of block wall damping prior to initial 

cracking was not required since elastic response of this wall occurs in the high 

frequency range of the floor spectra. The primary source of conservatism in the 

HCLPF capacity is the selection of a conservative modulus of rupture. The 

Standard Review Plan permits the use of ACI 531 allowable block wall stresses 

with certain increase factors for extreme load combinations. However, no increase 

is permitted for the modulus of rupture perpendicular to the bed joint. This is 

considered to be overly conservative since ACI 531 itself permits a one-third 

increase in working stress allowables when stresses due to earthquake loading are 

included. Review of available test data indicates that the increase factor should 

be at least 1.33. The modulus of rupture perpendicular to the bed joint was 

conservatively based upon the ACI 531 allowable value increased by a factor of 

1.33. The HCLPF capacity using the CDFM method was determined to be 0.67g. 

This value is greater than the capacity of 0.57g found by the fragility analysis 

method. 

5.4.2.3 Refueling Water Storage Tank 

Major differences between the CDFM and fragility analysis methods in the 

determination of HCLPF capacities for the refueling water storage tank were 

associated with the following quantities: 

o Damping 

o Shell buckling stress 

o Material and component strengths 

o Effective ductility 

A conservative damping value of 5% was estimated for the horizontal impulsive 

mode response. This is the value recommended by NUREG/CR-0098 for welded 

steel structures at or near the yield point. The same analytical model developed in 

the fragility analysis to determine the resistance against overall seismic base 

moment was used. As in the fragility analysis, input to this analytical model 

~consisted of the shell buckling stress, bending capacity of the anchor bolt chair top 

plate, and fluid holdown force on the tank bottom plate. The shell buckling stress 

was determined using the NASA design coefficients which have a 900Jo confidence 

of exceedance. Plastic moment capacities for the anchor bolt chair top plate were 

based on an effective yield stress taken as the average of the minimum specified 

yield and tensile strengths. This value corresponds to an equivalent elasticperfectly plastic representation of the material stress-strain relationship. 

Additional conservative bias was introduced into the top plate capacity to account 

for uncertainty in the location of the loading imposed by the anchor bolt nut and 

washer. Additional seismic capacity implied by nonlinear behavior of the tank as 

it uplifts and yields the bolt chairs prior to shell buckling was conservatively 

neglected. A HCLPF capacity of 0.2lg was determined for the RWST using the 

CDFM method. This value is the same as the capacity determined by the fragility 

analysis method. 

[Manos, 1986] has developed an empirical method for the seismic design of 

unanchored tanks. Manos' eomparison of his approach with data compiled on 

damage and undamaged tanks from actual earthquakes indicates that this method 

is CDFM equivalent. HCLPF capacities for the RWST, DWST, and PWST based 

5-55 
.. : . ~ 

upon Manos' approach are listed in Chapter 7. For example, a capacity of 0.24g is 

determined for the RWST. This capacity is greater than the 0.2lg HCLPF capacity 

found by tlie fragility analysis and CDFM methods. Intuitively~ the seismic 

capacity of an anchored tank would be expected to be greater than that of an 

unanchored tank. However, the mechanics of seismic resistance of unanchored 

tanks is not well understood at this time. Because the nature of seismic resistance 

of anchored tanks may be much different than the source of resistance for 

unanchored tanks, it is not possible to derive firm conclusions on the HCLPF 

capacities of the Maine Yankee tanks based upon Manos' work. 

5.5 HCLPF Capacity of Plant 

5.5.1 Accident Sequences 

Following the review of plant information~ a list of the components that make up 

the front-line and support systems required to perform the plant safety functions 

(Group A functions) was developed. This list is given in Chapter 7 and provided 

the basis for the remainder of the margin review. After the plant walkdowns and 

subsequent analyses by the systems analysts and fragility analysts, the list of 

components was reduced by screening out those components that were found to 

have HCLPF capacities greater than 0.30g pga. The components for which 

fragilities and. HCLPF capacities were determined are discussed in Section 5.4. 

These remaining components were used in the development of the event trees and 

fault trees for the seismic induced core damage accident sequences as described in 

Volume 2. The event trees and fault trees were analyzed to determine the cut sets 

for each accident sequence that could lead to core damage. From these cut sets, 

the Boolean expressions for the accident sequences were developed. The 

component failures that are significant to these accident sequences are given in 

Tables 5.5~1 and 5.5-2. Table 5.5-1 gives the seismic induced failures along with 

the fragility parameters used to represent their capacity. Table 5.5-2 gives the non 

seismic failures and their failure probability. Note that the component items and 

the non seismic failure events have been numbered consecutively; the missing 

numbers represent the items that were screened out in the final pruning of the 

event and fault trees. 

The Boolean expressions for the two dominant accident sequences are: 

Small LOCA 

= 4 + 7 + 20 

No LOCA 

= ( 4 + 20 ) * ( 8 + 15 + 17 + 22 ) 

+ 8 * ( 14 + 16 ) + 7 * 15 ' 

(5.5~1) 

(5.5-2) 

where each number in these expressions corresponds to the failure of a component 

given in Table 5.5-1 or Table 5.5-2. In the above expressions, the notation "+" 

5-56 
'•' . • ... ,. :t, 

Table 5.5-1 Component Seismic Fragility Parameters 

Item 

No. 

4 

7 

8 

20 

21 

Item 

Transformers 

RWST 

DWST 

Circu1a ting Water 

Pump house 

PWST* 

• See Page 5-67 

~R f3u 

0.84 0.30 0.32 

0.45 0.20 0.25 

0.36 0.20 0.26 

0.69 0.24 0.27 

0.57 0.20 0.26 

HCLPF 

Capacity(g) 

0.30 

0.21 

0.17 

0.30 

0.27 
Item 

No. 

10 

II 

12 

13 

14 

15 

16 

17 

22 

Table 5.5-2 Probabilities for Nonseismic Failures 

Description 

Opera tor Failure to Close 

PCC Isol. Valves 

Random Failure of DG-1 B 

Random Failure of DG·1A 

Operator Failure to Place AFW 

Pump Train B in Service Locally 

Nonseismic Common Cause 

Failure of DGS 

Nonseismic Common Cause 

Failure of AFW 

Opera tor Failure to Refill DG 

Fuel Tanks by Opening Valve 

or Running P·33A,B 

Operator Failure to Place AFW 

Pump Train B in Service from 

MCR 

Random Failure of the Turbine 

Driven Aux. Feedwater Pump 

Median 

Unavailability 

(per demand) 

8.0E-02 

4.2E-02 

4.2E-02 

l.SE-01 

1.6E-03 

1.2E-04 

8.0E-03 

4.0E-02 

3.0E-02 

* Error factor equals (95% Confidence Value/Median Value). 

S-58 

Error 

Factor* 

2 

5 

5 

2 

s 

5 

3 

3 

5 
denotes probabilistic addition (union) and "*" indicates probabilistic multiplication 

(intersection). 

Because the impulse lines inside the containment could not be inspected to confirm 

that small LOCA cannot occur, the Boolean expression for small LOCA and No 

LOCA were combined in two different ways. 

The first method uses split fractions to express the conditional probability of a 

seismic induced small LOCA given the seismic event. The two Boolean Eqs. (5.5·1) 

and (5.5-2) are combined using the split fraction p, and sensitivety studies are 

performed on this split fraction. 

Core Damage 

= p. [Small LOCA] + 

( 1 - p ) . [No LOCA] (5.5-3) 

The second method uses an additional term in the small LOCA Boolean expression 

that represents the initiating event of seismic induced small LOCA and the other 

terms in Eq. (5.5·1) represent the failure of mitigating systems. The core damage 

Boolean expression is then the logical combination of the Boolean expressions for 

two accident sequences i.e., [Small LOCA] • (Small LOCA Boolean Expression] + 

[No LOCA] • [No LOCA Boolean Expression]. The HCLPF capacity of the plant 

against core damage is a function ofthe HCLPF capacity against small LOCA 

(initiating event) if this capacity is less than 0.2lg. The plant level core damage 

HCLPF capacity is governed by the HCLPF capacity of RWST. If the small LOCA 

HCLPF capacity is larger than 0.21g, the plant level HCLPF capacity is controlled 

by the small LOCA HCLPF capacity. The HCLPF capacity against small LOCA 

was not determined since we could not access containment. However, the plant 

level HCLPF capacity is expected to be greater than 0.2lg. 

In the following, the quantification of these Boolean expressions for the purposes 

of deriving the plant HCLPF capacities is discussed. 

5.5.2 Probabilistic Method 

Following the rules of Boolean algebra, the component fragilities are combined 

using the numerical procedure proposed by Kaplan (1981). For this purpose, the 

component fragilities are discretized into a family of fragility curves with a 

subjective probability estimated for each curve as discussed in Section S.4.1. The 

probability assigned to each curve is developed based on the uncertainty 

distribution on the median capacity. Each fragility curve is assumed to be 

completely described by the median capacity and the value of ~R· The fragility 

curve is not truncated in either the lower or upper tail. 

The discrete probability distribution (DPD) for a component, say 4 , is expressed 

as: 

4 4 

{ <Qi , f i , (a) > } (S.5-4) 
and is shown graphically in Figure 5.5-1. Here, a is the peak ground acceleration, 

f i is the conditional probability of failure of the component under acceleration a 

and qi is the subjective probability that the value of f i is in the range f i - ~f j/2 

and fj + ~fi/2. For example, the aggregation of fragilities for two components 4 

+ 7 ts given by a convenient numerical integration scheme [Kaplan, 1981]: 

where 

( .•! l .. '.j• .• : 

0.75 

0.50 

0.25 

4+7 

{ <Qik 

4+7 

,fik ,(a)>}, 

4+7 4 7 

qik = qi qk 

4 7 4 7 

1 - (1-fi) (1-fk) or max {fi, fk }. 

·------ -... / 

/ / 

/ / 

/ @ I / I ql I I I @; I 

I q2 I 

I I qCf) 

1 

I 

I 

I I 

' I 

I 

I 

I I 

I 

I I 

I I 

I I 

I 

I 

I 

I 

I 

I 

f. +t~ f. /2) 1 1 " 

f.-f:..f./21 

1 1 / 

/ 

/ 

I 

I 

I 

I 

I 

I@) 

1 qn 

Subjective 

/ 

I 

/ 

/ 

/ 

I 

I 

I / I / 

/ / 

/ 

I 

/ 

..-" 

/ 

/ 

/ 

/ 

Pro ba hili ty 

Distribution of 

Failure Probability f. 

' 1 

--

Peak Ground Acceleration, a 

Figure 5.5-1: Seismic Fragility Curves of n Component 

5-60 

(5.5-5) 
··."···. 

In the expression for fik' two bounds are given corresponding to perfect 

independence or perfect dependence between component failure events. These 

bounds are used to assess the significance of correlation between failure events. 

Expressions similar to Eq. (5.5-5) are derived for the case of joint events 4 * 8 

By taking two components at a time, the plant level fragility is developed 

p p { <qi , fi , (a)>}. (5.5-6) 

Figure 5.5-2 shows a plot of the plant level (small LOCA core damage) fragility 

curves in which the family of fragility curves is reduced to the 5%, 50%, and 95% 

confidence fragility curves. From this plot, the HCLPF capacity (defined as peak 

ground acceleration corresponding to 5% probability of failure at 95% confidence) 

for small LOCA core damage is obtained as 0.2lg. 

Inspection of the small LOCA core damage Boolean expression [Eq.5.5 .. 1] given 

above indicates that the dominant components are the singletons. The singleton 

component with the lowest HCLPF capacity is the refueling water storage tank 

(RWST) with a HCLPF capacity elf 0.2lg. Failure of this tank results in no coolant 

being available for reactor vessel injection following a LOCA. 

The other singleton components have HCLPF capacity equal to 0.30g. The capacity 

of the transformer is estimated based on the proposed upgraded condition. 

The seismic fragility of the circulating water pumphouse is conservatively 

estimated. However, its failure has a less significant effect on the small LOCA 

core damage HCLPF capacity because of the smaller ~ R and ~u· 

The HCLPF capacity for No LOCA core damage accident sequence was estimated 

using the Boolean expression given in Equation 5.5 .. 2 as 0.32g (Figure 5.5-3). The 

higher capacity against this sequence compared to the small LOCA sequence is 

because of the absence of low capacity singletons in the expression. Although 

DWST, i.e., component 8, with its HCLPF capacity of 0.17g appears in this 

sequence, its failure has to occur simultaneously with one of the higher capacity 

components i.e., transformer, and circulating water pumphouse. 

Core Damage HCLPF Capacity 

As explained above, the core damage HCLPF capacity calculation requires a 

knowledge of the split fraction between the two accident sequences (i.e., small 

LOCA and Transient). For different assumed split fraction values, the core 

damage HCLPF capacities were obtained as shown in Table 5.5-3. 

The conclusion regarding the dominance of R WST failure in the HCLPF capacity 

estimation (displayed in the small LOCA accident sequence) is a function of the 

split fraction assumed. If the plant HCLPF capacity needs to be increased, it is 

not necessary to concentrate only on R WST. A walkdown and review of small 

impulse lines within the containment may be performed to estimate their fragilities 

in order to assign a realistic split fraction. By this procedure, the plant HCLPF 

capacity may be concluded to be higher without the necessity of any upgrading of 

the components. 

5-61 
LiJ 

([ 

::J 

] _J CONF. 

H .05 <( .50 LL .95 

LL 

0 

>-

tH 

.....J 

H 

m 

<( 

CD 

0 

a: 

0. 

.....J 

<( 

z 

0 

H 

tH 

0 

z 

0 

u 

)( 110 0 

PEAK GROUND ACCELERATION (g) 

Figure 5.5-2 Fragility Curves for Small LOCA Core Damage. 

5-62 
:..u 

cr 

:J 

'·1 _; t....;..Nf. 

H .05 

< .!5,;) 

LJ... ,95 

I• 

0 ·sr I 

/ 

>-

I rH 

_j 1 H 

en 

< 

en 

0 I 

a: I .J I a.. I ! / I 

4 I 

z .J 0 

H 

1:-

u 

H 

0 

0 

z 

.0 I I ~-------- ~-I --~-----+ 1.0 0 .0 .2 .4 ,6 .8 

)( 1~ 

PEAK GROUND ACCELERATION (g) 

Figure 5.5-3 Seismic Fragility Curves for No LOCA Core Damage. 

5-63 
Table 5.5-3 Summary of Plant Level HCLPF Capacities 

Case Description HCLPF Capacity (g) 

Small LOCA 0.21 

- Independent Seismic Failures 

2 Small LOCA 0.21 

- Dependent Seismic Failures 

3 No LOCA 0.30 

• Independent Seismic Failures 

with Nonseismic Failures 

4 No LOCA 0.30 

- Independent Seismic Failures 

without Nonseismic Failures 

5 Core Damage 

- split fraction p = 0.01 0.30 

p = 0.10 0.28 

p = 0.50 0.23 

6 Small LOCA 

- RWST 0.26 

Reduced Fluid Level 

5-64 
Effect of Nonseismic Failures 

The overall plant HCLPF capacity was calculated using the Boolean expressions for 

core damage accident sequences which contained both seismic and nonseismic 

failures. In the following we describe the reasons for including the nonseismic 

failures in this calculation. 

There are two kinds of unavailability. The first kind is that the component fails 

to start on demand due to random failure, common cause failure or operator error. 

The other is that the component is unavailable due to maintenance or failure to 

run. The first is expressed as failure rate per demand; the second is in terms of 

unavailability per unit time (e.g.~ year). 

For the seismic event, the component fragilities may be considered as failure rate 

per demand (varies with earthquake level). Hence component fragilities and 

nonseismic failure on demand rates may be added probabilistically. For the 

nonseismic unvailabilities, the technique is to convert them into conditional 

probabilities by multiplying the failure rate by a time interval during which the 

component is needed to perform. 

Example is that the diesels are expected to start on seismic induced loss of offsite 

power and continue to operate for some hours. 

Probability of failure of diesel system = P fs + )...t + P f(a) 

where Pfs = probability of failure to start 

X = rate of failure to run 

t • required time for the diesels to be running 

Pf(a) = seismic fragility of diesels 

The probabilities for nonseismic failures given in Table 5.5~2 include the 

probability of failure to start, operator error and the probability of failure to run. 

The uncertainties in the unavailability (per demand) is expressed in terms of the 

error factor which is equal to the ratio of the 95% confidence value to the median 

unavailability. The unavailability is modeled by a lognormal probability 

distribution. 

Need for Consideration of Nonseismic Failures in Seismic Margin Studies 

In the seismic margin studies, we are interested in estimating the largest 

earthquake that the plant can withstand with a high degree of confidence. A 

particular component may be able to withstand this earthquake. Assume that the 

plant will not suffer core damage if any one of the two components survives this 

earthquake. Conservatively, we may treat the component failures as statistically 

dependent. Hence, the plant level HCLPF capacity may be considered to be equal 

to the higher of the two component capacities. However, the plant HCLPF 

capacity will be different if one of the components is affected by high nonseismic 

failure rate. One component may fail in an earthquake, yet the other component 
·,, 

may not be available for the plant to survive the earthquake. Therefore, it is 

important to consider nonseismic failures in estimating seismic margins. Yet, any 

nonseismic failure that does not occur simultaneously with a seismic failure of a 

component is not of particular significanc~ in a seismic margin study. 

The effect of including nonseismic failures is observed in the No LOCA core 

damage sequence in that the HCLPF capacity for the sequence changes from 0.33g 

(when the nonseismic failures arc not included) to 0.32g (when the nonseismic 

failures are included). Since the components with HCLPF capacities >0.3g have 

been screened out, these plant HCLPF capacities have been represented in Table 

5.5-3 as >0.3g. In the small LOCA sequence, there arc no nonseismic failures. 

Correlation between Seismic Failures 

The above calculations were performed assuming perfect independence between 

seismic failures of different components: i.e., the seismic capacities are assumed to 

be statistically independent both in randomness and uncertainty. This is a realistic 

assumption because the components involved in the core damage Boolean 

expression are yard tanks (R WST, and DWST), transformer, and circulating water 

pumphouse. These are dissimilar items of..structures and equipment, their locations 

in the plant and within the structures are different, and their dynamic 

characteristics are also different. Hence correlation in the seismic responses and 

seismic capacities of these components is judged to be minimal. 

It is realistic to expect some correlation in the component failures. Assumption of 

perfect dependence in both uncertainty and randomness is an extreme case. 

Assumption of perfect dependence in the uncertainties of different component 

fragilities means that the median ground acceleration capacities of all components 

are known if the median ground acceleration capacity of one component is known. 

Since the uncertainty arises from insufficient understanding of structural material 

properties, approximate modeling of the structure, inaccuracies in the 

representation of mass and stiffness, and the use of engineering judgment in lieu 

of plant specific data, it is expected that all components will be affected to some 

degree by these uncertainties. Therefore, some probabilistic dependence between 

component median capacities may be expected. Perfect dependence in uncertainty 

is however an extreme assumption. 

Dependence in the randomness arises from a common earthquake generating the 

responses in different components and common structural/material properties. 

Assumption of dependence in the randomness means that if the fragility 

(conditional probability of failure) of a component for a given peak ground 

acceleration is known, the probability of failure of the other components would be 

somewhat modified by that knowledge if it were known. 

The plant level fragility, and therefore, '-the HCLPF capacity depends on the degree 

of dependence in randomness and uncertainty between the component failures. 

However, the degree of dependence is difficult to estimate. One approach is to 

bound the core damage HCLPF capacities by assuming perfect dependence as 

opposed to the case of perfect independence. This calculation was performed for 

the small LOCA core damage Boolean expression given above. The small LOCA 

core damage HCLPF capacity is estimated to be 0.2lg i.e., governed by the capacity 

5·66 
of RWST. The plant level HCLPF is determined to be approximately 0.2lg. When 

the Boolean expression is dominated by singletons, the assumption of perfect 

independence is more severe than the assumption of perfect dependence between 

failures if the fragilities are approximately equal; if there is a single component 

with a very low capacity compared to the rest of the components in the Boolean 

expression consisting of singletons, both the assumptions give about the same plant 

level HCLPF capacity. 

The question of dependence between failures is important when there are similar 

components experiencing common seismic excitation. The case in point is the yard 

tanks (i.e., R WST, DWST, and PWST). By reviewing the cut sets, it was found that 

a tripleton cut set could lead to core damage. It is 

DWST • PWST • R WST · 

The HCLPF capacity for this cutset would be higher than the highest of the three 

tank HCLPF capacities (i.e., 0.27g for PWST). 

5.5.3 Deterministic Method 

This approach is based on the assumption that the HCLPF capacities of components 

estimated in Section 5.4 are true lower bound values. The HCLPF capacity of the 

plant is obtained directly by studying the Boolean expressions for small LOCA and 

no LOCA: 

Small LOCA 

= 4 + 7 + 20 

NoLOCA 

= ( 4 + 20 ) • ( 8 + I S + 1 7 + 22 ) 

+ 8 • ( 14 + 16 ) + 7 • l.S · 

For calculating the plant level HCLPF capacity in this method, the nonseismic 

failures are ignored. Also, the cutset that includes a low probability nbnseisrilic 

failure is also omitted from this estimation. Therefore, the simplified·· Boolean 

expression is: 

NoLOCA 

= ( 4 + 20 ) • 8 

In this method the HCLPF capacity of a "doubleton" cut set is calculated as the 

higher of the two component HCLPF capacities. The HCLPF capacity of a 

"tripleton" cut set is calculated as the highest of the three component HCLPF 

capacities. The HCLPF capacity of a union of singleton cut sets is estimated to be 

the lowest of all the component HCLPF capacities. Using this procedure, the 

HCLPF capacities of the core damage accident sequences are: 

S-67 
.· I,' I 

' : ~ . 

HCLPF capacity against small LOCA core damage 

= min [ 0.30, 0.21, 0.30 ] · 

= 0.2lg 

HCLPF capacity against No LOCA core damage 

= max [ min ( 0.30, 0.30 ), 0.17 ] 

= max [ 0.30, 0.17] 

= 0.30g 

Note that the failure of DWST with its HCLPF capacity of 0.17g is not governing 

the plant level HCLPF capacity because DWST has to fail simultaneously with one 

of two nonseismic failures to l~ad to core dafage. Since these failures have 

median probabilities of 6 x 10" and 8 x 1 o- , per demand respectively, it is 

appropriate to ignore the DWST failure in the plant level HCLPF capacity 

calculation. 

5.5.4 Sensitivity Studies 

Two sensitivity studies were performed to assess the effect of certain assumptions 

on the plant HCLPF capacity. These sensitivity studies addressed the following: 

1. Effect of shearwall stiffness reduction on structure response and 

equipment seismic input. 

2 Reduction of RWST fluid level. 

On-going scale model testing being conducted by the NRC has indicated potential 

reductions in the stiffness of concrete shear walls of up to a factor of four, from 

elastically calculated values due to cracking. This would imply a reduction in the 

elastic structure frequencies of up to 50%. 

For the case with an assumed small LOCA, the R WST is the dominant contributor 

to the plant HCLPF capacity. R WST constructed of stainless steel and is located in 

the yard on its own foundation. Its HCLPF capacity is not affected by the 

potential structural frequency shift. For the case No LOCA, the 4160/480 V 

transformer is the dominant contributor to the plant HCLPF capacity. The 

transformer has a 12-13 Hz fundamental frequency which corresponds to a spectral 

acceleration on the downward slope of the floor response spectra. A reduction of 

the building frequencies would result in a reduction of the seismic input to the 

transformer with a correspodig increase in its HCLPF capacity. As a further check 

on the effect of the building frequency shift, a similar evaluation was made on 

each of the components within the final Boolean expression (Table 5.5-1). The 

potential building frequency reduction did not lower the HCLPF capacity for any 

of these components. Thus, for the purpose of the Maine Yankee seismic margin 

study, the shear wall stiffness reduction and resulting building frequency shift 

does not affect the plant seismic margin. 
Because the seismic capacity of the RWST controls the plant HCLPF capacity for 

the small LOCA case, the effect of reducing the fluid level within the tank was 

investigated. A reduction of the fluid level to a total height of 33 ft, which is 

approximately equal to 10% reduction from the current level of 37 ft, was assumed. 

This change leads to a reduction in the effective fluid weight, mass centroid 

height, and overall tank seismic loads. The R WST HCLPF capacity is increased to 

0.28g with a corresponding increase of the small LOCA HCLPF capacity to 0.26g. 

5-69 

:· '' 

CHAPTER 6 

COMMENTS AND RECOMMENDATIONS ON SEISMIC MARGIN REVIEW 

METHODOLOGY 

6.1 Selection of Review Earthquake Level 

In the course of this trial application of the seismic margin review methodology, it 

was found that sufficient guidance was not included in NUREG/CR-4334 and 

NUREG/CR- 4482 for the selection of the review earthquake level. NUREG/CR4334 states that: 

11The Panel has focused its efforts on earthquakes that could 

occur in the eastern part of the U.S., specifically east of the 

Rocky Mountains. Because of limited data on large 

magnitude events, the assessment of component capacities is 

limited to earthquakes of less than a Richter magnitude of 

about 6.5, which are characterized by three to five strong 

motion cycles with a total duration of 10 to 15 seconds. As 

the Richter magnitude increases above 6.5, the Panel 

recommendations may be slightly non-conservative. The 

frequency content of the earthquake round motion is 

assume to be represented by median broadband response 

spectra, and the structures are assumed to have fundamental 

frequencies above 1 Hz. Note that for high frequency, high 

acceleration, low magnitude earthquakes, the Panel's 

recommendations are overly conservative." 

The report goes on to state that: 

"The margin review must begin with a target earthquake, in 

order to provide focus. It is assumed for the purposes of 

the review activity that some external source (perhaps the 

NRC staff, or the utility) has designated the earthquake 

level (in terms of peak ground acceleration, pga) which is 

the level at which the review is aimed". 

The Panel in the NUREG/CR-4482 has stated that: 

"The choice of the review level is a critical one since it is 

used as a basis for screening out components. The review 

level should be specified in terms of pga and enough 

spectral information to assure the applicability of the 

material of Chapter 5 in NUREG/CR-4334 which is 

partially given in Table 2-1 of this report. If the review 

earthquake spectral content is not consistent with the 

assumptions made in NUREG/CR-4434, then this difference 

needs to be taken into account. Table 2-1 was 

6-1 
constructed to cover most spectra genera ted by magnitude 

6.5 earthquakes or less. In addition, spectral information 

will be needed to calculate HCLPF values." 

For the seismic margin review of Maine Yankee, the NRC specified that the 

review earthquake level would be 0.3g with a median Newmark~Hall ground 

response spectrum as defined in NUREG/CRH0098 for rock sites. Since this 

spectrum is applicable generically for all rock sites, it was concluded by the NRC 

that the selected spectrum is a 84% confidence spectrum for Maine Yankee (See 

Chapter 2). The requirements on the review earthquake level are also somewhat 

implicit in the Panel's recommendations. Yet, for future applications, it is 

suggested based on this trial plant review that more explicit guidance be provided. 

6.2 Use of Screening Guidelines 

6.2.1 Extent of Review Needed 

As discussed before, this study has two primary objectives: to confirm the 

appropriateness and adequacy of guidelines given in the Expert Panel reports and 

to estimate the seismic margin for Maine Yankee. Because of this dual objective, 

the study team had to expend many engineering hours in reviewing the components 

that were supposedly to have been screened out in the initial screening, i.e., those 

components identified as having seismic capacities generically higher than the 

review earthquake level and denoted by letter "C". For example, the review by the 

study team included components such as valves, diesel generators, pumps, HV AC 

ducting~ cable trays and cabling although the Panel screening guidelines state that 

these components have seismic capacities larger than 0.30g pga. It may be pointed 

out that the Panel screening guidelines are invariably conservative in that the 

specific components in a plant may actually have capacities larger than indicated 

by the Panel. Also, the collective experience and judgment of the Expert Panel in 

developing the guidelines cannot be expected to be matched by any single team 

performing the seismic margin reviews in future. Hence,· in keeping with the 

primary objective of minimizing the review effort needed in a seismic margin 

review, it is recommended that the components identified as having capacities 

larger than the review earthquake level be screened out with no further review in 

future seismic margin studies. 

6.2.2 Additional Screening Guidelines 

The Expert Panel reports have not explicitly discussed the seismic capacities of 

steel structures. In the trial plant review on Maine Yankee, three steel structures 

were analyzed to assess their HCLPF capacities. These structures do'· have 

capacities in excess of 0.30g. At the same time, they could not be screened out 

based on a walkdown review because of unusual connection details and structural 

arrangement. 

Therefore, some guidance is necessary in the margin review methodology for steel 

structures. 

6-2 
' ' ~ . 

6.2.3 Difficulties of Reviewing Certain Components 

The majority of the componen·ts identified for review in the Maine Yankee seismic 

margins study were reviewed during the plant walkdowns. Component design 

reports and seismic qualification analyses were also reviewed on a majority of the 

Maine Yankee equipment as a part of the fragility assessment. Difficulties in 

reviewing certain Maine Yankee components fell into one of two categories: 

o Components whose qualification reports were available only from 

the NSSS supplier 

o Components located inside the containment structure 

6.2.3.1 Reactor Internals and CEDM 

The reactor internals and the control element drive mechanism (CEDM) are two 

components requiring fragility assessments for the Maine Yankee margin study. 

The utility (MY AC) did not have access to the seismic qualification documentation 

for two reasons: 

1. This information is considered proprietary by the NSSS vendor 

2. Due to the vintage of the Maine Yankee Plant, seismic 

qualification was not required for the CEDM and for portions of 

the reactor internals 

The NSSS supplier, Combustion Engineering (CE), is the only credible source of 

capacity /response information on the internals and the CEDM. Discussions 

between CE, MYAC and EQE resulted in a contract being set up through LLNL for 

CE to collect the qualification data from their files and to meet with EQE and 

Maine Yankee representatives. CE was very cooperative during this meeting and 

provided the following: 

o Summaries of the qualification report results 

o Similarity assessments for components like the CEDM which had 

no previous seismic qualification 

This interfacing with the NSSS supplier will generally be required on future 

seismic margin studies, just as it has been required on almost ail of the previous 

seismic PRA's. In addition, for plants where the utility does not take such an 

active role in the design/margin review, meetings with the architect-engineer to 

obtain plant/component information may also be required. 

6.2.3.2 Equipment Within Containment 

Due to radioactivity concerns, critical components inside the Maine Yankee 

containment are accessible for review only at the time of a plant outage. The 

margin study schedule precluded the possibility of one of the plant walkdowns 

coinciding with the March 1987 plant refueling outage. This fact necessitated the 

use of previous photographs and design drawings for fragility derivation purposes. 

6-3 
Fortunately, Maine Yankee personnel have maintained a relatively complete file of 

component photographs based on previous seismic integrity studies. The 

experiences at Maine Yankee point out the need in future margin studies to plan 

for the review of components within containment. Difficulties in reviewing plant 

components will be more severe for plants which do not maintain organized 

documentation on the components within containment. 

Impulse line failures were assumed to be the source of a small LOCA at Maine 

Yankee. This conservative assumption was required due to the tremendous number 

of hours which would be required to walk down each of these impulse lines and 

assess potential system interaction problems. These lines originate from the 

primary pressure boundary (i.e., RPV, steam generator, pressurizer, primary coolant 

loop piping, etc.) and are field routed to instrument racks inside containment. The 

amount of work required to demonstrate the seismic margin in each of these lines 

plus the fact that a walkdown of these lines would have to take place during a 

plant outage necessitated the assumption of a small LOCA as an initiating event. 

6.3 Availability of Qualification Data 

In-structure response spectra for the median NUREG/CR-0098 ground response 

spectrum were generated for use in the trial plant review by Maine Yankee during 

the course of this study. The structure building models were developed recently 

for other Maine Yankee programs. The availability of in-structure response 

spectra generated by current techniques can reduce the amount of effort necessary 

to quantify component fragilities since correction or regeneration of responses was 

not necessary. Adequate floor spectra may not always be available for older 

plants. In such cases, it may be necessary to develop new dynamic models and 

perform dynamic analysis to define the seismic input to equipment. This will 

increase the amount of time and funding necessary to perform seismic margin 

studies. 

Most of the Maine Yankee Group A structures were screened out, so detailed 

seismic load distributions to the structural members were not necessary for the 

trial· plant review. For the screened in structures, seismic load distributions were 

either available from recent dynamic analyses, or could be estimated with 

relatively little effort due to simplicity in structure load paths. Consequently, 

determination of seismic load distributions was a minor consideration. However, 

this would not have been the case had it been necessary to rely upon load 

distributions from the original design calculations. The original structure design 

calculations for Maine Yankee comprised a large volume of information. 

Determination of the seismic loads and verifying their accuracy within this body 

of material would have been time-consuming and tedious. Typically, original 

structure design calculations have not been used in past PRAs for this reason. 

When necessary, approximate load distributions have been independently 

developed. Generation of structure load distributions with the level of accuracy 

appropriate for a seismic margin review could require additional effort. 

Detailed information on the Maine Yankee block walls was available. These data 

were very useful in conducting the trial plant review since the data provided 

locations of all block walls in the plant and identified any safety related 

components that could be affected by their failure. In particular, this latter set of 

6-4 
information was valuable since it permitted quick identification of several Group 

A block walls. It is not known if comparable information exists for most older 

plants. 

Seismic qualification information for Maine Yankee equipment was available for 

selected critical components. Maine Yankee engineers and their consultants have 

conducted specific qualification tests and analysis as the need to do such an 

analysis has been identified. There is very little seismic qualification analysis 

available from the original plant design documentation. This fact stems from the 

fairly relaxed requirements in the seismic qualification area at the time of the 

Maine Yankee plant construction. Seismic qualification in many cases consisted of 

simply a letter from the vendor that his equipment component had adequate 

seismic capacity without any supporting documentation. 

Examples of equipment components for which seismic qualification information 

was available included the diesel generator day tank, battery racks, RHR heat 

exchanger, emergency service water pump, component cooling heat exchanger, 

demineralized water storage tank and the diesel generator remote control panel. 

Seismic anchorage calculations were also available for all of the electrical 

equipment which had their anchorage/supports upgraded from their original design 

(e.g., inverters, motor control centers, control room cabinets, transformers, etc.). 

Combustion Engineering provided seismic analysis data for some of the reactor 

internal components, but the remaining reactor internals structures and the control 

element drive mechanism did not have qualification data because there was no 

detailed seismic analysis requirement for Maine Yankee vintage plants. Equipment 

for which seismic qualification data was not available included the majority of the 

valves, fans, dampers, heat exchangers, small tanks, pumps, and air conditioners at 

Maine Yankee. 

6.4 Walkdown Procedures 

Based upon the trial plant review, suggestions on improvements to the review 

methodology in the area of walkdown procedures would include the following: 

o Third walkdown 

o Access to personnel with specialized expertise 

o Data sheets 

o Level of walkdown for bulk items 

During the course of the trial plant review, certain components with relatively low 

HCLPF capacities were identified. The utility elected to develop conceptual 

modifications to be installed in the next outage. Plant HCLPF capacities reported 

for this study reflect the increased component capacities based upon the conceptual 

modifications. A third walkdown will be necessary to confirm that these 

modifications are installed, and that they provide seismic capacity at least 

comparable to the conceptual retrofits upon which the calculated component 

capacities were based. 

6-5 
During the walkdowns, plant and utility personnel having expertise or specialized 

knowledge in particular fields were interviewed. Specific areas of expertise 

included: 

o Fire water systems 

o Electrical 

o HVAC 

o Instrumentation and control 

o Control room personnel 

o Block walls 

The insight gained from discussions with these individuals proved to be very 

useful in assessing the overall state of the plant and resolving any particular 

seismic issues. Scheduled meetings with plant experts in each of these areas allow 

the fragility and systems analysts to gain in-depth understanding of plant specific 

parameters such as design criteria, construction practices, equipment locations and 

functions, and operator actions in emergency events. 

Walkdown data sheets (Section 4.3.3), while not noted in the review guidelines, 

were developed prior to the first walkdown. These data sheets served as a 

checklist of items to review during the walkdown and facilitated the gathering of 

information necessary for HCLPF evaluation. While no set format need be 

established for future margin reviews, comment on the use of data sheets in the 

review guidelines should be considered. 

Bulk items such as cable trays, piping, and ducting extend throughout the plant, 

and local configurations and support details can vary. Detailed walkdown of these 

bulk commodities would be time-consuming. The extent of walkdown required for 

review level earthquakes less than 0.3g is typically specified in general terms. As 

an example, the Expert Panel recommends that example cable trays be inspected to 

verify that they are adequately anchored and braced and that specific 

vulnerabilities do not exist. In the trial plant review, a general survey of the cable 

tray systems was performed and supplemented by detailed inspection of a typical 

system. More guidance in the seismic margins methodology on the level of 

walkdown for bulk items would be useful. 

6.5 Guidance on CDFM Capacity Calculation Procedures 

The second report by the Expert Panel [Prassinos et al., 1985] recommends two 

candidate methods for calculating the HCLPF capacities for components: the 

conservative deterministic failure margin (CDFM) method and the fragility 

analysis (FA) method. The fragility analysis method was used in this study as was 

done in over 20 seismic PRAs. This method, although very familiar to the present 

analysis team, calls for subjective judgments to estimate parameters such as Am, 

~ and ~U for each variable needed in the seismic capacity estimation. The 

CDFM method, on the other hand, prescribes the parameter values and procedures 

to be used in calculating the HCLPF capacities. It is conceptually very attractive 

because it aims to avoid subjective judgments on the part of the analysts: first, the 

analysts may not be equipped to make such judgments and second, there may be 

inconsistencies between the subjective fragility assignments by different analysts 

with varied background and expertise. For the sake of consistency and 

6-6 
repeatability, we need a prescriptive method to estimate the seismic capacities of 

structures and equipment. Ideally, the CDFM method would fit the bill when it is 

fully developed. The guidance on the use of CDFM method in NUREG/CR~4482 is 

not sufficient for this purpose. 

In the course of this study, five example components (i.e., refueling water storage 

tank, steel structure, diesel day tank, inverter, and block wall) were analyzed using 

the two candidate methods (i.e., CDFM and FA). Our experience was that several 

judgmental decisions had to be made in arriving at the parameters of the CDFM 

method. In each case we were not sure whether we met the intent of the method, 

i.e., conservative estimation of the capacity, yet more liberal than the SRP 

requirements: In some cases we may have been overly conservative as was pointed 

out by the Peer Review Group. The difficulties arise because of two factors: 

o The CDFM method has not been fully developed for all structures 

and· equipment items. 

o The parameters of the CDFM method such as damping, material 

strength, static capacity equations, system ductility, and methods 

for floor spectra generation are not explicitly specified; even 

where they are specified they may be overly conservative. Also, 

the appropriate conservatism in the selection of the CDFM 

parameters needs to be determined using calibration methods. 

Development of deterministic procedures such as the CDFM method may be 

accomplished using probabilistic models in such a way that over a large number of 

components both the CDFM and FA methods yield identical HCLPF capacities. 

Once such a calibration is achieved, the CDFM method may be confidently used in 

future seismic margin reviews. The calibration should start with a fragility 

evaluation of a representative set of components using the models and parameters 

agreeable to the group (i.e., Seismic Margins Expert Panel). The parameters of the 

CDFM method are to be selected such that over this representative set of 

components the HCLPF capacities derived using the CDFM method and the 

fragility analysis method are approximately equal. 

We recommend that such a calibration study be performed. 

6.6 Staffing Requirements and Schedule 

6.6.1 Staffing Requirements 

NUREG/CR-4482 gives estimated staffing requirements for a seismic margin 

review; for a plant founded on rock with a review earthquake level of 0.30g pga, 

the estimate to perform the fragility analysis is 19 engineering-months. The actual 

engineering~months expended in the present study is about 25% more than the 

Panel's estimate. Several factors need to be considered in this regard: 

o The dual objective of the study, i.e., seismic margin review of 

Maine Yankee and a trial application of seismic . margin review 

methodology, required additional review efforts and interface 

meetings not to be expected in future studies. 

6~7 
o Maine Yankee is an older plant designed and constructed before 

the development of quality assurance programs and seismic 

qualification methods presently existing in the nuclear industry.· 

Hence, the qualification reports on certain equipment items were 

not available. Because of the reevaluation efforts, additional 

information and new structural models have become available. 

However, the extent of information available is not comparable 

to that normally available for a modern nuclear power plant 

(e.g.,NTOL plant). The staffing requirements in the NUREG/CR4482 should therefore be revised to reflect the availability of 

design and qualification data in the particular plant. 

o The utility performed the seismic response analysis for structures 

in the plant and developed the floor response spectra for the 

selected review earthquake. The engineering hours spent on this 

effort should be included in the estimates for future margin 

studies. 

o The plant design information was provided by the utility with 

some assistance from the NSSS vendor. Also, the utility personnel 

assigned for the interface were extremely familiar with the plant 

arrangement, systems and design. Typically, the interface will 

extend to the plant architect-engineer and constructor. The level 

of effort to collect plant design information and perform the 

plant walkdown may be much larger for other plants as more 

organizations become involved. Our seismic PRA experience 

indicates that the Maine Yankee interface experience is very 

fortunate and not typical. 

6.6.2 Schedule 

This study has been conducted over an eight-month period; the plant walkdowns 

have also been conducted as per the schedule outlined in the Panel's report 

[Prassinos et al., 1985]. The schedule appears to be reasonable. 

6.7 Applicability of Methodology for Other Plants 

The methodology has been developed using the insights obtained from seismic 

PRAs on six PWRs supplied by Westinghouse and Babcock & Wilcox and a BWR 

Mark II. This study on Maine Yankee has added to this collective experience in 

that a PWR supplied by Combustion Engineering has been analyzed. From a 

fragility standpoint, the CE reactor internals and the control rod drive mechanism 

have augmented the PRA fragility information base. The screening guidelines may 

be revised to reflect this additional information. 

The trial plant review has been for a plant on a rock site and for a review 

earthquake level of 0.30g. The methodology, the review guidelines, and the 

staffing requirements have not been verified for other conditions, i.e., soil 

conditions and higher review earthquake level. Also, the methodology has not 

beendeveloped for screening systems in a BWR. In a BWR, the effect of 

6-8 
t' 

high·frequency hydrodynamic loads on the equipment needs to be considered. The 

systems and components to be reviewed in a BWR would be different and the level 

of effort and staffing requirements would vary. The applicability of the Panel's 

recommendations to these reactor/site/review levels needs to be demonstrated. 

>, 

•' 

CHAPTER 7 

RESULTS AND CONCLUSIONS 

The primary objective of this study has been to apply the seismic margin review 

methodology developed by the NRC Expert Panel on a trial basis in order to 

confirm the applicability of the methodology for future seismic reviews and to 

provide input to the Panel for modifications in the methodology and review 

procedures. The other objective of the study has been to determine the seismic 

margin of the Maine Yankee Atomic Power Station. This report focuses on the 

fragility aspects of the seismic margin review. The report has described the plant 

review, screening of the components based on their generic high seismic capacities, 

plant walkdowns and calculations of the HCLPF capacities of the components, 

systems, accident sequences, and the plant. During this review, potential seismic 

vulnerabilities have been identified and for some of them the utility has 

proposed/incorporated certain modifications. 

7.1 Screened Out Systems and Components 

Based on the screening guidelines of the Expert Panel, systems (both front-line and 

supporting sytems) not supporting Group A functions were screened out by Energy 

International Inc. These systems are discussed in Volume 2 of this Report. Among 

those components supporting the Group A function systems, some of them were 

screened out based on their seismic capacities as generically larger than 0.3 g pga, 

the selected review level earthquake. Table 7.1-1 lists the Group A structures, the· 

screening procedures used, and the HCLPF capacities. Table 7.1-2 provides a 

summary of Group A block walls; included are the wall ID number, the Group A 

components that may be affected by the wall failure and the bases for screening 

and HCLPF capacity. Table 7.1-3 lists the equipment in Maine Yankee by generic 

categories, the systems they support, their location, the screening procedures used 

(i.e., screening tables in NUREG/CR-4334, walkdown review, calculation review) 

and the HCLPF capacity. 

7.2 HCLPF Capacities of Screened-in Components 

The definition of screened-in components depends on the stage of the reiew. 

Screening was done at four stages: 

0 

0 

0 

0 

Preliminary screening based on the plant design information and 

using the Panel guidelines 

Preliminary screening confirmed by first walkdown and 

additional items screened out based on walkdown observation 

Screening based on conservation capacity calculations using 

~implified methods 

Screening based on final capacity calculations that show the 

HCLPF capacities to exceed 0.30g 

7-1 
j:· 

Table 7. 1·1 Suanary of Group A Structures· 

Screeni!ll 

NUREG Walkdown Fragility HCLPF 

Structure 4334 Review Evaluation Capacity Caaaents 

Containment structure c >0.3g Reinforced concrete structure. 

Contairwent internal structure c >0.3g Reinforced concrete structure. 

Priaary auxiliary bui I ding/fuel bui I ding c >0.3g Reinforced concrete structure. 

Circulating water ptllllhouse X X >0.30g Collapse capacity judged to be >0.3g 

based upon past earthcp~ke perfor'lllf1Ce, 

test data, and fragility evaluation. 

Turbine/service building, concrete portions c Reinforced concrete structure. 

Turbine/service building, steel fr&llled X 0.38g HCLPF capacity based upon detailed 

-J floor at EL 391 ·0" fragility evaluation. I 

tv 

Tunbine/service building, other steel X >0.3g HCLPF capacity >0.3g based upon 

framed portions conservative fragility evaluation. 

COntair-.ent spray ~ouse c >0.3g Reinforced concrete structure. 

Ventilation equipment room c >0.3g Reinforced concrete structure. 

Main steam valve hoUse, exterior concrete c >0.3g Reinforced concrete structure. 

structure 

Main steam valve house, interior steel X >0.3g HCLPF capacity >0.3g based upon 

structure conservative fragility evaluation. 

M.C.C. room c >0.3g Reinforced concrete structure. 

Aux feed pumphouse, purge air exhaust area c >0.3g Reinforced concrete structure. 

Fuel oil pumphouse c >0.3g Reinforced concrete structure. 
-..J 

I 

IJ.l 

Wall ID No. 

c ·2-1, 

c 20-2, 3 

c ·2-2, 3 

c 0~5-1, 

c 20·1 

c 46·1, 2 

c 61·1 to 3 

SB 21·7 

SB 21·17 

SB 21-18 

SB 35·1 

·~ - ~# 

Table 7.1-2 Summary of Group A Block Walls 

Group A C~ts 

Assorted safety class piping· and 

equipwent 

Contairaent Hner 

Pressurizer instru~~mtation and 

tabing 

Steaa generator snmbers and tabing 

PCC and sec piping, PCC surge line 

Main control board (aux feed panels) 

Main control board (aux feed panels), 

aux logic panels 

Battery groups 3 and 4, cable trays 

HCLPF 

Screening Capacity 

Out 

Out 

In >0.3g 

In >0.3g 

Out 

In >0.3g 

In >0.3g 

In >0.3g 

Comnents 

These walls are loose blocks shields. 

Adjacent components are protected by 

shielding consisting of steel fra.ing. 

Breach of the containaent liner dUe to 

illp8Ct by the block wall is considered 

w.l ikely 

HCLPF capacity based upon sinpl Hied 

fragility evaluation. 

Seisaic retrofits are installed. HCLPF capacity 

based upon si111plified fragility evaluation. 

The PCC and sec lines are about 2411 in 

diameter. The PCC surge line is about 

611 in diameter. These piping systems 

are of welded constructiOn. Failure 

due to impact by the block wall is 

considered highly unlikely. 

Seismic retrofits are installed. HCLPF capacity 

based upon simplified fragility evaluation. 

Seismic retrofits are installed. HCLPF capacity 

based upon simplified fragility evaluation. 

HCLPF capacity judged greater than value for wall 

SB 35-4. 
~ 

I 

~ 

Wall ID No. 

SB 35-2 

SB 35·3 

sa 35-4 

sa 35·7 

sa 45-t 

S8 45·2 

SB 45·3 

SB 45-6 

SB 61·2 

Table 7.1·2 Summary of Group A Block Walls (Continued) 

Group A Coaponents 

Battery groups 3 and 4, cable trays 

Battery groups 3 and 4, cable trays 

Battery groups 3 and 4 

PCC surge line, PCC temperature 

controller 

125V DC distribution cabinets 1 to 4, 

Battery group 12. Inverters 11 and 12, 

Battery chargers 11 and 12, Bus 8, 

Battery gr~ 11, MCC SA, 480Y 

emergency switchgear 

Battery gr~ 11 

PCC surge line 

PCC surge tank TK-5 

HCLPF 

Screening Capacity 

In >0.3g 

In >0.3g 

In >0.3g 

In >0.3g 

In >0.3g 

In >0.3g 

In >0.3g 

OUt 

OUt 

Connents 

HCLPF capacity based on detailed fragility evaluation. 

HCLPF capacity based on detailed fragility evaluation. 

HCLPF capacity based on detailed fragility evaluation. 

HCLPF capacity judged grea~er than value for wall 

S8 35-3. 

Seia.ic retrofits are installed. HCLPF capacity based 

on simplified fragility evaluation. 

Seismic retrofits are installed. HCLPF capacity based 

on simplified fragility evaluation. 

Seis.ic retrofits are installed. HCLPF capacity based 

on simplified fragility evaluation. 

The PCC surge line is shielded from the 

wall by a seis.ic retrofit consisting 

of two steel angles ruming parallel 

with the pipe. Failure of both the 

angles and the surge line clJe to i..,act 

by the block wall is considered highly 

wtl ikely. 

The PCC surge tank is shielded from the 

bloek wall by a seismic retrofit that 

consists of welded wire fabric spanning 

between steel framing. The shielding 

is judged to be sufficient to restrain 

the blocks from inpacting the tank. 
Table 7.1~2 Summary of Group A Block Walls (Continued) 

HCLPF 

Wall 10 No. Group A Components Screening capacity Conments 

SB n-2 PCC surge tank TK-5, sec surge tank TK-59 OUt The surge tanks are shielded from the 

block wall by a seismic retrofit that 

consists of welded wire fabric spanning 

between steel framing. The shielding 

is judged to be sufficient to restrain 

the blocks from impacting the tanks. 

TB 21·2 PCC heat exchangers E·4A and 4B OUt The heavy construction of the heat 

exchangers is judged to be capable of 

withstanding i...,act by the block wall. 

No other vulnerable components are 

exposed. 

-.....1 

I VE 21·1 Fans 44A and 44B. ducting and filter tn >0.3g Seismic retrofits are to be installed Vl 

in the next outage. Detailed fragility 

evaluation of conceptual modification 

sketches indicates that the HCLPF capacity 

is >0.3g. 

VE 21-4 sec lines In >0.3g HCLPF capacity based upon simplified 

fragility evaluation. 
Table 7.1·3 Slftlary of Maine Yankee Equipment Screening and HCLPF Capacities 

Equipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equipment I tern System Elevation 4334 Review Review Capacity Cooments 

TANKS 

1. Refueling Cavity Water Storage Tank HPS( Yd. + 20' X FA 0.21g (1) 

TK·4 CDFM 0.21g (2) 

(0.24g) (3) 

2. Primary Component Cooling Surge Tank PCC SB + 6P X >0.3g Conservative Fragility Calculation 

TK·5 Detenmined HCLPF Capacity is >0.3g. 

3. Primary Water Storage lank AFW Yd. + 20' X 0.27g (1) 

Tk-16 (0.25g) (3) 

...,J 4. Demineralized Water Storage Tank AFW Yd. + 201 X 0.17g (1) I Tk·21 (0.25g) (3) 01 

5. Air Start Receiver Tank AFW VA + 21' X >0.3g Conservative Fragility Calculation 

Tk-25 Determined HCLPF Capacity is >0.3g. 

6. Auxiliary Fuel Oil Supply Tank (buried) FO APR + 21 1 c >0.3g HCLPF Capacity Judged >0.3g 

Tk-28A for Buried Tanks. 

7. Auxiliary Fuel Oil Supply Tank <buried) FO APR + 21 1 c >0~3g HCLPF Capacity Judged >0.3g 

Tlc-288 for Buried Tanks. 

(1) FA - HCLPF capacity calculated per the Fragility Analysis approach. 

(2) CDFM - HCLPF capacity calculated per the Conservative Deterministic Failure Margin approach. 

(3) HCLPF capacity for unanchored vertical fluid storage tanks per Manos• approach. 
Table 7.1-3 SUIIIIary of Maine Yankee Equipment Screening ard HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

ard NUREG Walkdown Calculation HCLPF 

Equipment Item System Elevation 4334 Review Review Capacity Colrments 

TANKS (Continued) 

8. Spray Chemical Addition Tank HPSI Yd. + 20 1 X >0.3g Conservative Fragility Calculation 

Tk-54 Determined HCLPF Capacity is >0~3g. 

9. Secondary ~t Cooling Surge Tank sec SB + 70 1 X >0.3g Conservative Fragility Calculation 

Tk-59 Determined HCLPF Capacity is >0.3g. 

10. Emergency Diesel Day Tank FO AB + 21 1 X FA 0.43g Reflects the Modified Condition. 

Tk·62A CDFM 0.54g 

11. Emergency Diesel Day Tank FO AB + 21 1 X FA 0.43g Reflects the Modified Condition. 

-....] Tk-628 CDFM 0.54g I 

-....] 

12. DG-1A Compressed Air Tank DG AB + 21 1 X >0.3g Conservative Fragility Calculation 

Tk-76A1 Determined HCLPF Capacity is >0.3g. 

13. DG-1A Compressed Air Tank DG AB + 21 1 X >0.3g Conservative Fragility Calculation 

Tk-76A2 (Same supt. as A1) Determined HCLPF Capacity is >0.3g. 

14. DG-1A Compressed Air Tank DG AB + 21 1 X >0.3g Conservative Fragility Calculation 

Tk·76A3 (Same supt. as A1> Determined HCLPF Capacity is >0.3g. 

15. Diesel Starting Air Receiver 1A DG AB + 21 1 X >0.3g Conservative Fragility Calculation 

Tk·76A4 Determined HCLPF Capacity is >0.3g. 

16. Diesel Starting Air Receiver 1A DG AB + 21 1 X >0.3g Conservative fragility Calculation 

Tk·76A5 (Same supt. as A4> Determined HCLPF Capacity is >0.3g. 
Table 7.1-3 Summary of Maine Yankee Equipment Screening and HClPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equipment I tern System Elevation 4334 Review Review Capacity Cocrments 

TANKS (Continued> 

17. Diesel Starting Air Receiver 1A DG AB + 21 1 )( >0 .. 3g Conservative Fragility Calculation 

Tk·76A6 (Same supt. as A4) Determined HCLPF Capacity is >0.3g .. 

18. Dg·1B Compressed Air Tank DG A8 + 21 1 X >0 .. 3g Conservative Fragility Calculation 

Tk-7681 Determined HCLPF Capacity is >0 .. 3g .. 

19. Dg-18 Compressed Air Tank DG A8 + 21' X >0.3g Conservative Fragility Calculation 

Tk-7682 (Same supt .. as 81) Determined HCLPF Capacity is >0.3g. ~ 

20. Dg·18 Compressed Air Tank DG A8 + 2,. X >0.3g Conservative Fragility Calculation 

-..J Tk-7683 (Same supt. as 81) Determined HCLPF Capacity is >0.3g. t 

00 

21. Diesel Starting Air Receiver 18 DG AB + 21 1 X >0.3g Conservative Fragility Calculation 

Tk-7684 Determined HCLPF Capacity is >0 .. 3g. 

22. Diesel Starting Air Receiver 18 DG A8 + 21 1 )( >0.3g Conservative Fragility Calculation 

Tk-7685 (Same supt. as 84) Determined HCLPF Capacity is >0.3g. 

23. Diesel Starting Air Receiver 18 DG A8 + 21 1 )( >0.3g Conservative Fragility Calculation 

Tk-7686 (Same supt. as 84) Determined HCLPF Capacity is >0.3g. 
Table 7.1-3 Surmary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipnent Qual iff cation Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equipment J tem System Elevation 4334 Review Review Capacity Connents 

PUMPS 

1. Prhnary CoqJOt~ent Cooling Puzp PCC TB + 21' c >0.3g HCLPF Capacity Judged >0.3g 

P·9A for Horizontal Pumps. 

2. Primary CoqJOnent Cooling Puzp Pee TB + 21' c >0.3g HCLPF Capacity Judged >0.3g 

P-98 for Horizontal Pumps. 

3. Secondary Coq:JOnent Cooling Plllp sec TB + 21 1 c >0.3g HCLPF Capacity Judged >0.3g 

P·10A for Horizontal Pumps. 

4. Secondary Coq:JOnent Cooling Puzp sec TB + 21' c >0.3g HCLPF Capacity Judged >0.3g 

....,J P-108 for Horizontal Pumps • I 

\0 

5. Charging P~ HPSI PAB + 21 1 c >0.3g HCLPF Capacity Judged >0.3g for 

P-14A Vertical Plllp W/Shaft Length <20 1 • 

6. Charging PUlp HPSI PAB + 21 1 c >0.3g HCLPF Capacity Judged >0.3g for 

P-148 Vertical Pump W/Shaft Length <20•. 

7. Charging Pl.JIP HPSl PAB + 21 1 c >0.3g HCLPF Capacity Judged >0.3g for 

P·14S Vertical Puzp W/Shaft Length <20 1 • 

8. Emergency Feed Pl.JIP P-25A AF\1 AF + 2,. c >0.3g Horizontal Pl.JIP, HCLPF Cap. >0.3g. 

Oil Cooler (Skid mounted) E-86A X >0.3g Oil Cooler HCLPF Cap. Judged 

>0.3g Assessed on \lalkdown Review. 

9. Emergency Feed Turbine PUlp P-258 AF\1 VA + 21 1 c >0.3g Horizontal Pl.JIP, HCLPF Cap. >0.3g. 

Oil Cooler (Skid mounted) E-868 X >0.3g Oil Cooler HCLPF Cap. Judged 

>0.3g Assessed on \lalkdown Review. 
-. 

Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG Valkdown Calculation HClPF 

Equ i pnent Item System Elevation 4334 Review Review Capacity Conments 

PUMPS (Contiooed) 

10. Emergency Feed Pulp P·25C HPSI AF + 21' c >0.3g Horizontal Pump~ HClPF Cap. >0.3g. 

OH Cooler (Skid mounted) E·86C X >0.3g Oil Cooler HClPF Cap. Judged 

>0.3g Assessed on Walkdown Review. 

11. Service Water P~ sw cw + 7• c X >0.3g Conservative Fragility Calculation 

P·29A Determined Motor to Pump Bolts 

HClPF Capacity is >0.3g. 

12. Service Water Pump sw cw + 7• c X >0.3g Conservative Fragility calculation 

P-298 Detenmined Motor to Pump Bolts 

-l HClPF Capacity is >0.3g. I 

-

0 

13. Service Water Pump sv cw + 7• c X >0.3g Conservative Fragility Calculation 

P·29C Determined Motor to P~ Bolts 

HClPF Capacity is >0.3g. 

14. Service Vater Pump S\1 cw + 7• c X >0.3g Conservative Fragility Calculation 

P·29D Determined Motor to P~ Bolts 

HClPF Capacity is >0.3g. 

15. Auxiliary Fuel Oil Transfer P~ FO Yd. + 21 1 c >0.3g HCLPF Capacity Judged >0.3g for 

P·33A (Vertical Pump over Tk·28A) Vertical P~ V/Shaft length <20 1 • 

16. Auxiliary Fuel Oil Transfer Pump FO Yd. + 21' c >0.3g HClPF Capacity Judged >0.3g for 

P·33B (Vertical P~ over Tk·28A) Vertical Pump V/Shaft length <20'. 

17. Containment Spray P~ cs cs + 141 c X >0.3g Conservative Fragility Calculation 

P·61A Determined Motor to P~ Bolts 

HCLPF Capacity is >0.3g. 
.,: 

Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equipment Item System Elevation 4334 Review Review Capacity Conments 

PUMPS (Contin.aed) 

18. Containnent Spray P~ cs cs + 141 c X >0.3g Conservative Fragility calculation 

P·61B Detennined Motor to Pump Bolts 

HCLPF Capacity is >0.3g. 

19. Contairaent Spray ~ cs cs + 141 c X >0.3g Conservative fragility Calculation 

P-61S Oetenmined Motor to Pump Bolts 

·HcLPF capacity is >0.3g. 

HEAT EXCHANGERS 

...,J 

' 1. Resicl.lal Heat .Removal Heat Exchanger * PCC cs + 141 X >0.3g conservative Fragility Calculation 

E-3A Oetenmined the Supports and Anchor 

Bolts HCLPF Capacity is >0.3g. 

2. Residual Heat Removal Heat Exchanger * sec cs + 141 X >0.3g Conservative Fragility calculation 

E-3B Detennined the Supports and Anchor 

Bolts HCLPF Capacity is >0.3g. 

3. Primary Coq»nent Cooling Heat Exchanger PCC TB + 21 1 X >0.3g Conservative Fragility Calculation 

E·4A Determined the Supports and Anchor 

Bolts HCLPF Capacity is >0.3g. 

4. Primary Coq:lonent Cooling Heat Exchanger PCC TB + 2,. X >0.3g Conservative fragility Calculation 

E·4B Detenmined the Supports and Anchor 

Bolts HCLPF capacity is >0.3g. 

5. Secondary C~t Cooling Heat sec TB + 21 1 X >0.3g Conservative Fragility Calculation 

Exchanger E·SA Determined the Supports and Anchor 

Bolts HCLPF Capacity is >0.3g. 
Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Sui lding 

and NUREG IJalkdown Calculation HCLPF 

Equi J:lllent l tern System Elevation 4334 Review Review Capacity Cooments 

HEAT EXCHANGERS (Continued) 

6. Secondary Component Cooling Heat sec TB + 211 X >0.3g Conservative Fragility Calculation 

Exchanger E·5B Determined the Supports and Anchor 

Bolts HCLPF Capacity is >0.3g. 

7. Fuel Pool Heat Exchanger * PCC FB + 21' X >0.3g Conservative Fragility Calculation 

E·25 Determined the Supports and Anchor 

Bolts HCLPF Capacity is >0.3g. 

8. Seal Water Heat Exchanger * * PCC PAB + 11 t --·· Valves PCC·M-90 & 219 Have a HCLPF 

E-34 Capacity >0.3g and will Isolate 

-J 

I this C~ent. 

-

N 

9. Reactor Containment Air Recirculation * PCC RC + 46• X >0.3g Conservative Review Determined 

Cooler E-54-1 - the Cooler Supports and Anchorage 

HCLPF Capacity is >0.3g. 

10. Reactor Containment Air Recirculation * PCC RC + 46' X >0.3g Conservative Review Determined 

Cooler E-54·2 the Cooler Supports and Anchorage 

HCLPF Capacity is >0.3g. 

11. Reactor Containment Air Recirculation * PCC RC + 461 X >0.3g Conservative Review Determined 

Cooler E-54-3 the Cooler Supports and Anchorage 

HCLPF Capacity is >0.3g. 

12. Reactor Containment Air Recirculation * PCC RC + 461 X >0.3g Conservative Review Determined 

Cooler E-54·4 the Cooler Supports and Anchorage 

HCLPF Capacity is >0.3g. 

13. Reactor Containment Air Recirculation * PCC RC + 46• X >0.3g Conservative Review Determined 

Cooler E-54·5 the Cooter Supports and Anchorage 

HCLPF Capacity is >0.3g. 
' ,l ~: 

Table 7.1-3 SLIIIII8ry of Maine Yankee E~ipnent Screening and HCLPF Capacities <Continued) 

E~ipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

E~ipment Item System Elevation 4334 Review Review Capacity Conments 

HEAT EXCHANGERS (ContinJed) 

14. Reactor Containment Air Recirculation * PCC RC + 46' X >0.3g Conservative Review Detennined 

Cooler E·54·6 the Cooler SUpports and Anchorage 

HCLPF Capacity is >0.3g. 

15. Reactor Coolant Regenerative Heat * * HPSI RC + 21 1 ···- Valves PCC·M·90 & 219 Have a HCLPF 

Exchanger E·67 Capacity >0.3g and will Isolate 

this Coftponent. 

16. Safeguards P~ Seal leakage Cooler * sec - cs - 021 X >0.3g Cooler HCLPF capacity Judged 

E·91A >0. 3g Assessed on Wa l kdown Review. 

~ 

' -\.J.) 17. Safeguards P~ Seal Leakage Cooler * PCC cs + oo• X >0.3g Cooler HCLPF Capacity J~ed 

E-918 >0.3g Assessed on Walkdown Review. 

18. Charging Pump Seal Leakage Cooler * sec PAB + 111 X >0.3g Conservative Fragility Calculation 

E·92A Detenmined the Supports and Anchor 

Botts HCLPF Capacity is >0.3g. 

19. Charging Pump Seal Leakage Cooler * PCC PAB + 11 1 X >0.3g Conservative Fragility Calculation 

E·92B Determined the Supports and Anchor 

Bolts HCLPF Capacity is >0.3g. 

20. Seat Water Heater * * HPSI PAB + 11 1 -.-- Valves PCC-M-90 & 219 Have a HCLPF 

E-96 Capacity >0.3g and will Isolate 

this COiq)Onent. 

21. Auxiliary Charging Purp P-7 Lube Oil* CH PAB + 111 X >0.3g Cooter HCLPF Capacity Judged 

Cooter <on skid} >0.3g Assessed on ~alkdown Review. 
Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Item System 

HEAT EXCHANGERS (Continued) 

22. Containment Penetration Cooling Lines* PCC 

(line integrity only) 

23. LPSI PUlp Coolers LPSI * 

(coolers are coils mounted on the pumps) 

24. Containment Spray P~.~~p Coolers * CS 

(coolers are coils mounted on the P-61 Pumps) 

~ISCELLANEOUS COMPONENTS I 

.s::.. 1. 

2. 

3. 

Diesel Generator DG·1A DG 

Diesel Generator Heat Exchanger E-82A * 

DG 1A Engine Control Panel 

Diesel Generator DG-18 DG 

Diesel Generator Heat Exchanger E-828 * 

DG 18 Engine Control Panel 

Seal Water Supply Filter 

FL-348 

* * HPSJ 

'Building 

and 

Elevation 

various 

cs + 21 1 

cs + 141 

AB + 221 

AB + 22 1 

unknown 

Equipment Qualification Screening 

NUREG 

4334 

c 

c 

\lalkdown 

Review 

X 

X 

X 

Calculation 

Review 

HCLPF 

Capacity 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

>0.3g 

C011111ents 

Cooter HCLPF Capacity Judged 

>0.3g Assessed on Walkdown Review. 

Cooler HCLPF Capacity Judged 

>0.3g Assessed on walkdown Review. 

Cooler HCLPF Capacity Judged 

>0.3g Assessed on Walkdown Review. 

HCLPF Capacity is Judged >0.3g 

for the Diesel Gen. and 

Peripherals. 

HCLPF Capacity is Judged >0.3g 

for the Diesel Gen. and 

Peripherals. 

Valves PCC·H·90 & 219 Have a HCLPF 

Capacity >0.3g and will Isolate 

this C~t. 
-~ 

Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipnent Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equipment ltelll System Elevation 4334 Review Review Capacity Ccnnents 

ELECTRJCAL DISTRIBUTION SYSTEMS 

1. 4160V Emergency Bus Elec SB + 46• X >0.3g Conservative Fragility Calculation 

Bus 5 Determined the Anchorage 

HCLPF Capacity is >0.3g. 

2. 4160V Emergency Bus Etec SB + 46• X >0.3g Conservative FragilHy Calculation 

Bus 6 Determined the Anchorage 

HCLPF Capacity is >0.3g. 

3. 480V Emergency Bus Elec SB + 46• X >0.3g Conservative Fragility Calculation 

Bus 7 Determined the Anchorage 

-...J HCLPF Capacity is >0.3g. I 

-Vl 

4. 480V Emergency Bus Elec SB + 46' X >0.3g Conservative Fragility Calculation 

Bus 8 Determined the Anchorage 

HCLPF Capacity is >0.3g. 

5. 480V Emergency Motor Control Center Elec SB + 461 X >0.3g Conservative .Fragility Calculation 

MCC·7A Determined the Anchorage 

HCLPF Capacity is >0.3g. 

6. 480V Emergency Motor Control Center Elec RMC + 21' X >0.3g Conservative Fragility Calculation 

MCC·7B Determined the Anchorage 

HCLPF Capacity is >0.3g. 

7. 480V Emergency Motor Control Center Elec cs + 20 1 X >0.3g Conservative Fragility Calculation 

MCC-781 Determined the Anchorage 

HCLPF Capacity is >0.3g. 
Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

-Building 

and HUREG Walkdown Calculation HCLPF 

Equipnent Item System Elevation 4334 Review Review Capacity Connents 

ElECTRlCAL DISfR.IBUJlON SYSTEMS <Continued) 

8. 480V Emergency Motor Cant rol Center Elec sa + 46' X >0.3g Conservative Fragility Calculation 

MCC·8A Determi Oed the Ancho.rage 

HCLPF Capacity is >0.3g. 

9. 480V Emergency flotor Control Center Elec RMC + 36 1 X >0.3g Conservative Fragility Calculation 

MCC·88 Determined the Anchorage 

HCLPF Capacity is >0.3g. 

10. 480V Emergency Motor Control Center Elec cs + 20' X >0.3g Conservative Fragility Calculation 

MCC·8B1 Determined the Anchorage 

-...l HCLPF Capacity is >0.3g. t 

0\ - 11. 120V AC Vital Bus (Wall mounted) Elec SB + 21 1 X >0.3g Bus HCLPF Capacity Judged >0.3g 

Bus 1A Assessed During Walk.down Review. 

12. 120V AC Vital Bus (Wall mounted) Elec sa + 21 1 X >0.3g Bus HCLPF Capacity Judged >0. 3g 

Bus 2A. Assessed During Walk.down Review. 

13. 120V AC Vital Bus (Wall mounted) Elec sa + 21' X >0.3g Bus HCLPF Capacity Judged >0.3g 

Bus 3A Assessed During Walkdown Review. 

14. 120V AC Vital Bus (Wall mounted) Elec sa + 2,. X >0.3g Bus HCLPF Capacity Judged >O.lg 

Bus 4A Assessed During Walkdown Review. 

15. 125V DC Bus Distribution Cabinet Elec sa + 46' X >0.3g Conservative Fragility Calculation 

Bus 1 Determined the Anchorage 

HCLPF Capacity is >0.3g. 
-, 

Table 7.1·3 Sunary of Mah-.e Yankee EquiF8!f"t Screening and HCLPF capacities (Contii"A2d) 

Equipment I tal 

ELECTRICAL DISTRIBUTION SYSTEMS (tontinued) 

-..,J 

I 

--..,J 

16. 12SV DC Bus Distribution Cabinet 

Bus 2 

17. 125V DC Bus Distribution Cabinet 

Bus 3 

18. 125V DC Bus Distribution Cabinet 

Bus 4 

19. Station Battery No. 1 (1) 

New Lead Catch• Batteries 

20. Station Battery No. 2 (Lead Antimony) 

21. Station Battery Mo. 3 (1) 

New lead Calcha Batteries 

22. Station Battery No. 4 (Lead Antimony) 

Syst• 

Elee 

Elee 

Elee 

Elee 

Elee 

Elee 

Elee 

Building 

and 

Elevation 

sa + 46• 

sa + 46• 

sa + 46' 

sa + 46• 

SB + 46• 

sa + 35 1 

sa + 35 1 

ECf,li~t Qualification Screening 

IIUREG 

4334 

Walt4own 

Review 

Calculation 

Review 

X 

X 

X 

X 

X 

X 

X 

(1} New Lead Calcium Batteries being installed during the next MYC outage (Spring 1987), see Section 4.4. 

(2) Existing MYC Lead Antimony batteries not scheduled to be changed out during the next outage. Lacking data 

KCLPF 

capacity 

>0.3g 

>0.3g 

>0.3g 

>0 .. 3g 

(2) 

>0.3g 

(2) 

on the seismic performance of lead antimony batteries precludes the determination of a HCLPF Capacity, see Section 4.4. 

C~ts 

Conservative Fragility Calculation 

Oetenained the Anchorage 

HCLPF Capacity is >0.3g. 

Ccnservative Fragility Calculation 

Detenained the Anchorage 

HCLPF Capacity is >0 .. 3g. 

COnservative Fragility Calculation 

Oetel'lli ned the Anchorage 

ICCLPF Capacity is >0.3g. 

Conservative Fragi l i.ty Calculation 

Oetennined the Anchorage 

HCLPF capacity is >0 .. 3g. 

Conservative Fragility calculation 

Determined the Anchorage 

HClPF Capacity is >0.3g. 
Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

sui lding 

and NUREG Walkdown Calculation HCLPF 

Equipnent Item System Elevation 4334 Review Review Capacity Connents 

ELECTRICAL DISTRIBUTION SYSTEMS (Continued) 

23. Battery Charger No. 1 Elec sa + 46• X >0.3g Conservative Fragility caLculation 

8C·1 Determined the Anchorage 

HCLPF capacity is >0.3g. 

24. Battery Charger No. 2 Elec S8 + 46• X >0.3g COnservative Fragility Calculation 

BC-2 Deten~i ned the Anchorage 

HCLPf Capacity is >0.3g. 

25. Battery Charger Jlo. 3 Elec SB + 46• X >O.]t Conservative Fragility tateutation 

BC-3 Detenli ned the Anchorage 

.... HCLPF capacity is >0.3g • I 

~ - 26. Battery Charger Jlo. 4 Elec S8 + 46• X >0.3g Conservative Fragility calculation 

BC-4 Det-ensi ned the Anchorage 

HCLPF Capacity is >0.3g. 

27. Inverter No.1 Elec S8 + 46• X FA 0.82g 

INVR-1 CDFM 1.13g 

28. Inverter No.2 Elec SB + 46• X FA 0.82g 

INVR·2 CDFM 1.13g 

29. Inverter No.3 Elec SB + 461 X FA 0.82g 

INVR-3 CDFM 1.13g 

30. Inverter No.4 Elec SB + 46 1 X FA 0.82g 

lNVR-4 COfM 1.13g 
Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equipment Item System Elevation 4334 Review Review Capacity Coaments 

ELECTRICAL DISTRIBUTION SYSTEMS (Continued) 

31. Station Service Transfonner Elee SB + 46' X 0.30g HCLPF Calculated for Modified 

X·507 (located adjacent to Bus 7 & 8) Anchorage Configuration. 

32. Station Service Transformer Elee SB + 46• X 0.30g HCLPF Calculated for Modified 

X·608 (located adjacent to Bus 7 & 8) Anchorage Configuration. 

33. Diesel Generator 1A 480V Distribution Elee AB + 22 1 X >0.3g Panel HCLPF c·apacity Judged 

Panel (wall mounted) >0.3g Assessed During Walkdown 

Review. 

...... 34. Diesel Generator 18 480V Distribution Elee AB + 22 1 X >0.3g Panel HCLPF Capacity Judged t 

- Panel (wall mounted) >0.3g Assessed During Walkdown \0 

Review. 

35. Diesel Generator Control Panel 1A Elec AB + 22 1 X >0.3g Conservative Fragility Calculation 

Determined the Anchorage 

HCLPF Capacity is >0.3g. 

36. Diesel Generator Control Panel 18 Elec AB + 22 1 X >0.3g Conservative Fragility Calculation 

Determined the Anchorage 

HCLPF Capacity is >0.3g. 

37. Distribution Cabinet 1 (wall mounted) Elec SB + 21' X >0.3g Cabinet HCLPF Capacity Judged 

DC-1 >0.3g Assessed During Walkdown 

Review. 

38. Distribution Cabinet 2 (wall mounted) Elec SB + 21' X >0.3g Cabinet HCLPF Capacity Judged 

DC-2 >0.3g Assessed During Walkdown 

Review. 
';' · .. 

Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG Walk down Calculation HCLPF 

Equipment I tern System Elevation 4334 ·Review Review Capacity C00111ents 

ELECTRICAL DISTRIBUTION SYSTEMS (Continued) 

39. Distribution Cabinet 3 (wall mounted) Elec SB + 211 X >0.3g Cabinet HCLPF Capacity Judged 

DC-3 >0.3g Assessed During Walkdown 

Review. 

40. Distribution Cabinet 4 (wall mounted) Elec SB + 21 1 X >0.3g Cabinet HCLPF Capacity Judged 

DC-4 >0.3g Assessed During Walkdown 

Review. 

41. Main Control Board Elec SB + 211 X >0.3g Conservative Fragility Calculation 

120V AC Vital Bus 1·4 Detenmined the Anchorage 

"-.J HCLPF Capacity is >0.3g. I 

IV 

0 

42. Electrical Control Board Elec SB + 21 1 X >0.3g Conservative Fragility Calculation 

DG·1A & 18 Start 1 & 2 Circuits and Control Power Detenmined the Anchorage 

HCLPF Capacity is >0.3g. 

43. Auxiliary Logic Cabinets Elec SB + 21' X >0.3g Conservative Fragility Calculation 

Determined the Anchorage 

HCLPF Capacity is >0.3g. 

44. ESF Auxiliary Panels A & B Elec SB + 21 1 X >0.3g Conservative Fragility Calculation 

Determined the Anchorage 

HCLPF Capacity is >0.3g. 

45. Air Condition Control Panel ACCP Elec SB + 21 1 X >0.3g Conservative Fragility Calculation 

Determined the Anchorage 

HCLPF Capacity is >0.3g. 

46. Safety Parameter Display System Cabinets Elec SB + 21 1 X >0.3g Conservative Fragility Calculation 

Determined the Anchorage 

HCLPF Capacity is >0.3g. 
. ; _ _, .. 

Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG ~allcdown Calculation HCLPF 

Equipnent 1 tem System Elevation 4334 Review Review CapacHy Connents 

HVAC 

1. DG-1A Room Exhaust Fan HV AB + 31' X >0.3g Fan HCLPF Capacity Judged 

FN-20A >0.3g Assessed During ~alkdown 

Review. 

2. DG·1B Room Exhaust Fan HV AB + 31' X >0.3g Fan HCLPF Capacity Judged 

FN·20B >0.3g Assessed During ~alkdown 

Review. 

3. Computer Room Air Conditioner * sec SB + 391 )( 0.38g HCLPF Calculated for Modified 

AC-1A Support Configuration. 

.....,] 

I 

N 4. Computer Room Air Conditioner * PCC SB + 391 X 0.38g HCLPF Calculated for Modified 

AC-18 SUpport Configuration. 

5. Lab Air Conditioner * sec SB + 391 X 0.38g HCLPF Calculated for Modified 

AC-2 Support Configuration. 

6. Diesel Generator Air Intake & Exhaust DG AB + 3,. X >0.3g Daq:>er HCLPF Capacity Judged 

Daq>ers >0.3g Assessed During ~alkdown 

Review. 

7. Containment Spray Fan sec cs + 20 1 X >0.3g Conservative Fragility Calculation 

FN-44A Determined the Anchorage 

(Modified Anchorage) HCLPF Capacity is >0.3g. 

B. Containment Spray Fan sec cs + 20 1 )( >0.3g Conservative Fragility Calculation 

FN·44B Determined the Anchorage 

(Modified Anchorage) HCLPF Capacity is >0.3g. 
Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equipment I tern System Elevation 4334 Review Review Capacity Conments 

VALVES 

1. Aux. Feedwater Regulating Valve AOV AFW AF + 23 1 c >0.3g Valve within bounds 

AFW·A·101 of experience data. 

2 .. Aux. Feedwater Regulating Valve AOV AFW AF + 23 1 c >0.3g Valve within bounds 

AFW·A-201 of experience data. 

3. Aux. Feedwater Regulating Valve AOV AFW AF + 23 1 c >0.3g Valve within bounds 

AFW·A·301 of experience-data. 

4. Block Valve for AFW·A·101 AOV AFW AF + 23 1 c >0.3g Valve within bounds 

-..J AFW·A-338 of experience data. I 

t-V 

t-V 

5. Block Valve for AFW·A-201 AOV AFW AF + 23 1 c >0.3g Valve within bounds 

AFW·A·339 of exped ence data. 

6. Block Valve for AFW·A-301 AOV AFW AF + 23 1 c >0.3g Valve within bounds 

AFW-A-340 of experience data. 

7. HPSI Pump B Discharge to charging header HPSI PT + 13 1 c >0.3g Valve within bounds 

CH·A-32 AOV of experience data. 

8. HPSI Pump A Discharge to charging header HPSI PT + 13 1 c >0.3g Valve within bounds 

CH·A·33 AOV of experience data. 

9. Inlet to Charging Header AOV ** HPSI PAB + 181 c >0.3g Valve within bounds 

CH·F·38 of experience data. 

10. VCT Discharge to HPSI Pumps MOV CH PAB + 241 c >0.3g Valve within bounds 

CH·M-1 of experience data. 
Table 7.1-3 SUrmary of Maine Yankee Ec:,Jipnent Screening and HCLPF CapaciUes (Continued) 

Equipnent Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Ec:,Jipment Item System Elevation 4334 Review Review Capacity C01111e11ts 

VALVES (Continued) 

11. VCT Discharge to HPSI Pumps MOV CH PAB + 24 1 c >0.3g Valve within bounds 

CH-M-87 of experience data. 

12. Containment Spray Header Isolation Valve cs cs + 19 1 c >0.3g Valve within bounds 

CS-M-1 MOV of experience data. 

13. Containment Spray Header Isolation Valve cs cs + 191 c >0.3g Valve within bounds 

CS·M·2 MOV of experience data. 

14. cs P~ Containment Suction MOV cs cs - 08• c >0.3g Valve within bounds 

-....l CS·M-91 of experience data. I 

N 

VJ 

15. cs PlJI't) Containment Suet ion MOV cs cs - 08• c >0.3g Valve within bounds 

CS·M·92 of experience data. 

16. HPSI Discharge to Loop 1 MDV HPSI PAB + 23 1 c X >0.3g Valve exceeds experience 

HSI·M-11 data bounds by 17%. 

17. HPSI Discharge to Loop 1 MOV HPSI PAB + 231 c X >0.3g Valve exceeds experience 

HSI·M·12 data bounds by 17%. 

18. HPSI Discharge to Loop 2 MOV HPSI PAB + 23 1 c X >0.3g Valve exceeds experience 

HSI·M·21 data bounds by 17%. 
J it" 

Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG \lalkdown Calculation HCLPF 

Equipment Item System Elevation 4334 Review Review Capacity Comnents 

VAlVES (Continued) 

19. HPSI Discharge to Loop 2 MOV HPSI PAB + 23 1 c X >0.3g Valve exceeds experience 

HSI·M-22 data bounds by 17%. 

20. HPSI Discharge to Loop 3 MOV HPSI PAB + 23 1 c X >0.3g Valve exceeds experience 

HSI·M-31 data bounds by 17%. 

21. HPSI Discharge to loop 3 MOV HPSI PAB + 23 1 c X >0.3g Valve exceeds experience 

HSI·M·32 data bounds by 17%. 

22. HPSI Discharge to Sl Header MOV HPSI PAB + 23 1 c X >0.3g Valve exceeds experience 

-..1 HSI·M-40 data oounds by 17%. t 

tv 

~ 

23. HPSJ Pump Discharge NOV HPSI PAB + 231 c X >0.3g Valve exceeds experience 

HSI·M-41 data oounds by 17%. 

24. HPSI Pump Discharge MOV HPSI PAB + 23 1 c X >0.3g Valve exceeds experience 

HSI·M·42 data oounds by 17%. 

25. HPSI Discharge to Sl Header MOV HPSI PAB + 231 c X >0.3g Valve exceeds experience 

HSI·M-43 data bounds by 17%. 

26. HPSI Suction from R\IST MOV HPSI Yd + 21 1 c >0.3g Valve within bounds 

HSI·M·SO of experience data. 

27. HPSI Suction from R\IST MOV HPSI Yd + 211 c >0.3g Valve within bounds 

HSI·M·51 of experience data. 
-~ 

Table 7.1·3 Sli!IDary of Maine Yankee Equipnent Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equ i pnent Item Systert Elevation 4334 Review Review Capacity Connents 

VALVES (Continued) 

28. CS Discharge to HPSI Pump MOV cs cs + 19• c >0.3g Valve within bounds 

HSI·M·54 of experience data. 

29. CS Discharge to HPSI Pump MDV cs cs + 191 c >0.3g Valve within bounds 

HSt·M·55 of experience data. 

30. RWST Discharge to LPSI MOV cs Yd. + 281 c >0.3g Valve within bounds 

LSI·M-40 of experience data. 

31. RWST Discharge to LPSI MOV cs Yd. + 28 1 c >0.3g Valve within bounds 

-1 LSl·M-41 of experience data. I 

N 

VI 

32. Decay Heat Release Valve AOV ASDA VA + 43 1 c >0.3g Valve within bounds 

MS·A·162 of experience data. 

33. AFW Pump B Turbine Throttle Valve AFW VA + 2,. c >0.3g Valve within bounds 

MS·A-173 of experience data. 

34. Decay Heat Release MOV SPC VA + 43 1 c >0.3g Valve within bounds 

MS·M-161 of experience data. 

35. Auxiliary Steam Supply Valve MOV MS VA + 43 1 c >0.3g Valve within bounds 

MS·M-255 of experience data. 
) 

Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Blfi lding 

and NUREG Walkdown Calculation HCLPF 

Equi FDent I tern System Elevation 4334 Review Review Capacity Conments 

VALVES (Continued) 

36. Turbine Steam Supply Pressure Control AF\1 VA + 21 1 c >0.3g Valve within bounds 

MS·P-168 of experience data. 

37. Turbine Steam Supply Temperature Control ASDA VA + 411 c >0.3g Valve within bounds 

MS·T-163 AOV of experience data. 

38. PCCW Supply to P·12A Seal Cooler ArN PCC PAB + 04 1 c >0.3g Valve within bounds 

PCC·A-53 of experience data. 

39. Return from Penetration Coolers AOV PCC PT + 12 1 c >0.3g Valve within bounds 

-.J PCC·A-216 of experience data. I 

N 

a-, 

40. Return from Penetration Coolers AOV PCC PT + 121 c >0.3g Valve within bounds 

PCC·A-238 of experience data. 

41. Return from CEA Air Coolers PCC RC + 01' c >0.3g Valve within bounds 

PCC·A·270 of experience data. 

42. DG·1A Cooling Water OUtlet AOV PCC AB + 21 1 c >0.3g Valve within bounds 

PCC·A·493 of experience data. 

43. PCC\1 OUtlet from RHR Heat Exchanger PCC cs + 02 1 c >0.3g Valve within bounds 

PCC·M·43 of experience data. 
Table 7.1·3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipment Qualification Screening 

Building 

and NUREG \lalkdown calculation HCLPF 

E~i pment Item System Elevation 4334 Review Review Capacity Conments 

VALVES (Continued) 

44. PCC\1 Isolation to BR & L\1 Coolers MOV PCC PAB + 11 1 c >0.3g Valve within bounds 

PCC-M-90 of experience data. 

45. PCC\1 Isolation to Letdown Heat Excttangers PCC PAB + 21' c >0.3g Valve within bounds 

PCC·M-150 MOV of experience data. 

46. PCCW Isolation to Containnent PCC PAB + 11' c >0.3g Valve within bounds 

PCC-M-219 MOV of experience data. 

47. Cooler Supply Temperature Control AOV PCC TB + 37 1 c >0.3g Valve within bounds 

-J PCC·T-19 of experience data. I 

N 

-J 

48. Cooler Bypass Temperature Control ArN PCC TB + 371 c >0.3g Valve within bounds 

PCC-T-20 of experience data. 

49. Pressurizer Safety Valve SRV RC + 65 1 c >0.3g Vendor data review foam 

PR·S·11 valve within bounds of 

experience data. 

50. Pressurizer Safety Valve SRV RC + 65 1 c >0.3g Vendor data review found 

PR·S-12 valve within bounds of 

experience data. 

51. Pressurizer Safety Valve SRV RC + 65 1 c >0.3g Vendor data review found 

PR·S-13 valve within bounds of 

experience data. 
Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued} 

Equipment Qualification Screening 

Building 

and NUREG Walkdown Calculation HCLPF 

Equi pnent Item System Elevation 4334 Review Review Capacity Conments 

VALVES (Continued) 

52. Power-Operated Relief Valve PORV RC + 66• c >0.3g Vendor data review fOU'Id 

PR·S-14 valve within bounds of 

experience data. 

53. Power-Operated Relief Valve PORV RC + 66• c >0.3g Vendor data review fOU'Id 

PR·S-15 valve within bounds of 

expedence data. 

54. Power-Operated Block Valve MOV PORV RC + 64• c >0.3g Vendor data review fOU'Id valve 

-...J PR~M-16 within bounds of experience data. I 

IV 

00 

55. Power-Operated Block Valve MOV PORV RC + 64• c >0.3g Vendor data review fOU'Id valve 

PR·M~17 within bOunds'of experience data. 

56. Non-Seismic Return Header Stop Valve AOV sec TB + 21 1 c >0.3g Valve within bounds 

SCC·A-460 of experience data. 

57. Non·Seismic Return Header Stop Valve AOV sec TB + 431 c >0.3g Valve within bounds 

SCC·A-461 of experience data. 

58. Cooler Bypass Temperature Control Ar:JV sec TB + 371 c >0.3g Valve within bounds 

SCC·T·23 of experience data. 

59. Cooler Bypass Temperature Control Ar:JV sec TB + 371 c >0.3g Valve within bounds 

SCC-T-24 of experience data. 
Tabte 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

Equipnent Qualification Screening 

Building 

and llJREG Walkdown calculation HCLPF 

Equi p!lent I tell Syst• Elevation 4334 Review Review Ceplcity C~ta 

VALVES (Continued) 

60. DG·1B Cooler Inlet Te.perature Control sec AI + 23• c >O.Sg Valve within bounds 

SCC·T·305 AOV of experience data. 

61. RCP Seal Water Inlet AOV * * PCC SB + 23• c >0.3g Valve within bounds 

SL·P·3 of experience data. 

PIPING X >0.3g Piping HCLPF Capacity Judged 

>0.3g Assessed During Walkdown 

Review. 

-1 

I 

~CTUATIONS 

1. Instrument Racks various various X >0.3g Racks HCLPF Capacity Judged 

>0.3g Assessed During Walkdown 

Review. 

2. Cable Trays & Conduit various various X >0.3g Cable Tray & Conduit HCLPF Judged 

>0.3g Assessed During Walkdown 

Review. 

3. I~lse Lines (El) 

* Component whose failure may breach critical system pressure boundary. 

* *Valves PCC·M·90 and PCC·M-219 have been determined to have a capacity of 0.3 g•s or higher, thus, these components do not require a seismic 

review as the valves will isolate all components downstream from these valves. 
Equipment Item 

Legend: 

SYSTEM 

AFV 

ASDA 

CH 

cs 

.....J DG I 

(.;.) 

0 

FO 

HPSI 

HV 

PCC 

PORV 

sec 

Table 7.1-3 Summary of Maine Yankee Equipment Screening and HCLPF Capacities (Continued) 

System 

Auxiliary Feedwater 

Alternate Shutdown Decay Heat Removal 

Charging 

Containment Spray 

Diesel Generator Starting 

Fuel Oil 

High Pressure Safety Injection 

Area Heating and Ventilation 

Primary Celq)Onent Cooling 

Power-Operated Relief Valve 

Secondary Coqxment Cooling 

Building 

and 

Elevation 

Equipment Qualification Screening 

NUREG 

4334 

Walkdown 

Review 

SYSTEM 

Calculation 

Review 

SL Seal Water 

HCLPF 

Capacity 

SPC Secondary Pressure Control 

SRV Safety Relief Valves 

sv Se~ice Vater 

Connents 
Table 7.1 ·3 SUIIR8ry of Maine Yankee Equipment Screening and HCLPF Capacities (continued) 

Equi pnent J tem system 

Legend: (Contirw.aed) 

-....J 

I 

UoJ 

Building 

AB Turbine Building Auxiliary Bay 

AF Auxiliary Feed P~ House 

CS Contairment Spray P~ House 

cw Circulation water P~ House 

PAB Primary Auxiliary Building 

PT Pipe T1.n1el 

PV Purge Air Valve Room 

RC Reactor Coolant 

RMC Reactor Motor Control Center Room 

SB Service Building 

TB Turbine Building 

VA Steam and Feed Water Valve Area 

YO Yard 

Building 

and 

Elevation 

Equipaent Qualification Screening 

NUREG 

4334 

Walkdown 

Review 

Calculation 

Review 

HCLPF 

Capacity CCIIIIIet'ltS 
The screened-in components are the ones that are left in in the final event and 

fault trees using which the accident sequence Boolean expressions are derived. The 

seismic fragility parameters and the HCLPF capacities of these components are 

given in Table 5.5-1 in Chapter 5. The nonseismic unavailabilites are listed in 

Table 5.5-2 in Chapter 5. 

7.3 HCLPF Capacity of Plant 

Two dominant accident sequences that could lead to core damage were studied in 

this r,eview: small LOCA and transient. The Boolean expressions were derived for 

these ~sequences consisting of seismic failures and nonseismic unavailabilities. The 

seismic failures were quantified by the seismic fragilites reported in Table 5.5-1. 

For the accident sequence of small LOCA, the HCLPF capacity was determined to 

be 0.21 g pga. This is almost entirely governed by the failure of the refueling 

water storage tank. For the accident sequence of no LOCA, the HCLPF capacity 

was determined to be in the range of 0.32 g to 0.33 g depending on whether the 

nonseismic unavailabilities are considered or not. The component failures 

contributing to this sequence are the transformers, and circulating water 

pumphouse. 

These accident sequences could not be combined together to estimate the core 

damage HCLPF capacity since the split fractions for the two dominant accident 

sequences are not known. Parametric studies were done wherein the different split 

fractions were assumed. The resulting HCLPF capacities were seen to vary from 

0.23 g to 0.32 g for split fraction ranging from SO% to 1% for the small LOCA. 

7.4 Identification of Low Capacity Components 

During the course of this review, certain items were identifcd tobe potential 

seismic vulnerabilites at Maine Yankee. Anong these arc the aged lead antimony 

batteries, the internal seismic supports for transformers, vibration-isolation 

supports for containment spray fans, anchorage of diesel day tank, and block wall 

near the containment spray fans. The utility has proposed that certain 

modifications would be made for the components in the next outage in March 1987. 

The seismic capacities of these components have been estimated in this study in 

their upgraded condition. 

7.5 Conclusions 

This seismic margin review has been performed with the following assumptions 

and limitations: 

o The review earthquake level was specified by the NRC as the 

NUREG/CR-0098 median spectrum. We have interpreted this 

spectrum to be a 84 percent confidence site specific spectrum for 

Maine Yankee. The spectral values arc treated as corresponding 

to the higher ones from the two orthogonal horizontal directions. 

7-32 
o 'The structural models and the floor response spectra generated by 

Maine Yankee have been ·cursorily reviewed by EQE and judged 

to be adequate for the purposes of this margin review. 

o In general EQE is in agreement with the Panel's 

recommendations; where the Panel's recommendations were not 

specific or did not cover a particular. item, they have been 

identified and the review has been accomplished based on our 

own experience fropl past seismic PRAs, earthquake damage 

investigations, and judgment. 

o Since the analysis .. team could not perform the walkdown inside 

the containment, the ·seismic capacity of components inside the 

containment could not be determined. In some instances, Maine 

Yankee provided photographs for certain equipment items inside 

the containment. However, we could not confirm the absence of 

.potential system interaction effects that may make the impulse 

lines inside the containment vulnerable in earthquakes and lead 

to a small LOCA. 

o Our review and walkdowns concentrated on components 

.supporting Group A system functions; however, a review of nonGroup A system components (block walls, for example) led us to 

believe that there are no anomolies at Maine Yankee which would 

violate .the Expert Panel systems screening guidelines. 

o In keeping with the Expert Panel philosophy, the screening of 

components was performed using conservative procedures. For 

the screened-in components, the seismic capacities have been 

calculated using conservative methods since the scope did not 

permit detailed response analysis and thorough investigation of 

failure modes and their capacities. In all cases, the factors 

contributing to the seismic margin and their varilabilities are 

identified and quantified using procedures normally used within 

the state-of-the-art. 

0 The plant level HCLPF capacity was calculated based on two 

accident sequences: small LOCA and no LOCA. Since the small 

LOCA could not be dismissed at this review earthquake level of 

0.30g, it was postulated to occur: the contributing component to 

this sequence HCLPF capacity is R WST. Sensitivity studies 

indicate that the plant level HCLPF capacity is higher than the 

HCLPF capacity of the R WST. 

o The HCLPF capacity of the plant has been determined based on 

the seismic capacities of components in .their existing or 

. proposed modified conditions. Maine Yankee ·has proposed that 

certain modifications or replacements would be made for station 

batteries, transformer , internal core/coil assembly anchorage, 

vibration-isolation supports for containment spray fans and air 

conditioners, anchorage of diesel day tank, and block wall near 

7-33 
the containment spray fans. We recommend that a third 

walkdown of the plant be ·performed to assure that these 

modifications have been carried out and that the HCLPF 

capacities of these components calculated in the present study are 

still applicable. 

Based on this margin review and subject to the above restrictions, EQE confirms 

that the high confidence low probability of failure capacity of Maine Yankee 

plant is at least 0.2lg refered to the ground response spectrum specified above; it is 

also our judgment that the actual capacity of the plant is much higher since the 

assumption of a small LOCA occuring is conservative. 

7-34 
REFERENCES 

American Concrete Institute (1981 ), Building Code Reauirements for 

Concrete Masonry Structures (ACI 531-79) (Revised 1981) and Commentary -

ACI 531R-79. 

ANCO, Engineers, Inc., Generic Seismic Qualification Using Existing Test 

Data, 

Bernreuter, D.L., J.B. Savy, R.W. Mensing, J.C. Chen, and B.C. Davis 

(April 1985), Seismic Hazard Characterization of the Eastern United States 

Volume 1: Methodology and Results for Ten Sites, UCID-20421 Vol. 1. 

Lawrence Livermore National Laboratory, Livermore, California. 

Budnitz, R.J., et al., (July 1985), An Aporoach to the Quantification of 

Seismic Margins in Nuclear Power Plants, NUREG/CR-4334. 

CampbelL R.D., R.P. Kennedy, and R.P. Kassawara May 1986, Seismic Margin 

Review Methodology, Draft EPRI Report. 

Coats, D.W., May 1980, Recommended Revisions to Nuclear Regulatory 

Commission Seismic Design Criteria, NUREG/CR-1161. 

CYGNA Energy Services (April 1982), Dynamic Modeling of Class I 

Structures. Maine Yankee Atomic Power Station, Report BM-Y-MY-80006-5, 

Rev. 0. 

EQE Incorporated (October 1986), The Performance of Cable Tray and 

Conduit Systems In Actual and Simulated Seismic Motion, Report 

82001.05.001 Draft B, prepared for the Seismic Qualification Utility Group. 

Gerard, G. and H. Becker (1957), Handbook of Structural Stability, NACA 

TN 3783. 

Hall, D.H. (April 1981), Proposed Steel Column Strength Criteria, Journal of 

The Structural Division, ASCE. 

Haroun, M.A. and G. W. Hausner (April 1981), Seismic Design of Liquid 

Storage Tanks, Journal of The Structural Division, ASCE. 

Hashimoto, P.S., R.P. Kennedy, R.B. Narver, and D.A. Wesley (December 

1984), Conservative Seismic Capacities of the Maine Yankee Reactor 

Containment Including and Excluding Design Incident Pressure, NTS/SMA 

36001.01, prepared for Maine Yankee Atomic Power Company by Structural 

Mechanics Associates. 

Henries, W.E., and S.R. Boeschen (March 1986), Evaluation of AOVs and 

MOVs to a Greater Than Design Base Earthquake in the Maine yankee 

Nuclear Power Plant, Prepared for Maine Yankee Atomic Power Company. 

Kaplan, S. (1981), On the Method of Discrete Probability Distribution in Risk 

and Reliability Calculations, Risk Analysis, Volume 1. 

R-1 
' ' . ~ 

Kennedy, R.P. and M.K. Ravindra (May 1984), Seismic Fragilities for Nuclear 

Power Plant Risk Studies, Nuclear Engineering and Design, Vol. 79, No. 1, pp. 

47-68. 

Manos, G.C. (August 1986), Earthquake Tank-Wall Stability of Unanchored 

Tanks, Journal of the Structural Engineering, ASCE. 

National Aeronautics and Space Administration (June 1966), Shell Analysis 

Manual. 

Newmark, N.M. and W.J. Hall (May 1978), Development of Criteria For 

Seismic Review of Selected Nuclear Power Plants, NUREG/CR-0098. 

Prassinos, P.G., M.K. Ravindra, and J.B. Savy (October 1985), Trial Guidelines 

for Seismic Margin Reviews of Nuclear Power Plants, NUREG/CR-4482, 

Lawrence Livermore National Laboratory, Livermore, California. 

PRA Procedures Guide, NUREG/CR-2300, 2 volumes, January 1983. 

Ravindra, M.K., and R.P. Kennedy (August 1983), .. Lessons Learned from 

Seismic PRA Studes," Paper M6/4, Proceedings of the Seventh Conference on 

Structural Mechanics in Reactor Technology, Chicago, Illinois. 

Reed, J.W., R.P. Kennedy, and M.K. Ravindra (December 5, 1986), Memo to 

Fragility - Hazard Interaction Subgroup Joint Working Group on Seismic 

PRA, ASCE Dynamic Analysis Committee. 

Smith, G.V., An Evaluation of the Yield. Tensile, Creep. and Trupture 

Strengths of Wrought 304. 316. 321. and 347 Stainless Steels At Elevated 

Temperatures, ASTM Data Series DS 5S2. 

Stevenson, J.D., and W.E. Henries (January 1983), Seismic Review of the 

Maine Yankee Nuclear Power Plant, Volume I, prepared for the Maine 

Yankee Atomic Power Company. 

Structural Mechanics Associates (February 1983), Seismic Margin Review 

Midland Energy Center Project. Volume I. Methodology and Criteria, 

prepared for Consumers Power Company. 

URS Corporation/John A. Blume & Associates, Engineers, Seismic Anchorage 

Guidelines for Nuclear Power Plant Equipment. Volume 1: Development and 

Use of Guidelines, prepared for Electric Power Research Institute, Palo Alto, 

California. 

Weingarten, V.I., E.J. Morgan, and P. Seide (March 1965), Elastic Stability of 

Thin-Walled Cylindrical and Conical Shells Under Axial Compression, AIAA 

Journal. 

Wozniak, R.S. and W.W. Mitchell (May 9, 1978), Basis of Seismic Design 

Provisions For Welded Steel Oil Storage Tanks, American Petroleum Institute, 

43rd Midyear Meeting, Toronto, Ontario, Canada. 

R-2 
STRUCTURE 

APPENDIX A 

MAINE YANKEE ATOMIC POWER STATION 

ARRANGEMENT DRAWINGS 

CONTENTS 

Site Plan ........................................ ,................................................................................................ A-2 

Plot Plan ......................................................................................................................................... A-3 

Reactor Containment Plan, El. 46' .. 0" .............................................................................. A-4 

Containment Plan, El. 20'·0" ................................................................................................ A~S 

Primary ·Auxiliary Building Arrangement .................................................................... A-6 

Primary Auxiliary Building Arrangement .................................................................... A-7 

Turbine Building El. 21 '-0" ................................................................................................... A~8 

Turbine Building El. 39'-0" ................................................................................................... A-9 

Turbine Building Plan ........................................................................................................... A-1 0 

Service Building Addition .................................................................................................... A~ I I 

Circulating Water Pump House ......................................................................................... A-12 

·Turbine Building El. 21' .. 0" .................................................................................................. A-13 
--~- ··- ........... 

---

N 

I 

< 

! 

i ~ =t 

~~ ; ! !o 

7 :1 J. I I :,; . l ... p ~ I' " d . 

~ j~· ti pt : ... 

:2. !H .. : 

: k !: ! .-.~ \oo 

il': =~ ~~; : ... .,o It • i !·l i! ~~· 

1!1! • ~ 0 ~~l : ... ;: . iii 

i~ ~ ii~ ~l ~h !.i~ ,~~ ;.~ :~4 .:::;a i r~ ~ 

;t; '! 

t j•l 

(J -; 

I 

iH 

i 

r 

..... 

z 

1.1.1 

.~k :::E 

z ~:r..""'i-\':! < ·~~ ..... l···!·liil z 

0 

)\_ u 

a:: 

0 

\ 

!_:::::, ... -··-·--~ :t; ..... 

(.) 

ltl <( 

! ,-h r::; w 

c.:.r ;;: a: 13 ! . I 

I 

.1:!1.. 

-· "1=i-

• 0 

I 

~ 

'If 

....i 

LIJ 

z 

<( 

...J 

a.. 

.,, 

!,g 

ii! 

ih· ,:.1 .:!i 

•••• J!J! ···I •U. 

0 

u 

c • Q. 

u 

:i 

Cl 

lie 

"' ~ z 

: 

1-.1 z 

.c 

:1! 

"" 

~ 

'-1 

" 

I 

..,. 

I 

II( 

I 

I 

.I 
~ I 

J 

~~ ll 

. " ~ i :f.,. 

l ~ 

~ 

jl 

1,/) 

I 

C( 
!l 

;::~ 

~ 

" lL 

u!i! 

~~----

c:i 

u 

ri >- !lr 0::: A. 

<( ,_ 5l d (.!) 2 

C) X a: ... ::> 0:: C( 

<[ <[ ..... I ..... 

>- C!) :.: 

z I 0:: z ~ 

"' <'t i5 

"' :::!: -l ..... 

2 EE 5 !!: 

<I II. a. CD :E 

":> 

L..J 

co: 

10 

I 

< 
I 

>-t- 0 

+ 

a::~ u 

:!:!e> a:: ~~ 

;'rg fw ... :: ..JZ ~ 

6 oil -a:; (,.) %:Z ss XQ: ,. l :i ,; :E i: i 1 i i 1l s li 5 

~- j act ~ ! I § II 

!~ r .... ! ! J w "''l:J i )-C) hi LLJ 

: ·i i- N a;z :.::: 

l""' II"' . z 1':! .., ... : . ., ... fl') ct- S! ~'\~ lri ~Q _ _. w 

S2 a;- ;;; 1.1.. a..m <{ 

:I 

.. 

S! ,... 

< • 

"' '> 

~ • 

k:JotrtJ..,J.. I':~~UK).io 

rf 
I ... 0 

. 3~ 

r 

I 
., . 

t 

T·--------------------~:~~=~--~1 

01 

I 

< 
.. ....... .:2. .,. 

. .... ...... "- . ..... 

:a 

i 

' 

I • 

t 

I 

. :!. 

0 

r-1 

I 

< 
'J-

, 

I • ·-1-·-· 

I 

I 

,i 

... 

.... 

I 

.... 

... +-· 

l 

I 

! ' 

-l 

l 

, 

.!i 

·'""f IL. 

·---· 

r 1 

····"" 

<!) 

z 

a 

...J ~ 5 CD 

CD 

> 

w a: 

":' <t ... ~ :J III 0 x ~ ~ ;:) 

-~ ii: < 

(E ··- :0 ffr o:J '•L . 

z 

<( 

.J 

a. 

d 

u 

a: 

II 

Q, 

~ ~ 

loJ 

w 

:.: 

z 

~ M 

1.&1 .-I 

I li .;! 

"' =:! 

NRC FORM 336 

(2-84) 

U.S. NUCLEAR REGULATORY COMMISSION 

NACM 1102, 

3201, 3202 BIBLIOGRAPHIC DATA SHEET 

SEE INSTRUCTIONS ON THE REVERSE. 

2. TITLE AND SUBTITLE 

Seismic Margin Review of the Maine Yankee Atomic 

Power Station 

Volume 3. Fragility Analysis 

5. AUTHOR!S) 

M. K. Ravindra, G. S. Hardy, P. S. Hashimoto, and 

J. J. Griffin 

7. PERFORMING ORGANIZAl'ION NAME AND MAILING ADDRESS (Include Zip Code) 

EQE, Inc. 

1. REPORT NUMBER {Assigned bv T/DC. ildd Vol. No., if <rnvJ 

NUREG/CR-4826 

UCID-20948 

Vol. 3 3. LEAVE EILANK 

4. DATE REPORT COMPLETED 

MONTH YEAR 

February 1987 6. DATE REPORT ISSUED 

MONTH YEAR 

March 1987 6. PROJECT/TASK/WORK UNIT NUMBER 

Newport Beach, CA 92660 9. FIN oR GRANT NUMBER 

Under Contract to: Lawrence Livermore National Laborator; 

Livermore, CA 94550 A0461 

10. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (lhcludeZip Code! 

Division of Engineering Safety Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 

12. SUPPLEMENTARY NOTES 

13. ABSTRACT (200 words or lsss) 

1 1a. TYPE OF REPORT 

Technical 

b. PERIOD COVERED (Inclusive dates) 

June 1986-March 1987 

This Fragility Analysis is the third of three volumes for the Seismic Margin Review of 

the Maine Yankee Atomic Power Station. Volume 1 is the Summary Report of the first 

trial seism1c marg1n review. Volume 2, Systems Analysis, document the results of the 

systems screening for the review. The three volumes demonstrate how the seismic 

margins revi~w guidance (NUREG/CR-4482) of the NRC Seismic Design Margins Program 

can be applied. 

The overall objectives of the trial review are to assess the seismic margins of a 

particular pressurized water reactor, and to test the adequacy of this review 

approach, quantification techniques, and guidelines for performing the revtew. Results 

from the trial review will be used to revise the seismic margin methodology and 

guidelines so that the NRC and industry can readily apply them to assess the inherent 

quantitative seismic capacity of nuclear power plants. 

14. DOCUMENT ANALYSIS a. KEYWORDS/DESCRIPTORS 

Seismic Margins Seismic Analysis 

b. IDENTIFIERS/OPEN-ENDED TERMS 

RU,S. G.P.O. 1987-181-682160027 

15. AVAILABILITY 

STATEMENT 

Unlimited 

16. SECURITY CLASSIFICATION 

(This p11ge) 

Unclassified 

(This reporrl 

Unclassified 17. NUMBER OF PAGES 

18. PRICE