ML20210G027

Amend 1 to Action Plan for Performance Improvement

ML20210G027
Person / Time
Site:	Rancho Seco
Issue date:	09/30/1986
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
Shared Package
ML20210G008	List: ML20210G003 ML20210G027
References
PROC-860930, NUDOCS 8609250323
Download: ML20210G027 (497)

Text

SACRAMENTO MUNICIPAL UTILITY DISTRICT i

Rancho Seco amenament ;

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N 8609250323 DR 860'/13 ADOCK 05000312 SMUD """

September,198 SACRAMENTO MUNICIPAL UTILITY DISTRICT 6201 S Street, P.O. Box 15830, Sacramento CA 95852-1830 YOUR ELECTRIC SERVICE

RANCHO SECO l ACTION PLAN AMENOMENT 1 SEPTEMBER, 1986 1

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SUMMARY

OF CHANGES --

Page Change Reason for Change I change Appendix D title from. original Appendix D was for

" Assessment and Comparison of information, new Appendix D Test Programs" to " Action Plan provides index to commitments Commitment Summary" in Action Plan i thru- added detailed index uses page clarification i xit nu%ers 1 through 9, roman numerals not used 1-1 General: throughout document clarification and referencing 1 numbers have been provided and

, all commitments have been identified -

I with a unique 1 number 1-8 reverse order of items 2 & 3 to be consistent'with in 1 1.1 subsequent presentation 1-11 . restructure first sentence in clarity 1 1.1.3 1-11 item 1.1.3.d change STOP-TRIP to program name change

SPIP l-12 add description of QCI-12 program respond to NRC letter of 9/5/86 relationship to existing prcgrams item 4 1-12 1 1.1.4, add reference to section clarification 40 l

l-13 expanded 1 1.2 to address relation- respond to NRC letter of 9/5/86 ship between PP-MIP and MSRC/PRC item 11 1-14 to end of 3rd bullet in i 1.2.1, reflect process change ,

add "...and for system related items.to the System Engineer for information..."

l-15 General: " Assistant General title change Manager-Nuclear", to " Deputy General Manager-Nuclear";

change made throughout Action Plan 1

l-17 under "Long Term", add similar completeness 1 items from 1 2.3.1.3 i 1-18 update to reflect completion status change of task 1

Page Change Reason for Change 1-19 change Appendix E from entire entire list is several hundred QTS list of open items to be pages long--copies are a sampling available at Rancho Seco 1-20 Revise 1 1.4.1 to Reflect SRTP SRTP Development changes 1-21 revise 11.4.1 to reflect changes description of program change in SRTP l-21 1 1.4.2 changed wording clarification 1-22 change " Figure 1-3" to typographical error

" Figure 1-1" 2-2 Asst. General Manager-Nuclear to SMUD organizational change Deputy General Manager-Nuclear 2-6 Asst. General Manager-Nuclear to SMUD organizational change Deputy General Manager-Nuclear i

2-10 Asst. General Manager-Nuclear to SMUD organizational change Deputy General Manager-Nuclear 2-19 added 1's to criteria clarification 2-20 revised to reflect responsibility - consolidation of Tracking &.

for Tracking recommendations Closecut 2-22 1 2.3.4 change " deficiency" to clarification

" deficiencies" 2-23 same as page 2-6 title change 2-24 same as page 2-23 title change 2-25 wording in i 1, make list editorial of items 2-26 end of section 2.5, change consistency throughout document

" Prior to Restart" to "as promptly as practical" 3-3 same as page 2-23 title change

3-5 reworded sentence was not a sentence 3-9 same as page 2-23 title change

! 9.

11

p Page Change Reason for Change l

(J 4A-1 added 1 4A.a on Commitment lessons learned to long-term Systematic Assessment Program 4A-1 changed B&W Owners Group program title changed program name from STOP-TRIP to SPIP 4A-2 changed EM Cwners Group program title changed program name from STOP-TRIP to SPIP 4A-3 added 1 #'s plus new clarification and scope commitment, 1 e change 4A-4 4A.I.2, added comitment to assess comprehensiveness need for furthar reviews 4A-5 4A.2.2, added explanation of how respond to NRC letter of partial loss of systems was 9/5/86, item 8 and item 9 considered, and how the program considered in multiple events 4A-8 4A.2.3.4, add commitment to applicability of DFC approach evaluate IE Bulletin 79-27 by 0FC to issues within Bulletin d techniques 4A-9 installed new Section 4A.3 expands description of SPIP, thru which replaced earlier BHOG 4A-13 STOP-TRIP Program 4A-14 sections 4A.4.2 and 4A.4.3, status change and respond to update to reflect completion NRC letter of 9/5/86, item 13 of task and references to Appendix I (except from QCI-12) 4A-15 reworded sentence reflect current status 48-8 added description of QTS in 48.2.1.2 QTS satisfies commitment 48-9 "

... identified..." replaced use of correct word

... associated..." in 1 48.2.2.2 4B-15 revised discussion to explain respond to NRC letter of intent of commitment, consolidated 9/5/86, item 14 48.4.1.2, 3, added note 48-16 reworded sentence to provide purpose for ,

commitment ' '

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Page Change Reason for Change -

4B-17 added commitment #8 needed for comprehensive management of program 48-18 added T 4B.4.1.9 which states respond to NRC letter of role of V&V program 9/5/86, item 6 4B-20 48.5.1, added reference as to which clarification values were identified 48-21 48.5.2.6, "...in place..." replaces clarification

... installed..."

48-28 48.9 added discussion on role of respond to NRC letter of human factors in modification 9/5/86, item 5 process 48-28 48.9.1.1.b,"...handjack clarification of corrective position..." replaced "...value action commitment status..."

48-30 48.9.2.3, added information on B&W information assessment 48-31 added item 48.9.3.3 to ensure consistency of modifications to the control room 48-31 48.10.2.1, revised description of refinement of commitment the Nuclear Information System being designed in 1 48.11.1.2 1 48-31 1 4B.11.1 added reference to respond to NRC letter of NUREG 0737 9/5/86, item 7 48-33 inserted reference to V 48.11.1.1 clarification 48-34 reworded sentence in 1 48.11.2.1 clarification 4C-1 4C, add d..., addressing issues from clarification the 12/26 event...."

4C-3,4,5 4C.1., revised commitments reflect changes due to commitment to install EFIC 4C-7 revised commitments reflect changes due to commitment to install EFIC 4C-9 revised description of commitment provide current project status 4C-13 4C.2.a.la, revised wording clarification tv l

Page Change Reason for Change 7' 'S 4C-13 4C.2.a.l.b revised wording clarification 4C-15 added item 4C.2.b.l.4 moved commitment to inspect electrical terminations from 4C.2.a 4C-16 add section 4C.2.d, upgrade provide response to NRC letter SPDS to class I of 9/5/86, item 3 4C-17 1 4C.3.a "... prior to restart..." change in commitment advanced replaced "... cycle 8..." installation of EFIC 4C-18 revised section 4C.3.a/.3a.1 change in commitment advanced installation of EFIC 4C-20 4C.3.d.l.4, change "MSLFL" to MSLFL is replaced by EFIC

" Main Stream / Main Feedwater" 4C-28 4C.8.1.6, added words clarification

... priority 1..."

4C-29 4C.8.2.1, added words clarification

... priority 2 & 3..."

4C-34,35 revised section 4C.11 on MOV's describes expanded programs and commitments 4D-1 insert revised section 4D provide detailed description thru of developed System Review 4D-30 and Test Program B-2 added reference to closure clarification activities, B.1 B-3 added reference to closure clarification activities, B.2 B-5 added reference to closure clarification i activities, B.3 B-7 added reference to closure clarification activities, B.4 B-9 added reference to closure clarification activities, B.5 B-11 added reference to closure clarification activities, 8.6

% B-11 section B.6 amended provide response to NRC letter of 9/5/86, item 19(a) v

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Page Change Reason for Change i

B-14 added reference to closure clarification activities for B.7 B-15 section B.7 amended response to NRC letter of 9/5/86, item 19(b)

B-15 1 B.7.1 " Transition" replaces clarification

" Conversion" B-20 B.8, added reference to closure clarification activities B-21 B.9, added reference to closure clarification activities B-23 B.10, added reference to closure clarification activities B-23 B.10, deleted " Note" note was to wr1ter not to be a part of Plan B-23 B.10, revised paragraph provide reference to commitment in Action Plan B-26 add last 1 to B.11 specify nature of commitment B-29 B.13, added reference-to Action Plan clarification B-32 B.14, added reference to Action Plan clarification B-34+ B.15, added references to commit- references ments and closure actions B-50 B.15.c, added text which had not completeness been typed into original B-51 added comment about ICS Tuning clarification B-53 expanded response to include information SPIP efforts B-55 B.15.d, added references to commit- clarification ments in Action Plan B-55 revised District response to better describe actions taken FINDING B.15.d to resolve Finding i

B-55 added references to commitments clarification in Action Plan O

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Page Change Reason for Change -

B-56 B.15.e added references to commit- clarification ments in Action Plan B-57 B.15.f. changed March 29, 1984 to date of event March 19, 1984 B-58,59 B.15.g., expanded description of clarification 3/19/84 event and references to Action Plan B-61 deleted first page on finding duplicate entry B.15.h i

B-61 B.15.h, specified references within clarification

-Action Plan -

t B-62 B.16, "... Fan..." replaced typographical error

... Fair..."

B-63 added sentence clarification l B-64 B.17, provided reference to Action clarification -

Plan  !

B-65 B.18, provided reference to Action clarification Plan I i

B-67 B.19, added words completeness

...and control..."  !

B-68 B.20, provided reference to Action clarification l

Plan l

B-69 B.21, provided reference to Action clarification Plan B-70 B.22, provided reference to Action clarification Plan B-71 B.23, provided reference to Action clarification Plan l

B-73 B.24, provided reference to Action clarification Plan B-75 B.25, provided reference to Action clarification Plan l C-1 changed paragraph numbers to consistency reflect changes in text i

v11

Page Change Reason for Change C-12+ added references to NRC questions response.to NRC letter of 9/5/86 0-1 repalced old appendix with listing improve useability of document of Action Plan Commitments G-1 provide updated example System provide example of typical SSR Status Report for information I-l provide new Appendix on Interview response to NRC letter of Program 9/5/86, item 13 e

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INDEX Section Page 1.0 Introduction & Program Overview 1-1

.1 Post Event Review 1-3 l

.2 The objectives of the Action Plan 1-4 1.1 Input Process (Issue Identification) 1-8

.1 Department Managers Hardware and Programmatic 1-8 Recommendations

.2 Management Process Review 1-10 i

.3 Systematic Assessment Process (QCl-12) 1-11

.4 System Review and Test Program (SRTP) 1-12 1.2 Issue Evaluation and Disposition Procesi Management 1-13

.1 Process Overview 1-13

.2 The Recommendation, Review and Resolution Board (RRRB) 1-14 p .3 The PAG Committee Membership 1-15

.4 Activity Prioritization Criteria 1-16 1.3 Implementation of Action Plan Activities 1-18

.I Modifications and Maintenance Improvement Actions _1-18

.2 Management, Operations, and Administrative Process 1-19 Improvement Actions 1.4 System Review and Test Program 1-20  ;

.1 Systems Review and test Program 1-20

.2 System Functional Review 1-21 i

1.5 Issue Resolution Closeout 1-24 l 1.6 Independent Program Oversight 1-25 1.7 Document Purpose 1-26

.1 Organization - Body 1-26 l

.2 Organization - Appendices 1-27 1-29

( 1.8 Conclusion

Section Page 2.0 Management of the Action Plan 2-1 2.1 Independent Review Group (IRG) 2-2

.1 Purpose 2-2

.2 Hission 2-3

.3 Tasks 2-3

.4 IRG Findings 2-5 2.2 Restart and Implementation Organization 2-6

.1 Responsibilities 2-6

- The Deputy General Manager, Nuclear 2-6 -

The Restart / Implementation !!anager (RIM) 2-6 j The Outage Manager 2-7 The Test Program Director 2-7 l The Program Ofrector 2-8 The Manager, Nuclear Plant 2-8 -

The Manager, Nuclear Engineering 2-9 The Manager, Nuclear Projects 2-9 The Manager, Site QA 2-9 The Manager, Nuclear Training 2-10 The Manager, Nuclear Licensing 2-10  ;

.2 Qualifications 2-10 i 2.3 Action Plan Activity Management Process 2-17 :

.1 The establishmnet of the Recommendation Review and 2-17 '

Resolution Board (RRRB)

.1 Criteria for prioritizing restart schedule 2-19

.2 The establishment of the Performance Analysis Group 2-20 ,

.3 Implementation and Close-out 2-21 I

.4 Action Plan Activity Tracking and Reporting 2-22 Section Page 2.4 Review Meetings and Reports 2-23 Prior to Power operation Restart Report 2-23

.1 Internal 2-23

.2 External 2-24 2.5 Transition Actions 2-25 3.0 Performance Improvements Underway Prior to the 12/26 Event' 3-1 3.1 Projects Underway Prior to December 26, 1985 3-3

.I Staffing and Organization 3-3

.2 - Training Program 3-4 -.

1. Management Restructure 3-4

2. INPO Accreditation Effort 3-5

.3 Maintenance Program 3-6

.4 Quality Assurance 3-7

.5 Systematic Troubleshooting '3-8

.6 Root Cause Program 3-9 Activity Assessment 3-10 '

l

.7 4.0 Restart and Performance Improvement Action Plan 4-1 j 4A Systematic Assessment Program 4A-1 [

4A.1 Precusor Review Program 4A-2 4A.2 Deterministic Failure Consequence Analysis 4A-5 4A.3 B&W Owners Group Programs - Safety and Performance 4A-9 Improvement Program (SPIP) 4A.4 Plant Interviews 4A-14 48 Management, Operations, and Administrative Process Improvement 48-1 48.1 Management Effectiveness 48-2 ,

48.2 Quality and Quality Assurance 48-7

- - . , _ . - . . _ _ , _ _ . - . . - . , _ - . _ , . . . . _ . . _ - . , , _ . , _ - _ _ _ - - , - . - _ . , , - . . . . . . - _ , , , , _ _ , , . _ . . . . _ , , _ - _ _ _ , - - . , _ _ . _ . - . , , , . . - ~ _ . _ . , _ , ,

Section Page 48.3 Training 48-10

.1 Management Controls, Facilities and Resources 48-10

.2 Specific Instruction Actions Items - Operations 48-11 l

.3 Specific Instruction Action Items - Health Physics 48-13 )

.4 Specific Instruction Action Items - Emergency Planning 48-14 I 48.4 Operations and Operating Procedures 48  !

48.5 Maintenance Programs and Procedures 48-19 48.6 Health Physics and Radiological Controls 48-22 )

48.7 10CFR50 Appendix I Discharge Guidelines 48-23 -

48.8 Emergency Preparedness 48 48.9 Human Factors 48-28 48.10 Management Information System 48-31 4B.11 Commitment Management 48-33 48.12 Configuation Management 48-35 48.13 Materials Management 48-38 4C Plant Modifications and Maintenance Improvements 4C-1 4C.1 Integrated Control System (ICS) and Interfacing Systems 4C-2 1.a General Programmatic Actions 4C-2 1.b Actions to Raduce the Impact of Power Loss on ADV's 4C-6 and TBV's 1.c Actions to Address the Adequacy of the ICS Power Monitors 4C-7 1.d Actions to Improve the Status Indication on Loss of ICS 4C-8 g Power 1.e Actions to Evaluate the Failure Consequences of Various 4C-9 l

ICS Inputs, Outputs, and Components  ;

1.f Actions to Address the Adequacy of ICS Power Supply 4C-10 ;

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1 l

Section Page a l 4C.2 Non-Nuclear Instrumentation (NNI) 4C-12

(]N 2.a General Programmatic Actions 4C-12 ,

2.b Actions to Address the Adequacy of NNI Power Monitors 4C-14 2.c Actions to Address the Adequacy of Status Indication for 4C-15 NNI and Affected Instrumentation on Loss of NNI Power 4C.3 Feedwater and Steam Systems 4C-17 3.a Actions Associated with Emergency Feedwater Initiation 4C-17 and Control-3.b Action to Improve Auxiliary Steam Control Valve Operation 4C-18 on Loss of ICS Power 1 . .

3.c Actions to Reduce Main Feedwater Contributions to Reactor 4C-19 Trips 3.d Actions from Control Room to Ensure Capability to Isolate 4C-20 Main Steam System 4C.4 Emergency Diesel Generator Reliability 4C-21 4C.5 Reactor Coolant System and Pressurizer 4C-22 5.a Upgrade Pressurizer Relief Valve Discharge Piping Supports 4C-22 4C.6 Enhance the Post Accident Sampling System (PASS) Operability 4C-23

! 4C.7 Actions to Enhance Control Room /TSC and NSEB HVAC - Operability 4C-25

. and Reliability 4C.8 Instrument Air System. Reliability 4C-28 4C.9 Reactor Building Purge Flow Rate Measurements 4C-30 4C.10 Fire Protection Systems 4C-31 10.a Fire Alarm Systems 4C-31 r

10.b Separation of NSEB Damper Controls and Equipment 4C-32 10.c Water Leakage through Floors Following Actuation of Fire 4C-32 Protection Systems 4C.11 Motor Operated Valves 4C-34 i

4C.12 Critical Pumps Fatlure on Loss of Suction 4C-36 l

4C.13 Maintenance Programs and Actions 4C-37

Section Page 4C.14 Once Through Steam Generators (OTSG'S) 4C-40 40 System Review and Testing Program 40-1

1. Purpose 40-2

2. Organization 40-3

3. Responsibilities 40-5

a. System Review and Test Program Director 40-5

b. Test Review Group 40-7

c. Performance Analysis Group 40-8

4. -System Selection for Systems Review 40-8 -

5. System Review Report Summary 40-15

a. System Status Report Overview 40-15

b. System Investigation Report Overview 40-16

c. Details of Report Components 40-17

6. Restart Testing Program 40-22 4D.7 Comparison Between Rancho Seco and Davis-Besse Programs 40-25

.1 System Selection 40-25

.2 Problem Identification 40-26

.3 System Review 40-28 l l'

.4 Test Program Development 40-29

.5 Restart Testing 40-30 APPENDIX A A-1 District Board of Directors' Policy Statement on Performance A-1 Improvement at Rancho Seco APPENDIX B 8-1 Specific District Responses to NUREG--1195 Findings B-1 i

O Section Py

, Findings B-1 B.1 Power Supply Monitor B-1 B.2 Repositioning of ICS Controlled Valves or Loss of ICS B-3

. B.3 PM Program for Manual Valves B-4 B.4 Procedural Guidance for loss of ICS B-6 B.5 Feedpump Trip Criteria B-B B.6 Priority of PTS or Pressurizer Level B-10 B.7 Training on Loss / Restoration of ICS Pcwer B-14 B.8 Recognition of ICS Power Condition B-20 -

B.9 Damaged. Hand Operator on AFW Valve B-21 B.10 Radiological Controls and Emergency Preparedness B-23 B.11 Installation of EFIC B-25 B.12 Reactor Vessel Thermal Shock B-27 B.13 Reactor Vessel Integrity B-29

B.14 Use'of ICS in FSAR Design Basis Events B-30

. B.15 Precursors to 12-26 Event B-34 B.15.a Power Supplies to ICS/NNI B-35 B.15.b January 5, 1979 Loss of ICS Power B-36

B.15.c ICS Reliability Study, BAW-1564 B-37 B.15.d Olstrict Response to IE Bulletin 79-27 B-54 i B.15.e District Response to February 1980 Loss of NNI at B-56 Crystal River B.15.f Olstrict Response to NUREG-0667 B-57 B.15.g Significance of Partial Loss of NNI at Rancho Seco B-58

March 1984 B.15.h Appilcability of " Reference Plant" studies to B-60 Rancho Seco O

Section Page B.16 Timely Identification of Loss of ICS Power Condition B-62 B.17 Usefulness of Annunciator Procedures Manual B-64 B.18 Performance of ICS Following Restoration of Power B-65 B.19 Control Room Indicators which Fall to Mid-Scale B-66 B.20 Adherance to Radiation Protection Requirements B-68 B.21 Programmatic Efforts to Desimminate Lessons-Learned and B-69 Plant Changes B.22 Operating Crew Minimum Required Staffing B-70 B.23 Role of STA B-71 B.24 Application of Systematic Troubleshooting B-72 B.25 IIT Requests for Information B-74 B.26 Appilcability of Generic PTS Analysis to Rancho Seco B-75 APPENDIX C Cross Reference Action Plan to NRC Open Items C-1 A. Areas of Concern Relating to December 26, 1985 C-1 Overcooling Event (NRR and Region V)

8. Region V Additional Items C-7 C. Licensing Areas (NRR) Not Required for Restart, Per C-8 the NRC D. Response to NRC Letter of September 5, 1986 C-9 APPENDIX D Action P1an Comnitment Summary D-1 APPENDIX E Sample Portion of Action Plan Activitiv Tracking Report E-1 APPENDIX F Major Milestone Schedule Restart of Rancho Seco Unit 1 F-1

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J 2 Section Page l' APPENDIX G I Example System Status Report G-1

! APPENDIX H l

! Sample Test Specifications H-1 i i t I, APPENDIX I l Plant Staff Interview Program I-1

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INDEX OF FIGURES NO. NAME PAGE 1-1 Post-Trip Window 1-S 1-2 Issue Evaluation and Disposition Process 1-7 1-3 Identifying the Issues and Problems Associated I-9 with Plant Operating Performance 2-1 Restart / Implementation Organization 2-16 4A-1 B&W Owners Group Safety and Performance 4A-13 Improvement Program 40 System Review and Test Program Organization 40-4 -

40-2 System Reviews and Reporting 40-24 l

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1.0 INTRODUCTION

& PROGRAM OVERVIEW The December 26, 1985 overcooling incident at Rancho Seco has prompted a comprehensive investigation that looks far beyond the specific problems directly associated with that incident. The steadily degrading performance of the plant and its staff are symptomatic of more serious deficiencies than those associated with the December incident. ,

While plants.similar to Rancho _Seco have achieved performance levels consistent with the better performers in the industry, the operating record of the Rancho Seco Nuclear Plant, as measured by plant performance statistics (capacity factor, etc.), Systematic Assessments of Licensee Performance (SALP), and INP0 performance indicators has not been satisfactory. Subsequent to a 1984 evaluation by consultant LRS, the Sacramento Municipal Utility District (SMUD) Board of Directors decided to take action to improve the Rancho Seco performance.

To achieve this objective each of the areas affecting plant performance was investigated or studied. These areas included plant hardware, management and administrative systems, organizational structure and staffing, maintenance, training, personnel, and physical facilities. To implement and achieve observable results from the changes indicated by these studies requires time and significant financial ccmmitment. The 01 strict Board of Directors decided in 1984 that the existing investment, combined with the projected electric demands in the SMUD service area, and the benefits 1-1

t to be derived from achieving a higher level of performance, justify the additional investment required to achieve the designed results.

They also realized that reliability improvement is closely coupled with safety improvement which has always been a first priority.

Before actions associated with this decision to improve performance could take effect, the importance of the program was reinforced by a number of undesirable operating experiences in 1985. The most significant of these events occurred on December 26, 1985, when a loss of power to the plant's integrated control system led to a plant overcooling. The cooldown rate specified to limit the stresses induced in reactor systems heavy metal ccmponents was exceeded.

While subsequent analyses determined that no serious stresses were induced, the significant potential of this event is not to be understated.

Following the event, the District and the NRC independently conducted reviews to determine the nature and extent to which management, programmatic and hardware deficiencies contributed to this--and previous--incidents. The District has documented its findings in the IAG Root Cause Report 85-41 and the NRC Incident Investigation Team has documented their findings in NUREG-il95. The conclusions and recommendations identified and contained in NUREG-1195 are consistent with those reached by the Olstrict. In general, these findings are:

a. The trip and the associated rapid cooldown was caused by the failure of Rancho Seco to implement design changes in a timely manner which would have compensated for known design weaknesses.

1-2

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b. The failure of Rancho Seco to implement adequate compensatory

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measures for the design weakness, such as procedural guidance and training, contributed to the significance of the event.

c. Maintenance program deficiencies contributed to the inability to f

mitigate the severity of the cooldown transient.

d. Non-ccmpliance with existing procedures contributed to the overcooling event and caused additlonal unnecessary complications.

e. Manufacturing defects in the electrical terminations of -

pa'rticular control cabinets (ICS) initiated the event.

1.0.1 Post Event Review The post event reviews by the District and the NRC identify the specific actions to minimize the potential of this event occurring again and those necessary.to assure the event did not degrade or impact the ability of the plant to operate safely and reliably.

These pertinent documents and actions include the following:

IAG Root Cause Report NUREG 1195 Equipment Failure Investigations

- ICS

- ICS Controlled Valves

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- Makeup pump

- Radiation Monitor Effects of overcooling transient on components Operational Review (including adequacy of procedures)

Human Factors Evaluation Adequacy of Training Thermal / Hydraulic Response of the Reactor Coolant System Health Physics Emergency Preparedness Based on these findings, SMUD has modified, expanded, and accelerated the implementation of its performance improvement action plan.

1.0.2 The objectives of the Action Pla, are to:

1. Reduce Reactor trips

2. Reduce challenges to safety systems

3. Assure the plant remains in the post-trip window (The allowed ranges of reactor coolant system pressures and temperatures immediately following a reactor trip, see Figure 1-1).

4. Assure compliance with license requirements l 5. Minimize the need for operator actions outside the control room.

6. Improve the reliability and availability of the plant O

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gagree mta., a 0 triaa Suitte *t.egear9ng. Je Figure.1-1 Post-Trip Window Normal plant power operations and Reactor Post-trip response will provide temperature / pressure conditions wnich are within the above " Post-trip" window.

Conditions which are outside the window may involve challenge to safety systems and result in events such as~over cooling, under cooling and loss of subcooling.

\

1-5

The Action Plan has been structured to achieve these cbjectives through the implementation of a number of individual program 1.

elements. A schematic representation of these Action Plan Program elements and their relationships is contained in Figure 1-2. In N general, these program elements are structured to. '

a) assure that issues or deficiencies in plant design, operations and operating procedures, management and management processes, training, etc., which have the potential to contribute to an ,

event such as the December 26, 1985 event, or. negatively impact the performance of the Rancho Seco power station are identified and input to the action plan for evaluation and resolution.

b) assure that each of these issues receives a thorough evaluation and is properly dispositioned. ,

Y c) assure that actions are implemented in an efficient, effective, and timely manner consistent with their importance to safety and reliability of olant operations. .

d) assure that closure of the acticn items is complete, addresses the issue adequately and that the actions are taken in accordance with the approved plant procedures.

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1.1 INPUT PROCESS (ISSUE IDENTIFICATION)

The process is designed to assure that the rev'bw of the various f l

areas which can impact plant performance (i.e. 'menagement and ,'

management processes, plant design, operations and operating procedures, maintenance, training, and other support activities) i s adequate to identify any deficiencieb. The process consists of activities which review these impact areas from four perspectives, -

1 illustrated in Figure 1-3: (1) the top down department managers hardware and programmatic reccmmendations; (2) the management process ' i review;'(3) the bottom-up systemat'ic' assessment program elements; (4) and the system review and test program. Each of these ,

perspectives has advantages and disadvantages relative to its i

effectiveness and efficiency in identifying deficle'acles and )

' 1 .

developing improvement actions to be taken.

By inepsporating key [

featu'res of each, we have tailored an action plan which is diverse and broad in scope, directly addressing the type of deficiency which has contributed to the poor ;erformance record at Rancho Sego and the December 26, 1985 event. A description of these Issue Identification

! Action Plan elements is as follows:

5 1.1.1 Department Managers Hardware and Programmatic Recommendations 1

An assessment of the plant design, management, operations, and administrative system deficiencies was conducted based on the functional organizations

• knowledge and existing documents of previous evaluations by others. The following important sourc'es were used as input to thi s al sessrter l t.

1-8

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r-LRS Management Audit INPO Audit Reports Commitment Lists American Nuclear Insurers (ANI) Open Items NUREG 1195 IAG Report 85-41 This assessment led to the development of the management, operations, and administrative process improvement action plans (see Section 48) and the plant modification action plans (see Section 4C).

1.1.2 Hanagement Process Review ,

The management process review was conducted by the management process review group which has:

a. Reviewed previous management audits and assessments from the last five years to date.

b. Reviewed SMUD responses / commitments to these documents.

c. Conducted a direct status assessment of current management processes and functions.

d. Abstracted management assessments from the other Plant Ferformance and Management Improvement Program (PP&MIP) investigations.

1-10

The near-term and long-term action plans to improve management

. effectiveness which resulted from these actions are contained in the

^

management, operations, and administrative process improvement action plans (see Section 48). The MPRG team will facilitate the implementation of all management effectiven'ess action plans.

1.1.3 Systematic Assessment Process (OCI-12)

I The Plant Performance and Management Improvement Program (PP&MIP) is l 4

, a comprehensive broad-scope, systematic assessment program. This program was developed and is being implemented to perform a detailed review of the plant design and experience as well as the appropriate industry experience. This program will confirm the action plan-

[N elements based on the department manage,rs' assessment, and identify any necessary enhancements to these plans to address any deficiencies identified through this program. Specific input areas includs:

a. Precursor Reviews

b. Plant Staff Interviews

c. Deterministic Failure Consequence Reviews

d. 8&W Owners Group Safety and Performance Improvement Program (SPIP) a e. December 26 Event and NUREG-Il95 items

f. Selected projects 1

This process develops recommendations, which must be resolved, and l (N provides a structured program with prioritization criteria to ensure  :

\.,_-  !

that each item is appropriately resolved. The actual disposition of a  ;

I l-11

recommendation will be done utilizing the established administrative and programmatic processes in place within the nuclear organization.

As an example, a recommendation, which when dispositioned, requires a modification to the Control Room will, as a part of the configuration control and modification programs, receive Human Factors review against the CRDR criteria and commitments. In addition, it will be reviewed by Operations for E0P and Training Impacts, and may well require that a Verification and Validation (V&V) exercise be performed prior to incorporation. The Action Plan adds no steps to these programs: it does add significantly to the number of items being processed through these programs.

1.1.4 System Review and Test Program (SRTP)

The system review and test frogram is a key element in the implementation process. The systems engineers implementing this program are in a position to obtain an overview perspective of the issues and recommendations. This overview perspective combined with the implementation of the system review process may lead to the identification of additional issues and recosendations.

The details of this comprehensive System Review and Test Program are provided in Section 40, which includes definition of the role of the System Engineers.

O l-12

r~N 1.2 ISSUE EVALUATION AND DISPOSITION PRJCESS MANAGEMENT The process to address and manage the re' solution of issues is described in detail in a special Plant Procedure, QCI-12. An overview of the steps of this process as well as a description of the RRRB, PAG, the prioritization process and the oversite role of the independent review group is as follows.

The Plant Performance and Management Improvement Programs (PP&MIP) as defined within QCI-12, is intended as a one-time effort, although specific elements will be incorporated into the administrative and operational processes. The program  !

I' identifies issues and provides for their prompt review and -

] disposition. The disposition implementation is accompl1shed by utilizing the existing administrative processes, procedures, and programs. While these are themselves subjects for the PP&MIP, they are not suplanted by the PP&MIP. As examples, the Plant .

Review Committee (PRC) continues to review and approve all procedures, while the Management Safety Review Committee.(MSRC) performs its duties in addressing nuclear safety concerns and 1

issues. Items involving safety issues, whether initiated by the PP&MIP or otherwise, are addressed and resolved before these committees.

• 1.2.1 Process Overview Issues, along with recommended solutions, are identified by the L special task or input groups.

l 1-13 l

These issues and recommendations are sent to the Review, Recommendation, and Resolution Board (RRRB) for validation and acceptance.

Recommendations involving systems under review are sent to the system engineer and others are sent directly to the Performance Analysis Group (PAG) (made up of the Nuclear Department Managers). All recommendations, both valid and invalid, are dispositioned to the satisfaction of the PAG. This grouo confirms the disposition of the priorities and valid issues and

~

sends them on to the appropriate department for implementation.

The invalid issues and recommendations are sent to QA for ,

independent review and formal'close out and for system related i items, to the System Engineer for information.

Once implementation action on valid recommendations is completed a close out record is prepared and sent to QA for verification of content and validation that the implemented action is consistent with intent of the issue and recommendation.

1.2.2 The Recommendation, Review, and Resolution Soard (RRRB) is a nine member multidisiplined group of individuals witn nuclear experience and training drawn from SMUD, another utility with a B&W NS$5, NSSS Vendor, and the plant Architect Engineering firm.

As described in QCI-12, the RRRB is to determine the " validity" 1.e.,

the correctness and uniqueness, of each received recommendation.

They consider only tne technical merits, not the cost, time, or 1-14

l

't j .

resources available. Once they process a recommendation, it is passed to the System Engineer or the Performance Analysis Group (PAG) for evaluation and disposition.' '

l.2.3 The PAG membership is composed of the Managers of the Nuclear

Departments or their designees. Thisfgroup has knowledge of the 1

competing priorities, needs, commitments, and resources available to i' resolve each' recommended action. The member nuclear departments are:

Nuclear Projects, Chairman Nuclear Licensing '

Nuclear Quality Nuclear Plant Nuclear Engineering IO Nuclear Training i

In addition to receiving input from the RRRB, the PAG also receives l Inputs from the Management Process Review Group, the Systems i

. Engineers for System Review and test issues, and recommendations from the department managers themselves. All input is evaluated against 4

the objectives of the program as defined previously in this section.

The PAG then determines the disposition and assigns the appropriate

.i f priority to each item. ~

i

. Implementation of the dispositioned recommendations after approval by 3 .

the Deputy General Manager, Nuclear, fal.ls to the line organizaticn .

i i l which utilizes the processes and controls governing these

!V 1 l-15

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activities. This includes resource allocation and scheduling of plant affecting work, configuration control processes, training programs, and operations.

1.2.4 Activity Prioritization Criteria The criteria by which activities to resolve issues are prioritized and placed in the appropriate disposition period is as follows:

Restart - Actions to be completed prior to restart or completion of the Restart Test program which will:

a. assure the plant remains within the post-trip window

b. assure compliance with technical specifications

c. minimize the need for Operator Action outside the control room within the first ten minutes of an event.

Near Term - Actions to be initiated as promptly as practicable, schedule developed, resources assigned and maintained until completed which will:

a. enhance the ability to remain in the post trip window (e.g., auto action vs. operator action)

b. reduce reactor trips

c. reduce challenges to safety systems

d. produce near-term programmatic benefits.

1-16

Long Term - Actions to be programmed for the longer term which will support the achievement of the 1990 INPO plant performance objectives, i.e.

l

a. Improve reliability f

b. improve availability

c. major programmatic enhancements f

4 1

1-17

1.3 IMPLEMENTATION OF ACTION PLAN ACTIVITIES The majority of the actions to evaluate the impact of the December 1985 event and to preclude this event from recurring have been completed by the District with many receiving concurrence of closure by tne NRC. The implementation of additional actions to address the broader performance improvement issues have been identified by the District's Nuclear Department Managers.

These implementation action plans, while broad, comprehensive, and likely to address any significant deficiencies, cannot be finalized until the evaluations of the systematic assessment program (PP&MIP)

-have been completed. The two areas of implementation are; a) plant maintenance and modification actions and b) management, cperations, and administrative process improvement actions. Each of these areas is described below.

1.3.1 Modifications and Maintenance Improvement Actions A description of the specific major modifications and maintenance improvements is provided in Section 4C along with the prioritization for each. Most of these items include more than a single recommendation or disposition to accomplish their implementation. As such, Appendix E is proviaed, which lists by system, examples of the l

valid recommendations being developed and processed. The entire list i is available at Rancho Seco. It is expected that new items will continue to be identified and processed through the Systematic Assessment Program until the end of September 1986.  !

l-18

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f 1.3.2 Management, Operations, and Administrative Process Improvement Actions i The major management and administrative process areas identified for 1

improvements are described with the associated prioritized l 4 l

j. dispositions in Section 48. The QCI-12 tracking system (QTS),

j 1

typlified by . Appendix E, lists the valid programmatic improvement i

l actions associated with management, operations, and the administrative processes.

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i 1-19

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1.4 SYSTEM REVIEH AND TEST PROGRAM There are two system review programs to be conducted as part of this action plan.

The first program is modeled af ter the Davis-Besse program and is a key element in the restart of Rancho Seco. This program is structured to provide a systems review of issues, and recommendations as well as to develop a restart test program. This action will provide additional assurance that these systems have retained their FSAR/USAR functional basis, or have adequate analysis to justify differences, and that they have been adequately tested. f The second program is a longer term program modeled after the NRC's Safety System Functional Inspection. This program provides a more detailed look at the reliability and component design criteria.

1.4.1 System Review and Test Program 1

The System Review and Test Program (SRTP), which is modeled after the r

l Davis-Besse program, is a key element in the implementation process.

It consolidates all system related issues with the relevant recommendations from the Systematic Assessment Program, for coordinated resolution. This consolidation is done by an assigned System Engineer who performs supplementary tasks such as walkdowns, functional criteria development, review of other corrective action systems for system issues needing resolution prior to restart, and develops an integrated program for resolution of system issues. The 1-20

}

_(x systems engineer then defines any necessary testing and functions as the test engineer to accomplish it. The result is a determination  !

that the system is ready for plant operation. ,

l The systems engineer program being implemented at Rancho Seco is j modeled after the'INPO G000 PRACTICE . In this program the system engineer is responsible, among other things for: (1) the development '

of system functional requirements; (2) assuring coordinated and effective disposition of system deficiencies; (3) assure the adequacy J

of system testing; and (4) to develcp and maintain a set of test i

requirements which assure the material readiness and operability of each system.

A selection criteria has been established to divide the major systems into two categories, " Selected" and " Additional". In both cases the system functional criteria are developed in combination with the problem statements and presented to the-Performance Analysis Group  !

(PAG). Pending their concurrence, the recommendations of the System Engineer are implemented, and those systems in the " Selected" category tested (prior to or during) the restart.

1.4.2 System Functional Review A review of reports on Reactor Trips at B&W plants was performed as part of the Precursor Review, an element of QCI-12. This review  !

showed that over 40% of these transients were due to mismatches between reactor heat generation and secondary side heat removal.

6 1-21 f

l_

I I

Consequently, systems critical to secondary side heat removal will undergo a long term and more extensive system review than that described above. ,

This process, which is modeled after the safety system functional inspections by the NRC, consists of: (a) Design Basis reconstitution; (b) reliability assessment of the system; and (c) the evaluation of -

individual component design criteria to assure that the individual comporants support the system design basis.

A reliability assessment of these systems will also be conducted as part of this program to determine which components of these systems are critical to the prevention of reactor trips and which are critical to assure that, immediately after a reactor trip, the transient remains in the post-trip window (see Figure 1-1). The system i

I surveillance tests, where appropriate, will be evaluated to assure l

l they adequately demonstrate the operability of the system and/or 1

components to meet their design basis requirements. The five systems selected for-this comprehensive review are:

.1 Main Feedwater System

.2 Auxiliary Feedwater System l

.3 ICS/NNI

.4 Pressure Control functions of the Main Steam System

.5 Instrument Air O

1-22

This comprehensive review of these five selected systems will be initiated prior to restart and will be completed prior to coming out of the cycle 8 refueling outage.

1 l

1 1

1 i

i l-23 i

1.5 ISSUE RESOLUTION CLOSE0VT A formal process will be implemented to assure the effective and complete closecut of action plan items. This is accomplished through the Quality Assurance department which is charged with the responsibility to verify that actions were taken in accordance with the plan and existing procedures and to validate that the actions taken meet the intent of the original recommendation and resolve the original issue.

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1-24

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! 1.6 INDEPENDENT PROGRAM OVERSIGHT .

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! Independent oversite of the Action Plan is provided by an independent  !

1  ;

review group (IRG). This group consists'of. senior persons with

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} significant experience in the management and oversite of nuclear ,

power plant operations,-design, and regulations.  !

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1.7 DOCUMENT PURPOSE The purpose of this document is to communicate the District's planned actions and program status, both internally and externally. The body of the document contains a description of the program elements and a general description of the actions to be taken. The document's appendices provide descriptions of the detailed actions to be taken, and their status, scheduler.information and supplemental information such as responses to the NUREG 1195 conclusions and recommendations.

This particular structure was selected to accommodate the unique features of the District's program.

1.7.1 O_rjanization - Body The body of the report is organized as follows:

a. Section 2.0 Management of the Action Plan - describes Management processes and features of the program to assure that i

appropriate actions are being taken to effectivel'y, efficiently, l

and thoroughly identify, prioritize, control, and implement changes to resolve the type of deficiencies which contribution I

to the December 26 event and the poor performance record of Rancho Seco. This includes:

The Independent Review Group The Restart and Implementation Organization (RIO)

- The Action Plan Tracking, Reporting and Close Out process The Action Plan Adjustment process The Transition Plan

! l-26

\ . _ . _ _ _ _ -

b. Section 3.0 Improvements Prior to the December 26, 1985 Event -

describes those actions which the Olstrict had 'taken prior.to the December 26, 1985 event.

c. Section 4.0 Action Plan - provides a: description of-the actions to be taken to address the concerns raised by the December 26, 1985 event and those associated with- the poor performance record of Rancho Seco. This information is contained in the following subsections.

1) The systematic assessment process elements.

2) The management, operations, and administrative process Improvement Action Plans.

\

3) The Plant Modifications Action Plans.

4) The system review and test program.

1.7.2 Organization - Acpendices i

! The appendices of the report are organized as follows:

l l

i a. District Board of Directors Performance Improvement Policy Statement.

b. The District's Response to the NUREG 1195 Findings and Conclusions, i

1-27

c. Cross Reference of NRC Open Items to Action Plan Sections,

d. Action Plan Commitment Summary.

i

e. Sample Action Plan Activity.

f

f. Status Report Test Program Objectives.

g. Action Plan Schedules.

h. Example System Review and Test Report.

O l

l l

O l-28

. . . . = . -.

1.8 CONCLUSION

( ;i 1

lV

~As the District management team shaped this program, it became clear that there was a need for policy direction regarding.the long term future of the Rancho.Seco Nuclear Generating Station. This policy direction is considered essential to provide a foundation for long-term planning. The policy direction has been prepared and was unanimously endorsed by the Board of Directors. A copy of this Board Policy Statement is included as Appendix A.

In summary, the Rancho Seco Action Plan is an enhanced and

^

accelerated version of a program already underway at the time of.the December 26, 1985 event. This action plan is intended to p re-establish a dedication to excellence which will be the basis for regaining the confidence of regulators, county and state officials, investors,"and customers. The Action Plan constitutes a cc plete reassessment of management's role in establishing an environment in which excellent performance is expected and in which any deviation from excellence is cause for prompt and aggressive corrective action. While the number and scope of activities contained in this i plan are significant, a large number of activities important to l

l restart and performance improvement at Rancho Seco have been completed. The SMUD team is dedicated to bringing about this performance improvement in an orderly, safe, and effective manner and believes this Action Plan to be the vehicle for such change.

l l-29

2.0 MANAGEMENT OF THE ACTION PLAN I This section describes the actions the District has taken and the processes which have been established to assure that the Action Plan is thorough and comprehensive, and can be implemented in an effective, efficient, and timely manner.

To accomplish this the District has taken the following steps:

1) Established an Independent Review Group (IRG). -

b) Estab11shed a temporary Restart and Implementation Organization (RIO).

c) Established a management process to identify, track, control, implement, and close out deficiencies associated with the management and operation of the facilities.

d) Developed a transition strategy for the long term which incorporates appropriate features of the Action Plan into the line organization to achieve and maintain a high level of plant i

performance.

O 2-1

.-___._m -, , - . , , . - . , ,- - -. . -_ _ _ - _ - _ _ _ _

2.1 INDEPENDENT REVIEW GROUP (IRG) 2.1.1 Purpose An IRG, made up of Messrs. John Jackson, Richard De Young, James O'Hanlon, and Arthur Gehr, has been established to periodically provide the District's General Manager and Deputy General Manager, Nuclear with assessments as to the effectiveness of the Plant Performance and Management Improvement Program, and the readiness of the plant to restart and operate safely and reliably.

These four individuals, who comprise the IRG, encompass a broad range of pertinent experience.

John Jackson - Many years as a quality expert in design, construction, operation and training, as well as the~ analysis and oversight of quality functions. ,

Richard De Young - A distinguished career with key assignments in nuclear regulation and inspection and enforcement, as well as current experience in the analysis of nuclear management.

James O'Hanlon - Significant responsibilities in plant maintenance, and operation with current oversight responsibilities at Davis Besse.

O 2-2

l Arthur Gehr - A recognized expert in nuclear law and regulation who played a significant role in the development of the nuclear programs at Commonwealth Edison and Arizona Public Service Company.

2.1.2 Mission The mission of the IRG is to assure the General Manager and the Board of Directors that'the Action Plan for Performance Improvement at Rancho Seco is thoughtfully developed and appropriately implemented such that Rancho Seco is properly prepared for restart and ongoing -

~

operation with a reasonable likelihood that it will operate safely and reliably.

2.1.3 Tasks The following tasks will be performed:

Interview personnel

1. Key SMUD managers and supervisors

2. NRC personnel on-site, in Region V, and Washington

3. Rancho Seco staff personnel

4. Others Carl Andognini (Consultant to the Board)

J. Mattimoe (Former SMUD General Manager)

E. Wilkinson (Former President INPO) .

I

! 2-3 1

l

O INPO (all key managers)

The SMUD Board of Directors Attend key meetings NRC Region and Headquarters meetings Special Davis-Besse Review Group Attend Formal Presentations given by the following organizations to relate the progress in each of the functional areas over the course of the Action Plan implementation.

Engineering

• Quality Assurance Purchasing Maintenance Licensing (USAR)

• Training Operations

• Regulation 1

Review Technical Presentations IIT Report SALP Report history LRS INPO l

PAT Reports Participate in Plant Tours O

2-4

Review Organizational Issues t Organization Functional responsibilities Personnel Delegation of authority Perform reviews of key operating and admialstrative procedures.

i#;

2.1.4 IRG FINDINGS The IRG is to make findings in four areas.

Are the responses to the immediate implications of the December

~

26, 1985 overcooling event appropriate and complete?

Is the plan to identify programmatic deficiencies and the 4

broader implications of the plant's past history and operating performance adequate and are the planned corrective actions sufficient to reasonably expect safe and reliable operation?

Are the priority of action criteria appropriate to assura that those actions taken before restart are sufficient to reasonably assure safe operation?

Is the executive direction and the management plan appropriate toprovideproperimplementationoftheActi5nPlan?

2-5

2.2 RESTART AND IMPLEMENTATION ORGANIZATION -

The Restart and Implementation Organization is a temporary organization designed to augment and assist the normal line organization in the restart and implementation of improvement action items. This organization is shown in Figure 2-1. This organization will remain in effect until disbanded by the Deputy General Manager, Nuclear.

2.2.1 Responsibilities The Deputy General Manager. Nuclear provides management oversight to the implementation of the Action Plan and assures an effective interface with the normal line organization. He is responsible for approval of the content and the schedule for implementation. He has the authority to abolish the temporary organization when, in his judgement, the additional resources are no longer required to ensure the timely implementation of the action items.

The Restart / Implementation Manager (RIM) is responsible to the DGM, Nuclear for directing, through the Restart and Implementation Organization, the implementation of the restart and improvement programs. In this capacity, he is accountable for scope, schedule and incremental costs for the activities which must be completed to facilitate restart and to comply with the commitments of the action plan. The RIM has the authority to adjust schedules and resource applications within pre-approved action item categories without the O

2-6 l.

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3 AGM's prior approval to' assure efficient app! ation of resources and ' '

D consistency of plant, system, and component conditions. i

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The RIM provides functional and administra.tive direction to the Test i

,(, #

Prcgram Director and Program Dire ~ctor. He provides functional' direction for restart and improvement activities to the Outage f g Manager and the Nuclear Department Managers. Administratively, the '

Nuclear Department Manager's reporting relationship remains unchanged.

The RIM has authorization to issue instructions as needed to further J

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, '\

define the Restart and Implementation Organization or to control s 1 3'

interfaces to effect the action plan. ,

i s.

The Outage Manager is responsible to the RIM for execution of!the 1] 5 physical work items required for resta I.' Ti.e Outage Manager also *

-s i provides scheduling'for the entire Res~ tart and Implementation ,

, ( s Organization. He has the authority t,o rearrange sequence of work within limits of the plant technical specification requirements, plantoperatingproceduresandschgdulecommitments. Changes which < .

would require extension of schedule. commitments,,0r piutt procedure' ,

3: .

, '. *1:

changes, shall only be made with the concurrence of' the' RIM, and fer the latter case,xthe Manager, Nuclear Plant. \ ^

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s l

The Test Program Director is respouiible to thf RIM for_ develdpinq -

and implementing the ted progrard ident'ified in the action plan. _

This includes: Identification of organization and resource needs;.  ? .

development of specific' test objectives and acceptance criteria; 2-7 1

' ~

.l/

M development of special test procedures (as required); perform special test procedures; coordinating (with the Outage Organization and Nuclear Operations) the performance of surveillance tests; working with the Outage Manager in the development of a detailed test schedule; and evaluation of test results. All procedures used in the te;t program shall be reviewed, approved and used in accordance with AP. 2, " Review, Approval and Maintenance Procedures", AP. 302, s

"Special Test Procedures", or AP. 303, " Surveillance Program".

The Program Director is responsible to the RIM for oversight and -

coordination of the implementation of those programmatic commitments in the action plan. In this capacity, he works with the Nuclear

. Department 'lanagers or their designees to ensure that the proper

\ interfaces and priorities are maintained and that resources are used effectively. This position will work with Nuclear Department j Managers and the outage organization to develop a detailed schedule.

In addition to those commitments contained in the action plan, the l \ Program Director shall also ensure implementation of commitments l \ contained in the coordinated commitment list required for restart.

O u,i The Manager, Nuclear Plant is responsible to the RIM for providing the required operations, maintenance, health physics and technical support to accomplish the scheduled activities. This includes timely t S review of procedures and test results by the PRC and MSRC. He is responsible for the preparation of reports as required by the RIM.

In addition, he is responsible for assuring that programmatic e

I commitments in the action plan are not only met in a timely fashion, L .

,, 2-8

^

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l but are coordinated through the Program Director to ensure ~

l Q

consistency within the overall programmatic improvement plan.

The Manager, Nuclear Engineering is responsible to the RIM for providing the required engineering and construction support for design modifications, engineering studies and other engineering support required to accomplish the schedule activities. In addition, he is responsible for assuring that progrcmmatic commitments in the action plan are not only met in a timely fashion, but are coordinated through the Program Director to ensure consistency within the overall -

programmatic improvement plan. He is re.ponsible for the preparation of reports as requested by the RIM.

The Manager, Nuclear Projects, in the role of Performance Improvement Manager, is responsible to the RIM for the activities of the Management Process Review Group and those review activities conducted under the auspices of QCI-12. He is also responsible for preparation of reports as required by the RIM.

The Manager, Site OA is responsible to the RIM for providing appropriate surveillance and quality engineering to assure the scheduled activities are conducted in accordance with the District's legal commitments and quality programs. He shall also be responsible for developing and implementing a method for verification of completion of action plan items. Corporate QA will retain the l

independent audit function.

2-9

The Manager, Nuclear Training is responsible to the RIM for providing the required training support to accomplish the scheduled activities. This includes the identification of needs, in concert with appropriate line managers; development of training materials; presentation, evaluation and documentation of the training. In addition, he is also responsible for assuring that programmatic commitments in the action plan are coordinated through the Program Director to ensure consistency within the overall programmatic improvement plan.

The Manager, Nuclear Licensing is responsible to the RIM for providing the required licensing and engineering planning support for the scheduled activities. This includes the establishment of licensing strategies, in concert with the line managers; preparation of submittals in a timely manner, coordination of meetings with NRC and other agencies; and support for Training in preparation and presentation of emergency preparedness training. In addition, he is responsible for assuring that programmatic commitments in the action plan are coordinated through the Program Director to ensure consistency within the overall programme. tic improvement plan.

2.2.2 Qualifications

• Deputy General Manager, Nuclear - John E. Ward 1

Mr. Ward is an experienced chief executive-level, nuclear-experienced manager with 34 years proven performance in Ol t 2-10

planning, directing, and analyzing complex operational and

' O '

Q engineering projects and organizations. He has completed naval reactors training and has advanced degrees in nuclear physics.

He is a recognized expert in the area of the utility industry regulatory processes. Mr. Hard has authored numerous papers presented at meetings of the AIF, ANS, ASCE, and PHI as well as being published in Electrical World and Public Utilities Fortnightly on the subject of nuclear regulation and utility management. He is a registered professional engineer in the

~

fields of Mechanical and Nuclear Engineering in the State of -

California.

Restart / Implementation Manager - Dan C. Poole Mr. Poole has 25 years of experience in the nuclear field. This includes serving as Plant Manager of the Crystal River Nuclear Plant, with responsibility for all aspects of operation, maintenance, and technical support. In addition, he has also.

served in the capacity of Assistant Manager of Operations and Maintenance of the Callaway Plant, and the capacities of Superintendent Operations, Superintendent Technical Support, and Training Supervisor at the J. M. Farley Nuclear Plant.

Test Director - Jim Field Mr. Field is the Nuclear Technical Support Superintendent at Rancho Seco. He has over 11 years experience in the Rancho Seco 2-11

. _. .. - . - . - _ _ _ ~ - . _ . .. . .__ __ _ _. . . _ - _ _ _ .

Technical Support Group. In addition, he has recently headed up the group which performed the Deterministic Failure Consequence Analysis described in Section 4.A.

• Program Director - T. C. Lutkehaus Mr. Lutkehaus has over 15 years of nuclear power plant experience, which includes 12 years experiento at the Crystal River Nuclear Plant. The Crystal River Nuclear Plant is a

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Babcock & Wilcox designed PHR very similar to the Rancho Seco Plant. In this time he held management positions in maintenance, technical support and was the Assistant Plant Manager and Plant Manager.

Outage Manager - J. R. Shetler Mr. Shetler is the Manager, Nuclear Scheduling at Rancho Seco.

He has 15 years of Babcock and Wilcox PHR experience. This has included system design and procurement, startup/ test support, and outage maintenance and coordination' activities. His outage coordination activities have spanned some ten outages over the last ten years including the role of outage manager.

Manager, Quality - S. R. Knight Mr. Knight has over twenty-five years experience in construction and operation of power plants with initial experience in design, 9

2-12

project engineering and test engineering of U. S. Navy nuclear g power plants. His recent experience has been in design, licensing, construction, test and operation of commercial generating plants including design reviews, safeguards, and 1

waste management, QA/QC, programs for inventory, material, and maintenance con, trol.

Manager, Nuclear Plant - G. A. Coward Mr. Coward has 19 years with the District and has over 16 years -

i experience with the Rancho Seco Plant. He has held positions of Senior Mechanical Engineer, Supervisor fluclear Maintenance Division and Nuclear Plant Superintendent.'

i Manager, Nuclear Projects - J. V. Vinquist Mr. Vinquist has 12 years of Nut..ar Power Plant experience in increasingly responsible roles. He performed assignments as I&C Start-up Engineer, Assistant Electrical Maintenance Supervisor, Electrical Maintenance Supervisor, Maintenance Engineer, and Assistant Plant Manager - Technical support. He also obtained l

and maintained SR0 license and periodically performed in

! capacity of Shift Supervisor. Prior to joining SMUD, he was a consultant assigned as Technical Staff Assistant to AGM, Nuclear at SMJD.

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_ _ _ . _ . _ _ _ _ _ _ _ _ __ _ _ . _ . _ - _ _ _ _ _ ._. __. _ . _ . _ . _ _ __ _______1

Manager, Nuclear Training - P. Turner O

Mr. Turner has worked within the training and nuclear fields for over 20 years. In addition to training assignments at the Tennessee Valley Authority, and the Institute of Nuclear Power Operations, Mr. Turner was Manager of the Nuclear Training Department at Kansas Gas and Electric company.

• Manager, Nuclear Engineering - D. Gillispie Mr. Gillispie is an INPO employee on loan to the District. He has more than 20 years experience in the Nuclear Power field and approximately 16 years of commercial nuclear experience. Prior to this current assignment, he was Manager of the Technical support Department, and directed the avaluation of technical support activities at nuclear power plants and at the corporate level. He also served approximately 5 years as Manager of INP0's Events Analysis Department.

Manager, Nuclear Licensing - R. Ashley Mr. Ashley has 30 years of experience in atomic energy and nuclear power plant design and operation, with responsibilities in engineering licensing and project management. He has directed special licensing activities for two major nuclear plants and assisted on technical, scheduling, and licensing matters for several others. He has participated in developing O ,

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t j and implementing the restart programs for two nuclear plants _

.that received NRC shutdown orders. He is a registered professional Nuclear Engineer in the State of California.

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. . , ~ , . . . . . - _ . . - , - - . , _ _ - . , , . - . . _ - . . _ . . . _ . . _ _ . . , - - . . . . . . . . _ - - - . . . . - , . , _ _ , . . - _ . , - - - -

Figure 2-1 Restart / Implementation Organization AGM NUCLEAR l J. E. Ward S. Knight Restart / Manager Implementation Site 0A

. Manager D. C. Poole l_ _ _ _ _ _ _ _ _ _ _ _,

I I I I I I J. Field T. C. Lutkehaus System J. R. Shetler J. Vinquist D. Gillispie G. Coward R. Ashley P. Turner Program Review Outage Manager Manager Manager Manager Manager Director and Test Manager Nuclear Nuclear Nuclear Nuclear Nuclear Program Projects Engineering Plant Licensing Training Director Performance _ _ _ _ _ Indicates functional accountability Improvement for restart and improvement program Action Plan responsibilities.

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4 2.3 ACTION PLAN ACTIVITY MANAGEMENT PROCESS m

The unique nature of the District's Action Plan combined with the time frame of implementation has necessitated the development and implementation of a process to accommodate adjustments to the plan while assuring the general objectives and direction of the program are maintained. To accomplish this, the 01 strict has established the necessary organizational structure, assigned appropriate authority and responsibility and developed the necessary guidelines to assure the program accomplishes its intended near and long term objectives. -

The organizational elements instituted to meet these requirements include:

O 2.3.1 The establishment of the Recommendation Review and Resolution Board (RRRB)

The Board is a nine member multidisciplined group of individuals with nuclear experience and training drawn from SMUD, another utility with a B&W NSSS, the NSSS Vendor, and the Plant Architect Engineering firm. The functions of this Board are: (a) to screen recommendations for clarity and duplication, (b) evaluate issues and recommendations, and (c) recommend the appropriate disposition and priority for the recommendation based on its technical merits.

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To guide the Recommendation, Review, and Resolution Board in making these technical assessments and prioritizing the implementation actions, the following guidelines have been established:

Hould implementation of the proposed recommendation:

a) Reduce reactor trips b) Reduce challenges to safety systems c) Remain in nominal post-trip window d) Assure compliance with license requirements e) Minimize the need for operator action outside the control room within the first 10 minutes of an event f) Indicative of programmatic deficiency g) Significantly improve reliability /ava;! ability If the recommendation meets any of these guidelines and is determined to be valid, the RRRB provides initial prioritization in accordance with engineering judgenent as follows:

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.1 CRITERIA FOR PRIORITIZING RESTART SCHEDULE O

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.1 ACTIONS TO BE COMPLETED PRIOR TO RESTART l 1

l OR COMPLETION OF THE RESTART TEST PROGRAM Assure plant remains in post-trip window Assure compliance with technical specifications Minimize the need for operator action outside the control room within the first 10 minutes of an event

.2 ACTIONS TO BE INITIATED AS PROMPTLY AS ,

PRACTICABLE, SCHEDL'LE DEVEL0?ED, RESOURCES ASSIGNED AND MAINTAINED UNTIL COMPLETED (IT IS THE INTENT TO INITIATE THESE ACTIONS AS SOON AS PRIORITY ONE ACTION COMPLETIONS MAKE RESOURCES AVAILABLE)

Enhance ability to remain in post-trip window automatically Reduce reactor trips Reduce challenges to safety systems Produce near-term programmatic benefits f

.3 ACTIONS TO BE PROGRAMMED FOR THE LONGER TERM Improve reliability Improve availability Major programmatic enhancements 2-19 i

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In addition to screening and validating recommendations, the RRR8 inputs each recommendation to the master data base which records and tracks each recommendation through its life cycle.

Once the RRR8 has finished the validation process, the recommendation (valid or invalid) is forwarded to different organtrations based on its characteristics for action. The alternate paths are: (a)

Progr?vnatic recommendations are sent directly to the Performance Analysis Group (PAG) for disposition, and (b) system related recommendations are sent to the Systems Engineer, who in turn develops an integrated implementation and test plan for the system.

This system plan is then sent to the PAG for disposition. The PAG with the approval of the DGM, Nuclear, determines the course of action for these recommendations and sends them to the appropriate departments fcr implementation through the Action Plan or the development of justification as to why the recommendation is invalid.

l 2.3.2 The establishment of the Performance Analysis Group i

This group is made up of the Nuclear Department managers or their designees that report directly to the Deputy General Manager, l

Nuclear. The function of this group is to review and determine the appropriate disposition, from a management perspective, of the recommended actions of the Recommendation, Review, and Resolution Board.

O i 2-20

During this process, the recommendation is reviewed in light of _

existing program activities to determine whether the disposition of the recommendation (actions and priority) can be accommodated through existing programs or whether adjustments are necessary to assure the overall program objectives are met. This group also determines the department which will have the responsibilit'y for implementing the necessary actions to satisfy the finding and recommendation.

The Performance Analysis Group is also charged with monitoring the implementation of the Action Plan. The need for and approval of -

changes to the plan to assure the near and long term objectives are met will be developed by the Nuclear Projects Manager and approved by the Performance Analysis Group and the Deputy General Manager, Nuclear. This includes changes to the priority of individual action items.

l j 2.3.3 Implementation and Close-out 1

Approved actions are implemented by the appropriate line organization in accordance with the approved Quality Assurance manual and the existing approved department policies and procedures. The final step in the implementation process is the development of a closure

. package. This package contains sufficient information to describe the specific actions taken to implement the recommendation and where appropriate contains the actual implementation documentation. The closeout document is approved by the implementing department manager and forwarded to the Quality Department for final verification and closecut.

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Quality Assurance performs a verification of the implementation of the recommendation. This verification audit can apply sampling techniques where appropriate but will be of sufficient depth to assure the objectives of the recommendation have been met. When the determination is made that implementation is complete, the Quality Department documents this conclusion in the closure package and forwards it to the Performance Improvement Manager for logging and filing.

2.3.4 Action Plan Activity Tracking and Roporting The site Quality Department is responsible for the tracking of the Action Plan items. The tracking system employed by the site Quality Department has the necessary features to correlate these items or deficiencies it is being implemented to address. This is particularly important since each issue or deficiency may require several actions to accomplish closure, and a particular action may be required as part of the resolution of more than one issue. The status of each activity will be maintained, monitored, and reported on a weekly basis during program implementation prior to Restart and on a monthly basis following Restart Power Ascension testing.

The Outage Manager is responsible for maintaining the schedules for the implementation activities which impact plant hardware and management or programmatic issues. These schedules will be updated frequently to satisfy the internal needs of plant outage management but will be updated at least weekly to meet the external interface requirements.

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A 2.4 REVIEW MEETINGS AND REPORTS v

The DGM, Nuclear will provide status reports to internal and external groups on a regular basis. Prior to Power operation, a special Restart Report will be prepared documenting the critical findings, and the implementations of the associated actions taken to address these findings during the implementation of the Action Plan. The periodic reports and meetings to communicate internally and externally are described below.

2.4.1 Internal On a monthly basis, the DGM, Nuclear will meet with the Rancho Seco Implementation Committee of the Board of Directors to review in detall, progress of the action plan. At the subsequent-full Board meeting, he will present an overview of his report.

On a monthly basis, the DGM, Nuclear will meet with the General Manager, Assistant General Managers and relevant staff to review in detail progress of the restart and performance improvement plan.

Informally, the DGM, Nuclear will provide the General Manager with daily updates.

On a monthly basis, the DGM, Nuclear will provide a summary status report for 01 strict employees.

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2.4.2 External O

On a monthly basis, for the period beginning August 1986 and extending to six months after the beginning of restart heatup, the DGM, Nuclear will meet with NRC staff and Region V to provide a formal status report on the progress of the Action Plan.

On a monthly basis, for the period August 1986 and extending to six months after the beginning of restart heatup, the DGM, Nuclear will ineet with the Independent Review Group to review in detail, progress of the Action Plan. These meetings will continue on a quarterly basis for a one year period.

On a monthly basis for the period August 1986 and extending to six months after the beginning of restart heatup, the DGM, Nuclear will send a monthly written repor~t of Action Plan progress to the following organizations: '

l American Nuclear Insurers Institute for Nuclear Power Operations B&W Owners Group Executive Committee Supervisors, Sacramento County I l

Supervisors, Amador County Supervisors, San Joaquin County Chairman, California Energy Committee O

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2.5 TRANSITION ACTIONS

\

a The District's 1990 Plant Performance objectives are to achieve

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., performance levels which will place Rancho Seco among the top performing nuclear plants in the United States.

The District will modify the Nuclear Organization as required to achieve and maintain performance at this level by incorporating beneficial features of the systematic assessment program. Required changes to the organization will be in place prior to the disbanding -

't of the temporary Restart and Implementation Organization (RIO). The termination of RIO is planned after the restart power ascension testing is complete and sufficient actions have been taken on the long term performance improvement actions to assure they can be managed and implemented by the normal line organization. A f transition plan will be developed to assure that this evolution in the organizational structure is orderly and all required actions are 1

4 anticipated and managed.

The long term objectives will be achieved through the implementation 4

of the performance improvement items identified in this document and

! the implementation of supplemental programs. In particular, the Olstrict intends to implement the programs necessary to:

a) identify the components critical to power production.

, b) to monitor, trend, and evaluate unavailability contributors.

f c) to undertake a plant specific risk assessment.

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d) to implement those features of the precursor review program necessary to achieve a high quality lessons learned program.

The transition of the special NRC and IRG oversite activities for this Action Plan to those consistent with normal industry practices should occur when the restart actions are complete and those near term actions to be initiated as promptly as practical and continued to completion have been initiated and a plan developed in sufficient depth to assure their completion on the projected schedule.

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l 3.0 PERFORMANCE IMPROVEMENTS UNDERWAY PRIOR TO THE 12/26 EVENT O

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In late 1984, the' Board of Directors of the Sacramento Municipal Utility District recognized that the Olstrict had a significant number of challenges facing them. Foremost among those challenges was the need for improvement in the operation of the Rancho Seco' Nuclear Generating Station. The Board recognized that the Rancho Seco problems had developed over a number of years through the joint attitudes and performance of the Board,- the Staff, and plant personnel.

To a large degree, these failures were made evident by the overburden the District's staff felt in responding to the large number of changes required to implement the THI-2 lessons learned in areas including plant modifications, personnel performance, management systems, analytical capability, training, and organizational structure. The Board also recognized that the dynamics of the public power arena contributed significantly to the District's arrival at its current situation. The Board was taking corrective actions on these issues when a transient occurred on December 26, 1985 at Rancho Seco which emphasized the need for further action.

During 1985, in recognition of the above situation, the Board embarked on an overall program to upgrade the Olstrict's organization and operations thereby improving the effectiveness and reliability of plant performance. The thrust of the improvement program was to deal 3-1

with a large spectrum of management and organizational issues including: ,

Establishment of a commitment to excellence in performance at Rancho SeCo, including strengthening the technical competency of the people and the organization. ,

The effectiveness of interface activities within upper management and between departments.

Organ 12ational streamlining, staff enhancement and other organizational improvements. -

Effective attention to detall.

Upgrading of the . training organization, training facilities', and training programs Establishment of a clearly defined maintenance program.

Establishment of an effective systematic troubleshooting program Development of a comprehensive root cause analysis program The intent of dealing with these issues was to elevate these areas of the operation to the level of excellence consistent with the charter of the Board and the expectations of the regulators, industry, and public.

Detailed actions to implement the Board's desire for overall improvement were developed and each was assigned to a specific individual for completion on a specified schedule. Consistent with these plans a number of activities were initiated, beginning in the Spring of 1985, and the program continued through the summer.

3-2

)

r 3.1 PROJECTS UNDERWAY PRIOR TO DECEMBER 26, 1985 k

v 3.1.1 Staffing and Organization Significant to the conditions determined in the 1984 study was the recognition of the sizable growth in plant staff .without a corresponding restructuring or expansion of the management staff. A new organizational structure was approved with_six Nuclear l

Departments reporting to the DGM, Nuclear:

1

1. Nuclear Operations

2. Nuclear Engineering ,

3. Quality

4. Nuclear Training

5. Nuclear Licensing

6. Nuclear Projects The last three departments listed had previously been elements within the' Nuclear Operations and Engineering Departments.

.Tne Board approved this structure, and to provide the staff to fill-new management and technical positions, nationwide recruiting efforts were mounted. All new key positions were staffed by early 1986.

Numerous other structural changes occurred within the various departments. The purpose and effect has been to reduce the diverse managerial requirements upon individual supervisors and 3-3

r superintendents by establishing new divisions and alignments which bring similar functions under the directi.on of a single manager.

Middle management is now better able to cope with the demands of the groups for whom they are responsible. We are seeing them spend more time on details while interacting with the personnel and projects coming under their purview. As a number of these people have been at Rancho Seco for less than a year, there has been a considerable injection of new concepts and methods within the organization. This, coupled with the traditionally responsible and professional attitude of the Rancho Seco staff, has resulted in an overall attitude which is receptive to the programmatic approach and committed to attention-to-detail and accountability.

3.1.2 Training Program

1. Management Restructure Previously, the Nuclear Training Department was an organization under the Operations Department Manager. It was recognized that this reporting level was inappropriate for the expanded importance of the training function and that it ought to be elevated to departmental status to ensure top management involvement.

As of June 1985, the training organization became a department answering directly to the Assistant General Manager, Nuclear.

The position of Training Manager was established and filled from 3-4

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f, outside the District organization, bringing a new perspective and experience level to the department.

i 2

i Hith the recent transfer of onsite Emergency Planning training to the training department, all of the plant training programs are now under the training department with only one exception.

This exception is the Fire Brigade training program which will I

also be transferred to the Training Department as soon as

quallfled personnel can be hired.

These management and structural changes, together with the training procedure and policy revisions identified in Section 4.C.3, will result in management recognition of the training g function as an integral part of plant operations and ensure b ~

effective coordination of the training function with all other nuclear organization functions.

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2. INP0 Accreditation Effort i

i The District has been committed to INP0 Accreditation for its training programs for the past several years in an effort to I

improve the overall training program. The District is using a

, phased approach for this effort. The focus on obtaining accreditation for the various department functions is on a sequential basis, starting with operations.

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! 3-5

_ _ _ _ - - _ . _ _ _._.__,_~__._:___.,_______-_.._ -

The first phase, consisting of Senior Reactor Operator (SRO),

Reac:or Operator (RO), Shift Technical Advisor (STA), and Non Licensed Operator (NLO), received accreditation in April 1986.

The remaining six training programs, which involve maintenance training, chemistry and radiological protection technician training, and technical support staff and managers training programs have been submitted for accreditation in June 1986.

The accreditation process will ensure that the shortcomings identified in maintenance training are corrected.

3.1.3 Maintenance Program Prior to the December 26, 1985 event, the District had initiated a number of actions designed to improve its maintenance program. In addition, at the time of the event, several specific maintenance program enhancement actions were underway. These included:

Search for an experienced individual from the industry to fill the Maintenance Manager position.

Increased staffing levels authorized for maintenance in the 1986 Budget. This included allowances for Preventive Maintenance Suoervisors and dedicated Preventive Maintenance Crews in the Mechanical, Electrical, and I&C Groups.

Dedicated Planning Personnel authorized for the Electrical, Mechanical, and I&C Maintenance Groups.

3-6

, -~s . Formation of a centralized scheduling group to provide overall

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N_s/ crioritization and integration of maintenance work with other plant activities.

A full-time consultant was reviewing and refining the Mechanical Preventive Maintenance Program.

4 i 3.1.4 Quality Assurance

] Rancho Seco has undergone a number of changes in the quality program and organization in the 1985 and early 1986 time period. These t

changes have been implemented in response to the-independent analysis of an outside organization (LRS). The two areas which have been impacted the most are Quality Control and Quality Engineering.

Quality Control Inspectors from Nuclear Engineering Const'ruction and l- Nuclear Operations were combined in mid 1985 and transferred to the Quality Orgar.ization.

4 The Quality Control section consists of a Quality Control Supervisor with two QC Coordinators reporting to him. A total of twenty two qualified inspectors that cover the area of I&C, electrical,

). mechanical, civil, NOE and concrete structures are now on board as i

! Olstrict employees. For outages, the inspectors are augmented as needed by contract personnel. The QC inspectors are involved in j source inspection, receipt inspection, construction and maintenance activities.

)

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Quality Engineering was established as a new section in early 1985 and staffing to the current level of 10 professional engineers was completed in January,1986. They were recruited from various A/E's and nuclear generating utilities. Quality Engineering is involved in the day-to-day maintenance operations, both corrective and preventive maintenance. The design engineering review group monitors -the design control procedures and assures that quality requirements are added to both design specifications and purchase requisitions. The addition of Quality Engineering has increased the capability of the organization and, therefore, it's ability to perform its intended function.

A major revision of the QA Manual was undertaken in 1985. An extensive effort was made to upgrade the manual and separate it into two sections consisting of (1) 18-point policy section and (2) tne Quality Assurance Procedures. The manual was reviewed and approved by the MSRC and NRC via the 10CFR50.54(a) requirement. The policy section will become part of the USAR.

3.l.5 Systematic Troubleshooting In March 1985, Rancho Seco was shutdown for a scheduled 90-day refueling for cycle 7. The plant was returning to service 94 days later when a RCS vent line cracked, causing a shutdown due to excessive loss of primary coolant. The subsequent investigation and repairs led to a greatly expanded IE Bulletin 79-14 pipe support program, which did not allow restart until late September. Between 3-8 L

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-s e thenandtheend-of-the-yearthreereactortripsfoccurred. Upon each n;

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occurrence, a special systematic troubleshooting program (based upon ,

the Davis-Besse NUREG 1154 Appendix B methods and criteria) was '

1:b implemented. This was in addition to programs that were already in A place such as Root Cause Analysis, the sup~ port work done by the B&W

. ,e Owners Group Transient Analysis Program (TAP) team, anc the Nuclear

• 4 .6 % '

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• Operations Trip Report investigation. ,

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, In each case, the implementation of the Systematic Troubleshooting method led to a greater understanding of the event and determination h \ ' ,,

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< of corrective actions to preclude re-occurrence. This method wss i again instituted on December 26, 1985 asafirstresponsetot$e ,

'n .

December 26, 1985 overcooling event.

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• l-1 3.1.6 Root Cause Program '

y Early in 1985, the Incident Analysis Group was established to provide independent analysis of events and activities to determine the /

programmatic root cause of each. They reported their finding to the *

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Management Review Team which is made up of the Nuclea? Department

• Manage's and OGM, Nuclear. This program was quite successful during s

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I its first year in providing the independent, multi-disciplinary O?f me x e analysis necessary to produce useful root causes and programmatic ,

determinations. '

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3.1.7 Activity Assessment It can be seen that basic problerrs had been recognized and initial actions taken before the December 26, 1985 incident. None of these programs had achieved the momentum to affect plant performance, lf' though all have appropriately become key elements of this Action Plan. Many actions associated with the District's initial program and the December 26, 1985 event analysis were completed prior to the ,

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development and submittal of this Action Plan.

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_ , . . ._- -. . - = . - . . . . . - -- .-- . .- -.

J f' 4.0 RESTART AND PERFORMANCE IMPROVEMENT ACTION PLAN

. This section provides a description of the actions being taken to i ' address the concerns raised by the December 26, 1985 event and the performance record of Rancho Seco.

Section 4A provides a description of the systematic assessment processes.

i Section 48 provides a description of the management, operations, and administrative process improvements for each of the Rancho Seco functional areas as developed by the respective department managers.

Section 4C provides a description of the plant modifications and mal'ntenance improvements developed by the' functional organizations. ,

I Section 40 provides a description of the systems review and test program.

1 I

Note that a significant number of the actions committed to be completed prior to Restart have been acccmplished and are awaiting systematic closure by the QA prccess.

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4-1

4 Og 4A SYSTEMATIC ASSESSMENT PROGRAM h

This section provides an overview of the special tasks established to acccmplish the systematic review of the physical plant, it's i operating procedures, training, maintenance, and related areas impacting performance, including a look at industry and plant history. The details of these special tasks, and the procedural guidance for their implementation, are contained in QCI-12 " Plant Performance and Management Improvement Program (PP&MIP)". The objective of these retrospective tasks is to assure that plant affecting deficiencies are identified and brought to management attention such~that necessary corrective actions can be implemented.

, These special tasks include:

o i

• Precursor Review Deterministic'. Failure Consequence Analysis B&W Owners Group - SPIP Program .

Plant Personnel Interviews i

! 4A.a Commitment I The icng-term benefits of having effective programs'to accomplish syster.atic assessments is clearly in the best intent of Rancho Seco.

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Prior to restart, these programs will be established as a part of the administrative process at Rancho Seco, to ensure that the benefits they provide will be applied to future issues, modifications, changes,- and conditions.  ;

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! 4A-1 i - _ _ , _ _ _ _ _ _ _ . _ . . _. _- _ ._ _ _.-. _ _ _ . ~ . _ , . _ _ _ . - _

4A.1 PRECURSOR REVIEW PROGRAM 4A.1.1 Objective The objectives of the Precursor Review Program are to systematically review historical documents and recommendations for events or conditions and to determine their significance to Rancho Seco. Frcm the events and conditions that are judged to be applicable and significant to Rancho Seco, a specific recommendation will be made to improve the affected plant area (design, operations, maintenance, etc.) to either preclude the occurrence or minimize the effect of the event or condition at Rancho Seco. The identified issues and improvemer.t recommendations will be input to the Recommendations Review, and Resolution Board (RRRB) for disposition.

4A.I.2 Scope of Work j

The scope of work to be performed in the Precursor Review Program is divided into two parts.

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1. Review of Past Trios and Transients on B&W-Designed Plants The review of transients on B&W-designed plants (transients are defined in the S&WOG SPIP Program) consists of the following:

a. All Transient Assessment Program (TAP) Category C transients will be evaluated and investigated for their applicability and impact on Rancho Seco.

4A-2

O b. All Category B TAP events will be reviewed to determine if U

any of the recommendations made are applicaole to Rancho Seco and to determine whether, because of plant differences, the transient could have been more severe at Rancho Seco. ,

c. 'All recommendations for Category A TAP transients will be reviewed to determine their applicability to Rancho Seco;

, d. All Rancho Seco transients, starting from the Rancho Seco  ;

" light bulb 9 event in~1978, will be reviewed.

.e. Review NUREG-0667 " Transient Response of B&W Designed l

Reactors" for open recommendations.

These_ events will be reviewed with recommendations or concerns identified and passed on to the RRRB. The review program described above.will be completed before plant restart.

2. Other Document Reviews

In addition to the review of the TAP data, the following documents wil'1 be reviewed by a multi-discipline experienced i team

i.

a. Rancho Seco Licensee Event Reports and Occurrence Description Reports Ov dA-3 1

- .-- .. , . . , _ . , _ _ _ _ _ =

. . _ , _ . _ ,_ _ _ . . - - , . . , , . - . _ . . _ _ , _ . . . , . . . . ~ . - , . _ - - _ . . . - . _ _ . _ _ ~-

b. Significant Operating Experience Reports (SOER) issued by the Institute of fluclear Power Operations

c. Bulletins issued by the NRC Office of Inspection and Enforcement

d. flotices/ Circulars issued by the NRC Office of Inspection and Enforcement ,

e. Babcock and Wilcox Reports (Preliminary Safety Concerns, Site Instructions, and other relevant BAN reports)

This program provides for a reverse chronological review starting frcm 1985. Prior to start-up, documents dating back to March 1978 will be reviewed. The Precursor Review team will access and make a .

li recommendation as to the need and scope of any further reviews.

I 4A.l.3 Criteria and Methoaology for Precursor Evaluation Each document will be reviewed to determine whether issues are applicable to Rancho Seco. For each document a Precursor Review Checklist will be completed. For those issues which produce a reccmmendation, tne Precursor Review Reccmmendation Form will be completed and forwarded to the RRRB.

4A.l.4 Schedule Evaluations of all recommendations will be ccmpleted and prioritized before plant restart.

4A 4

' [^ 4A.2 DETERMINISTIC FAILURE CONSEQUENCE ANALYSIS 4A.2.1 Purpose The objective of the Deterministic Failure Consequence Analysis is to determine the consequences of failures of systems-on power operation or post-trip response capability, and to evaluate related procedural guidance provided to the operators. The intent of the analysis is to identify areas where failures of plant systems or procedural inadequacies could potentially result in unnecessary reactor trips, unsatisfactory post-trip response, undue challenges to the operators, or challenges to the safety systems. Recommendations will 'be developed which improve plant reliability, post-trip response, anc operator performance when or where inadequacies or enhancements are identified. '

4A.2.2 Program Scope The effect of loss of electrical power, instrument air, and control power will be evaluated for impact on plant operations. These systems were chosen because failures in these systems closely approximate the consequences of most postulated plant system failures. The analysis will identify af'fected systems-which challenge or adversely effect the capability to mitigate transient conditions. No attempt has been made to analyze every combination of f failures which could occur, yet by starting with the assumed loss of an active ccmponent, and ccmcounding the effect by assuming concurrent ,

u-s

failure of components with common motive power or controls, a very wide range of likely partial loss of systems has been considered.

Whi!e the methcdology was applied, considering for the most part the individual loss of power, Instrument Air, or ICS/NNI, the results can be evaluated in light of the assumption that all three conditions exist simultaneously. Such a situation is the basis for the decision to install a class 1 Emergency Feedwater Initiation and Control (EFIC) system, plus class 1 air supplies to certain valves utilized in this event. This event would challenge the plant safety systems, but with these new features, it should not cause adverse transient conditions or, control problems.

4A.2.3 Methodology Each system will be reviewed as described below:

.1 Loss of Electrical Power t

Teams will analyze each 480V bus, its source and loads. Each l

team.will review electrical elementaries beginning at the end loads. Each breaker off the Motor Control Center (MCC) or panel will be " failed" individually. (Note: The " failure" is an assumption, no physical positioning at circuit breakers, etc. ,

will be attempted.) The consequence of failure of each load can 9

JA-6

)

N then be determined. The process is repeated for all MCCs and panels off a common 480V bus. Once the consequence of loss of the individual loads is evaluated, the loss of the" source (s)

will be analyzed.

l

~A similar analysis will be performed on the 120/125V buses, except that the failure will be assumed to include the inverter, battery, and alternate supplies.

Upon completion of the analysis ,of the individual 480V buses, the loss of the 4160V bus and loss of the individual transformers to off-site power will be analyzed. Finally, the loss of off-site power will be analyzed.

Electrical elementary drawings will be " yellow lined,"

identifying the breakers " opened" and affected components to ,

~

ensure each load is addressed.

.2 Loss of Instrument Air i

An evaluation of the effects of loss of instrument air will be performed. Individual components (loads) on the Instrument Air System will be " failed" and the effect upon the plant

~

determined. The entire system will then be " failed" to 1 determine the effect on the plant. The P& ids will be " yellow lined" to ensure that each component and/or header is addressed.

V 4A-7

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.3 Loss of ICS/NNI The loss of ICS and NNI power supplies will be evaluated to determine failure states and resultant actions or suggested modifications necessary to establish 2. known safe state with little or no operator action. Appropriate drawings will be

" yellow lined" to ensure each component or parameter is addressed.

.4 IE Bulletin 79-27 t

Due to the number and scope of. changes to ICS/NNI (the subject of this Bulletin) andthenatureofthequestionsitasks(Ehich are similar to the OFC approach), this Bulletin was evaluated as a part of the DFC scope, and utilized the OFC approach.

i l 5 l 4A.2.4 Process Recommendations developed during the analyses will be submitted to the RRRS. Specific notes of those systems affected by the " failure" which lead to the recommendation will be made. These notes will be forwarded to the relevant system engineer.

4A.2.5 Schedule This project, including the evaluation of reccmmendations generated from tne review, will be completed and crioritized before plant restart.

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(~q 4A.3 B&W CWNERS GROUP PROGRAM - SAFETY AND PERFORMANCE IMPROVEMENT PROGRAM

• .N' ' (SPIP) 4A.3.1 Background In January 1986, the B&WOG initiated a new concept in their management of B&WOG activities. Instead of having projects undertaken by individual committees, task forces, or-working groups, which were directed only at the group's goals, it was decided that projects would be focused around a few major programs. These programs have since been idertified to be:

Trip Reduction and Transient' Response Improvement

,/]

• Availability . Improvement

(

/

• Regulatory Commitments

[

Economic Benefits ,

Each of these Programs has an organization to carry cut its specific goals.

Safety and Performance Imorovement Program (SPIP) is the major i

element of the Trip Reduction and Transient Response Improvement Program.

l During the development of the Rancho Seco PP&MIP, reference was made to the S&WOG "$ TOP-TRIP" program. The "SPIP" and "STCP-TRIP"

(' '}

programs are the same, with SPIP being tne crograms finally adopted l

name.

1 1

JA-9 l

4A.3.2 Cbjective The SPIP was established to improve B& HOG plant safety and performance. The program's stated objective is to:

Reduce the number of trips and complex transients on B&WOG plants and ensure acceptable plant response during those trips and transients which do occur.

The specific goals of the program are:

1. By the end of 1990, the average per plant trip frequency will be less than two per year.

2. By the end of 1990, the number of complex transients, as classified by measurable parameters (Category "C") will be reduced to 0.1 per plant per year based on a moving three year average. (Note: SPIP Category "C" is similar to PP&MIP Priority "1" criteria).

The District is aggressively participating in the SPIP, as it does in the other B&WCG programs. Such involvement will ensure a broad perspective is taken witn respect to plant irrprovements, as well as allow other B&W owners to benefit from the Rancho Seco Plant Improvement Program.

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[~'N 4A.3.3 Program Scope l

The B&W Owners Group 'SPIP Program is similar in many ways'to the -

District's Plant Performance Improvement Program. However, it is designed as an extension and expansion from previous B&W Owners Group (B&WOG) activities aimed at reducing the number, and severity of reactor trips which occur. Recently, the program scope has been expanded to collect information and develop the response to the NRC Assessment of the " Sensitivity" of the B&W Designed NSS. Thus, a number of programs are already underway and recommendations being prepared for evaluation and implementation by the member utilities.

In addition, the B&WOG program is designed as an ongoing activity and thus will still be providing recommendations for plant improvement after the Rancho Seco Action Plan. ..

The SPIP process includes the following major elements:

1

• Define concerns Prioritize concerns Integrate, schedule and perform projects Issue project reports with recommendations B&WOG Steering Committee approve and issue recommendations to owners

• - Track implementation status with Reccmmendation Tracking Syste'm i

I The most important aspect of the SPIP is the implementation of the recommendations at the B&WOG plants.

4A-Il

4A.3.4 Methodoloav/ Procedure  ;

To ensure effective SPIP implementation at Rancho Seco, the Olstrict has assigned a SPIP Coordinator (SPC) who provides the interface between the District and the B&WOG SPIP. Additionally, the District's coordinator is a member of the SPIP Management Team. This latter role is a unique opportunity to provide programs leadership and gain detailed insight that will -ald implementation at Rancho Seco.

The coordinator's role is shown schematically on Figure 4A-1 and is summarized as follows:

Ensure SPIP recommendations are promotly input to the RRRB Follow up to ensure SPIP intent correctly interpreted Regularly report implementation status to B&WOG and to PAG Provide results of other PP&MIP actions to the B&WOG for consideration of generic applicability and integration with SPIP 4A.3.5 Schedule-Time of Performance In recognition of the fact that the SPIP will extend beyond Rancho Seco restart, any recommendations issued by the SPIP after startup I will be addressed through the long-term program for similar treatnent l

l described in Section 4A.a. The District shall continue to fully support the objective, goals, and activities of the SPIP. ,

l O

4A-12 l

I

\ FIGURE 4A-1 B&H'0HNERS GROUP SAFETY AND PERFORMANCE IMPROVEMENT PROGRAM (SP'IP)

SPC SUBMIT SPIP RECOMMENDATIONS TO RRRB SUPPORT RRRB AND PAG REVIEH, MONITOR RESULTS REPORT DISTRICT' STATUS KEEP PAG INFORMED REPORT STATUS OF,NON-l_ CN SPIP RECOMMEN0ATIONS OF DISTRICT STATUS SPIP DISTRICT RECOM-TO B&WOG VIA'RECOMMEN- __

IN RTS MENDATIONS TO B&WOG DATION TRACKING SYSTEM (RTS)

FOLLCHING RESTART, SUB-MIT FUTURE SPIP RECOM-MENDATIONS TO COMMIT-MENT TRACKING SYSTEM 4A-13

4A.4 PLANT INTERVIEWS 4A.4.1 Purpose The interview program is to surface previously unresolved, but "known", problems which can (1) cause reactor trips and/or contribute to the severity of transients, and (2) degrada plant reliability or the optimal performance of the operating personnel. On this basis l the result of the plant staff interviews were processed through the QCI-12 PP&MIP. ,

4A.4.2 Project Scoce I

This program interviewed personnel from key plant and nuclear support functional groups. These persons were encouraged to identify systems, components, or operational problems and concerns of which they are aware and provide recommendations on how to resolve them.

One hundred and fifty-seven volunteers were needed to satisfy the representative selection criteria based upon MIL-ST0-105D. This requirement was met in all areas and a total of 180 volunteers have been interviewed.

dA.4.3 Interview Methodology The interview program covered a cross section of plant personnel and was intended to encourage personnel to identify issues or concerns of which they are aware that, when resolved, can contribute to optimal, reliable, or improved operation. Intervie.vs were conducted utilizing ,

4A 14

people who are skilled in the interview process. For the most part, I

] j d these same people accomplished the Rancho Seco CRDR interview program, and they were assisted in preparing for the PP&MIP by consultants who are experts in industrial human relations. The

. details of this program are provided in Appendix I, which is extracted from QCI-12. [

An introduction and question form was prepared and presented to each interviewee prior to the interview. Each interviewee is asked for info *mation about his/her background and is briefed as to the purpose of the interviews. The q0estions on the forms are then discussed one-by-one. Each interviewee is asked to expand on each answer until the interviewers' feel no further meaningful information is available.

The Interview Project Coordinator consolidates the ccmpil'ed list of concerns / recommendations. He forwards the recommendations to the RRRB using the Recommendation / Resolution sheets. The Interview Program Coordinator assures that the concerns and recommendations l 3

have been acted upon and dispositioned to the RRRS.

4A.4.4 Schedule The program interviews, including evaluation and disposition of the recommendations will be ccmpleted prior to plant restart.

Approximately 180 volunteers were interviewed. Some -1600 recommendations were developed prior to consolidation to eliminate duplication. These recommendations have all been crocessed through v the RRRS.

JA-15

4B MANAGEMENT, OPERATIONS, AND ADMINISTRATIVE PROCESS IMPROVEMENT This section identifies the program enhancements developed by the Olstrict's department managers which are being implemented to address the deficiencies contributing to the performance record and the December 26, 1985 event. These programmatic actions form the broad framework for the implementation of the findings from the systematic review process. Each programmatic action area is discussed in general, followed by specific commitments which address the lieficiencies related to the programmatic area. -

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48-1

48.1 MANAGEMENT EFFECTIVENESS O

Management and management process effectiveness have a major impact on the ability to operate the Rancho Seco Nuclear Generating Station in a safe and reliable manner.

The objectives of these actions are as follows:

a. To develop guidelines and agreements by which the SMUD Board, as

~ -

the governing entity, can improve its effectiveness in directing and monitoring the District's activities and obligations relating to the Rancho Seco Nuclear Generating Station.

b. In light of significant reorganization and managerial changes, monitor the status of corporate headquarters management improvements and provide an assessment to the Deputy General Manager, Nuclear, General Manager and Board.

c. To enhance the management process in support of the safe and reliable operation of the Rancho Seco Nuclear Generating Station.

48.1.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program O

48-2

1. Review current executive level management practices and attitudes to ensure that executive level management processes support the safe and reliable operation of the Rancho Seco Nuclear Generating Station. The Management Process Review Group will obtain an opinion from the Independent Review Panel.

48.1.2 ' Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained until Completed

1. Board of Directors / General Manager Improvement -

a. Establish within the Board, guidelines and agreements by which the Board, as an entity, can more. effectively set policy and direction.

(

b. Establish written performance measurement criteria, and a performance review process, for the General Manager (GM).

.c. Clarify the Board / General Manager working relationship in I

writing, including the reporting desired by the Board from the General Manager.

2. Corporate Management Improvement O

48-3

a. Assess current corporate-support interfaces with the Nuclear Organization and make recommendations to the Deputy General Manager, Nuclear and the General Manager regarding improved management of interorganizational working relationships.

3. Nuclear Program Management Improvement

a. Develop and implement a Rancho Seco Business Plan for use by the Board of Directors.

b. Establish a comprehensive, cohesive and clearly understandable set of GM and DGM, Nuclear policies and practices which provide upper tier direction for similar efforts at the functional manager and supervisory levels.

l

c. Establish up-to-date functional organization charters and position descriptions which accurately reflect responsibilities authorities, and accountabilities for all organization functions and job classifications.

d. Upgrade management programs and practices in the areas of functional planning, decision making, problem solving and 1

Interdepartmental collaboration.

O 1

48-4 i

1 1

e. Establish appropriate management monitoring and control Q systems to ensure that all levels of department management are kept informed on important department performance trends or problem areas on a timely basis. At the same time, ensure that excessively burdensome administrative control systems are not perpetuated or introduced.

f. Develop an employee communications program originating from the office of the DGM, Nuclear to ensure that all department employees are kept informed of District

~

concerns, departmental priorities and performance progress on a timely basis and encouraged to feel that they are an important part of the Rancho Seco team.

g. Develop a program for improving communications skills of Nuclear Department managers in presentations to the Board j of Directors, the public, and staff.

i

h. Establish a department Human Resource Management program which includes:

l

1) identification of priority management l development / training needs and the appropriate means for addressing each; O

48-5

2) identification of departmental priorities in terms of current vacancies and/or pipeline concerns;

3) engage more department management collaboration with the District's Human Resources organization in the recruitment / selection process.

1. Improve Department media and community relations by establishing a more pro-active media / community outreach

~ -

program.

j. Improve Nuclear Department interfaces with all other Departments in the District by instituting additional interdepartmental communication and problem-solving processes on a regular basis.

k. Develop a Rancho Seco Facilities Master Plan.

O 48-6 l

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48.2 QUALITY AND QUALITY ASSURANCE An effective quality program at Rancho Seco is important to achieve the near and long-term performance standards and objectives of'the' District.

The objectives of these actions are to improve the overall effectiveness of the Quality Assurance effort and to assure the ,

benefits of Quality Assurance _are realized in the near and long-term.

48.2'.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Reorganize the Quality function at Rancho Seco to enhance the Site Quality Assurance Department, providing increased focus in the following areas:

a. Quality Engineering

b. Quality Control i

l i c. Surveillance 1

i

d. Vendor Qualification and Source Inspection

e. Nuclear Program Audits O

48-7

2. Develop and implement the procedures and processes necessary to independently verify the effective closure of the actions identified for the Action Plan. A Quality Tracking System (QTS) has been developed and is in operation to aid in accomplishing this task.

3'. Institute interim measures to strengthen the materials control at the Rancho Seco site. This action will provide additional assurance that materials being installed are properly documented and in compliance with the applicable codes and standards. -

4. Institute interim measures to enhance the integration of QC planning with maintenance and construction instructions and activities. This action will assure effective and efficient quality inspection hold points are identified and implemented.

5. Increase the Site QA Department staff to assure the added demands of the Action Plan and changes in responsibilities can l

l be effectively implemented.

4B.2.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed

1. Update and modify the Quality program policies and procedures to enhance the effectiveness of the Quality Program, particularly I those dealing with material control, engineering, quality surveillance, and maintenance.

O l

48-8 l

-2. Identify and develop enhancements to the QA program to address any programmatic and management areas that have identified deficiencies from a quality perspective. 1 48.2.3 Actions to be Programmed for the Longer Term

1. Develop and implement the necessary policies and procedures to establish a more proactive quality program, which will also improve the effectiveness of audits.

i k

i 48-9

48.3 TRAINING In addition to subdividing training improvement actions by pricrity, O

it is also useful to subdivide into improvement areas. The actions presented here are divided into: management controls, facilities and resources; and specific instructional (Operations, Emergency Planning, Other).

4B.3.1 Management Controls, Facilities and Resources These improvement actions are those required to bring the 01 strict's Nuclear Training Programs up to a state-of-the-industry condition.

The intent of this action is to provide long term results and no items have been identified requiring completion prior to restart.

Actions to be Initiated as Promp'fy as Practicable, Schedule O

1. t Developed, Resources Assigned and Maintained Until Comoleted l

a. Continue the upgrade of Non-Licensed Operator Training to maintain INP0 Accreditation.

l l b. Initiate the process of achieving INPO Accreditation for l

the maintenance training area.

c. Develop the plan for installation of a computerized Training Informatica Management System.

O 48-10

V

.s .

d. Develop plans for centralized and secure storage of ,

/ training records. V

. 4 4

' g'i A, s

' I c, " 'p 'h ' ,:

Develop or purchase a Rancho Seco Siniulator Baseline Data

e. '

% .i k' Information and Tracking System consistent kitti the l' .

\

q J.s , >j Simulator now being purchased.

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Incorporate short term training items (lessonsilearnedF 3  ; ? ?' /

f. t y i. 1 1 -t

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into permanent training materials. ,

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2. Actions to be Programmed for the Longeir Term , '/;' j'

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t t . /

Complete the purchase and. InstaMation of a plant spec'1 A lc ;

a.

.I simulator. t x

, ;b

b. Complete the staffing of the Training Department vith SMUD a

employees. ,

1

~

c. Complete and maintain 1NPO accreditation for the;frem4f nder of the Training Programs.

48.3.2 Specific Instruction Actions' Items - Operations I

The specific instruction action items for Operations are detailed -

below:

48-11

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! 1. Actions to be Completed Prior to Restart or Completion of the Restart Test Program

a. Train licensed operators on Emergency Operating Procedures,

? including the changes resulting from the December 26, 1985

~ '

event, those revisions resulting from the EOP/AT0G review,

- and all completed recent design modifications.

,q .

b. Train licensed operators on the loss of ICS/NNI, including s ,

those procedures added or revised as a result of the

%, ' :t s

t December 26, 1985 event and related design modifications.

c. Train operators on valve applications and operation including 0]T on local and/or manual operators. This includes limits and precautions such as use of valve wrenches. Incorporate into requalification training.

d. Train operators on the specific lessons learned from the December 26, 1955 event. These include such items as the makeup pump failure, overfilling the makeup tank, cooldown rates, reactor vessel nead bubble formation, and the functioning of valve actuator controls.

e. Train operators on watch standing principles, including command and control training for shift supervisors, role i

and function of STA, equipment monitoring.

i O

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48-12 l

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f. Retrain operators on health physics requirements associated with their job responsibilities.

g. Train operators on their job related functions associated with startup testing.

48.3.3 Specific Instruction Action Items - Health Physics l

The specific instruction actions for Health Physics are summarized below:

1. Actions to be Completed Prior to Restart or Completion of the Restart Test Program

a. Train Health Physics Technicians and Operators on the procedure (s) for entry into areas of unknown radiological conditions.

b. Train Health Physics Technicians and Operators on proper response to radiological emergencies.

c. Train Health Physics Technicians and Operators on evaluation of radiological effluent discharges.

! 48-13 a w..-, ., ,. - ,-----,,-w,, . . . - , . . . - - - - -- - - - , - - - . , . - , - - - . - , _ , . - - - - - ,- -

48.3.4 Specific Instruction Action Items - Emergency Planning The following specific action items are to improve emergency response.

O

1. Actions to be Completed Prior to Restart or Completion of the Restart Test Program

a. Train assigned maintenance personnel on the maintenance of the Interim Data Acquisition and Display System (IDADS).

b. Train operators on the operational use of IDADS.

c. Update Emergency Preparedness Training Instructor Guides, Student Guides, and visual aids to support the October 1986 Drill.

d. Train assigned personnel on the revised Emergency Plan Procedures.

e. Provide management guidance to the operating crews (through training) on prioritizing multi-casualty events.

2. Actions to be Programmed for the Longer Term

a. Develop and implement an improved process for continuing Emergency Response Organization training.

O 48 14

48.4 OPERATIONS AND OPERATING PROCEDURES Improvement in operations is brought about by improving management controls such as procedures,-organization, policies, and through improved training of personnel. Improvements in training are addressed in Section 48.3. The actions te improve management controls are addressed here.

48.4.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program -

1. Issue a procedure defining the policy for procedural compliance and procedural guidance. This procedure will provide a direction on what constitutes " procedural compliance" and

" procedural guidance".

2. Correct procedural deficiencies identified during the review of the December 26, 1985 transient, and-those identified by the PP&MIP as requiring correction prior to restart.

3. Review, and revise as required, administrative control of E0P changes to assure verification and validation is performed.

i Conduct V&V on changes to the E0P't made subsequent to the original E0P implementation program, if not previously done.

4 The recommendations of IE Information Notice 86-64:

" Deficiencies in Upgrade Programs for Plant Emergency Operating Procedures" will be considered.

k 48-15

4. Compare Plant E0P's to the Generic B&W ATOG Technical Basis Document. Incorporate identified improvements into E0Ps.

Note: The Generic B&W Technical Basis Document (TBD) was issued after the E0P's, based upon the previously developed Rancho Seco specific ATOG Document, were implemented in accordance with the Rancho Seco Procedures Generation Package. The generic TBD replaces the plant specific ATOG, which embraces and Improves. This upgrade to the Rancho Seco E0P's, -

based upon the TBD, was being planned before the December 26, 1985 event.

5. Make the necessary modifications to the design change process to assure that design changes are incorporated into all operating procedures in a timely manner.

6. Assure operating procedures address the recommended topics of Regulatory Guide 1.33 Sections 11sted below. Implement 1

procedures which may be required.

1

a. Section 3 Procedures for Startup, Operation, and Shutdown of Safety Related Power System.

b. Section 6 Procedures for Combating Emergencies and Other Significant Events.

1 0 48-16 I

7. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification of important secondary mechanical systems.

Initiate corrective action for inconsistencies which are identified. This compliments the long term Configuration Management actions defined in Section 48.12. Systems included are:

a. Air Ejector / Gland Seal

~ -

b. Auxiliary Feedwater

c. Auxiliary Steam

d. Component Cooling Water

e. Instrument Air

f. Main Circulating Water

g. Main Condensate

h. Main Feedwater

1. Nitrogen Gas

j. Plant Cooling Water

k. Service Water 1

1. Turbine Electro Hydraulic Control

m. Turbine Lube 011 l

8. Establish an Administrative Procedure to insure that " Systematic Troubleshooting" is accomplished on requisite future events.

Guidance for this program is provided in memo GAC 85-1001, Rev.

2.

s 48-17

9. For those plant modifications, done during this outage, which impact the Control Room, the E0P's, or Operator Training, implement the Verification and Validation Program as appropriate to each. This program utilizes reviews, Table-Top exercises, Halk-Throughs, Simulator Exercises, and Plant Operation to accomplish the V&V to closeout modifications which impact the Control Room.

48.4.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed -

1. Develop a revised Nuclear Operations organization and begin staffing at the management level.

2. Develop a staffing plan and schedule to meet the needs of the revised Nuclear Operations organization. This will include the needs for licensed operators identified as rotational / transfer assignments.

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j 46-18 l

A 48.5 MAINTENANCE PROGRAMS AND PROCEDURES O The quality of the maintenance programs has a direct impact on the material condition and reliability of systems and components throughout the entire plant. The actions described below are intended to provide District Management with assurance that the material condition of Rancho Seco's safety systems, and those systems required for normal control as well as post-trip. control, are such that safe operation may be resumed.

48.5.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Inventory Calibrated Test Equipment (CTE) and calibrate and/or control use to prevent use of uncalibrated CTE.

2. Identify and assure current calibration of all in-plant instrumentation used in the performance of surveillance testing.

i

3. Rework the nt!keup pump and return to service.

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4. Complete the in-progress battery replacements (A, B, C, D, E, F).

5. Perform refueling interval surveillance of snubbers.

O 48-19

6. Complete rework of terminations in the Bailey Cabinets in the Control Room (NNI/SFAS/RPS/ICS).

7. Perform biennial Diesel Generator Inspection and replace turbochargers.

8. Define the critical items to be included in the PM program.

(This is to be an accelerated portion of the planned PM Program Upgrade.) As a minimum, this will include the Manual Limitorque Operated Valves (105), the Manual Non-Limitorque Operated Valves

~ -

(135), other Manual Valves important to process flow control in Class 1 and steam generator heat removal applications (143),

plant instrumentation required for surveillances, safety related HVAC and the Control Room normal HVAC system.

9. Complete Preventive Maintenance (PMs) on selected manual valves O

identified in 48.5.1.8 above.

48.5.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed

1. Develop a departmental procedure hierarchy and writer's guide for Maintenance Procedures.

2. Identify and prioritize procedures for generation and/or revisien.

O 48-20

1 4

3. Achieve authorized staffing levels within the PM organizations .

and activities.

4. Develop and/or revise the required programmatic procedures for the PM program to: assign responsibilities, authortty and accountabilities for the program; establish criteria and define the scope of the program; and define the interface with other work control processes.

5. Review existing PM tasks and frequency for critical equipment. -

Revise and augment as required by programmatic selection criteria.

I i 6. Perform Laboratory Failure Analysis of the ICS S1/S2 switches and ICS Power Supply Monitor which were in place on December 26, 1985.

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t 48-21

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48.6 HEALTH PHYSICS AND RADIOLOGICAL CONTROLS O

Improvements in Health Physics or radiological controls is brought about by improvements in organization, procedural controls, and training. Training is addressed in Section 48.3. Organizational and procedural actions include:

48.6.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Relieve operators of special HP duties.

2. Prepare a procedure, for Health Physics Technicians to use for entry into unknown radiological conditions.

O

3. Revise setpoints for plant gaseous effluent monitors to ensure unambiguous indications.

4. Issue a Radiological Event Directions Manual to provide more l

guidance for abnormal situations.

5. Issue new manuals to separate event and instrument procedures from the Radiation Control Manual.

O 4B-22

48.7 10CFR50 APPENDIX I DISCHARGE GUIDELINES The 10CFR50 Appendix I Guidelines for radioactive cesium can be exceeded under certain circumstances when the current lower limits of detection are used and applied in conjunction with technical specification requirements.

The objective of these actions is to upgrade the analysis and controls to provide confidence that discharges and the cumulative ,

impact of discharges will satisfy the objectives of the environmental i i discharge requirements.

48.7.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Evaluate the current radioactive waste analysis methods and sensitivity relative to their ability to support operation needs and the performance objectives of this task.

2. Develop and implement the changes in Radiochemistry methods and controls necessary to achieve the objectives of this task.

3. Review and revise the off-site discharge calculation manual to incorporate these changes.

l 48-23

4 48.7.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed

1. Evaluate the design of plant systems with the intent to improve the ability to operate within Appendix I Guidelines when operating with primary to secondary leakage. Implement plant improvements as appropriate.

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O 4B-24

1 48.8 EMERGENCY PREPARE 0 NESS O An effective Emergency Preparedness program is essential to the assurance of the health and safety of the pubile should an environmentally impacting event occur at Rancho Seco. Several weaknesses were apparent in the Emergency Preparedness Plan, and its implementation during the December 26, 1985 event.

l The objective of these actions is to upgrade the Emergency Preparedness program to assure that it le efficient and effective. -

48.8.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Meet with NRC (Region V) to review / critique Emergency Plan Action Plan.

i

2. Update Emergency Plan and Implementing Procedures to address December 26 event Lessons Learned.

3. Establish independent meteorological assessment capability.

4. Integrate EP commitments into commitment tracking program.

5. Evaluate notification / communication system and implement upgrades related to December 26 event lessons Learned.

48-25

6. Simplify Control Room dose calculation procedure.

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7. Implement new PASS procedures including core damage assessment.

8. Initiate " mini-drills" program.

9. Coordination and support of the Training Group per 48.3 including:

a. ERO identification

b. Plan / procedure update information

c. Facilities identification

d. Instruction materials upgrade

e. Scheduling

f. Tracking and documentation

g. Data base management

h. Management support of ERO participation 48.8.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed

1. Establish separate onsite and corporate plans.

2. Consolidate / cross index EP and Central files.

3. Complete multi-parametric data base including EP records and schedules.

O 48-26

4

4. Provide positional analyses for ERO versus District staff.

1 i 5. ~ Define / implement public education program enhancements for emergency response.

6. Integrate Emergency Plan Implementing Procedures and plant

{ operating procedures.

1-I 7. Complete installation of notification / communication system i

• including verification,-training and drills program. -

8. Redefine Emergency Preparedness maintenance program.

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l 48-27

48.9 HUMAN FACTORS Human Factors reviews of the transients in October and December,1985 have identified the need for some plant modifications and procedure changes.

The objective of these actions is to correct the identified deficiencies and enhance human interface with the plant.

It is important to recognize that the Rancho Seco Human Factors -

program, which developed the CRDR Response and program, is an integral component of the plant configuration control program. As such, modifications which result from recommendations developed by the Action Plan PP&MIP, are being reviewed for Human Factors considerations and compliance with CRDR requirements and commitments.

48.9.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Implement modifications and procedure changes resulting from the post trip Human Factors reviews:

l l a. Provide operator training associated with AFH valves FV-20527 and FV-20528.

b. Provide accurate local hand jack position indication for AFW valves FV-20527 and FV-20528.

48-28 l

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.c . Improve interface between Security and Control Room personnel

d. Install long cord on red phone.

e. Relabel ICS power supply breakers SI/S2.

t 48.9.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed. Resources Assigned and Maintained Until Completed

1. Modify Control Room access doors such that one door is used for exit and one for entrance.

2. Modify red phone power supply to eliminate spurious ringing following power supply transfers.

q

3. Assess capability to accelerate implementation of CRDR l modifications (Mod 142) currently scheduled for Cycle 8 and

! cycle 9 refueling outages. If possible accelerate implementation. Assessment by B&W is underway.

48.9.3 Actions to be Programmed for the Longer Term i

1

1. Participate in the INP0 Human Performance Evaluation program.

l 48-29

2. Implement CRDR modifications (Mod 142). These modifications were identified in the District's submittal to the NRC in December, 1985, which documented the results of the Rancho Seco control room design review.

3. Review modifications incorporated following 12/26 event for impact and/or compliance with CRDR criteria / program.

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O 48-30

48.10 MANAGEMENT INFORMATION SYSTEM b

U Plant Records and Database information are very important to the efficient and effective conduct of support activities for plant design, design modifications, operation, and maintenance. The records and data provide historical information on all the various activities at -the site and must be effectively managed to support improvement programs and ongoing programs.

The objective of these activities is to enhance the records and data -

management activities to effectively and efficiently support the Nuclear Department's needs.

48.10.1 Actions to be Completed Prior to Restart or Completion of the Restart

[ Test Program

1. Review implementation of NRC Generic Letter 83-28 commitments and develop plan for program enhancement.

4

2. Provide site information systems support for implementation and records management activities needed for restart.

I 48.10.2 Actions to be Initiated as Promptly as practicable, Schedule i Developed, Resources Assigned and Maintained Until Completed

1. Prepare a Nuclear Information Systems General Design, identifying and prioritizing improvement projects for implementation within three years.

4B-31 f

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2. Implement Vendor Data Program enhancements identified to achieve the program objectives.

'48.10.3 Actions to be Programmed for the Longer Term

1. Establish program and policy for development of integrated data management system design and implementation.

a. Complete Nuclear Information Management System (NIMS)

~

evaluation.

b. Establish on-site facilities and organization to support NIMS hardware / software.

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c. Implement NIMS program / data.

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48,11 COMMITMENT MA.1AGEMENT

.b Commitment management is an important aspect of the Olstrict's interface with "outside" agencies as well as for the management of the District's day-to-day activities.

Improvement in commitment management is aided by improvements in management controls, such as implementing procedures-and tracking commitment systems. The actions to improve the commitment management programs are identified below. -

48.11.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program O 1. Revise commitment management procedure to include tracking and y/

compliance features.

2. Install a new commitment tracking system per 48.11.1.1 above.

3. Develop system / user documentation for the new commitment tracking system.

4. Verify the commitment tracking system database with respect to current known commitments, such as those addressing NUREG 0737 issues and requirements.

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5. Verify all ccmmitments required prior to restart are ccmplete.

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48-33

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! 48.11.2 Actions to be Initiated as Promptly as Practicable. Schedule Developed. Resources Assigned and Maintained Until Completed

1. Verify the comprehensiveness of the commitment tracking system database.

2. Each Nuclear Department will establish specific milestones for reducing its backlog of open commitments.

48.11.3 Actions to be Programmed for the Longer Term

1. Integrate the commitment tracking system with the Nuclear Information Management System.

O 48-34

48.12 CONFIGURATION MANAGEMENT n/

~

w Configuration management identifies, controls, status and verifles the plant document and hardware configurations.

This is an important program which is necessary to assure actions on the part of the various organizations responsible for interacting with the plant do so based on a consistent set of documents which are also consistent with the plant hardware configuration.

The objective of tLese actions is to take the, steps necessary to improve the effectiveness of the configuration management system at Rancho Seco.

4B.12.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Verify that control room drawings are current, in accordance with existing procedures.

2. Provide Nuclear Engineering support to plant operations to address ano expedite configi. ration management issues.

48,12.2 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed l

i 48-35 f

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1. Review and evaluate all temporary modifications and close out all existing abnormal tags that need to be converted to permanent plant modifications.

2. Reduce the backlog of DCNs.

3. Develop the System Design Basis documents for NEP manuals on S0P related systems.

48,12.3 Actions to be Programmed for the Longer Term

1. Establish management direction for a Configuration Management program for Rancho Seco consisting of:

a. Policy D. Specifications

c. Computer hardware and software

d. Implementing procedures

e. Training

2. Establish or upgrade existing equipment and supporting documentation identification systems needed for total plant configuration control.

3. Upgrade change control packages that control modification from change request through close out. This includes identification of all affected documentation such as procedures, training plans, and simulator upgrades.

46-36

l 1

4. Reorganize drafting into a designldrafting organization.

5. Train Nuclear Engineers to utilize these designers to reduce the engineering work load.

6. Develop new, or simplify existing procedures to clearly define the review process for drawings.

7. Conduct a cost / schedule review to determine whether a CAD system

~

can be justified for Rancho Seco. -

8. Develop or upgrade existing systems to provide verification that i

configuration documentation reflects the true hardware configuration.

9. Develop or upgrade existing systems to provide the status of all

, documentation and equipment in a timely and accurate manner.

i j 10. Develop a work package system for all facility changes.

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48-37 i

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48.13 MATERIALS MANAGEMENT O

The Materials Management program is a key element which supports the achievement of reliable and safe plant operations. This program assures that correct parts are available for the repair of plant components and obtains and manages the materials necessary to support plant modifications.

The objective of these actions is to enhance the materials management program at Rancho Seco to assure the long-term plant performance goals can be achieved.

48.13.1 Actions to be completed prior to restart or completion of the restart test program:

Conduct a review of the current Materials Management Program.

O 1.

2. Develop and implement an interim program to provide materials in support of the restart program.

O 48-38

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I- 4C PLANT MODIFICATIONS AND MAINTENANCE IMPROVEMENTS i

i This section ' identifies specific plant modifications, and l l t l maintenance improvements, which are being implemented to resolve the

lessons-learned, recommendations, and programmatic deficiencies 1

Identifled by the Restart and-Performance Improvement Program. Each subject area, addressing issues from the December 26 event, is

! discussed in general, followed by specific commitments relative to i

i that issue.

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4C.1 INTEGRATED CONTROL SYSTEM (ICS) AND INTERFACING SYSTEMS The Integrated Control System is an important element in the control of the plant during normal operations. Several changes to the ICS and interfacing systems have been identified by the District and the B&W Owners Group which will improve the operators ability to maintain the plant within the post trip window and reduce the potential for reactor trip as a result of failures of tne ICS. The general programmatic actions and specific design modifications to upgrade the Rancho Seco ICS and associated systems are described in this section.

4C.I.a General Programmatic Actions Changes to the ICS, which offer potential to improve the reliability of the ICS, improve the operators ability to maintain the plant within the post trip window, and reduce the potential for reactor j

trips are being developed or have been developed by various l organizations.

The objective of this general programmatic action is to implement those changes which are judged to enhance safety cr operability.

4C.I.a.1 Actions to be Comoleted Prior to Restart or Comoletion of the l

Restart Test Program 1 O a:-2

. __ _ _ . ._ _ _ . _ = . _ . _ _ . _ ._.__ __ _ _._ . _ _____ _ . _ ___ _ _. . _ _ . __

4 ,

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, The following specific modifications /or actions will be  !

implemented:

. a. Improve ICS Reliability 4

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1. Improve reliability of ICS power supplies -

l a) Provide dedicated AC Inverters q

b) Provide larger DC capacity i 2. ' Implement Upgrades

!- r j a) Reduce MFHP runback rate to 257./ min. ,

i-b) Substitute for RC flow the RCP Status, i.e., number of RCP's 1

c). Remove SU FW flow correction to the MFW flow signal i

d) Replace modules with improved Bailey revised modules i

e) Remove FH temp correction from total"FW demand signal i

f) Revise RCP Runback rate to 25%/ min.

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' i g) Revise Asymetric CR0 Runback rate to 37./ min.

b AC-3 l

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h) Improve post maintenance testing and surveillance

1) Replace all S1 and S2 switches with new switches j) Rewire PSM to delete the "Daisychain" k) Test / Inspect all cabinet wirewraps and lugs

b. Improve Status Information

1. Upgrade ICS/NNI annunciation to remove ambiguities

2. Provide computer inputs for ABT (AC power supply) status logic

3. Add open/close status lights for ADVs, TBVs, MFW, and SUFH valves

4. Label ICS/NNI indicators and recorders as to source of power i

! S. Add first-out MFP trip monitor I

^

6. Provide adequate parameter trending independent of ICS/hNI t

l 7. Provice clear labels for all S1 and 52 switches O

i 4C-4 l

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j 8. Add all hot shutdown related parameters to SPDS, independent of ICS/NNI d

c. Improve Plant operability by assuring that plant will go to

. known. safe state on loss of.ICS/NNI q

1. Revise casualty procedures for ICS/NNI failure l

2. Replace ICS AFH'and ADV control with Class 1 EFIC J

3. Trip ICS on loss of NNI-X, Y, or Z power (AC or DC) i

4. Trip MFP's on loss of ICS power (AC or DC) i j 5. Provide auto closure cf TBV's on loss of ICS Power i

6. Provide Class I bottled air supply for ADV's, AFW, MFW,  ;

. and SUFW Valves 4

j 7. Modify Auxiliary Steam Reducing Station to fall at setpoint

d. Improve ICS restoration l 1. Revise restoration of power procedures to' address new 4

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modifications i-i I

2. Evaluate (and correct as necessary) control element failure position and power restoration position i

s 4C-S

4C.l.a.2 Actions to be Initiated as Promptly as Practicable, Schedule Develooed, Resources Assigned and Maintained until Comoleted

1. Develop and implement those design changes or enhancements that were not required to be implemented prior to restart, consistent with their assigned priority.

4C.l.a.3 Actions to be Programmed for the Longar Term

1. Actively participate in the B&W Owners Group efforts to upgrade or replace the Integrated Control System.

4C.I.b Actions to Reduce the Impact of Power loss on ADV's and TBV's The loss of power to the ICS during the Decerher 26, 1985 event l

l resulted in the ADV's and TBV's falling to mid-stroke due to the 1 .

bi-polar nature of the ICS. This resulted in an overcooling 1

transient following the reactor trip.

l l The objective of these acticns is to implement plant features which are effective in addressing this deficiency. To accomplish this objective, the following actions will be taken.

O 4C-6

-f 3 4C.I.b.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

}

1. Provide controls in the Control Rocm from which the operator j can cperate the TSV's, and which will cause these valves to remain closed on loss of ICS (DC) power. Note: ADV's will be controlled by EFIC.

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j 4C.I.c Actions to Address the Adequacy of the ICS Power Monitors The DC power supply monitor for the integrated control system was identified following the December 26, 1985 event as a single failure point which could contribute to the loss of ICS. ,

The objective of these actions is to evaluate the potential impact of the power supply monitor and to implement changes which will effectively improve the reliability of the power supply.

4C.I.c.1 Actions to be Completed Prior to Restart or Comoletion of the Restart Test Program

! 1. Determine the contribution the ICS Power Supply Monitor had in December 26, 1985 transient.

2. Evaluate the potential improvement to ICS reliability if redundant PSM's are installed.

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= .. , , _ . - _ _ _ . - . . . _ - . .

3. Determine the potential benefits to be obtained through the installation of independent PSW s.

4 Implement design improvements or document justification for not implementing PSM modifications.

4C.I.d Actions to Improve the Status Indication on Loss of ICS Power The status indicator window logic for DC power status to the ICS was not adequate to provide the operators with sufficiently clear indication of the ICS DC power status to provide for prompt operator recognition of the cause of the loss of ICS.

The objective of these actions is to implement design changes whicn are effective in addressing this deficiency. To accomplish this objective the following actions will be taken.

4C.I.d.1 Actions to be Completed Prior to Restart or Comoletion of the Restart Test Program 1

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1. Provide separate windows for the status indication of the following:

L

a. ICS Trouble (fan failure, power supply failure).

b. ICS Failure (Loss of DC Buss).

l 4C-3 l

, /s . 4C.1.e Actions to Evaluate the Failure Consequences of Various ICS Inputs, Outouts, and Components The ICS has undergone a number of changes since a generic failure / consequence analyst's"was performed in (approximately) 1980, BAW1564. These changes include not only physical out philosophical design basis changes.

The objective of these actions is to update and evaluate the impact the changes have had to the overall functional performance of the ICS relative to individual failures of various ICS inputs, outputs,

, and components. To accomplish this objective, the following actions j willbetaken.

\- 4C.1.e.1 Actions to be Initiated as Promptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed

a. Participate in B&W Owners Group efforts to perform a generic I

failure / consequence evaluation of the Model 820 ICS. This  !

effort was initiated August 1, 1986.

8 4C.I.e.2 Actions to be Programmed for the Longer Term

a. Evaluate results and applicability of B&W Owners Group evaluation findings and recommendations to Rancho Seco.

G JC-9

b. Conduct supplemental evaluations as required to achieve Rancho Seco specific information.

c. Develop and implement applicable modifications to Rancho Seco ICS.

4C.I.f Actions to Address the Adequacy of ICS Power Supply The December 26, 1985 transient was initiated when the ICS DC power was interrupted by acticn of the single Power Supply Monitor.

Operator action 26 minutes later reclosed the breakers providing power to the DC Power Supplies. During the power outage, the ICS Hand / Auto stations were inoperable and devices controlled by the ICS responded as when receiving a zero volt CC control signal.

The objective of these actions is to evaluate, develop, and implement beneficial design changes to reduce the likelihood that this event would reoccur. To accomolish this objective, the following actions will be taken.

4C.l.f.l Actions to be Programmed for the Longer Term

a. Evaluate need for DC Bus battery backup based on reliability of power supplies, recent modifications, etc.

b. Evaluate Hand / Auto station backup power.

O 4C-10

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' c. Participate with B&W Owners Group to enhance ICS and related power supplies.

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d. Develop and implement modifications identified as necessary during evaluations. .

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4C.2 NON-NUCLEAR INSTRUMENTATION (NNI)

The Non-Nuclear Instrumentation is an important element in the control of the plant during normal operating evolutions. Several changes to the NNI have been identified by the Distr:ct and the S&W Owners Group which improve the operators ability to monitor and control plant parameters. The general prog'ammatic and specific actions to upgrade the NNI are described in this section.

4C.2.a General Programmatic Actions Changes to NNI, which offer the potential to improve the reliability of the NNI, and improve the operator's ability to minimize reactor trips and maintain the plant within the post trip window, are being developed or have been recommended by various organizations.

The objective of this general programmatic action is to assemble, review, and implement those changes which are judged to be safety or operationally beneficial. To accomplish this oojective, the following actions will be taken:

4C.2.a.1 Actions to be Initiated as Promotly as Practicable, Schedule Develooed, Resources Assigned and Maintained Until Comoleted The following specific modifications /or actions will be implemented, as a minimum, to achieve the objectives noted:

O

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's a. Improve NNI

! 1. Improve reliability of NNI power supplies

2. Upgrade NNI Modules to latest'B&W models.

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3. Update NNI drawings to correct _ discrepancies identified in 0

the deterministic failure analysis and the District's ongoing efforts to upgrade plant performance.

b. NNI Status Information l .

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.l. Upgrade annunciation 4

2. Provide computer inputs for ABT status i

f 3. Tag control room instruments with relevant NNI power or

! signal channel identification 9

4. Provide adequate NNI parameter trending in IDADS i-1 (including recorder replacement) i i i l c. Improve Plant operability by assuring that plant will go to i- known safe state on loss of NNI

1. Provide automatic trip of ICS power on loss of NNI-X, Y, j or Z power 1

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2. Revise procedures for loss of NNI

3. Review NNI restoration procedures for adequacy in event of NNI failure 4C.2.a.2 Actions to be Programmed for the Longer Term

1. Actively participate in the B&W Owners Group efforts to upgrade or replace the Non-Nuclear Instrumentation equipment.

4C.2.b Actions to Address the Adequacy of NNI Power Monitors As a result of the' December 26, 1985 event, the ICS power supply monitor (PSM) was identified as a single point which could cause a loss of ICS DC power. The NNI has much the same power monitoring configuration as the ICS. The objective of these actions is to evaluate the poter.tial impact of these power supply monitors and to implement changes which will effectively improve the reliability of the power supplies.

4C.2.b.1 Actions to be Ccmoleted Prior to Restart or Comoletion of the Restart Test Program

1. Evaluate the potential improvement to NNI reliability if redundant PSM's are installed.

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i s 2. Determine the potential benefits to be obtained through the s- installation of independent PSM's.

3. Implement design improvements or document justification for implementing PSM modification.

Inspect electrical terminations within NNI cabinets. l

4. -

4C.2.c Actions to Address the Adeauacy of Status Indication for NNI, and

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Affected Instrumentation on loss of NNI Power l

Analysis (DFA) has determined that a loss of NNI-X, -Y, or -Z power

> will re'sult in the loss of critical signals to the ICS and cause control room indication to become ambiguous. This situation results

,)

s/

m in a reactor trip and loss of indication necessary to maneuver the plant. The following modifications together mitigate the consequence and reduce the dependence on operator actions.

4C.2.c.1 Actions to be Completed Prior to Restart or Comoletion of the s

Restart Test Program

1. Tag NNI-affected indicator / recorder I

2. Trip ICS power on loss of NNI-X, -Y, or -Z power

3. Make modifications to ICS to provide autcmatic control of ICS/NNI power.

4C-15

4 Provide separate annunciator windows to indicate:

a. Loss of NNI-X (DC)

b. Loss of NNI-Y (DC)

c. Loss of NNI-Z (DC, switching supply)

d. NNI trouble (fan failure, single power supply failure) 4C.2.d Actions to Ensure Adequacy of Safety Parameters in the Event of  ;

Loss of NNI As a consequence of a Loss of NNI event there is likely to be some confusion as to which instruments are still functioning, even with the improvements identified in 4C.2.a above. To preclude such impacts on the timely response to transient events, and to ensure that the operator will have high quality and reliable information l for monitoring safety parameters, and the " Post-Trip" window, in f acccrdance with the plant Emergency Operating Procedures (EOP's);

the SPDS will be upgraded to class I. Although the operator can still follow and control an event without the SPDS (by plotting RCS Pressure and Temperature on paper per the ECP's) the upgrading of the SPDS P/T display will reduce the likelihood it would be necessary.

4C.2.d.1 Actions to be Comoleted Prior to Restart or Completion c? the Restart Test Program I

4 8 4C-15

A i

- p 4C.3 FEEDWATER AND STEAM SYSTEMS kj) -

The design features and performance of the Feedwater and Steam Systems are important elements in determining the ability of the operators to perform normal and abnormal operating evolutions in a manner which minimizes the number of reactor trips, challenges to safety systems and maintain the plant parameters within the post trip window. The actions associated with these systems are described in this section.

4 4C.3.a Actions Associated With Emergency Feedwater Initiation and Control Rancho Seco developed a design and established a schedul'e for the

. implementation of an Emergency Feedwater Initiation and Control 1

System (EFIC) in response to industry lessons learned. This system provides:

a) Safety Grade Auxiliary Feedwater initiation and control.

i b) Safety _ Grade Atmospheric Dump Valve (ADV) control.

c) Safety Grade Main Steam Failure Logic.

4 The schedule for the implementation of this modification is prior to restart. This system would have reduced the severity of the cooldown transient of. December 26, 1985.

4C-17

These changes will provide for control of ADV's, TBV's, and AFW  !

control valves, provide independent of the ICS. To achieve this objective, the following act'ons will be taken.

4C.3.a.1 Actions to be Completed Prior to Restart or Comoletion of the Restart Test Program

1. Install EFIC and provide a control room panel as an extension to the H1SS panel.

2. Implement control grade modifications to close TBV's on loss of ICS power.

4C.3.b Action to Imorove Auxiliary Steam Control Valve Ooeration on Loss of ICS Power Loss of ICS power during the December 26, 1985 event opened the auxiliary steam pressure control valve to mid-position, causing overpressure in the auxiliary steam header and lifting of one of the headers two relief valves.

The objective of these actions are to effectively address this plant design condition.

4C.3.b.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program O

AC-13

T I'~ a. Develop and implement plant modifications to the auxiliary

.I steam controls to assure valve control on loss of ICS power.

4C.3.c Actions to Reduce Main Feedwater Contributions to Reactor Trips The main feedwater system is a significant contributor to plant trips and transients. The objective of these actions is to reduce the number and severity of feedwater induced trips and transients.

4C.3.c.1 Actions-to be Completed Prior to Restart or Completion of the Restart Test Program -

1. Review operation of Main Feedwater Pumps and status of

' incorporating lessons learned from October 2, 1985 event.

, J

2. Retune the ICS.

i

3. Validate setpoints and proper initiation / interface with Auxiliary Feedwater System.

4C.3.c.2 Actions to be Initiated as Promotly as Practicable, Schedule l Developed, Resources Assigned and Maintained Until Completed i

1. The recommendations contained in the B&WOG Availability i_ Committee Report, 47-1159449-00, "MFH Pump Trip Reduction l Program Final Report" will be evaluated for applicability to

' Rancho Seco and implemented as approoriate.

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4C-19

4C.3.d Actions from Control Room to Ensure Capabifity to Isolate Main Steam System l The two 507. capacity steam lines run from the Reactor Building to the Main Turbine. There are numerous penetrations into tnese lines which provide plant steam services and steamline protection. All service connections, greater than two inch alameter, are provided with remotely operated valves. The controls for these diesel power backed electrically operated valves are located in the Control Room. Safety and relief valves are not provided with upstream isolation valves.

4C.3.d.1 Actions to be Comoleted Prior to Restart or Comoletion of the Restart Test Procram i

1. Halkdown the Main Steam Lines and verify that each service connection,' greater than two inch diameter, is provided with capability to Isolate from the Control Room.

2. Tag control rocm switches to clearly indicate valves associated with A- or S-Steam Generators.

3. Evaluate Main Steam Line Supports for effects of flooding.

4. Increase 10 ADS sample frequency for Main Steam / Main Feedwater !

Parameters. ,

O 40-20

4C.4 EMERGENCY DIESEL GENERATOR RELIABILITY O

The reliability of emergency Diesel generators is important to the mitigation of certain events.

L The objective of these actions :is to enhance the reliability of the Rancho Seco Emergency 01esel Generator systems. To accomplish this objective, the following actions will be taken.

4C.4.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Replace Turbochargers.

4C.4.2 Actions to be Initiated as Promotly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed

1. Evaluate performance history of the Bruce-GM Emergency Diesel i Generators and develop recommendations for reliability

• i l improvement.

l l 2. Determine and implement identified modifications.

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l 3. Enhance the preventive maintenance program.

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JC-21 l - .. . - . ,. . . - , . _ , , . - . _ _ . . - - , _ . - . ~ _ - - , - _ . _ - - . . - . . _.... - . .

4C.5 REACTOR CCOLANT SYSTEM AND PRESSURIZER 4C.5.a Upgrade Pressurizer Relief Valve Discharge Ploing Succorts The subject piping was reanalyzed for dynamic loads including 2-phase and liquid flow as part of the response to TMI lessons learned. SMUD provided the results of this analysis and a justification for continued operation with the existing configuration by letter to the NRC on July 29, 1983. As committed, these supports are now being upgraded to restore their design margin.

4C.5.a.1 Actions to be Comoleted Prior to Restart or Ccmoletion of the Restart Test Program

1. Issue ECNs for new supports and support modifications.

2. Inspect, reanalyze, and redesign.(as required) ring structure anchoring supports to Pressurizer.

3. Construct new supports and modify existing supports and ring structure (if required).

4. Construct pressurizer support structure modifications and modifications to existing work platforms required to resist new pipe support loads.

O JC-22

I 4

4C.6 ENHANCE THE POST ACCIDENT SAMPLING SYSTEM (PASS) OPERABILITY The objective of these actions is to enhance the operability of the PASS system to meet its design objectives.

4C.6.1 Actions-to be Completed Prior to-Restart or Completion of the Restart Test Program t

i

1. Complete the SCAS panel. rebuild.

2. Complete associated peripheral equipment upgrades.

3. Document the compensating equipment in the environmental lab.

4. Ccmplete work required to solve H 2 monitoring heat tracing-problems.

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l 5. Revise operating procedures and complete training on revised system and conduct system functional test.

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T 4C.6.2 Actions to be Initiated as Promotiv as Practicable, Schedule Developed, Resources Assigned and Maintained Until Comoleted i

) 1. Replace Dionex program controller.

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2. Install R-15044 Sample Cryers.

4C-23

..-..--,-,-c-- ,.-r__- - - - - ,.m.-w -- - . - --,m-., ue--...-ym,o ,--,cy.--.---yy.v .,,yyw-.y,, ...,...w-,. --,v fTrv-'-'*** ---a*v-* W w-w

4C.6.3 Actions to be Programmed for the Longer Term Complete PASS decay heat valve replacement during the cycle 8 outage.

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O JC-24 '

e 4C.7 ACTIONS TO ENHANCE CONTROL ROCM/TSC AND NSEB HVAC - OPERABILITY AND RELIABILITY i

i Since the start-up and turnover of the essential HVAC' systems for the Control Room /TSC and the NSEB during the 1985 refueling outage, 3 ,

several maintenance and operational problems with these systems have been identified. Most items had been identified and addressed by -

mid-December of 1985 by ar, interdepartment task force, with several additional items identified as a result of studies following the December 26 transient.

The objective of these actions is to ensure operability of'these systems in-accordance with the original design bases,' improve.

reliability of the systems, and facilitate the maintenance and ,

operability or the systems.

! 4C.7.1 Actions to be Completed Prior to Restart or Completion of the

Restart Test Program

1. Excessive HVAC noise affecting Control Room Operator

ccmmunications L. ?

a. Prepare _and implement a detalled action plan / testing program to identify source (s) and propose modifications.

I

b. Develop and implement those design and procedural changes

} needed to reduce noise levels to allowaole limits.

! 4C-25 i

_ . , _ . .. ... ,____.._, _m . . _ _ . . . . _ _ . _ _ _ . _ . , _ , , _ , _ _ . - _ _ . _ . . . . . . _ _ . _ . . . _ . . . _ _ . , . _ . _ . . . -

4C.7.2 Actions to be Initiated as Promotly as Practicable, Schedule Develoned. Resources Assigned and Maintained Until Completed

1. Evaluate and implement design changes if necessary to improve balancing capabilities.

2. Develop and implement the changes to install flow meters to facilitate surveillance testing of the Control Room /TSC HVAC filter units.

3. Develop and implement the changes identified as necessary to facilitate maintenance cf Control Room /TSC HVAC equipment (i.e., replace Air Handler Unit access doors, modify the lube manifold to condenser fans, etc.)

4. Fire Damper upgrade actions

a. Develco and implement the changes necessary to add dampers through the TSC ceiling.

b. Develop and implement the changes necessary to upgrade

dampers in the wall between Control Room and TSC.

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5. Control Room /TSC essential filter unit flow control

a. Investigate flow control through filter units and recommena imorovements.

l JC-26 i

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6. Essential HVAC compressor motors i

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a. Investigate replacement of existing motors.

7. NSEB essential air handler air tiow

a. Develop improved methods to adjust air flow.

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8. If required, develop and implement design change to add a sample manifold to-filter banks in each of two units. ,

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A 4C-27

-,. - . - - , . , . , - - - - - . . - - - , , - - - - . .e,.n- ,.-_--,--,_e-,---.--. ,--,--,..,,,n., . , , . , . - . n- , , ,, . --- . . . - - - - - - - . - - . . - .- , - -. . - - - - - .. , _ . . _ -

4C.8 INSTRUMENT AIR SYSTEM REl.IABILITY Rancho Seco has experienced two transients in the past due to a loss, or partial loss, of instrument air. Shutdown to cold conditions is possible without the IAS but it is not a normal shutdown and may require the manual operation of important valves.

The objective of these actions is to improve the reliability of the instrument air system to minimize its contribution to plant transients and reduce the potential impacts to the cooldown transient.

4C.8.1 Actions to be Comoleted Prior to Restart or Cempletion of the Restart Test Program

1. Complete IAS system review to identify hardware modifications required to improve system reliabilit'y.

2. Replace leaking letdown filter valve operators.

3. Add diesel-driven air compressor.

i 4 Provide bottled air backup to critical valves.

5. Perform IAS walkdown to identify additicnal air leaks and any l

L P&ID discrepancies.

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6. Develoo and initiate priority i modifications identified during l

system review, i

'C-23

l I

i 4C.8.2 Actions to be Initiated as Promptly as Practicable, Schedule

)

Developed, Resources Assigned and Molntained Until Comoleted f

1. Develop and implement priority 2 and 3 modifications identified in IAS system review. .

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s

' 4C-29 i

, - , - - + , - - - ,,--,,-n~,,n- ---,..---ne--- ,,--a.., ~,n.n,,,--_n-,-

4C.9 REACTOR BUILDING PURGE FLOH RATE MEASUREMENTS The ability to determine' containment purge flow' rate is currently ircpacting the off-site dose calculations curing containment purges.

The objective of these actions is to enhance this calculational capability.

aC.9.1 Actions to be Initiated as Prceatly as Practicable. Schedule Developed, Resources Assigned and Maintained Until Completed

1. Perform engineering evaluation of flow measuring systen deficiencies and identify aapropriate modificaticns.

2. Install and test identified modifications.

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40-30 i

4C.10- FIRE PROTECTION SYSTEMS Ranc'ho Seco has completed the Appendix R safe shutdown analyses and has committed to specific additional. modifications that are shewn in the living schedule. Additional changes'and modifications which are-desirable to ' improve the operability of the itarious systems have been identified as a result of the December,1985 transient and the

. District's continuing efforts to improve plant. performance and are shown below.

4C.10.a Fire Alarm Systems The objective of these actions is to enhance the performance of'the fire alarm system.

4C.10.a.1 Actions to be Initiated as Promptly as Practicable, Sche ~dule Developed, Resources Assigned and Maintained Until Completed

1. Develop and implement modifications to provide for manual operator override of a trip of the Auxiliary Building ventilation fans.

i l

2. Upgrade the control logic and schematic diagrams for the fire protection system.

4C-31 l

3. Develop and implement appropriate upgrades to prevent spurious signals on power transients.

4C.10.b Separation of NSEB Damaer Controls and Equipment The current control circuitry and equipment is to be upgraded to closecut an action item identified in the safe shutdown review and to support the operability of the NSEB emergency ventilation system.

4C.10.b.1 Actions to be Initiated as Prcmptly as Practicable, Schedule Developed, Resources Assigned and Maintained Until Completed

1. Develop and implement fire alarm and HVAC panel upgrades to separate the centrol circuitry and equipment for the Train A and Train B Dampers.

4C.10.c Water Leakage through Floors Following Actuation of Fire Protection Systems The fire detection system in the NSES Cable Shafts and Tunnels needs to ce upgraded to close-out an item identified in the safe shutdown analysis and to support the EFIC system installation.

4C.10.c.1 Actions to be Initiated as Promptly as Practicable, Schedule Developed. Resources Assigned and Maintained Until Completed l

l 1. Identify vital areas of cotential imcact due to leakage.

4C - 3.?

t - _ .

, _ .._. _ . _ __ _ _. _ ..- _ . _ . . _ _ _ . _ . _ . . _ _ _ . _ . _ _ _ . . . _ _ _ _ _ . . _ _ _ _ . _ _ . . - ._m.. . . . _ _ _ _

A i

f-I

.2. Evaluate effect of impact of potential leakage on safe' shutdown l equipment.

3. Inspect all vital electrical equipment areas for potential

) . leakage paths.

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4. Evaluate the results of the inspection and. identify recommended. l i

.. corrective actions.

1 ,

i.

j. 5. Develop and tmplement changes necessary to address recommended ~

? . ,

corrective actions'. '

b 6. Review and upgrade as necessary preventative maintenance -;

procedures to maintain drain lines clear of obstructions.  !

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l 4C-33 .

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4C.ll MOTOR OPERATED VALVES Rancho Seco motor operated valves are being evaluated for the

• l potential to have common mode failures during plant transients due to improper switch settings. This issue was identified by NRC IE Bulletin 85-03. Other NRC bulletins and notices have referred to valve operation problems experienced at other utilities.

Additionally, INP0 has reported upon problems related to motor operators in their SCER and SER programs. SMUD has committed to develop and implement a program to ensure that motor operated valves function properly prior to restart.

As specifically addressed in IE Bulletin 85-03, thirty valve assemblies,will be refurbished and tested to ensure reliable operation of the AFH and HPI system. An additional seventy safety system valve assemblies will also be refurbished to ensure their proper functioning prior to start-up. The balance of M0V's on site

. (63) will be refurbished to ensure support system dependability.

This refurbishment program is underway. Parts and replacement components have been arriving since February 1986. Current procedures incorporating the latest vencor information and INP0 recommendations are being used to optimize each assembly. Dynamic monitoring of the motor operated valves to confirm the proper functioning of each is being accomplished with MOVATS testing equipment. At the completion of refurbishment, a baseline record of response will be recorded and be available for comparison at the

'O AC-34

1 i

k time of any future maintenance. The data will be recorded with a site computer software and hardware. Before return to service, each '

assembly will be satisfactorily checked against the' latest' input 4

from vendors, NRC, INPO, EQ inspectors, and the on-site Quality i

, Group.

The objective of the refurbishment program is to assure reliable operation of motor operated valves. Continued availability of the assemblies will be assured by a preventative maintenance program -

which addresses each MOV assembly.

i

4C.ll.1 Actions to be Completed Prior to Restart or Completion of the Restart Tes't Program s i V 1. Refurbish the 100 Safety Related MOV's.

4C.ll.2 Actions to be Initiated as Promptly as Practicable, Schedule f Developed, Resources Assigned and Maintained Until Completed i

1. Refurbish the 63 Non-safety Related MOV's. i i

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'(

i l 4C-35 yr y- - + - - - - - - -e-w - --e c r e- r, ce.-.w- a - w w-w-w --- w w, e- e r y- e -- r-.e--ee---m-e--r---=------,----ww--,--,,--,--,-w-m%. - w e -w+,ee w - . v ar. - - e - %r

• we --

4C.12 CRITICAL PUMPS FAILURE ON LOSS OF SUCTION During the Oscemoer 26, 1985 event, an operator error resulted in both sources of suction to the make-up pump being closed resulting in major pump damage.

The objective of these actions is to effectively address this issue to avoid reoccurrence.

4C.12.1 Actions to be Completed Prior to Restart or Completion of the

!iestart Test Program

1. Evaluate procedures and provide training to prevent recurrence.

4C.12.2 Actions to be Initiated as Promotly as Practicaole, Scnedule Developed, Resources Assigned and Maintained Until Comoleted

1. Engineering is to review design philosophy for suction valve interlocks and alarms on critical pumos and identify approcriate modifications.

O 4C-36

r 4C.13 MAINTENANCE PROGRAMS AND ACTIONS v .

'The quality of the maintenance programs has a direct impact on the material condition and reliability of systems and ccmponents throughout the entire plant. The actions described below are intended -to provide District Management with assurance that the material condition of Rancho Seco's safety systems, and those systems required for normal control as well as post-trip control, are such th'at safe operation may be resumed.

4C.13.1 Actions to be Completed Prior to Restart or Completion of the Restart Test Program

1. Inventory Calibrated Test Equipment (CTE) and calibrate and/or control use to prevent use of uncalibrated CTE.

2. Assure current calibration of all in-plant instrumentation used in the performance of surveillance testing.

3. Rework the makeup pump and return to service.

4. Complete the in-progress battery replacements (A, B, C, D, E, F).

5. Perform refueling interval surveillance of snubbers.

\

4C-37

6. Complete rework of terminations in the Bailey Cabinets in the Control Room (NNI/SFAS/RPS/ICS).

7. Perform bienr:'al Diesel Generator Inspection and replace turbochargers.

8. Define the critical items to be included in the PM program.

(This is considered to be an accelerated portion of the planned PM Program Upgrace.) As a minimum, this will include the Manual Limitorque Operated Valves (105), the Manual Non-Limitorque Operated Valves (135), other Manual Valves important to process flow control in Class 1 and steam generator heat removal applications (143), plant instrumentation required for surveillances, safety,related HVAC and the Control Room normal HVAC system.

9. Complete Preventive Maintenance (PMs) on manual valves selected due to their functional position, e.g., isolation of active equipment such as pumps, control valves, heat exchangers, l

cross-ties.

1 l

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10. Repair valves investigated as a cart of December 26, 1985 event troubleshooti ng. Includes FV-20527, P!-20528,7WS-063, FHS-064.

l l

l 4C.13.2 Actions to be Initiated as Promptly as Practicable, Schedule l

l Developed, Resources Assigned and Maintained Until Ccmpleted ,

O I JC-35 l

L

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l

1. Develop a departmental procedure hierarchy and writer's guide for Maintenance Procedures. .

2. Identify and prioritize procedures for generation.and/or revision.

~

3. Achieve authorized ~ staffing levels within the maintenance organizations.

4. Develop and/or revise the required programmatic procedures for the PM program to: as, sign responsibilities, authority and accountabilities for the program; establish criteria and define the scope of the program; and define the interface with other -

work control processes.

5. Review existing PM tasks and frequency for critical equipment.

l Revise and augment as required by programmatic selection 4

cr1teria.

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JC-39

4C.14 ONCE THROUGH STEAM GENERATCRS (OTSG's)

~

Since Rancho Seco is a Pressurized Water Reactor (PHR), tiie staff pays significant interest to the integrity of the OTSG tubes, whose condition can be inferred by analyzing secondary plant steam / water for traces of reactor coolant. Following the December 26 event, prolonged analysis was unable to conclusively establish whether or not a small leak existed within the OTSG's. As a result it was determined to take advantage of the outage to identify any degraded tubes and to take corrective action as required. Helium leak testing was done on both OTSG's which identified two tubes with very small leaks in the B-0TSG. Consequently, it was prudent to perform eddy-current inspection of the OTSG's to minimize the probability of a tube leak following restart. The program was comprehensive with approximately 5,000 and 7,000 tubes being inspected in the A and B OTSG's, respectively. Of these tubes, 12 in the A-0TSG and 52 in the B-0TSG were plugged. Since the majority of these tubes were

~

located in the " lane" regions of the OTSG's where previous programs had plugged tubes, B&W (OTSG. Manufacturer) recommended that the tubes in the " lane" regicn of both OTSG's be " sleeved" to minimize the probability of future leaks which would require down time and outages for plugging.

4C.14.1 Actions to be Comoleted Prior to Restart or Comoletion of the Restart Test Program

1. Ccmplete Helium Leak Test of cotn OTSG's.

JC-40

6

2. Perform Eddy-Current inspection of OTSG's to reccmmend tubes for plugging.

i

3. Plug those tubes identified in 1 and 2 above.

1

,1

4. Develop program and licensing documents necessary to sleeve tubes in lane region of OTSG's. '

l Install sleeves in selected tubes.

5.

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t 4C-41

4D SYSTEM REVIEW AND TESTING PROGRAM

\.s I Rancho Seco has instituted a Systems Engineer program modeled after INP0 Good Practice OP-209. Individual engineers are assigned specific systems for which they have the responsibility to know, among other things, the design basis, system limits and precautions, and to monitor the system condition. In addition, they aid other engineers, technicians, trainers and operators when interfacing with their assigned system. ,

I .

1 For the purposes of the Action Plan, the system engineer receives i all recommendations from the RRR8 which are relevant to their assigned system. The Systems Engineer considers the recommendations in light of their effect upon the system and presents an integrated system solution to the PAG. The System Engineer takes into .

consideration the system functional basis, maintenance trends and test records to assess the applicability of the issues in regards to the system performance history. From this review it is expected that further recommendations for system enhancements may be issued, i

including recommendations for specific testing. Testing will ,

l demonstrate that systems perform those functions important to the ,

safe operation of the station. Justification will be provided for those safety significant system functions not tested prior to  !

~

restart if further testing is not warranted.

40-1 1

, _ _ - . . _ - _ , _ . . . _ - - . . . _ - - - . - - . . ~ _ _ _ _ _ . _ _ _ . _ _ _ _._ . . _ _ . _ _ _ _ _ _ , , _ _ _ __ ._.._ ,_,_

Testing which is developed by this program is expected to demonstrate the material readiness of any system whose functional capability may be questioned. The testing will also provide an opportunity to refamiliarize the operating staff with system and plant operating conditions and procedures.

The District has sent qualified representatives to the Davis-Besse and Three Mile Island facilitates to review their restart testing programs. Lessons learned from these visits are being factored into

• the Rancho Seco System Review and Test Program (SRTP).

In addition, the results of the Davis-Besse SRTP are being reviewed for applicability to the Rancho Seco SRTP by the various system engineers. A comparison of the major features of the Rancho Seco I and Davis-Besse programs is included as section 40.7.

O 40.1 Purpose The Action Plan incorporates a System Review and Testing Program (SRTP) whose objective is to demonstrate that systems important to safety are ready to perform their required function when Rancho Seco l t

is ready to return to power. The questions this program was {

i

, developed to answer are.  ;

l l l

l' Which systems should be reviewed prior to restart? l i

I l l What system modifications and corrective maintenance are j t

required prior to restart? l O

40-2

What testing is required prior to restart to assure systems l

will perform functions important to safe plant operation?

i i

40.2 Oraanization k~ The SRTP Organization is shown in Figure 40-1.

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r 0

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n 40-3 ,

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1 REVISED 3 SEPTEMDER 86 l NUCLEAR '

PLANT PLANT REVIEW COMMITTEE l

^ .-

SYSTEM R$ VIEW TEST AND TEST ______ ggyggw PROGRAM DIRECTOR caour SYSTEM PROCESS TEST f At4D PROGRAM R EVIEW PROCRAM ,g DOCUMEt4TATIOtt A5515 tat 4T A5515 tat 4 T A55tSTAt4T l* DIRECTOR r.

otRECTOR o: RECTOR

?'- l g I I I 7'5".~"' "'".'.."'."' "l 'LOl.. 7."" ^"I '"m"""." .u-.

88""'*****

"surEnytsom

'UM" " .omnu ca sE a sEEn s$'e1m a sEUUn s 'U a SYSTEMS carsra, uooincAnoH SU R1ANCE SYSTEM SY3 TEM SYSTEM stAnt-ur SYSTEu srAnt-ur SCHMEKn EHCINEERS CHCINEERS Et4 Cit 4EEft3 [HCINE Ef t3 EtaCIHEElt3 F 40-1

40.3 Responsibilities

\s]

40.3.a System Review and Test Program Director The System Review and Test Program Director is responsible to the Plant Manager for the development, and implementation of the program to assure component, system and plant material conditions can support safe and reliable operation. In this capacity he works with the Nuclear Department Managers and the Scheduling Manager, and

• their designees, to ensure that the system reviews are thorough, -

test objectives are proper, and that the test procedures are developed, performed and evaluated in accordance with plant procedures. This position will work closely with the Restart Implementation Manager to develop and maintain a detailed schedule

( of work activities in support of the test program.

The primary purpose of this position is to structure a comprehensive system review and test program, including:

Defining the necessary organization and rescurces to accomplish the objectives.

Developing administrative controls for the system review and test program. As a minimum, these administrative controls will adoress responsibilities and authority, scope of system review, test procedure requirements such as review and approval cycle, content, format, rules for test conduct, and the evaluation and 40-5 i

reporting of results. Where practicable, existing procedures will be used, e.g., AP.2 for the development and review of procedures.

I Defining the test objectives for those components, systems and integrated systems to be tested. l Defining the test methodology to be used to accomplish the test  :

objectives. This includes a compilation of the procedures to j be used, and which procedure is being used to satisfy each test -

objective.

Preparation of test procedures.

Defining training requirements for test personnel.

f l

l Freparation of test results packages.

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l Preparation of schedules for procedure development and test  ;

i implementation in concert with the Implementation Manager.

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The System Revleu and Test Program Director is expected to provide the Plant Manager with weekly program status reports.

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40-6 l

i

40.3.b Test Review Group O The Test Review Group shall be established as a Test Subcommittee under Section VI.k of the Plant Review Committee Charter.

The Test Review Group shall contain the following members:

Nuclear Plant Department, Chairman

~

Nuclear Engineering Department Member -

Babcock & Wilcox Representative Member Bechtel Design Member Qual 1ty Assurance Department Member Training Department Member i

The System Review and Test Program Director will nominate members for the Test Review Group. The Plant Review Committee will confirm the members. The responsibilities of this Test Review Group include the following:

Review of the Test Identification section of the System Status Report to confirm that proposed testing will demonstrate all system functional requirements.

40-7

1 Review of all Test Specifications to assure that the scope and methodology demonstrate the system functions.

Review of related Special Test Procedures and new or revised Surveillance Procedures and recommend disposition to the Plant Review Committee.

Review of all restart related test results.

~*

The TRG will review all items in formal session and document -

their activities using meeting minutes and document review forms.

40.3.c Performance Analysis Group The responsibilities of the Performance Analysis Group (PAG) to the i 9

System Review and Test Program are encompassed by the duties I i

previously described in section 2.3.2. The PAG is responsible for  !

the review and approval of the system related modifications, and the maintenance and testing recommendations developed by the SRTP.

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l As described in subsequent portions of this section, the PAG will l review and approve: i i

The integrated program for resolution of system issues l

The system functional criteria t

The system restart testing program l

The system operational readiness determination l

40-8

40.4 System Selection for Systems Review In order to expedite the process of systems review, it was decided l

to identify those systems important to safety and begin a review in advance of completing the thorough and comprehensive investigative l

process of the Plant Performance and Management Improvement Program (PP&MIP) process.

i I

All systems at Rancho Seco are being investigated to some degree.

• These have been divided into two categories: Selected Systems - -

which comprises 27 identified systems, and Additional Systems -

which includes the 50 remaining systems. The results of the investigations will be documented in a System Status Report (SSR) for Selected Systems, and in a System Investigation Report (SIR) for  !

Additional Systems.

1 The criteria utilized to identify the Selected Systems by the SRTP  !

Director, which were then recommended to the PAG for treatment as i

" Selected Systems" as follows: i

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1. A history of significant or recurring problems. l

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2. The system was related, or contributed,.to the 12/26/85 event.

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3. The system is being significantly modified.

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4. The system has significant potential for initiating or g adversely affecting transients.

40-9

The criteria for selecting the Additional Systems are as follows:

O

1. The QCI-12 input phase has produced a recommendation for the system.

2. An open Work Request existed against the system as of 07/01/86.

The program allows for upgrading Addition Systems to Selected Systems. Based on the quantity and significance of issues raised by

' the PP&MIP process, additional systems may be upgraded by the PAG to Selected Systems status. f I

l' The selection mechanism for determining which systems require a ,

l systems review is the review process of the Plant Performance and  :

Management Improvement Program (PP&MIP) discussed in Section 4A.

This process identifies component and system problems based upon j extensive review of Rancho Seco performance history, condition, and design as well as industry experience. The system related output of l

the PP&MIP are fed through the RRRB to the appropriate System Engineers. On the basis of this evidence the System Engineer may i recommend Additional Systems for upgrading to Selected Systems -

l Status.

The criteria for upgrading systems is that significant problems have been identified or a large quantity of comparatively minor problems have been identified. This upgrading is initiated by the System O

40-10

4 Engineer making a recommendation in the System Investigation Report (SIR). This is submitted to the Performance Analysis Group (PAG) who makes the final decision on upgrading.

The Systems Review and Test Program Director has proposed, and the PAG has approved, the following lists of selected and additional systems. The current listing of selected systems is as follows:

1. Reactor Plant Reactor Coolant System 1

Decay Heat Removal System Seal Injection and Makeup System (including-High Pressure Injection)

Purification and Letdown System Nuclear Service Cooling Water System Nuclear Service Raw Water System f

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2. Steam Plant Steam Generator System Main Steam System Main Feedwater System Auxiliary Feedwater System

3. Power / Control Reactor Protection System (including ARTS) f Safety Features Actuation System EFIC (originally Main Steam Line Failure Logic) l 40-11 4

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Integrated Control System Non-Nuclear Instrumentation System 12S Volt DC Vital Power System 125 Volt DC Non-Vital Power System 120 Volt AC Vital Power System l 480 Volt AC System j 4160 Volt AC System 6900 Volt AC System i

~ 4. Auxiliary {

Control Room /TSC Essential HVAC System l Emergency Diesel Generator System Component Cooling Water Plant Air System Instrument Air System Auxiliary Steam System *  ;

• Upgraded by PAG direction on 9/11/86 The current listing of Additional Systems is as follows:

1. Reactor Plant Borated Hater System Core Flood System Containment Building Spray System Control Rod Drive System

• Pressurizer Relief Tank System Spent Fuel Cooling System i

40-12

Reactor Coolant Drain System

2. Steam Plant Air Ejector System Extraction Steam System Gland Steam Condenser System i Heater Drains and Vents System ,

High Pressure Turbine System Low Pressure Turbine System ,

Main Condensate and Makeup System l -

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3. Power and Control

{

Annunciator System Cathodic Protection System l Plant Communications System l

l Electro-hydraulle Control System Main Generator System i i Nuclear Instrumentation System l

Plant Computer System Plant Security System Radiation Monitoring System l l

Seismic Monitoring System i I'

Turbine Supervisory Instruments

4. Auxillary Adi!!!iffl%tiit/Sfit46 l Orainage and Sewage System l

l Cranes and Holsts

40-13 l

Carbon Dioxide System Diesel Fuel Oil System Demineralized Water System Fuel Handling System Fire Protection Water System Hydrogen Gas System HVAC - Other Lube Oil System Main Circulating Water System ,

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Nitrogen Gas System l

NSEB Essential HVAC Plant Buildings and Structures Plant Cooling Water System-Reactor Sampling System Radwaste System Secondary Chemical Addition System Seal Oil System Site Reservoir System .

Service Water System Turbine Plant Cooling Water System i

Turbine Plant Sampling System 1

Waste Gas System l

l 40-14

40.S System Review Report Summary n%j

' The review of the Selected and Additional systems under the System Review and Test Program will be documented in individual system reports. A System Status Report (SSR) will-be prepared for each Selected System and a System Investigation Report will document each Additional System Review. The process for the development, review

~

and approval of these reports is diagrammed in Figure 40.2.

40.5.a ' System Status Report Overview .

t The SSR for Selected Systems will be developed in three stages, with each stage documented in a separate revision (Rev. O, 1, and 2).  ;

1 '

The purpose of the Rev. O report is to initiate plant design and ,

C) modification work. The contents of the Rev. O report is an Executive Summary, a basic System Functional Description, and a  ;

listing of Problem Statements developed from the PP&MIP process and from the results of a review open Work requests. This Rev. O report is reviewed and approved by the Performance Analysis Group and the Deputy General Manager, Nuclear. Once it is approved, it is used by '

the system engineer to initiate design activities, maintenance l activities, and plant modifications through conduct of a " kickoff 3 meeting".

The Rev. 1 SSR is utilized to identify the testing necessary for each Selected System. It builds on the Rev. O SSR, providing a more detailed System Functional Description, additional problem O- statements frem a review of open ECN's, open NCR's, and outstanding [

i 40-15 g

I abnormal tags, and the identification of testing required to demonstrate functions important to safe plant operation to be conducted prior to restart. Rev. 1 is reviewed and approved by the Test Review Group, Performance Analysis Group, and Deputy General ,

1 Manager, Nuclear. Once approved, the preparation and implementation l of test specifications and procedures will begin.

A Rev. 2 SSR is prepared for each Selected System and is utilized for final system acceptance. This report contains everything from

'Rev. I plus additional problem statements documenting the results of the system walkdown, review of maintenance history trend investigation, and a review of the Davis-Besse SRTP results. This revision will also contain a summary of results of tests performed ,

to date and an operability statement by the System Engineer. This j report is reviewed and approved by the Test Review Group, Performance Analysis Group, and Deputy General Manager, Nuclear  ;

i prior to restart.

40.5.b System Investigation Report Overview i

System Investigation Reports (SIR) will be developed for all Additional Systems in two stages. The Rev. O SIR for each system will be utilized to get plant work started and to determine whether the system should be upgraded to Selected System status. The contents of the Rev. O report are an Executive Summary, a System Functional Oescription, and a listing of Problem Statements developed frcm the PP&MIP process and from the results of a review of open Work Requests, open NCR's, outstanding abnormal tags and 40-16

open ECN's. The report also contains a justification for upgrading '

the system to Select Status or for~ maintaining the Additional System status. This report is reviewed by the PAG and by the Deputy General Manager, Nuclear.

Rev. 1 of the SIR for each system is utilized for final system acceptance and for consideration for upgrading to Selected System status. The contents of the Rev. I status report contains the Rev.

O SIR plus any additional Problem Statements and an Operability  ;

• Finding. This report is approved by the Performance Analysis Group -

and the Deputy General Manager, Nuclear.

40.5.c Details of Report Components g The following provides additional details of the major components of the SSR's and SIR's. I i

i 40.5.c.1 System Functional Description i t

The System Functional description is a listing of the capabilities

that the system must provide in order to assure reliable operation l and effective accident mitigation. The functions determined to be i

important to safe plant operation will be identified for evaluation

! in the test identification process of the SSR.

40-17

The System Engineer will prepare the System Functional Description.

Source documents to be used are:

USAR Technical Specifications Emergency Operating Procedures B&W Guide Specifications Bechtel Rancho Seco Design Manual NEP 5400 (System Design Bases)

Vendor manuals Applicable design calculations.

Comparison will be made to the Davis-Besse function determination. l 40.5.c.2 Resolution of System Problems l

i The System Engineer will document all problems with the system via the Problem Statements. These problems are identified via the PP&MIP program (RRRB recommendations) and by several other processes ,

described below. Resolutions are proposed for each problem. l Priority is indicated as Restart, near term or long term. I The System Engineer considers the ccmbined effect of the resolution ;

on the system and determine if all known systen deficiencies are f corrected. He will also satisfy himself that no new issues are created.

O 40 18

In addition to problems identified through the PP&MIP the following processes will also provide input to the system engineer. Problems identified through these activities will be documented Problem Statements.

• Review of Open Work Requests Open Work Requests will be reviewed and those which should be completed prior to restart will be identified. Those Work Requests which may be deferred until after restart will also be

~

identified on a Problem Statement and prioritized accordingly. -

Review of Open Engineering Change Notices (ECN)

Open ECW s will be reviewed and be classified as requiring implementation either pre or post restart.

• Review of Open NCR's Open NCR's will be evaluated to identify those requiring disposition prior to restart.

• Review of Outstanding Abnormal Tags Abnormal Tags, used for temporary plant modifications, will be reviewed to identify those that should become permanent modifications and those which should be cleared prior to restart.

Review of Maintenance History Trends The System Engineer will review trends of maintenance history for indications of system deficiencies.

40-19

Review of Findings from Davis-Besse The System Engineer will review the system / component problems identified at Davis-Besse as documented in the Davis-Besse System Review and Test program for applicability at Rancho Seco.

i Material Condition Walkdown The Systems Engineer, System Design Engineer and a designee of the Operations Manager, will inspect the system for evidence of a deterioration in material condition. A checklist will be provided as a supplement to QCI-12 to assure consistency in the conduct of these walkdowns. A confirmation that IE Bulletin 79-14 walkdowns were adequately performed on safety related systems will also be a part of this process. Problems found during the walkdown will be reported in the System Status Report and resolved using normal corrective action processes such as Work Requests and Nonconformance Reports.

40.5.c.3 Test Identification The System Engineer will review the testing that is being performed on the system. A comparison will be made to determine if all of the functions important to safe plant operation defined in the System Functional Description are being demonstrated by current testing.

If the functional testing is fourd to be inadequate or incomplete, appropriate additional testing will be specified.

'O 40-20

The results of the test identification will be documented in the N

System Status Report Rev. I and will specify the scope of restart testing to be conducted. Tests used to demonstrate system function will be ' included or referenced.  !

I r Justification will be provided for those functions or portions not  ;

i to be tested prior to restart.

+

I 40.5.c.4 Operability Finding .

Based on the identifled testing program and the recommendations and  ;

I dispositions for system problems the Engineer will specify any l additional maintenance, modifications, procedure changes, and l testing that is needed to assure reliable system operation. He will Indicate how these open items are being tracked to completion.

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+

40-21 l

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4D.6 Restart Test Program Testin) to be performed prior to restart will be identified by the O

SRTP an'J by other normal testing programs. The test program identified in the SSR to demonstrate functions important to safe plant operation will integrate testing due to modification, some post maintenance testing, surveillance testing and special functional testing as required.

'All newly identified testing to support the SRTP will be conducted -

as Special Test Procedures (STP) by Test Engineers.

The System Review and Test Program Director will develop a training program for test engineers. This program will cover the generation of Test Specifications, the writing of procedures, the revision of procedures (both permanent and temporary changes), test conduct and the review of results.

i Although the majority of the restart test program will be defined by the ongoing SRTP process, several major periodic and special testing +

l activities have already been identified and include the following: l l

.1 Loss of ICS/NNI Test

.2 Loss of Instrument Air Test - Both a sudden and a gradual loss of instrument air (similar to reccmmendation in Reg. Guide '

l.68.3) will be simulated with the plant en steam generator O

40-22

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cooling (subject to the resolution of any unreviewed safety i

questions).

.3 Integrated Leak Rate Test of Reactor Building.

.4 Complete the Balance of Ten-year In-service Inspection (except for those inspections requiring removal of the reactor vessel I

head). This includes required periodic pump and valve testing.  ;

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.5 Integrated Engineered Safeguards Actuation Test i l

.6 Emergency Diesel Generator Biennial Inspections \

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! 40-23 4

SYSTEM REVIEWS AND REPORTJNG t m ir >

QCI I2 Input Process i f Valldit y Priorit y Data Base RRR8 Sort I

Systems Related Management issues

! Additional Systems h Selected Systems t t Rev O-SIR M & SSR-Rev 0 t t

. ,',',nie"', '":,', xIca-.1e u..tI -

1 o

! b h h

! 4 Rev I-SIR M H SSR - Rev 1 H L Test Spec = ; TRG PAG

+ Test

_ Procedures n =

+

Test Results 2 :

i f H SSR - Rev 2 t

i

= Restart Decision l

I

Figure 4D-2

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4D.7 Comparison Between Rancho Seco and Davis-Besse Programs This chart is a comparison between the Rancho Seco and Davis-Besse System Review and Test (SR&T) program. The significant difference between these programs is that Rancho Seco will do a partial system ,

i review for all remaining systems and consider them for possible '

upgrading to Selected Systems if they meet the criteria as established.

SYSTEM REVIEW AND TEST PROGRAM $ -

i RANCHO SECO/ DAVIS-BESSE COMPARISON l t

.1 System Selection y Rancho Seco Davis-Besse Selection criteria includes

• Selection criteria includes non-safety systems non-safety systems 27 systems initially selected

• 31 systems initially selected Process for inclusion of ad-

• During course of program ditional systems estabitsbed - 3 additional systems in-

- 4 probable system addition cluded in~ program identified to date Partial system reviews to be

• Systems outside program performed on all remaining deferred for subsequent i

systems with any identified review.

problem or outstanding Maintenance Work Orders. ,

4 40-25

i This chart provides a comparison between the Rancho Seco/ Davis-Besse problem identification process.

SYSTEM REVIEW AND TEST PROGRAM >

RANCHO SEC0/0 AVIS-BESSE COMPARISON f i

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.2 Problem Identification Rancho Seco Davis-Besse System based problem identification System based problem identification t

- Review of outstanding work orders - Review of outstanding work orders !

- Review of open ECN's - Review of open FCR's (ECN's)  !

- Review of maintenance history trends - Review of maintenance history trends

- System walkdown for material - System walkdown for material condition condition

- Input from day-to-day system - Input from day-to-day system responsibility responsibility

- Review of test results of comparable Davis-Besse system

- Review of Licensee Even Reports

- Review of significant DCROR - Review of significant DCROR Human Engineering Discrepancy Human Engineering 01screpancy Reports Reports

- Review of Transient Assessment Reports

- Input from Operations and Maintenance personnel i

40-26

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t 4

Rancho Seco (cont'd) 4 -

Programmatic based problem identification

! includes:

1

, - Precursor review i

- Plant staff Interview 1

- Deterministic failure consequences i - B&WOG Safety and Performance Improvement Program i

- Other PP & HIP activities l

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1 -

i t

40-27

The following chart provides a comparison between the Rancho Seco/ Davis-Sesee system review process.

SYSTEM REVIEW AND TEST PROGRAM RANCHO SEC0/ DAVIS-BESSE COMPARISON

.3 System Review  !

i Rancho Seco Davis-Besse System level functions to be System level functions to be  ;

identified for each system identified for each system Problem identification both Problem identification system and programmatically system based l based Corrective actions developed

• Problem prioritization criteria and recommended by System by System Engineers Engineers and PP & MIP Problem report review and

• Problem report review and i corrective action concurrence corrective action concurrence l by PAG by IPRC i O

40-28

I This chart provides a comparison between the Rancho Seco/ Davis-Besee test program development.

SYSTEM REVIEW AND TEST PROGRAM RANCHO SEC0/0 AVIS-BESSE COMPARISON

. .4 Test Program Development Rancho Seco Davis-Besse i . .

Identify new test requirements

• Identify new test requirements for functions not adequately for functions not adequately tested tested Identify retest requirements

• Identify retest requirements for functions previously tested for functions previously tested or provide justification for or provide justification for why retest not required why retest not required i

Identify new test requirements

• Identify new test requirements to support plant modifications to support plant modifications Review and approval of test

• Review and approval of test l

requirements by TRG and PAG requirements by IPRC 1

E 4

4 I

40-29

This chart provides a comparison between the Rancho Seco/ Davis-Besse restart testing program.

I SYSTEM REVIEW AND TEST PROGRAM r.ANCHO SECO/ DAVIS-BESSE COMPARISON

.5 Restart Testing Rancho Seco Davis-Besse Test specifications and test

• Test specifications and test procedures to be prepared procedures to be prepared under existing procedural under existing procedural controls controls Test specifications and test O

Test specifications and test procedures to be reviewed procedures to be reviewed by TRG by JTG Test leaders to be trained Test leaders to be trained and quallfled on station and qualified on station testing program requirements testing program requirements All test results required to All test results required to verify function to be reviewed verify function to be reviewed by TRG by JTG & IPRC O

40-30

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APPENDIX A i OISTRICT BOARD OF OIRECTORS' l POLICY STATEMENT ON PERFORMANCE t

IMPROVEMENT AT RANCHO SEC0 l e l

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APPENDIX A DISTRICT BOARD OF DIRECTORS' POLICY STATEMENT ON PERFORMANCE IMPROVEMENT AT RANCHO SECO I. INTRODUCTION On July 3,1986, the Board of 01 rectors of Sacramento Municipal Utility Olstrict voted unanimously to adopt the policy and the associated performance improvement goals pr!!sented below. This policy and related planning guidance are intended to provide direction to all persons who

,Q may become involved in this effort.

LI The policy and associated implementation guidance stated herein are appropriate in view of the increased internal and external emphasis being placed on plant performance improvement. The emphasis on performance improvement has special significance for Rancho Seco for two reasons; it is a B&W plant which is perceived by the NRC to be more sensitive to upset conditions than other PHR's, and among the B&W plants, it has the poorest overall performance record.

To put the Performance Improvement Policy Statement below in proper perspective, it is appropriate to restate the 01 strict's primary mission. That primary mission is to generate electricity safely, reliably, economically. Carrytr.g out this mission in a responsible e)

A-1

manner involves a continuous emphasis on safety. It includes strict adherence to the SMUD Quality Program. It also includes taking the necessary short-term actions to achieve and maintain a desired balance over the long-term. In carrying out this primary mission, the Olstrict is accountable to many parties. They are accountable to:

their customers-owners, their bond holders, the general public, their employees, the Nuclear Regulatory Commission, and other nuclear and non-nuclear regulatory agencies.

The management challenge in establishing an optimum balance and Teeting

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each of these responsibilities in an appropriate manner 12 significant.

This Policy Statement is intended to aid in carrying out the actions required to significantly improve the performance of the Rancho Seco Nuclear Generating Station in a manner which is consistent with the Olstrict's primary mission.

II. POLICY STATEMENT The District's Board of Directors is committed to acH eving a prcmpt and significant improvement in performance at Rancho Seco and to provide the support necessary to achieve standards of excellence in all aspects of nuclear activities. To reinforce that commitment, the Board has estaolished the following ocjectives related to performance improvement at Rancho Seco.

A-Z

l (N Near-Term District Objectives Accomplish those actions which will substantially reduce the likelihood of anotner significant transient at Rancho Seco, i

- Initiate longer term actions which will contribute to a sustained improvement in plant performance e.t Rancho Secc.

Long-Term District Objectives

- Accomplish those actions whic;1 will allow the District to achieve the 1990 Performance Improvement Goals for Rancho Seco.

3 I Performance Improvement Program Goals

,J The 1990 goals which have been committed to INPO include the following items:

Improved Equipment / Plant Availability Reduced Forced Outage Rate Reduced Reactor Scram Rate Reduced Unplanned Safety System Actuations Reduced Safety System Unavailability Improved Thermal Performance Reduced Low Level Waste Volume Reduced Personnel Exposures Reduced Industrial Accident Rate A-3

Iil. POLICY IMPLEMENTATION AND PLANNING GUIDANCE The Board considers the Performance Improvement Program at Rancho Seco to be the District's highest priority activity. The Board intends to closely monitor progress toward the Performance Improvement Program goals.

The near-term objectives shall be satisfied prior to the restart of P,ancho Seco in accordance with the approved Performance Imcrcvement Program. The long-term objectives shall be pursued in accordance with the approved Performance Improvement Plan.

District management shall develop, maintain and follow a detailed Program Plan which encompasses all projects related to the Performance Improvement Program.

The Board of Directors encourages District participation in the activities of industry groups wnere sharing of information and costs can be beneficial to the District. This is especially acorcpriate regarding participation in the B&W Owners Group activities which involve interactions with other utilities with plants of similar design.

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4a A me-& A ---. .-- _-.d_ 4 h_A-----hha-- -6Ae 2 m.--4 '

.M e_J a ---ah- - - -A..a- Am- -a4 - -w--- - . _..a_. --- . - - _ --

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APPENDIX 8 i  :

l SPECIFIC DISTRICT RESPONSES TO  ;

NUREG-1195 FINDINGS I

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~-~ev-e-,-w._,__

APPENDIX B i

i SPECIFIC OISTRICT RESPONSES TO i NUREG-1195 FINDINGS j The following are the District's responses to the findings and i

conclusions section of NUREG-Il95.

4 i

i FINDING - SECTION 8.1: ICS/NNI Power Supply Monitor i

i l 1. The December 26, 1985 overcooling transient was initiated by the l

l power supply monitor in the nonsafety-related ICS (tripping the 4

+/- 24 Vdc power). The most probable cause of the tripping was a design weakness that apparently made the circuit susceptible to erratic operation if " contact resistance" between the 24 Vdc l bus and the power supply monitor were to develop, and the development of a blgh resistance connection (i.e., a bad crimp connection) in the wiring between the + 24 Vdc bus and the power supply monitor which exposed the design weakness and caused the module to trip. (SMUD has agreed to further explore the cause of the failure of the power supply monttor by having an l Independent laboratory conduct additional analyses).

I 0! STRICT RESPONSE l

Specific plant modifications were engineered and installed to. address the identified lessons learned. Engineering Change Notices and

\j subsequent fleid work accomplished the following:

i B-1 l

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a. leads to the power supply monitor now go directly to the power supply bus on the ICS and NNI.

b. Inspection and correction of terminations (i.e., lugs) was completed.

c. ICS Power Supply Cabinet Bus wiring was replaced.

d. the original power supply monitor, and the S1/52 switches, have

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been sent to an independent laboratory for analysts. -

e. A new power supply monitor has been calibrated and installed.

The District voluntarily undertook a program to test and inspect the terminations throughout the ICS, NNI, SFAS, and RPS cabinets, supplied by Bailey Meter Company.

Engineering is investigating ways to improve the reliability of the power supplies for the ICS/NNI. Consideration is being given to redundant and/or independent power supply monitors. Pending completion of these investigations, no decisions have been made as to the nature or schedule implementation of proposed enhancements.

l l

Section 4C.1 describes the ongoing investigations and activities which will provide the basis for closure of this Finding.

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i B-2

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FINDING - B.2: Repositioning of ICS Controlled Valves or Loss of ICS

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2. Upon loss of ICS de power and the subsequent automatic repositioning of a number of valves in the plant, the design of the ICS also caused the loss of remote control of the affected

^

t valves from the control room which necessitates manual actions locally at the valves.

! DISTRICT RESPONSE Design changes have been completed, or are underway, to change the operation of those valves important to mitigating the effects of !oss of ICS Power. The Turbine Bypass and Atmospheric Dump valves now i

remain closed upon loss of ICS power, while control is passed to devices which are powered independently of the ICS. These valves can l

now be opened / closed by the operator from the control room.

I As described in Section 4C.1, the AFW flow control valves will be automatically controlled to maintain correct steam generator level on loss of'ICS Power. The operator will be able to take manual control from within the Control Room. Repairs and maintenance on these 2

valves are described in Section 4C.13.

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8-3 i

FINDING - 8.3: PM Program for Manual Valves An AFW manual isolation valve could not be shut by the operators

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3.

after the failure of the auxiliary feedwater (AFW) (ICS) flow control valve. The failure of the AFW manual isolation valve was the result of a lack of any maintenance on this valve during the operational life of the plant. The lack of a maintenance program resulted in the valve being inadequately lubricated, which caused the valve to seize. It appears that the lack of a

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maintenance program could affect the operability of other manual valves at Rancho Seco.

DISTRICT RESPONSE Troubleshooting identified a lack of lubrication and rusted yoke nut bearings on Auxiliary Feedwater Isolation Valve FWS-063. Reworking these components restored the valve to an operable condition. As an element of the troubleshooting effort, the similar valve on the OTSG-B line (FWS-064) was inspected, as were all similar valves in service on the AFW system. All were found serviceable with only normal closing torque required to operate through their full travel, although evidence of recent lubrication was missing. These valves l are maintenance isolation valves and are required to be " locked" in position during power operation.

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B-4

In recognition of the desirability of having certain manual valves _

readily operable, the Nuclear Operations Manager has identified 143 valves which will be verified operable prior to resumption of power operation. These manual isolation valves are characterized by their purpose which is to allow isolation of important active equipment such as pumps, valves, and heat exchangers. They were selected to include both primary and secondary plant systems necessary to power production or nuclear safety. Function, not class, was the selection criteria. The program involved actual stroking of the valves and, "where necessary, servicing with lubricants, packing, or adjustments -

was done and documented. Statir. tics are being collected and evaluated to establish a summary status of valves in similar service. A complete listire of actions to be completed to improve and access the material condition of the facility is included in Section 48.5. Section 4C.13 describes the relevant specific maintenance to be accomplished.

1 Significant changes are underway with respect to the Preventive Maintenance Program at Rancho Seco. Staff is being added for the specific purpose of expanding the scope, content, and quality of the preventive maintenance program. Specific procedures-detailing the l

PMs are being provided or expanded to give confidence in the i

condition and operability of the PM'd equipment. This expanded PM

! program includes the above identified valves, as a first phase in the expansion, in addition to those already receiving periodic maintenance.

O B-5

FINDING - SECTION B.4: Procedure Guidance for Loss of ICS O

4. Rancho Seco Emergency Operating Procedures (EOP) do not address the loss of ICS power. The lack of specific guidance seems to be a weakness in the plant-specific E0Ps available to the operators on December 26, 1985. The Rancho Seco Anticipated Transient Operating Guidelines (ATOG) supplied by the B&W Owners Group include an explicit procedure for a loss of ICS power and the ATOG directs operators to that procedure. However, this

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procedure was not included in the Rancho Seco E0Ps.

DISTRICT RESPONSE During the December 26 event those actions necessary to respond to the consequences of the Loss of ICS power were appropriately defined within the E0P's. The operating philosophy is to place the plant in a stable post-trip condition, and then begin the cause-of-event trouble shooting, e.g. ICS power restoration in this case. In this event, due to the difficulty in closing the AFW valves and reluctance to trip the AFH pumps, restoration of ICS power would have mitigated the event earlier. While a Loss of ICS Casualty Procedure would have been helpful in expediting the power restoration, that function 1

correctly should not be a part of the E0P's. The E0P's must address the condition when ICS Power cannot be restored, for whatever reason, and a specific Casualty Procedure has been implemented for this requirement.

O B-6

l A review of the ATOG determined that there were no other needs or _

requirements which were not incorporated in the E0P's.

Training, described in Section 48.3.2, has been, or will be, given to the appropriate plant staff on the lessons-learned and changes made revelative to the E0P's and plant operations.

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l B-7

FINDING - B.5: Feedpump Trio Criteria

5. The E0Ps at Rancho Seco direct the operators to trip the O

appropriate feed pumps to terminate flow if the feedwater flow cannot be isolated. This was not done during the December 26, 1985 incident. The operators were reluctant to stop the AFH pumps even when they had difficulty stopping flow to the once-through steam generators (OTSG) by valve operation. The operators had decided that they would stop the AFW pumps only if water started to flow into the main steam lines. However, the operators failed to adequately monitor OTSG water level and, as a result, water was introduced into the steam lines. Their reluctance appears to be the result of the substantial emphasis placed on the .GW system by NRC and others, and a lack of confidence in the reliability of the AFH pumps (i.e., fear that the pumps sould not restart if stopped).

DISTRICT RESPONSE The E0Ps did not contain specific parametric criteria such as RCS Temperature, Steam Generator Level, or Pressurizer Level for when to trip main and auxiliary feedwater pumps during an overcooling. Lack of specific criteria let the operators be influenced by perceptions of NRC concerns regarding auxiliary feedwater operability.

Procedures have been revised to specify when to trip main, condensate and auxiliary feedwater pumps. This is a significant improvement within the E0P's as it removes the obstruction presented to the O

B-8

operator, and replaces it with a preplanned response to the observed _

conditions. A detailed review of the starting and operating reliability of the Auxiliary Feedwater pumps was done which did not support a lack of confidence in this equipment by the operators. In the twelve year operating history (315 attempted starts) three instances of failure to start these pumps were noted.

In the first case, in early 1975, P-318 tripped on over current during a Surveillance Procedure Test start. All class I motor

'overcurrent settings were checked as a result. The second case -

occurred on P-318 in 1980 when it failed to start during a surveillance test. This could not be duplicated. The third occurrence was a failure to start P-318 on its turbine drive, also for a surveillance test. P-319 has always started. There has never been a time when there has been loss of function of the Auxiliary Feedwater Pumps.

. Training necessary to implement the above described changes to the l

! E0P's is described in Section 48.3.2.

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FINDING - B.6: Priority of PTS or Pressurizer Level

6. The operators had considerable difficulty reconciling the dichotomy between avoiding the pressurized thermal shock (PTS) region (e.g, reducing high pressure injection (HPI) flow] and regaining pressurizer level (e.g., increasing HPI flow in accordance with their E0Ps). Their training and procedures were not adequate to resolve this conflict and to some extent tended to provide conflicting indication 5 of the appropriate priorities.

DISTRICT RESPONSE The Emergency Operating Procedure (EOP) in place on December 26 included rules which specifically state the events or plant conditions which mandate throttling of HPI. flows. Under Section 2.2 it is clear that HPI flow should be throttled to prevent exceeding PTS limitations.

The training programs for all License' Training have been examined.

The conclusion being that HPI throttling, even with no pressurizer level, is adequately addressed through the following:

a. Both initial License Training and Senior License Training Programs address E0P rules and provide background information.

b. Many drill scenarios in the Simulator Training Program contain the application and use of rule 2.

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c. All E0P rules are required to be committed to memory by all control room operators and are tested during NRC License and Rancho Seco Requalification Examinations.

2 The " conflicting priorities" concern has been further addressed by the training given to the operators since the event, which emphasizes the purpose and requirements of the E0P rules and the hierarchy of impl,ementation. At the same time, the E0P's have been reviewed and enhanced to provide clear direction to the operator when faced with tomplex events.

Although improvements in the ATOG derived E0P's have been made as a result of this event, a comprehensive operational assessment of the f

l E0P's demonstrated that they are an adequate and viable way of responding to the spectrum of transient events. For events such as this overcooling, they were adequate for responding to PTS concerns.

The improvements have been cirected toward precluding conditions which precipitate PTS issues.

Trainingtoemphasizethese;issueshasbeengiventotheoperators andisdescribedinSection)48.3.2. A comprehensive report l

describingtheRestartTralijingProgram,.datedSeptember1986,has beenpreparedandisavailad,leforreview. In response to question 19(a)(1/2) in a September 5, 1986 NRC letter, the following is given:

e

% 1. Describe how HPI vs. PTS concerns are directly handled in the 1

o E0?s.

k B-ll

DISTRICT RESPONSE O

Emergency Operating Procedure Rule 2 "HPI Flow Control" has been

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revised to state explicitly that HPI should be throttled to prevent entering the PTS restricted region. Combined with the previously existing statement in Rule 2, that HPI can be throttled any time reactor subcooling margin is restored, there is no ambiguity regarding HPI vs. PTS concerns. Procedure steps which call for HPI initiation refer to this rule.

2. Why were the PTS guidelines exceeded during the December 26 event?

DISTRICT RESPONSE O

Operators observed the impending PTS violation but did not take prompt action to minimize excursions into the PTS restricted region.

Several factors contributed to exceeding the PTS guidelines:

a. E0Ps did not provide explicit.enough instructions to assure rapid termination of the cooldown, such as a RCS minimum temperature criteria for requiring feed pump trip.

b. E0Ps did not explicitly require cooldown termination from the control room, thus inplant actions were initiated to control the cooldown. Coordination of inplant efforts required attention of control room operators which detracted from timely recognition of impending PTS criteria violation.

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c. E0Ps did not explicitly contain direction to gain plant control l O prior to initiating efforts to regain ICS power. Control room personnel efforts to restore ICS power detracted from timely recognition and response to the impending PTS criteria violation.

d. The subcooling margin stopped increasing for a few minutes, prior to rapidly increasing into the PTS restricted region.

Operator observation of subcooling during this stable period could have lead the operators to conclude the situation was

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stablizing and allowed them to momentarily direct their -

attention to inplant efforts or ICS power restoration.

Revisions to the E0Ps addressing the above factors have been made and appropriate training conducted. Numerous simulator exercises with all crews have shown that these changes are effective in avoiding PTS.

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FINDING - B.7: Training on loss / Restoration of ICS Power O

7. The operators received neither classroom nor simulator training on the overall plant response to either the total loss of ICS de power or the restoration of ICS de power.

DISTRICT RESPONSE l

Training' programs have been revised, to address the lessons learned Trom the December 26 event, and expanded to include the plant l

l modifications and procedure changes which have occurred. See Section 48.3.2 for details. Operator simulator training time has been increased by 607. for this year and the operators are being scheduled for two weeks of simulator training in 1987. This is twice as much as was previously scheduled.

i The post-event training on the B&W simulator included the following items:

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a. Emergency Operating Procedures (EOP) Training including all steps necessary to terminate overcooling, or OTSG overfill from any cause, including the loss of ICS power.

l

b. Effect of changes to ADV, TBV, and AFW valve operation following l

l loss of ICS power.

c. Command and control training, including implementation of the Emergency Plan.

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d. Recovery from SFAS, i.e., restoring normal makeup and letdown flow.

e. Differences between the B&W Simulator and the facility (Operator.

traps, details of these differences follow).

f. HPI and AFW throttling and pump trip criteria.

g. PTS recovery actions.

h. Cooldown rate interpretation and tracking.

1. Transition from AFW to MFW flow.

N The District is in the final stages of procuring its own plant specific simulator. Not only will the installation of a simulator ?t Rancho Seco afford additional crew training, but the simulator will also incorporate the Control Room Design Review Human Factors-modifications.

The effects of Restoration of ICS Power are under investigation by l

l the B&W Owners Group. Modifications to the Turbine Bypass Valves, TBV's, means that following loss of ICS power, control will automatically transfer to independent controls. Procedures cause these to be in manual mode when attempting to repower the ICS. Since these controls are independent of the ICS, whatever demands the ICS issues will have no effect. Once stable ICS conditions are observed, B-15

the operator can return the TBV's to ICS control. Installation of I EFIC will provide ADV and AFH control independent of ICS.

j "0PERATOR TRAPS" - DIFFERENCES BETHEEN THE B&W SIMULATOR AND RANCHO SECO

1. MUT outlet valve on 'B' SFAS panel at the simulator has a control on each, with the 'A' panel normally powered up due to ,

1 the power source selection for this valve. l

2. HPI/MUP recirc valves have no controls on the SFAS panels at the simulator. These valves are modeled to auto-close on an SFAS initiation. To re-open, the simulator instructor must do them at the simulator control panel.

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3. The simulator doesn't model separate controls for the HPI/MUP lube oli aux 111 aries on HlRC. Operators must understand that procedures require these auxiliaries to first be started when starting an HPI/MUP.

4. The simulator allows the MUP to draw essentially all its suction from the BHST when SFV-25003 and the MUT outlet are both open.

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At the plant, the MUT pressure and level determine whether the MVP will draw water from the MUT or BHST or both. There is a check valve on the MUT outlet at the plant, but not at the simulator.

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5. Response of RCS (pressure and level) to changes in secondary pressure (e.g. cycling.TBV's during a loss of ICS power situation) at the simulator is not as great as what would be seen at the plant. Operators should expect a greater effect on the primary side if cycling steam valves at the plant.

Maintaining a narrower secondary pressure band is one way of minimizing this greater response at the plant.

6. The E0P's direct the operator to isolate the OTSG's (one or both) under certain circumstances and provide a list of valves

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5 that need to be closed to accomplish the isolation. The following valves are not modeled at the simulator, but must be i

closed at the plant in order to isolate an OTSG:

- Matti steam to AFH pump Main steam to reheaters

- Main steam to pegging steam ('A' OTSG only)

OTSG blowdown l l

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7. The AFH SFAS valve controls at the simulator are located on the operators console rather than the SFAS panels as at the plant.

l This places the burden of operation for these valves on the.

console operator who is normally not responsible for operating l

these valves.

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8. For HPI cooling, the E0P's direct the operator to initiate SFAS channels IA & 18 with their respective pushbuttons. To get the same equipment response at the simulator, the channel 1/2 and 5/6 initiation pushbuttons must be used.

9. If the operator has to open the 38 and 3C2 breakers to drop the rods at the simulator, the CCH pumps, air compressors and PCW pumps have to be restarted after the breakers have been i

reclosed. i -

Also the breakers are located differently on the electric control panel than at the plant.

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10. Certain steps within the E0P's direct the operators to re-open i the seal return isolation valves following their automatic closure on an SFAS. At the simulator these valves do not auto-close on SFAS.

l 11. The simulator " Loss of ICS Power" transient differs from that at the plant in the following ways:

a. There are no auto-close switches for the TBV's and ADV's on the operator's console. The " circuitry" for these switches is modeled, however, the simulator instructor must actuate from the simulator control console.

i B-18

b. There are no AFH valve controllers on H2PS. The AFH Bailey

) valves go to their designed pre-set position, however, manual control remains on the operators console Bailey stations.

c. The power history associated with the transient is relatively high (may be design decay heat level). Thus the transient is essentially an undercooling transient until the AFH valves are opened farther.

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d. There is no "ICS Power / Fan failure alarm" at the simulator.

e.' The reducer control' failure for the auxiliary steam system is not modeled at the simulator. At the plant, this will fall to the 50% position. The auxiliary steam HOV's on i HISS are also not modeled at the simulator.

'1 B-19

FINDING - B.8: Recognition of ICS Power Condition

8. The operators who investigated the loss of ICS power did not adequately understand the ICS power system configuration. When 120 Vac power is still available from the IC bus and the ICS dc power supplies de-energized, the most credible cause for the loss of ICS dc power was the opening of switches Si and S2.

However, the operators did not recognize this fact and, as a result, did not shut the switches until 26 minutes into the transient. The fact that several operators did not recognize

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that switches Si and S2 were OFF suggests that their training on this crucial system was not adequate. In addition, although simplified drawings of the non-nuclear instrumentation (NNI) power supplies were posted on the NNI cabinets, comparable drawings for the ICS power supply had not been provided.

DISTRICT RESPONSES Since the event Training has been conducted on the design and operation of the ICS/NNI power supplies. See Section 48.3.2. In addition, S1 and S2 labeling has been engineered and installed at the switch location. A one line ICS power supply diagram has been posted on the cabinet door to aid the operator in troubleshooting power supply problems. A specific causality procedure for the ICS has been developed and is referenced by the Emergency Operations Procedures for use once the operator has established a stable plant condition, following Loss of ICS Power.

O B-20

FINDING - B.9: Damaged Hand Operator on AFW Valve

9. It does not appear that nonlicense:1 operators properly operated the AFW (ICS) flow control valves. An operator applied excessive force with a valve wrench to close an AFW (ICS) flow control valve. He did so because he had not accurately determined the position of the valve while attempting to shut it completely. As a result of his actions, the valve was damaged, reopened, and the manual (local) capability to operate the valve

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was lost. These consequences suggest training weaknesses in the -

acceptable use of valve wrenches, the proper methods for manually operating and overriding air-operated valves, and the use of available and backup indications to determine valve positions. These weaknesses suggest areas where hands-on

g training rather than walk-through or talk-through training may be necessary.

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{ DISTRICT RESPONSE l

A specific policy preventing the use of the valve wrenches on gear-drive valve operators has been established. Formal training on this policy and on the proper operation of such valves has been provided to the operators. See Section 4B.3.2.1.c. The training and qualification requirements for operators have been changed to require that each Individual operate certain valves, such as the Auxiliary Feedwater Control Valves, so that they are familiar with the feel and characteristics of the valve. Included in the training and operating l

l B-21

policies are the requirement to utilize the available indications of valve position and the need to communicate with the control room.

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O B-22

FINDING - B.10: Radiological Controls and Emergency Preparedness

10. While the deficiencies in SMUD's radiological control and emergency preparedness programs during the December 26, 1985 incident did not jeopardize the public health and safety due to the relatively minor radiological consequences of this incident, they do indicate weaknesses in SMUD's program and the training of Rancho Seco personnel.

DISTRICT RESPONSES The District has undertaken a program to significantly enhance the Emergency Plan and the effectiveness of its implementation ~ . Meetings have been held with Federal, State, and County representatives to resolve details of procedures and hardware configuration which were identified as impediments'to effective implementation. The December 26 event highlighted communications and procedure adherence issues. ,

Proactive steps have been taken by the District to get combined training between the Rancho Seco communicators and their counter parts in the counties. Site visits, both ways, have occurred and are now a part of the program. The Emergency Plan itself has been i.

l revised to improve and simplify its use during emergency conditions.

The plant operations staff has received training, and simulator practice, on effective command and control of complex events. See Section 48.3.4. Management policy is clearly stated that the Shift Supervisor has overall plant responsibility and is primarily responsible for the effective implementation of the Emergency Plan.

O B-23

Plant and Emergency Plan duties are clearly stated and pre-assigned within the operating crew. The Shift Supervisor / Emergency Coordinators role is to provide the overview and direction to the crew. He is not to perform as a control room panel operator.

A longer term project is underway which intends to simplify the Emergency Plan in terms of making it more " user friendly" A major.

benefit is expected which will be its ability to successfclly mitigate complex scenarios which involve multiple casualties such as, radiation release coincidence with personnel injury and a fire.

Improvements in the Radiological Controls, as obsarved in the December 26 event, are addressed in a following finding.

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B-24

FINDING - B.ll: Installation of EFIC O 11. The NRC staff was led to believe that the emergency feedwater initiation and control (EFIC) system would be installed in 1984 in response to a number of NRC requirements, including TMI Action Item II.E.1.2. Apparently SMUD decided to install an alternate system in response to II.E.1.2. SMUD's intent to satisfy II.E.1.2 with this alternate design was not made clear to the NRC staff, was not approved by the staff, and may not have complied with the requirements of II.E.1.2. As a result, the EFIC system, some features of which would have reduced the severity of the December 26, 1985 incident, has not yet been installed at Rancho Seco.

DISTRICT RESPONSE The AFW/EFIC (II.E.1.2) scope and schedule changes had been provided

'to the NRC staff.

l The NRC issued Safety Evaluation Reports in January and September 1982, assuming EFIC installation. On October 22, 1982, the District l

indicated that it would install interim safety grade AFW modifications and that EFIC was separate and beyord the AFW upgrade requirements of NUREG-0737. The District indicated at that time that EFIC would be installed by Cycle 7. This schedule was confirmed by.

the District on December 14, 1982, at which time the Cycle 7 outage was expected to occur in the Fall of 1984. The schedule for the i

B-25

l interim safety grade modifications was specified in a conformatory order dated March 14, 1983.

On April 28, 1983, the District submitted a revised AFH system description describing the interim AFH upgrades. NRR confirmed their understanding in an SER on the status of the AFH system dated September 26, 1983.

Then in a series of living schedule submittals, the District informed the NRC that the EFIC installation was scheduled in two phases (Cycle 8 and Cycle 9).

This approach was understood, and acknowledged, by the NNR staff during a meeting in October 1985, at which time, the District committed to accelerate the EFIC installation schedule to accomplish as much as possible during the Cycle 8 outage, with the balance of the installation to be completed during the cycle 9 outage.

While it has been the District's position in the past that provision of safety grade auto start of AFH vid the SFAS was sufficient to meet j the intent of NUREG 0737 item II.E.1.2, the point is no longer germain due to the decision to immediately install EFIC and related l

1 AFH upgrades. The upgraded AFH with EFIC meets fully the requirements of NUREG 0737 item II.E.1.2 including all long term requirements to design to IEEE-279. This is underscored by the SER received for this concept in April, 1983.

O B-26

4 FINDING - B.12: Reactor Vessel Thermal Shock

12. Although the RCS temperature dropped 180*F in 26 minutes, it would have had to rapidly drop another 215'F (i.e., to an RCS temperature of.about 170*F) while pressure was maintained at approximately 1400 psig, in order to seriously threaten reactor vessel integrity.

DISTRICT RESPONSE The District agrees with this finding, based on calculations provided by B&W. In its evaluation for the District dated February 1986, B&W also calculated the effect of the December 26, 1985 event on the reactor vessel as a function of the number of cooldowns consumed in.

N the transient and the cumulative total of cooldowns versus the number designed into the reactor vessel.

The B&W evaluation of the effects of the December 26 event concluded that the event consumed "0.3 cycles" of the 240 designed. To date, all over cooling and abnormal events in combination, have consumed five cycles based upon NSS fatigue analysis. Normal cycles have totaled less than 35, leaving 200 heatup/ cool down cycles available.

This number provides sufficient margin for achieving the balance of the plant design life.

The Electric Power Research Institute (EPRI) applied a new nonmandatory ASME Code Section XI, Appendix XX, to the December 26 event and its effects upon the Rancho Seco Reactor Vessel.

B-27

The ASME evaluation procedura allows demonstration of adequate structural integrity of the reactor vessel beltline without doing further integrity analyses as long as the reactor coolant pressure has not exceeded design pressure (2500 psi) and Tc - RT NDTS has not been less than 55'F during the transient. Calculations show that the Rancho Seco December 26, 1985 transient met these criteria.

The reactor vessel calculations demonstrate, that the Rancho Seco reactor vessel beltline region has adequate structural integrity for return to service without further evaluation.

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B-28

FINDING - B.13: Reactor Vessel Integrity

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13. The December 26, 1985 overcooling incident does not appear to have seriously threatened the integrity of the Rancho Seco reactor vessel. However, the plant has had a number of overcooling incidents in its 12-year operating history. Each time this occurs, the potential exists for additional operator errors and equipment failures that might have exacerbated (sic) the event and seriously threatened reactor integrity. Thus, the

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significance of this incident lies in the fact that under alternate scenarios, more serious consequences could occur.

DISTRICT RESPONSE The issue of reactor vessel integrity was discussed in the response to Finding 12. The District agrees with the IIT that the December 26, 1985 event does not appear to have threatened reactor vessel integrity and that the District's programs should focus on eliminating the precursor events which provide the situations which l can lead to serious events.

The District's Plant Performance and Management Improvement Program (Section 4A of Action Plan) specifically addresses investigations which are of a retrospective nature and focus on preventing trips and thereby avoid challenging operators and/or safety systems. In this way we prevent a transient that could cause the post-trip response to leave acceptable pressure and temperature limits and begin to develop the characteristics of a serious or challenging event.

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FINDING - 8.14: Use of ICS in FSAR Deslan Basis Events

14. It is not clear that the overcooling transient was within the Final Safety Analysis Report (FSAR) analysis of the Rancho Seco plant. Although PTS has been addressed generically, the FSAR accident analysis for Rancho Seco does not address this issue.

The most comparable analysis in the FSAR is for the cooldown due to a main steam line break. However, this analysis included only 100 seconds of the transient. In addition, the Rancho Seco

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FSAR analysis of main steam line breaks appears to be flawed and nonconservative in that it assumes that the nonsafety-related ICS operates successfully to mitigate the consequences of the accident.

DISTRICT RESPONSE The Rancho Seco original FSAR description of the main steam line break (MSLB) consisted of two analyses:

• MSLB with ICS actions, and

• MSLB without ICS or operator actions The MSLB analysis with ICS actions is conservative with respect to maximizirg offsite doses. The analysis assumes 17. failed fuel with the technical specification steam generator tube leakage. The ICS actions are assumed to occur to maximize the cooldown time to decay O

B-30

heat removal system operation thus maximizing the releases via the intact steam generator.

The MSLB analysis without ICS or operator action is conservative with respect to maximizing the potential for a return to criticality and potential adverse effects on the fuel.

During the licensing phase of Rancho Seco, the issue of MSLB inside the reactor building was raised. This concern was addressed by installation of automatic feedwater isolation, performed by the Main i Steam Failure Logic (MSFL). .The MSFL is independent of the ICS, consists of redundant actuation channels and is battery backed, however, it is not safety grade. Reactor Building Containment integrity is not required for the MSLB, as the worst case for dose considerations is the MSLB outside the reactor building.

(

, The above design basis was clear in the original FSAR. The analysis for MSLB without ICS, or operator acticn, was contained in a response to an NRC question. During the compilation of the Updated Safety Analysis Report, this analysis was poorly worded when incorporated into the text. The separation of the "with and without ICS" analyses is not clear, and can lead to incorrect conclusions. The District is currently revising the description of the MSLB analyses in the USAR for clarity. This clarification was included in the USAR update submitted in July, 1986.

O B-31

The cooldown rate of the MSLB analysis, without ICS or operator action, bounded that of the December 26, 1985 event. The cooldown rate of the analysis was such that high pressure injection (HPI) was initiated in 23 seconds, followed by core flood tank (CFT) injection 47 seconds after event initiation. During the December 26 event, HPI/SFAS initiation occurred after 3 minutes and the pressure never reduced to that needed for CFT injection.

A review has been made of the other. design basis accident analyses in Chapter 14 of the USAR and it has been determined that ICS, or other nonsafety grade equipment action is not assumed in the mitigation of those accidents, with the exception of fuel handling accidents. For the fuel handling accident, the releases are assumed to be filtered through the auxiliary building filters, which are nonsafety grade.

Credit for these filters is appropriate as they are subject to technical specifications and the system must be operating during fuel handling operations. This design basis was clearly described and reviewed in the final Safety Analysis Report.

l The subject of PTS is addressed in Chapter 4 of the Rancho Seco USAR. The Babcox and Wilcox report BAW-1791, "B&W Owners Group Probabilistic Evaluation of Pressurized Thermal Shock - Phase 1 Report," June 1983, is referenced and described, Numerous PTS events are evaluated in BAH 1791 including events similar to the December 26, 1985 event. BAW 1791 is discussed further in Section 8.26 of this response.

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With the PTS rule making in December 1985,'the NRC established screening criteria for PTS. The District's response to the screening criteria has been submitted, and will..be incorporated into the USAR in the 1986 update scheduled-for July 1987 submittal.

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FINDING - B.15: Precursors to 12-26 Event There were a number of precursors to the December 26, 1985 incident at Rancho Seco. These precursors indicate that improvements in the reliability of the ICS and procedures to efficiently mitigate a loss of ICS power have not been developed or implemented at Rancho Seco despite numerous efforts on the part of the NRC staff to improve the reliability of the ICS and to ensure that the necessary procedures to efficiently mitigate such an event would be available to the operators. While the staff had raised these issues on a number of occasions over the past 6 to 8 years, SMUD personnel had not implemented the actions, and the NRC staff had not taken effective action to ensure that the improvements in reliability and the procedures were developed and implemented at Rancho Seco. The specific findings associated with these precursors include:

(specific responses follow.)

DISTRICT RESPONSE The District is committed to a permanent precursor review program, see Section 4A.l.3. The Precursor Review, currently under way as part of the Plant Performance Improvement Program, is resulting in the identification of recommendations not only to address specific precursors, but also to determine whether the District's previous analyses of precursors was too narrow in scope and therefore worthy of additional action.

O B-34 i

1' FINDING B.15. C Power Supplies to ICS/NNI

a. Although the ICS power supply is similar to the NNI power supply, particularly with respect to the role of the power supply monitor, SMUD's principal emphasis following the lightbulb incident in March 1978 was on the NNI rather than on the ICS. This emphasis seems to have biased SMUD's subsequent reviews of issues associated with the NNI and ICS.

i 6ISTRICT RESPONSE -

The District is currently reviewing the ICS and its power supplies.

The reviews being performed and the modifications identified are described in sections 4C.1 and 4C.2 of the Action Plan.

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4 B-35 l

FINDING B.15.b: January 5, 1979 Loss of ICS Power O

b. The loss of ICS power transient at Rancho Seco on January 5, 1979 was similar to the December 26, 1985 incident. However, it was not as severe as the "lightbulb incident" and did not receive the same level of attention. As a result, changes in the design of the ICS were not made and procedures for loss of ICS were not developed.

DISTRICT RESP 0 HSE The District is reviewing the ICS as identified in sections 4C.1 and 4C.2 of the Action Plan. A procedure for the loss of ICS has been generated and the necessary implementing training provided. See Section 48.3.2.

O B-36

FINDING B.15.c: ICS Reliability Study, BAW-1564 O c. In March 1979, B&W issued a report (BAW-1564) in which they analyzed the reliability of the ICS. Although the B&W analysis noted a number of changes that appeared to be warranted in the ICS, SMUD concluded that no changes were necessary. A subsequent analysis of the ICS by the Oak Ridge National Laboratory criticized the B&W analysis and noted that it was of limfted scope and did not appear to meet the requirements of the

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, original Order. The NRC staff concluded that no immediate changes were required at Rancho Seco as a result of the B&W analysis. The long-term issues associated with the B&W report were to be considered in Unresolved Safety Issue (USI) A-47,

" Safety Implications of Control Systems."

O DISTRICT RESPONSE i

-The following clearly establishes the status of BAW-1564. The six recommendations from the report are being input to the evaluation process described in Section 4A. The present status of each is discussed at the end of the following except from the ASLB decision.

The ICS reliability study, BAH-1564, was 11tiga'ted as to its completeness and adequacy before the Atomic Safety and Licensing Board which issued a decision (LBP-81-12) dated May 15, 1981. The section of that decision pertaining to the ICS is provided below.

O B-37

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II. FINDINGS OF FACT A. Integrated Control System

18. Board Question H-C 16:

Is the failure mode and effects analysis for the Rancho Seco integrated control system complete and adequate?

One of the long-term actions directed by the Commission in its Order of May 7,1979, was that

"[t]he licensee will submit a failure mode and effects analysis of the Integrated Control System to the NRC Staff as soon as practicable." 44 Red. Reg. at 27779 (1979). Such an analysis was performed by B&W for Licensee as part of B&W's study of the reliability of the integrated control system ("ICS"). The results of B&W's reliability study are contained in B&W Report BAW 1564, " Integrated control System Reliability Analysis." CEC Ex. 3.

19. In order to assess the completeness and adequacy of B&W's analysis, it is important first to understand the Rancho Seco ICS and the Staff's concerns regarding it. The ICS is an automatic control system whose basic function is to continuously match the unit's O

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power generation to its load demand. The ICS does this by coordinating the rate of steam generation and the steam flow to the turbine. NRC Staff Testimony of' Dale F. Thatcher Relative to the Integrated Control System (Board Question 16), following Tr.ll63

(" Thatcher ICS Testimony"), at 2.

20. During normal operations, the ICS provides proper coordination of the reactor, steam generator,

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feedwater control, and turbine. Proper coordination -

consists of producing the best load response to unit load demand within the limitations and capabilities of the plant equipment. Id. at 3.

21. The ICS includes four subsystems: unit load demand control,-integrated master control, steam generator control, and reactor control. Id. at 2. Each of these subsystems (except for the unit load demand 1

control) regulates and interacts with a number of other plant control systems, such as the control rod drive system and the feedwater pump and valve controls. Id. at 3. The ICS can maintain a constant average reactor coolant temperature at power levels between 15% and 100% of load and can maintain constant steam pressure at all loads. Id. at 3. During load changes and system upsets the ICS applies signals to

control major parameters (feedwater flow, steam B-39

pressure, reactor power, and reactor coolant temperature) in such a manner as to achieve optimum overall plant response without challenging the safety systems. Testimony of B. A. Karrasch and R. C. Jones, fol. Tr. 535 ("Karrasch-Jon s") at 7-9. It has been demonstrated that the ICS can reduce power from 100%

to 15% and maintain that level should the turbine trip without calling upon the reactor's protective systems (Karrasch-Jones testimony at 10), although presently

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an anticipatory reactor trip on turbine trip has been added so that the ICS can no longer perform this function. Id. The ICS was thus designed to keep the reactor on line during off-r.ormal conditions and enhance plant availability. Id. at 7; Tr. 1076. If, because of protective system actions, the reactor does shut down, the ICS will control steam pressure and maintain a preset steam generator level by controlling steam and feedwater, so long as either main or auxiliary feedwater is available. Tr. 1105, 1118, 1119.

CEC has emphasized, both in its cross-examination and in its Proposed Findings, the notion that it is the sensitivity of the B&W steam supply system to secondary side conditions which makes the ICS necessary and which, therefore, makes reliability of the ICS a very important matter. CEC Proposed O

B-40 1

Findings at 31-32; Tr. 1103-1105. Both Staff and i Licensee emphasize the similarity of the ICS to the systems.used at other power plants, including fossil-fueled plants. Staff's Proposed Findings at

- 11; Licensee's Proposed Findings at 24; Karrasch-Jones Testimony at 7. It appears that, in the days shortly after the TMI-2 accident, the Staff was concerned that I

the ICS could cause or contribute to an incident.

Thatcher ICS Testimony at 5; CEC Ex. 26 at 1-5, 2-9.

In particular, the Staff then believed that an ICS malfunction could prevent auxiliary feedwater (AFW) from being supplied during a loss-of-main-feedwater transient or could cause such a transient. Id.; Tr.

J 1270-72.

23. The first concern was addressed on a short-term basis in the Commission Order of May 7, 1979, by requiring Licensee to "[d] develop and implement operating procedures for initiating and controlling auxiliary feedwater independent of Integrated Control System control." 44 Fed. Reg. at 27779 (1979). The adequacy i of Licensee's compliance with this aspect of the May l

l 7, 1979 Order was established by the Staff by visiting i

the site and conducting examinations of the operators i

to verify the adequacy of their training. This evaluation included a walk-through of some of the procedural aspects of manually controlling AFW

{

l B-41

Independent of the ICS and a review of plant diagrams to verify that the valves that would be utilized for l

AFH flow control were indeed independent of the ICS.

Thatcher ICS Testimony at 4, 5; Tr.1386, 3730, 3731; Staff Evaluation at 13.

24. A permanent solution to the first concern has been provided by Licensee's safety-grade AFH control system independent of the ICS.. This modification will completely remove the operation of the AFH system from the ICS. Thatcher ICS Testimony at 5; Tr. 1273.

25. It was the second concern relating to the ICS that led the Staff to ask that a failure mode and effects analysis ("FMEA") of the ICS be performed. Since the Staff was' interested in the potential role of the ICS as the instigator of a transient, it sought to have an analysis made of the reliability of the ICS and the effects of failures of that system on the plant's operation. Tr. 648, 937-39; Tr. 1270-73. A FMEA is a systematic procedure for identifying the modes of failure of a system and for identifying their consequences. It seeks to determine if any single failure in a system can prevent the system's function. It is considered to be the first general step of a reliability analysis. Thatcher ICS Testimony at 6. Accordingly, an ICS FMEA was one of O

8-42

the long-term actions directed by the Commission in its Order of May 7, 1979. As a long-term action it was not a condition of restart.

26. B&W performed the FMEA as part of its reliability analysis of the ICS. It determined the expected effects upon the B&W steam system from single failures of ICS inputs, outputs, and internal modules. The Rancho Seco plant was chosen specifically as a

~ -

representative design for all the B&W units for the l analysis. The analysis was complemented with an i

evaluation of field data from all B&W operating plants i

i and a. computer simulation to confirm the effects of various ICS failures on associated equipment.

Karrasch-Jones testimony at 11; Staff Ex. 5 at 3. The

! analysis was made a part of our record as CEC Exhibit 3, " Integrated Control System Reliability Analysis,"

BAW--1564, August 1979, as was a review by Oak Ridge National Laboratory of the analysis (Board Exhibit 1). Also a part of the record is Staff Exhibit 5, the l Staff review of both reports.

l

27. Fundamentally, B&W's analysis of the reliability of the ICS thus consisted of three parts: the FMEA, a computer simulation used to study the effects of failures in more detail, (both of these specific to I

Rancho Seco), and a review of operating experience I

from all B&W operating plants. Board Ex. I at 5.

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8-43

28. The overall conclusion of the FMEA was that the reactor core remains protected throughout any of the ICS failures studied. For those postulated ICS failures which could cause reactor trip, the safety systems would operate independently of the ICS malfunction and they were assumed to operate properly. The overall conclusion from the operating experience evaluation was that ICS hardware performance has not led to a significant number of reactor trips. It was, in fact, concluded that the ICS has prevented more reactor trips than it has caused and, accordingly, its net effect has been a

reduction in the number of challenges to the Reactor l

Protection System. It was further concluded that the FMEA shows that no.ICS failure can prevent proper safety system functioning and that the operating l experience demonstrates that the ICS is a reliable l

l system with regard to preventing plant upsets.

Karrasch-Jones Testimony at 11-12.

29. The ORNL Review concluded that although the ICS and related control systems contain areas which can be potentially improved, the ICS itself has proven to have a low failure rate and it does not appear to precipitate a significant number of plant upsets.

Specifically, the examination of the failure statistics revealed that only a small number of ICS O

B-44 i

malfunctions resulted in a reactor trip (approximately 6 or 162). In its review, the ORNL concluded that the ICS is a "significant asset to plant safety and availability." Board Ex. I at 11.

30. While agreeing with B&W's findings and conclusions and with the recommendations made by B&W for further improvements in areas relating to the ICS, the ORNL Review pointed out a number of perceived deficiencies

~ '

in B&W's approach to the FMEA portion of the.

reliability analysis. Tr. 1706-07, 1774. Board Ex. I passim. The main criticism leveled at the FMEA by .

ORNL was that the scope of the FMEA was too limited, leading to results having only limited value. Board O Ex. 1 at 4. The scope limitations identified by ORNL were: (1) n'ot considering the interactions between plant safety and nonsafety systems such as ICS; (2) not including analysis of failures of plant systems external to the ICS; (3) not considering multiple system failures; and (4) utilization of functional versus component diagrams as the building blocks in the analysis. Board Ex. I at 3, 4, and 6 through 8.

l

31. It was, indeed, critical language from Board Exhibit I that formed the basis for this Board's inclusion of BQHC 16 in this hearing. In particular, such statements as:

O B-45

...the B&W analysis is more notable for what it does not include than for what it does include."

and

...Because of this limited scope, the results are of limited value."

(Board Ex. I at 3 and 4) would surely give or.e pause if taken out of context.

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He note, however, the following points about each of the four numbered limitations of scope set forth above:

Point (1): Interactions between safety and nonsafety systems such as ICS were not considered. That is true, but such analysis was not specifically required by the NRC's May 7, 1979, order. A study of such actions is underway for all plants as a part of the Staff's

" Integrated Reliability Evaluation Program" (IREP) which has as one of its objectives to identify the risk significance of systems interactions originating in the ICS of B&W plants. Thatcher ICS Testimony at 8.

Point (2): Failures in systems external to the ICS were not included. This is beyond the scope of the May 7 Order. Actually, the B&W analysis did include some such failures in that it O

8 46

included failures in the inputs to and outputs from the ICS. Tr. 681-83, 1083-86. .

Point (3): Multiple failures were not considered. They were not, nor is it usual to include multiple failures in a FMEA. Tr. 1083:

Thatcher ICS Testimony at 6-7. Such an analysis l 1s usually used to determine whether a single failure can prevent operation of a safety system. Id. The ICS has not been required to

~

meet the single failure criterton and was not -

previously analyzed; such analysis can, however, be used to identify failure modes which lead to undesirable consequences. Id. at 7. As we noted above, no such consequences were found.

Point (4): Functional block diagrams were used rather than component diagrams to analyze the ICS. By this we mean that only the general functions of the ICS were used and failures of each functional block were considered, rather than identifying each specific piece of equipment .

and considering its failure. Board Ex. I at 6,

10. It is possible that presently undisclosed interactions between functions might be revealed by examining specific component failures. Board Ex. I at 6. It is also possible that (if the failure rates of specific components were known) one might estimate the probabilities of various B-47

modes of failure by that method. Tr. 1086.

However, by taking the approach which they took the B&W analysts clearly met the requirements put upon them. Further, it is not clear to the Board that a component-based analysis and estimated failure rates would give a clearer picture of reliability than the " actual history" approach which B&W supplied in addition to the FMEA. He think, in fact, that the reverse is true.

32. He note that the first conclusion of the B&W analysis was that:

1. The'[Non-Nuclear Instrumentation) power sources (external to ICS cabinets) have been vulnerable to single failures and human errors that have led to reactor trips and plant overcooling. (CEC Ex.

3 at 2-2) and we note further that it was failure of the Non-Nuclear Instrumentation power supplies that initiated the incident. 3717-18.

33. In other areas identified by the study, Licensee is considering changes to increase the reliability of the reactor coolant flow Input signal to the ICS (Tr.

3703-04), and has developed procedures to improve the l

O B-48

" tuning" of the ICS to the balance of the plant, having trained operators further in ICS control. Tr.

3704-05, i

34. Thus the ICS itself is even better now than it was when the B&W analysis was performed. As to what that analyses showed, even Board Ex. 1, which was, as we noted, in some respects critical, says:

The manufacturer contends, and we agree, that (1) the system prevents or mitigates more upsets than it causes and (2) the system is generally superior to manual or fragmented control schemes. Board Ex. 1 at 15.

35. In sum we find that the FMEA was undertaken in response to certain Staff concerns, that the results of the analysis should allay those concerns, and that the FMEA was adequate and complete for its purpose.

He note that it raised other issues whose resolution would be expected to yield an even more reliable and 1

safer plant (para. 33 supra), and that those issues are being acted upon. Although the need to perform a broader study of the B&W control system and its role in the initiation and the mitigation of transients has been identified and it will be carried out in the IREP, we see no reason to believe that the Rancho Seco O

B-49

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plant would present a hazard to public health and safety during the ongoing investigations and upgradin;.

The District and the B&W Owners Group have embarked on a program to perform a comprehensive ICS evaluation. A portion of that evaluaticn will be a new ICS FMEA, which will supercede the existing effort reported in BAH 1564.

B&W report BAW 1564 identified six recommendations for further investigation. The status of the District's actions with respect to those recommendations is provided below:

l RECOMMENDATION 1 l

B&W recommends that the NNI/ICS power supply be reviewed for possible changes to enhance reliability and safety.

I i

DISTRICT RESPONSE I,

The NNI/ICS power supplies have been subject to reviews and subsequent modifications to enhance reliability to date. In addltion, the District is reevaluating the power supplies as l identified in section 4 of the Action Plan.

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0 B-50 l

RECOMMEN0ATION 2 B&W recommends that the input signals from the NI/RPS system to the ICS - specifically the RC flow signal - be reviewed for possible changes to enhance reliability and safety.

DISTRICT RESPONSE This recommendation is being implemented as described in section 4 of

~-

The Action Plan.

RECOMMENDATION 3 B&W recommends that the ICS/ BOP system tuning, particularly

feedwater condensate systems and the ICS controls be reviewed for possible changes to enhance reliability and safety.

DISTRICT RESPONSE The District had reviewed the ICS/B0P system tuning and had performed ICS tuning prior to the December 26, 1985 event. ICS tuning will i also be performed during restart testing. ICS tuning is a part of the District's ICS maintenance program.

l B-51

RECOMMENDATION 4 B&W recommends that main feedwater pump turbine drive minimum speed control be reviewed for possible changes to. prevent loss of main feedwater or indication of main feedwater to enhance reliability and safety.

DISTRICT RESPONSE The District is currently evaluating this concern as identified in section 4A of the Action Plan.

RECOMMENDATION 5 B&W recommends that a means to prevent or mitigate the consequences of a stuck-open main feedwater startup valve be reviewed to enhance reliability and safety.

DISTRICT RESPONSE The District is currently evaluating this recommendation per section 4 of the Action Plan. The installation of EFIC will resolve this issue.

O B-52

l RECOMMENDATION 6 i'

B&W recommends that a means to prevent or mitigate the consequences of a stuck-open turbine bypass valve be reviewed to 4 enhance reliability and safety.

DISTRICT RESPONSE The District is currently evaluating this recommendation per section i

~4A of the Action Plan. Separately, the B&W Owners Group SPIP effort.

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is investigating ways of reducing challenges to pressure controlling devices.

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. - - - - - .. ..~ - _ - . - - . - _ . - - _ _ . .

FINDING B.15.d: District Response to IE Bulletin 79-27

d. As a result of the loss of power to NNI and ICS at Oconee in November 1979, NRC issued Bulletin 79-27 describing a number of actions to be carried cut by licensees. Although the Bulletin raised significant concerns about the consequences of a loss of power to instrumentation and control systems, SMUD concluded that no additional design modifications were necessary and that event-oriented procedures to deal with such events were not it would appear that Bulletin 79-27 was initially necessary.

Intended to solicit detailed information from licensees that could form the basis for an in-depth review of the issues associated with control systems comparable to the review of safety-related systems conducted as part of an operating license review. Based on the initial scope of the review, the conclusion was reached that SMUD's response did not contain sufficient information and did not adequately address the concerns in the Bulletin. After the progressive narrowing of the scope of the review, it was decided that the SMUD response was adequate, despite what appear to be a number of weaknesses in the SMUD response. Thus, the conclusion was finally reached that SMUD had provided reasonable assurance that they had addressed the concerns in Bulletin 79-27, and that the long-term impilcations of Bulletin 79-27 would be addressed as part of USI A-47.

O 8-54

DISTRICT RESPONSE m

U The District is revisiting Bulletin 79-27 as a part of the precursor review program, Section 4A.1, to ensure that the concerns identified have been adequately addressed in a broad scope fashion. In addition, reviews of, and modifications to, the ICS/NNI are.

identified in sections 4C.1 and 4C.2 of the Action Plan.

The Precursor Review Group reviewed the District's original response "to IE Bulletin 79-27 and determined that the response was no longer -

l appropriate and that further technical review was required. The Deterministic Failure Consequence Analysis Group reviewed the bulletin for identification of instrumentation and failure consequences. As a result of this review, recommendations were made 3

for further evaluation of identified failures and determination that proposed modifications would resolve the failure concerns. The recommendations were made in a;cordance with Systematic Assessment Program, Section 4A, the Action Plan.

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FINDING B.15.e: District Response to February 1980 Loss of NNI at Crystal River

e. Following the February 1980 loss of NNI power at Crystal River, the NRC identified an issue about the failure mode of atmospheric dump valves (ADV) on loss of ICS power. SMUD's response to this issue did not include the other valves at Rancho Seco that repositioned on loss of ICS power (i.e., they confined it to the narrow issue associated with the ADVs). In addition, SMUD deferred this narrow issue to installation of the EFIC system, which to date has not been installed at Rancho Seco. The NRC foind this response to be acceptable.

DISTRICT RESPONSE Following the December 26, 1985 overcooling event, the District performed modifications to the controls of the atruospheric dump valves, turbine bypass valves (TBV), and the auxiliary feedwater (AFW) flow control valves. These modifications are discussed in sections 4C.1, 4C.2, and 4C.3 of the Action Plan.

O B-56

FINDING B.15.f: District Response to NUREG-0667

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f. Because of concerns about the transient response of B&W-designed reactors and the role of ICS as an initiator of-such transients,-

NRC conducted an extensive study and made 22 recommendations in NUREG-0667. However, it does not appear that these recommendations were sent to SMUD for action or that the recommendations that are relevant to the December 26, 1985 incident were implemented at Rancho Seco.

I DISTRICT RESPONSE l

The District has implemented many of the NUREG 0667 recommendations.

The balance of the recommendations are currently being evaluated by the District and/or the B&W Owners Group as elements of the systematic assessments described within Section 4A.

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FINDING B.15.g: Significance of Partial Loss of NNI at Rancho Seco March 1984

g. The March 19, 1984 partial loss of NNI power at Rancho Seco again demonstrated that the failure of nonsafety-related equipment at B&W-designed plants has the potential to cause plant transients and to challenge the operator's capability to mitigate the transient without overcooling and undercooling the primary system. Despite the fact that this event occurred

~

nearly 2 years ago, the December 26, 1985 incident demonstrates that neither SMUD nor the NRC staff has implemented effective actions to resolve this situation. In questions asked by the staff and responses provided by the B&W Owner's Group following the March 1984 loss of NNI power at Rancho Seco, the Team again sees strong evidence of a narrow focus on the incidents initiated by inappropriate control system actions in response to false inputs from the NNI. The questions in general do not refer directly to the ICS. As a result, the full significance of the loss of power to the ICS was not addressed.

DISTRICT RESPONSE The District has implemented modifications and is in the process of implementing additional modifications per sections 4C.1, 4C.2, and 4C.3 of the Action Plan, to ensure that the plant will go to a known state with the ability to control decay heat removal independent of the ICS, following a loss of ICS or NNI power. For perspective, the O

B-58

March 1984 event was a partial loss of NNI caused by the effects of _

the Exciter Hydrogen fire on the class II power supply. When that power was lost, the auto-transfer to the vital power supply was Interrupted due to a low setpoint on the over-voltage protection circuit. This occurred approximately one hour.following the Reactor shutdown. It resulted in an ADV opening which was immediately recognized and closed by the operator in the Control Room using manual ICS control. Approximately two hours later, while troubleshooting the NNI power supplies, there was a short interruption of NNI-X power which did not adversely effect the plant. -

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k B-59 l . . . - - . - . . - _ .- - . - . . . . . - - .-.. - .. _ _ - . - - .- __. ..._ ._ -.

FINDING B.15.h: Applicability of " Reference Plant" Studies to Rancho Seco

h. While the scope of the analysis performed under USI A-47 is broad, it appears that to date the actual study includes only those events with the potential to produce consequences outside the design basis of the reference plant. Such events are rare so the study does not appear to address substantive issues of the frequent challenges to protection systems and frequent

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abnormal operating occurrences, such as those identified in BAH-1564, Bulletin 79-27, and NUREG-0667. In addition, the analysis does not consider the events that are significant at other than the reference plant. Differences in plant design that could cause an event to be significant at another plant are not adequately considered. Therefore, it appears that the analysis performed to date under USI A-47 does not address the long-term issues raised in bulletin 79-27, BAH-1564, or NUREG-0667 that are relevant to the December 26, 1985 incident.

Thus, results of the resolution of USI A-47 are of quite limited applicability to B&H-designed plants beyond the reference plant that was studied. The results are not directly applicable to most other B&W-designed plants such as Rancho Seco because of the differences in the design of the ICS.

O B-60

1 DISTRICT RESPONSE .

The resolution of USI A-47 is a NRC staff action. However, the District has implemented a program, as described in Section 4A the Action Plan, to identify events th&t challenge the safety systems or the operators and take appropriate corrective actions. In addition, .

the B&W Owners Group has embarked on the SPIP Prograra, which is aimed at reducing the frequency and severity of transients at B&W plants.

A portion of this program includes an extensive evaluation of the ICS

'as a contributor to plant transients. This evaluation will take into "

consideration the differences in ICS designs between the various B&W plants.

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l B-61

FINDING - B.16: Timely Identification of Loss of ICS Power Condition O

1. It appears that the transient initiator (i.e., the loss of ICS de pow 3r) was not fully recognized by control room operators until two minutes after the power was lost. Although the "ICS and Fan Power Failure" alarm alerts operators about ICS power failures, it appears that its importance was somewhat obscured because it also acts as a trouble alarm for fan failure or for loss of one of the redundant ICS dc power supplies, neither of which requires immediate operator actions or initiates a '

transient.

DISTRICT RESPONSE .

The operators had determined the loss of ICS power prior to the reactor trip, i.e., within the first 15 seconds as a result of the annunciator alarm and the loss of ICS Controller Hand / Auto Indicator lights. Coupled with the immediate "under cooling" effects (due to the Main Feedwater pumps being runback) there was no ambiguity which suggested a " Fan Power Failure." In response to the potential for  ;

ambiguous annunciator alarms, Engineering Changes R-0517 and R-0580 have been implemented to provide discrete alarm windows for both ICS and NNI Power failures separate from any other alarms. Furthermore, as a long term effort a complete reasses: ment of the Annunciator Systems is to be accomplished to insure that annunciators are previded for all important parameters and that they are unambiguous .

to the operator. This is an elemen't of the CRDR Human Factors 9::

B-62

l committed upgrades which preceded the occurrence of the December 26, 1985 event.

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FINDING - B.17: Usefulness of Annunciator Procedures Manual O

2. The Annunciator Procedures Manual was not used by the operators following the "ICS or Fan Power Failure" alarm. Even if the Annunciator Procedures Manual had been used, it contained very limited guidance concerning the implications of this alarm and would have been of no value to the operators in recognizing or restoring the loss of ICS dc power.

DISTRICT RESPONSE The Annunciator procedure, along with the other operating procedures (EOPs, cps, 50P's, SP's), have been the subject of a comprehensive Operational Assessment following the event. As a consequence, a number of revisions have been made to incorporate the lessons learned in this event. In additidn, a long term annunciator procedure upgrade program is being implemented which will provide annunciator procedures consistent with the Human Factors /CRDR Annunciator upgrades which are also being develepad. Training on the changes made has aircady been given or is committed as described in Section 48.3.2.

O B-64

FINDING - B.18: Performance of ICS Following Restoration of Power

3. The ICS performance upon restoration of power is still not fully understood, especially because performance may depend on the duration of the power interruption. However, when ICS de power is restored, reactor operators regain remote control of plant equipment from the control room. (It is the Team's understanding that the B&W Owaers' Group is planning to conduct an investigative program that will include this matter.).

DISTRICT RESPONSE The B&W Owners Group is evaluating various aspects of the ICS l

including performance upon restoration of power. .This evaluation N

' will consider the results of testing performed at Davis-Besse on the ICS performance upon power restoration.

4 i

The District has installed modifications to the important ICS controlled values which provide power and controls in the control room which are independent of the ICS. This ensures that the demands from ICS during restoration of power will not cause a subsequent transient. Following repowering and stabilization of the ICS, the [

operator can return control of each device to the ICS satisfied that the transfer will be "bumpless". Use of the controls has been incorporated into the " Loss of ICS" procedure and training has been performed as described in Section 4B.3.2. Proper functioning of these new controls, and validation of' the new procedure for l repowering, will be demonstrated during the restart test program.

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B-65

FINDING - 8.19: Control Room Indicators which Fall to Mid-Scale O

4. Most of the indicators in the control room (both meters and recorders) are part of the NNI system; hence, they are generally independent of the ICS. However, there are exceptions that had not been recognized prior to the December 26, 1985 incident.

For example, the main feedwater (MFW) flow recorders are affected by the ICS. During the December 26, 1985 incident, the recorder failed to a value near mid-scale when MFW flow was

~

actually zero.

DISTRICT RESPONSE l

l A detailed review of the ICS drawings was confirmed by testing to observe the effects of Loss of ICS de power. The result was that the following indicators are affected:

i Main feedwater Flow Recorders, A Loop and B Loop Main Generator Electrical Frequency Error Electrical Frequency Error is a parameter used within the ICS and is for information only to the operator. The Main feedwater Flow Recorders are affected by the ICS as it is necessary to combine the Startup and Main flow signals to develop the total feedwater delivered to each steam generator. The mid-scale value which results, following loss of ICS power, has been determined to have no adverse effect upon operations should this event reoccur. The reason is that a loss of ICS power will cause the main feedwater pumps to O

B-66

runback to minimum speed while the feedwater control valves go to .

mid-position. The runback causes an undercooling which results in a

- reactor trip. Since the indication of Main T?edwater Flow remains high (mid-scale position) the operator would expect an overcooling to i

result. The Emergency Operating Procedures (ECP's) direct that the J associated pumps then be tripped, in this case the Main Feedwater 4

Pumps. Since they have already runback to minimum speed, the trip would have little effect on the transient. Section 4C.I.d provides a description of the detailed actions being implemented to enhance

Indications and control'upon loss of ICS power. m c

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B-67

FINDING - B.20: Adherence to Radiation Protection Requirements O

5. Because of a perceived sense of urgency, two nonlicensed operators made an emergency entry into the makeup pump room without respiratory protection or adequate protective clothing, neither of which was readily available. As a result, their clothing was contaminated and they were exposed to airborne radioactivity.

DISTRICT RESPONSE Since the event, the appropriate procedures have been revised and training has been completed in response to the lessons learned. See Section 4B.3.2/.3/.4. Furthermore, managements policies regarding radiation protection and procedure adherence have been clearly stated to insure that all personnel understand their responsibilities.

Additional protective equipment, including respirators, have been staged at locations more convenient to personnel requiring their use to minimize delays during emergency conditions. Significantly, another health physics technician has been added to each shift to provide operations with dedicated Health Physics support.

This has been beneficial by improving communications and mutual support between the Operations and Health Physics functions.

O B-68

FINDING - B.21: Programmatic Efforts to Deseminate Lessons-Learned and Plant Changes

6. The operators did not remember a recent modification had been made to permit the TBVs and ADVs to be closed from the remote shutdown panel (outside the control room) Independent of the availability of ICS power. This change was made to accommodate a fire in the control room. Although this modification had been incorporated in the control room fire procedures, SMUD did not

~

review other procedures to determine the applicability of this -

modification.

DISTRICT RESPONSE The Training provided at the time the shut down panel centrols were installed did address the use of those controls for events other than fire in the control room. That the operators did not remember these controls until after the event was terminated, suggests that additional efforts are needed to prevent reoccurrence. Remedial training has been provided on this specific modification, further programmatic efforts will be directed toward ensuring that all affected procedures are changed which are related to lessons learned or plant changes. See sections 48.1 and 4B.3 for programs and techniques which are being implemented for this purpose.

O B-69

e FINDING - B.22: Operating Crew Minimum Recuired Staffino

7. Additional staffing above that required by plant Technical Specifications and other SMUD regulatory commitments allowed operators to perform certain tasks simultaneously. With staffing at the minimum required level, the actions performed would have had to be performed sequentially, would have taken longer, and could have exacerbated the overcooling transient.

DISTRICT RESPONSE The actions required to control the plant following the loss of ICS power event on December 26, 1985, could have been performed by the minimum required staff. 'The overcooling could have been terminated at any point by simply performing the E0P step to " trip the appropriate pumps." It was a conscious decision by the operator to not perform that step. That decision process has been resolved by subsequent management policy, procedure changes and operator training (Section 48.3.1.1.f, 4B.3.2). Furthermore, the District has implemented modifications and procedural changes which would decrease I the demands upon the operators by providing controls in the control room which are powered independently of the ICS. See sections 4C.1, 4C.2, and 4C.3. This would eliminate the need to dispatch operators out into the plant.

el .

l B-70 i

FINDING - B.23: Role of STA h

a

8. Neither the operators nor the Shift Technical Advisor (STA) could identify an instance.of when the STA provided engineering

. expertise during the incident. Hcwever, the operators found the STA valuable as an extra person on shift to help out during the incident.

DISTRICT RESPONSE Section 4.5 of NUREG-il95 indicates that the STA participated in the decision not.to trip the Auxiliary Feedwater Pumps. That statement suggests that the STA did provide engineering input to~the operators decision-making process. However, the District has reaffirmed the role of the STA in the decision-making process to ensure that the STA Functions in an independent overview role and that the operators have engineering expertise available when needed. See Section 48.3.2.

The STA provides valuable support to the operators and the District intends to continue to support their utilization in this role. They are vital members of the operating crew and can significantly enhance nuclear safety.

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FINDING - B.24: Application of Systematic Troubleshooting O

9. It appeared to the Team that SMUD personnel found the process of troubleshooting in a highly controlled, systematic, and well-documented manner, as proposed by the Team, to be quite different from their usual maintenance practices. This difference contributed to the difficulty that the Team experienced in reviewirg the troubleshooting program.

DISTRICT RESPONSE Immediately following a reactor trip on October 2, 1985, the District instituted a systematic program for analyzing the event and resolving root causes. The program was based on NUREG-1154 Appendix B, which described the Davis-Besse systematic troubleshooting program. The District's Transient Analysis Program was again implemented following a trip on December 5, 1985.

Following the December 26, 1985 event, the District again implemented the systematic Transient Analysis Program for troubleshooting. The project was in full effect when the IIT arrived on site several days later. The IIT performed a line by line comparison of the District's program with NUREG-1154 Appendix B without considering procedural and organizational differences between the two organizations. Asp result, the District's program was twice revised to incorporate IIT wording which, in the District's opinion, did not constitute substantive changes affecting the outcome of the troubleshooting or the effectiveness of the program.

B-72

(

The District had a highly controlled, systematic, and well documented troubleshooting program in place prior to, and followirg the '

December 26, 1985 event. Further enhancements in this effort will be ,

t accomplished as described in Section 48.4.1.9.

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0 B-73

FINDING - B.25: IIT Requests for Information

10. Throughout the Team's review of the December 26, 1985 incident, SMUD personnel had considerable difficulty providing information in the detail that the Team requested. Thus, SMUD personnel repeatedly summarized data, analyses, and plans without including the actual data and analyses. As a result, the Team had to request the detailed underlying data and analyses, which subsequently were provided. This iterative process delayed the Team's on site investigation.

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DISTRICT RESPONSE The District's investigations following the December 26, 1985 event were broad in scope and exceeded that identified by the IIT. As the IIT increased their knowledge of the plant and the event, they expanded their areas of interest. Often the District already had an investigation underway in the area of question. (This may have led the IIT to perceive that the District was not providing information in the detail requested.) When requested the detailed information was provided, as stated in Finding 10 above.

The District did have difficulty-in anticipating the areas where the IIT would desire detailed information. Significantly, the detailed information was available when requested indicating that the District had independently implemented an effective troubleshooting program.

Ol B-74

FINDING - B.26: Applicability of Generic PTS Analysis to Rancho Seco O

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11. In June 1983, the B&W Owner's Group reported (BAH-1791) the results of an analysis which predicted an overcooling transient caused by a loss of ICS power could occur at B&W-designed reactors with a high probability (about 4x10-2 per reactor year). If this probability were applicable to all eight B&W-designed operating reactors, such a transient could occur at-some B&W-designed plants approximately every 3 years. Thus, it

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would appear that this analysis predicts that events comparable -

to the December 26, 1985 incident would occur approximately once every third year even if the EFIC system were installed at all B&W-designed plants. In addition, the report notes that one B&W-designed plant has a combination of components that cause the transient frequencies to be even higher. The Team deduced that the plant was Rancho Saco. Finally, the generic B&W PTS analysis (BAW-1791) is not directly applicable to Rancho Seco because it assumes that the EFIC system is installed.

DISTRICT RESPONSE I

The B&W Owners Group has reviewed 8AW 1791 since the December 26, 1985 event. Overall, the report was found to still be valid. During the short term (prior to EFIC installation), the report underpredicts l the frequency of occurrences for some overcooling events. The report becomes more representative with the installation of EFIC. This review was accomplished per Section 4A.3.

8-75 l

APPENDIX C CROSS REFERENCE ACTION PLAN TO

' NRC OPEN ITEMS 1

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3 APPENDIX C O

CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments A. Areas of Concern Relating to December 26, 1985 Overcooling Event (NRR and Region V)

. A.1~ Loss of DC Power to ICS/NNI i 1 l Root Cause RJR 86-75 See Root Cause Report 85-41 i j

(VII.1) Note: RJR 86-75 is a reference to a letter Summary Report on 12-26-85

- Control Room Instrumentation failure B.16,B.17,8.19 sent by SMUD R.J. Rodriguez to

! (e.g., Main Feedwater Flowmeter J. B. Martin, NRC Region V 507. readout) Administrator on February 19, 1986

)

FMEA 4C.I.e

- Power Supply Monttors 4C.1.c l

Integrity of Electrical Terminations 4C.I.a.2.k.

4C.2.b.1.4, 4C.13.1.6 i

i A.2 Plant Response on Loss of ICS/NNI l

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ADV and TBV 4C.I.a.1

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Startup and MFH Control Valves, 4C.3 RJR 86-75 HFP Speed 4C.3 (V.1.4.b)

Control Independent of ICS for AFH 4C.3 l Control Valves, ADV and TBV 4C.I l

Secondary Steam System Valve 4C.3.d RJR 86-082  ;

! Isolation Halkdown Check Capability (VI.4.1) .

of Isolation from Control Room Response to Return of ICS Power B.7 and B.18 l

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments A.3 . Makeup /itPI Pump Failure Root Cause RJR 86-75 See Root Cause Report 85-41 (VII.3)

Interlocks on Water Supply Sources 4C.12 Makeup Pump Repair 4C.13.1.3 A.4 Overcooling Effects and Reactor Vessel and Steam Generators Analysis of Internal Transient on RJR 86-75 Vessel and Steam Generators (V.6)

License Amendment to Clarify Include in Restart Report Cool Down Rate Technical Basis for PTS Guidelines Include in Restart Report Potential for Core Lift Include in Restart Report A.5 Radiation Monitor Damage Root Cause RJR 86-75 (V.5)

Effects of Containment Isolation Include in Restart Report Systems required following ESFAS Include in Restart Report

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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, APPENDIX C i CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments A.6 Flooding of Main Steam Headers Evaluate Steam Header Supports 4C.3.d.1.3 Engr. Analysis Complete. Lines OK.

j Include in Restart Report A.7 Plant Maintenance and Testing I - Evaluate Maintenance Program 4C.13.1.8

, 4C.13.2.1 4C.13.2.4 Manual Valve Maintenance 4C.13.1.9 Preventive Halntenance and Other 4C.13.2.4 i Deficiencies I -

Troubleshooting for Root Cause Include in Restart Report j Determination Recommendation 21.0007.H Repair Damaged Valves 4C.13.1.10 Work Complete i -

Systems Test Program 4D i .

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• Subject Areas A, 8, and C were provided to SMUD by NRC via Telefax May 21, 1986.

C-3

APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments A.8 Training and Operator Performance Adequacy of Licensed Operator and 48.3 RJR 86-75 Non-Licensed Operator Training (V.9) o ICS Off Normal Operations 4C.) & 4C.3 o Operation on Makeup System 48.3.2 o AFH Throttling 48.3.2 o AfH Pump Trip Criteria 4B.3.2 o Emergency Plan 4B.3.4, 4B.8 o Communications 4B.8 o ADV, TBV Operation 48.3.2 o Differences Between B&H 48.3.2 RJR 86-75 Simulator and Rancho Seco Plant (V.9.8) o Operation of Manual Valves 48.3.2.1.c o Plant Hodifications, Procedural 4B.3.2 Changes and Additions Minimum Staffing Requirements B.22 Security / Safety Interfaces RJR 86-75 (V.7)

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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4 s APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments A.9 Normal and Emergency Procedures Event Related Procedure (Loss of 48.4.1- RJR 86-75 i ICS Power) (V.9.4) j -

Adequacy of ATOG for PTS B.6

! - Adequacy of Health Physics and 48.3.3 Emergency Procedures B.10, and B.20 i - Adequacy of Annunciator Procedures B.17 l Manual i

- Methodology for Ensuring Changes Management Issue Section.

Properly Reflected in Procedures

$ A.10 Human Engineering Deficiencies Breaker Position Indication 48.9.1.1.e i 4C.I.a.1.b.7 and B.8 1

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- ' Valve Position Indication on TBV, 48.9.1.1.b ADV and AFH Flow Control Valves B.7 and B.9 "ICS" and " Fan Failure" Alarm 4C.I.d. 4C.2.c, and B.16 i

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• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

C-5

APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments i A.11 Retrospective Issues Adequacy of FSAR Analysis B.14 EFIC System B.11 o Installation Delay o Current Schedule o NUREG-0737 II El.2 Justification

- Modifications and Improvements of B.15 Other Facilities Re.iability of ICS B.15 o Bulletin 79-27 o CR-3 Feb/80 Loss of NNI Power o NUREG-0667 o Partial Loss of NNI at Rancho Seco March /1984 Probability of PTS Events o BAH-1791 B.26

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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i APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• I Action Plan Other

Issue Reference Reference Comments i B. Region V Additional Items j B.1 Post Accident Sampling System ~4C.6 1

System Modifications 1

Procedures and Training i

! - Testing l B.2 Control Room /TSC Emergency HVAC System 4C.7. 48.4.1.3 1

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Adequacy of Design and Installation

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Modifications l

l Loading A Train on Diesel Generator j B.3 125V DC Station Batteries 1

Adequacy of Batteries Include in Restart Report i

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Replacement 4C.13.1.4 l

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• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

C-7

APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments B.4 Radioactive Liquid Effluent Releases 4B.6 and 48.7 Offsite Contamination

- Technical Specification Deficiencies Long Term Resolution B.S Emergency Plan 4B.8 Meterology Program Improvements Training 48.3.4 Procedures and Dose Assessment 4C.9 Addresses Reactor Building Purge Only C. Licensing Areas (NRR) NOT REQUIRED FOR RESTART, PER NRC C.1 Regulatory Guide 1.97 Living Schedule Implementation INSTALL BULK IN CYCLE 8 REFUELING OUTAGE, SOME IN CY9R.O.

IN RESPONSE TO NRC RAI

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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APPENDIX C v

CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments C.2 Control Room Design Review (DCRDR) 1 Living Schedule Implementation 48.9.2.3 STUDY COMPLETE END OF 1986

{ MODS INSTALLED CY9R.O.

C.3 SPDS

! - Living Schedule Upgrade to Safety Grade HILL BE CLASS 1 H/E OF CRTs, SEISMIC CRTS CY9R.O.

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Isolation Devicer I&C HILL ADDRESS (RECENT ISSUE) l Modifications (format) UNDER REVIEH BY HUMAN FACTORS

)

l C.4 Spent fuel Pool Cooling System Upgrade of Cooling Return Line CY8R.O. (LIVING SCHEDULE ITEM)

C.5 Inadequate Core Cooling Instrumentation Implementation CY8R.O. (LIVING SCHEDULE ITEM)

C.6 Class IE Electrical System TBI Diesel Generator Qualification CY8R.O. (TESTING THIS SUMMER) j . LIVING SCHEDULE ITEM l

Complete Class IE Electrical System CY8R.O. (LIVING SCHEDULE ITEM) i in NSEB i

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

t C-9

APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS' Action Plan Other Issue Reference Reference Comments C.7 Low Temperature Overpressure Protection Install Bypass Valve in Makeup CY9R.O. (LIVING SCHEDULE ITEM)

C.8 Security Modifications NSEB, Diesel Generator Building CONSIDERED BY NRC TO BE TIED TO DGB/NSEB OPERABILITY, L.S. INDICATES APRIL 1986 DATE.

C.9 Relief and Safety Valve Test Install Pressurizer Safety Valve CY8R.O. (LIVING SCHEDULE ITEM)

Supports C.10 Reactor Coolant Pump Seal Damage Automatic Initiation of Seal CY9R.O.(LIVING SCHEDULE ITEM)

Injection on Loss of Offsite Power C.ll Appendix R Complete Alternate Shutdown CY8R.O. (LIVING SCHEDULE ITEM)

(Additional Isolation Switches)

Automatic Fire Suppression NSEB CY8R.O. (LIVING SCHEDULE ITEM)

(Areas 81 & B2)

Circuit Separation of NSEB HVAC CY8R.O. (LIVING SCHEDULE ITEM)

Cable Reroute HPI Train A CY8R.O. (LIVING SCHEDULE ITEM)

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• O j

Action Plan Other Issue Reference Reference Comments C.12 Seismic Assessment of Foothills Fault NRC ACTION REQUIRED System C.13 Compilance With 10 CFR 50.46 (NUREG-0737 B& HOG EFFORT, DUE JULY 1986 (CCL ITEM)

II.K.3.31) l C.14 Diesel-Generator Reliability Technical TO BE SUBMITTED THIS SUMMER Specifications (GL 84-15) i l C.15 Revised Technical Specifications for- NRC ACTION REQUIRED, SUBMITTED FE8. 1986 l Containment Isolation Valves 1

C.16 10 CFR 50.62 Requirements for Reduction NRC ACTION REQUIRED ON B& HOG DESIGN SUBMITTAL,

SCHEDULED FOR CY9R.O., ON LIVING SCHEDULE of Risk from ATHS Events i

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• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986. '

j C-11

APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments D. This section lists the quest.ons from Enclosure 1 of the September 5, 1986 letter from Stolz to Hard, " Action Plan for Performance Improvement - Request for Additional Information." References are made to relevant Action Plan section which provides the requested information.

D.1 Provide further discussion of System 4D.5.c.1 Review done by System Engineer in SRTP.

D.2 Staff recommends a gradual loss of 40.6 instrument air test per Reg.

Guide 1.68.3 0.3 How will post-trip window be monitored 4C.2.d Pressure / Temperature data is when SPDS is unavailable. Independently available for plotting.

D.4 List functional steps in Action Plan 1.1.3 which result in CR modifications, compare to DCRDR process.

D.S List CR modifications identified by 48.9 Action Plan, for each discuss role of t fluman Factors which led to mod.

D.6 Discuss how Control Room will be vali- 4B.4.1.10 dated for responses to emergencies and abnormal events following modifications.

D.7 In Section 4A.1.2.2 include NUREG's, 4A.I.2.1.e Discusses NUREG-0667.

e.g., NUREG-0667, NUREG-0737, etc. 48.11.1 Discusses NUREG-0737.

Other NUREG's were not systematically reviewed for open items as the existing open items commitment list is being reviewed for status and priority.

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• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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O APPENDIX C C O CROSS REFERENCE TO NRC OPEN ITEMS *

! Action Plan' Other l

Issue Reference Reference Comments D.8 Section 4A.2.2 should include functional 4A.2.2 Revised loss of part of system.

D.9 Section 4A.2.3, methodology should 4A.2.2 Revised j; include loss of ICS/NNI together with a loss of offsite power, loss of j instrument air.

D.10(a) Clarification of how the output of 4D.4 4

PPEMIP will be used in SRTP.

! D.10(b)

Clarification of how the Rancho Seco 4D Section 4D.7 discusses specific i SRTP is modeled after Davis-Besse SRTP,- similarities / differences. Section j and how results will be factored into- 4D provides overall process.

Rancho Seco SRTP.

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D.10(c) How will Independent Review of SRTP be 4D.3.b/c perforined? 4D.5.a D.ll Describe how RRRB and PAG responsibill- 1.2 Revised ties relate to MSRC.

! D.12(a) Section 1.4.1 does not appear to' 1.4.1 and i.

] sununarize the SRTP in Section 40. 4D Revised i 4 l l D.12(b) Clarify how the five systems were 1.4.2 Revised l

} selected in Section 1.4.2. j i

j D.13 Section 4A.4 - Provide more detall on 4A.4 Revis.ed  ;

Interview methodology and process.

Appendix I added l .D.14(a) Why were only specific procedures 48.4 Revised ,

j targeted for review and. correction?

( D.14(b) Has VLV done on E0P's per the Rancho 48.4 Revised l Seco Procedures Generation Package?

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• Subject Areas A, B, and C were provided to SMUD.by NRC via Telefax May 21, 1986.

! C-13 i

APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments D.15 Section 48.5 - Why is there no program Response will be included in no program element to V&V upgraded Restart Report. Classic "V&V" maintenance procedures? applies to Operating Procedures, nature of maintenance procedures does not require formal V&V as a program element.

D.16 Section 4D, System Review and Test Program (a) llow are reviews and test results 4D.3.b documented?

(b) Clarify the systems selection 4D.4 process.

(c) Hill system reviee identify design 4A weaknesses which affect reoccurring operational problems?

(d) Will SR confirm system configura- 40.5.c.2 SR will be current for system ticns following neds since functions, walkdowns being done for IEB 79-14 walkdowns? material condition 79-14 criteria written into QC/QA programs address mods which involve 79-14 issues.

(e) Referenced figure 40-2 is missing. Entire Section 4D is rewritten in Action Plan Amendment 1, figure is now provided.

(f) Hill SR verify adequacy of SP's and 40.5.c.3 ISI/IST Procedures?

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

C-14 O O . O

O O APPENDIX C O

CROSS REFERENCE TO NRC OPEN ITLt:1*

Action Plan Other Issue Reference Reference Comments (g) How will results of Davis-Beese SRTP 4D.S.c.2 d

be factored into RS Testing Program?

4 (h) How will results of material 4D.5.c.2

? condition walkdown be factored into SSR?

l (1) Why will TRG only review the Test 4D.5.a/b 4 History Review section of SSR? 40.3.b/c

(j) Does completion of 10-year ISI 4D.6 I consist of periodic testing of

] pumps & valves?

(k) Hhat is " appropriate" system func- NA Issue resolved in Amendment I by i tioned test to confirm adequacy and specifying that all Selected effectiveness of Action Plan as Systems will be tested.

discussed in Section 1.f.3?

I I (1) What is meant by normal testing NA Issue resolved in Amendment I by

! with the plant performance monitor- specifying that all Selected i ing program, Section 1.f.3? Systems will be tested.

I (m) When is training for test engineer- 40.6

,' ing to be performed?

D.17 Table 1-A NA Table deleted in Amendment I  !

What is the basis for "...the regular

refueling outage system tests will provide complete functional testing of l RPS."

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! D.18 Table 1-C NA Table deleted in Amendment I

! Why is verification of system opera-

! bility per Tech Specs accompIlshed by tests other than TS Surveillances for some systems?

• Subject Areas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

C-15 t

APPENDIX C J I

CROSS REFERENCE TO NRC OPEN ITEMS

• Action Plan Other Issue Reference Reference Comments l 0.19 Appendix B (a) Finding B.6 (1) Describe how HPI/ PTS concerns B.6 Revised handled by E0P's (2) Why were PTS Guidelines B.6 Revised exceeded?

(3) Rationale for using overly conservative PTS guidelines.

(4) llow has Operator Training been B.6 Revised Training based on (1) & (2) revised to resolve conflicting accomplishes this item.

priorities within E0P's.

(b) Finding B.7 (1) What simulator was used for operator training.

(2) Provide details of differences-  !

between B&H Simulator and Rancho Seco.

D.20 Appendix G {

(a) Clarify process for PAG acceptance 4D.5.c.4 Note that Appendix G was replaced of system for restart. by latest revision in Amendment 1.

(b) Why do not all SSR's provide restart, near term, or long term recommendations?

• Subject treas A, B, and C were provided to SMUD by NRC via Telefax May 21, 1986.

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APPENDIX C CROSS REFERENCE TO NRC OPEN ITEMS

• 4 Action Plan Other Issue Reference Reference ' Comments D.21 Appendix H (a) Why are exceptions to the prereq- NA Test Specification provided as an utsites not required to be evaluated Example Only.

! for effect on specified test method?

(b) Why did the test specification for NA Test Specification provided as an the fine protection valve drain- Example Only, like modification not have a

! prerequisite requiring construc-i tion complete?

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! ' Subject Areas A, B, and C were.provided to SMUD by NRC via Telefax May 21, 1986.

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l APPENDIX 0 ACTION PLAN COMMITMENT

SUMMARY

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APPENDIX 0 ACTION PLAN COMMITMENT

SUMMARY

The following items are summarized from the Rancho Seco Action Plan for Performance Improvement. This summary is being included as Revision number 1 to the Action Plan.

The purpose of this summary is to provide a cross reference tool to assist in the closure of all Action Plan items.

This summary is formatted to provide a section overview of commitments (when necessary), the Action Plan paragraph number and a description of the commitment. "

Those individuals submitting restart recommendations per the QCI-12 procedure shall use this summary to cross reference their recommendation to the Action i Plan.

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1.4 SYSTEM REVIEW AND TEST PROGRAM 1.4.2.1 Main Feedwater System functional evaluation including O

reliability assessment will be performed. (PRIORITY 2) 1.4.2.2 Auxiliary Feedwater System functional evaluation-including reliability assessment will be performed.

(PRIORITY 2) 1.4.2.3 ICS/NNI functional evaluation including reliability assessment will be performed. (PRIORITY 2) 1.4.2.4 Pressure Control functions of the Main Steam System functional evaluation including reliability assessment will be performed. (PRIORITY 2) 1.4.2.5 Instrument Air System functional evaluation including reliability assessment will be performed. (PRIORITY 2)

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'l 1-4.A SYSTEMATIC ASSESSMENT PROGRAM  !

i9 t 4.A.a. Prior to restart, systematic assessment programs will i

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be established as'a' part of .the administrative process

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l of Rancho Seco. (PRIORITY 1)

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4A.1 PRECURSOR REVIEW PROGRAM 4A.1.2.1.a. All Transient Assessment Program (TAP) Category C O

transients will be evaluated and investigated for their applicability and impact on Rancho Seco. (PRIORITY 1) 4A.1.2.1.b. All Category B TAP events will be reviewed to determine if any of the recommendations made are applicable to Rancho Seco and to determine whether because of plant differences, the transient could have been more severe at Rancho Seco. (PRIORITY 1) 4A.1.2.1.c. All recommendations for Category A TAP transient will be reviewed to determine their applicability to Rancho Seco. (PRIORIlY 1) 4A.I.2.1.d. All Rancho Seco transients, starting from the Rancho Seco

" light bulb" event in 1978, will be reviewed.

(PRIORITY 1) 4A.1.2.1.e. Review NUREG-0667 " Transient Response of B&W designed Reactors" for open commitments.

4A.1.2.2.a. Rancho Seco Licensee Event Reports and Occurrence Description Reports will be reviewed. (PRIORITY 1) 9

-D4-

4A.l.2.2.b. Significant Operating Experience Reports (SOER) issued by the Institute of Nuclear Power Operations will be reviewed. (PRIORITY 1).

4A.I.2.2.c. Bulletins issued by the NRC Office of Inspection and Enforcement will be reviewed. (PRIORITY 1)

I 4A.l.2.2.d. Notices / Circulars issued by the NRC Office of Inspection and Enforcement will be reviewed.

(PRIORITY 1) 4A.l.2.2.e. Babcock & Wilcox Reports (Preliminary Safety Concerns, Site Instructions, and other relevant BAW reports) l will be reviewed. (PRIORITY 1) i, I

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4A.2 DETERMINISlIC FAILURE CONSEQUENCE ANALYSIS 4A.2.3.1 An evaluation of the effects of loss of Electrical O

Power will be performed. (PRIORITY 1) 4A.2.3.2 An evaluation of the effects of loss of Instrument Air will be performed. (PRIORITY 1) 4A.2.3.3 The loss of ICS and NNI power supplies will be evaluated to determine failure states and resultant actions or suggested modifications necessary to establish a known :afe state with little or no operator action. (PRIORITY 1) 4A.2.3.4 IE Bulletin 79-27 (Loss of Non-Class lE Instrumentation and Control Power Systems Bus During Operation) will be evaluated using a Deterministic Failure Consequences Analysis approach.

(PRIORITY 1) 9 1

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f 4A.3 B&W OWNERS GROUP PROGRAMS - SAFEIY AND PERFORMANCE. IMPROVEMENT PROGRAM (SPIP) i l

l' 4A.3.2 The District will fully participate in the Safety and Performance Improvement Program.

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4A.4 PLANT INTERVIEWS O

4A.4.2 A minimum number of interviews will be established to identify systems, components, or operational problems and concerns of which they are aware and provide recommendations on how to resolve them. (PRIORITY 1)

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48. MANAGEMENT, OPERATIONS AND ADMINISTRATIVE PROCESS IMPROVEMEN1 fk 4

48.1 MANAGEMENT EFFECTIVENESS 1

48.1.1.1 Review current executive level management practices and attitudes to ensure that executive level management processes support the safe and reliable

]

I operations of Rancho Seco. (PRIORITY 1) i 4

48.1.2.1.a. Establish within the Board of Directors, guidelines and agreements by which the Board, as an entity, can j more effectively set policy and direction.

i l (PRIORITY 2) l, 1 , 48.1.2.1.b. Establish written performance measurement criteria, and a performance process, for the General Manager (GM). (PRIORITY 2)

i i 48.1.2.1.c. Clarify the Board / General Manager working relationship f in writing, including the reporting desired by the Board from the General Manager. (PRIORITY 2) 1 48.1.2.2.a. Assess current corporate-support interfaces with the

! Nuclear Organization and make recommendations to the l

AGM, Nuclear and the GM regarding improved management l

of interorganizational working relationships.

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48.1. 3. a . Develop and implement a Rancho Seco Business Plan for use by the Board of Directors. (PRIORITY 2) 48.1.3.b. Establish a conprehensive, cohesive and clearly understandable set of GM and AGM-Nuclear policies and practices which provide upper tier direction for similar efforts at the functional manager and supervisory levels. (PRIORITY 2) 48.1.3.c. Establish up-to-date functional organization charters and position descriptions which accurately reflect responsibilities authorities, and accountabilities for all organization functions and job classifications.

(PRIOR 11Y 2) 48.1.3.d. Upgrade management programs and practices in the areas O

of functional planning, decision makir.g, problem solving and interdepartmental collaboration.

(PRIORITY 2) 48.1.3.e. Establish appropriate management monitoring and control systems to ensure that all levels of department management are kept informed on important department performance trends or problem areas on a timely basis. At the same time, ensure that excessively burdensome administrative control systems are not perpetuated or introduced. (PRIORIlY 2) 0

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48.1.3.f. Develop an employee communications program originating

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from the office of the AGM-Nuclear to ensure that all i department employees are kept informed of District concerns, departmental priorities and performance progress on a timely basis and encouraged to feel that they are an important part of the Rancho Seco team.

(PRIORITY 2) 48.1.3.o. Develop a program for improving communications skills of Nuclear Department managers in presentations to the Board of Directors, the public, and staff.

(PRIORITY 2) 1 5

48.1.3.h. Establish a department Human Resource Management program which includes: (PRIORITY 2)

(

48.1.3.h. 1) identification of priority management development / training needs and the appropriate

, means for addressing each; i

48.1.3.h. 2) identification of departmental priorities in terms of current vacancies and/or pipeline concerns; 48.1.3.h. 3) engage more department management collaboration

~w ith the District's Human Resources organization in the recruitment / selection process.

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48.1.3.i. . Improve Department media and community relations by establishing a more proactive media / community outreach program. (PRIORITY 2) 48.1. 3. .i . Improve Nuclear Department interfaces with all other Departments in the District by instituting additional interdepartmental communication and problem-solving processes on a regular basis. (PRIORITY 2) 48.1.3.k. Develop a Rancho Seco facilities Master Plan.

(PRIORITY 2) l I

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48.2 QUALITY AND QUALITY ASSURANCE i

e~

48.2.1.1 Reorganize the Quality function at Rancho Seco to enhance the Site Quality Assurance Department, providing increased focuses in the following areas:

(PRIORITY 1) 48.2.1.1.a. Quality Engineering 48.2.1.1.b. Quality Control 48.2.1.1.c. Surveillance 48.2.1.1.d. Vendor Qualification and Source Inspection 48.2.1.1.e. Nuclear Program Audits g ) 48.2.1.2 Develop and implement the procedures and, processes necessary to independently verify the effective closure of the actions identified for the Action Plan. The QTS tracking system has been developed to aid in completing this task. (PRIORITY 1) 48.2.1.3 Institute interim measures to strengthen the materials control at the Rancho Seco site. This action will provide additional assurance that materials being installed are properly documented and in compliance

with the applicable codes and standards. (PRIORITY 1) 1

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48.2.1.4 Institute interim measures to enhance the integration of OC planning with maintenance and constructicn instructions and activities. (PRIORIlY 1) 48.2.1.5 Increase the Site QA Department staff to assure the added demands of the Action Plan and changes in responsibilities can be effectively implemented.

(PRIORITY 1) 48.2.2.1 Update and modify the Quality Program policies and procedures to enhance the effectiveness of the Quality Program, particularly those dealing with material control, engineering, quality surveillance, and maintenance. (PRIORITY 2) 48.2.2.2 Identify and develop enhancements to the QA program to address any programmatic and management areas that have identified deficiencies from a quality perspective. (PRIORITY 2)

48.2.3.1 Develop and implement the necessary policies and procedures to establish a more proactive quality program, which will also improve the effectiveness of audits. (PRIORITY 3) l l

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48.3 -TRAINING O

V 48.3.1.1.a. Continue the upgrade of Non-Licensed Operator Training to maintain INPO Accreditation. (PRIORITY 2)

48. 3.1.1. b . Initiate the process of achieving INPO Accreditation for the maintenance training area. (PRIORITY 2) 48.3.1.1.c. Develop the plan for installation of a computerized Training Information Management System. (PRIORITY 2) 48.3.1.1.d. Develop plans for centralized and secure storage of training records. (PRIORITY 2)

( 48.3.1.1.e. Develop or purchase a-Rancho Seco Simulator Baseline

( ,

Data Information and Tracking System consistent with the Simulator now being purchased. (PRIORITY 2) 48.3.1.1.f. Incorporate short term training items (lessons learned) into permanent training materials.

(PRIORITY 2) 4 48.3.1.2.a. Complete the purchase and installation of a plant specific simulator. (PRIORITY 3) i 48.3.1.2.b. Complete the staffing of the Training Department with SMUD employees. (PRIORIfY 3)

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43.3.1.2.c. Complete and maintain INP0 accreditation for the remainder of the Training Programs. (PRIORITY 3) 48.3.2.1.a. Train licensed operators on Emergency Operating Procedures, including the changes resulting from the l December 26, 1985 event, those revisions resulting l from the E0P/ATOG review, and all completed recent design modifications. (PRIORITY 1) l 1

48.3.2.1.b. Train licensed operators on the loss of ICS/NNI, including those procedures added or revised as a result of the December 26, 1985 event and related design modifications. (PRIORITY 1) 48.3.2.1.c. Train operators on valve applications and operation including 0]T on local and/or manual operators. This includes limits and precautions such as use of valve wrenches. Incorporate into requalification training.

(PRIORI 1Y 1) 48.3.2.1.d. Train operators on the specific lessons learned from the Occember 26, 1985 event. These include such items as the makeup pump failure, overfilling the makeup tank, cooldown rates, reactor vessel head bubble information, and the functioning of valve actuator controls. (PRIORIIY 1) 9

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48.3.2.1.e. Train operators on watch standing principles, including command and control training for shift supervisors, role and function of STA, equipment monitoring. (PRIORITY 1) 48.3.2.1.f. Retrain operators on health physics requirements associated with their job responsibilities.

(PRIORITY 1) 48.3.2.1.a. Train operators on their job related functions associated with startup testing. (PRIORITY 1) 48.3.3.1.a. Train Health Physics Technicians and Operators on the procedure (s) for entry into areas of unknown radiological conditions. (PRIORIfY 1) , ,

j 48.3.3.1.b. Train Health Physics Technicians and Operators on proper response to radiological emergencies.

! (PRIORITY 1) i 48.3.3.1.c. Train Health Physics Technicians and Operators on evaluation of radiological effluent discharges.

(PRIORITY 1) i I

. 48.3.4.1.a. Train assigned maintenance personnel on the i

maintenance of the Interim Data Acquisition and j Display System (IDA05). (PRIORITY 1)

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48.3.4.1.b. Train operators on the operational use of IDADS.

(PRIORITY 1) 48.3.4.1.c. Update Emergency Preparedness Training Instructor Guides, Student Guides, and visual aids to support the October 1986 Orill. (PRIORITY 1) 48.3.4.1.d. Train assigned personnel on the revised Emergency Plan Procedures. (PRIORITY 1) 48.3.4.1.e. Provide management guidance to the operating crews (through training) on prioritizing multi-casualty events. (PRIORITY 1) 48.3.4.2.a. Develop and implement an improved process for continuing Emergency Response Organization training.

(PRIORITY 3) 0

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48.4 OPERATIONS AND OPERATING PROCEDURES O 48.4.1.1 Issue a procedure defining the policy for procedural compliance and procedural guidance. This procedure will provide a direction on what constitutes

" procedural compliance" and " procedural guidance."

(PRIORITY 1) 48.4.1.2 Correct specific procedural deficiencies identified during the review of the December 26, 1985 transient.

(PRIORITY 1) 1

48.4.1.3 Review and upgrade the CR/TSC HVAC operating l

procedures. (PRIORITY 1)

'O J

48.4.1.4 Verify technical correctness of E0P changes made since l May 1985. (PRIORITY 1)

. 48.4.1.5 Compare E0Ps to ATOG Technical Basis. Incorporate identified improvements into E0Ps. (PRIORITY 1) t 48.4.1.6 Make the necessary modifications to the design change process to assure that design changes are incorporated into all operating procedures in a timely manner.

(PRIORITY 1) l l

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48.4.1.7 Assure operating procedures address the recommended topics of Regulatory Guide 1.33 Sections listed below. Implement procedures which may be required.

(PRIORITY 1) l 48.4.1.7.a. Section 3 Procedures for Startup, Operation, and Shutdown of Safety Related Power System.

48.4.1.7.b. Section 6 Procedures for Combating Emergencies and Other Significant Events.

48.4.1.8.a Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Air Ejector / Gland Seal System. (PRIORITY 1) 48.4.1.8.b. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Auxiliary feedwater System. (PRIORITY 1) 48.4.1.8.c. Perform a valve walkdown to verify the consistency of as-built conditions, P&I0s, procedural lineups, and component identification for the Auxiliary Steam System. (PRIORilY 1)

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48.4.1.8.d. Perform a valve walkdown to verify the consistency of A)

(G as-built conditions, P& ids, procedural lineups, and component identification for the Component Cooling Water System. (PRIORITY 1) 48.4.1.8.e. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Instrument Air System. (PRIORITY 1) 48.4.1.8.f. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Main Circulating Water System. (PRIORITY 1)

! 48.4.1.8.a. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Main Condensate I

System. (PRIORITY 1) i

48.4.1.8.h. Perform a valve walkdown to verify the consistency of '

as-built conditions, P& ids, procedural lineups, and i

component identification for the Main feedwater System. (PRIORIfY 1) 48.4.1.8.1. Perform a valve walkdown to verify the consistency of as-built conditions, P&I0s, procedural lineups, and component identification for the Nitrogen Gas System.

(PRIORITY 1)

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48. 4.1. 8. .i . Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Plant Cooling Water System. (PRIORITY 1) 48.4.1.8.k. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Service Water System. (PRIORITY 1) 48.4.1.8.1. Perform a valve walkdown to verify the consistency of as-built conditions, P& ids, procedural lineups, and component identification for the Turbine Electro Hydraulic Control System. (PRIORITY 1) 48.4.1.8.m. Perform a valve walkdown to verify the consistency of O

as-built conditions, P& ids, procedural lineups, and component identification for the Turbine Lube Oil System. (PRIORITY 1) 48.4.1.9 Establish an Administrative Procedure to ensure that

" Systematic Troubleshooting" is accomplished on requisite future events. Guidance for this program is provided in memo GAC 85-1001, Rev. 2. (PRIORITY 1) 48.4.2.1 Develop a revised Nuclear Operations organization and begin staffing at the management level. (PRIORITY 2)

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___..___.____.______._____.__-..__.__y___.-_____.___.____.______ "

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48.4.2.2. Develop a staffing plan and schedule to meet,the needs i ofthe-revisedNuclearOperationsorganizdion. This j~ will include the needs for licensed operators i

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, identified as rotational / transfer assignments. -

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48.5 MAINTE. NANCE PROGRAMS AND PROCEDURES I 4B.5.1.1 Inventory Calibrated Test Equipment (C1E) and O 1 j

i calibrate and/or control use to prevent use of uncalibrated CTE. (PRIORITY 1) 48.5.1.2 Identify and assure current calibration of all in-plant instrumentation used in the performance of surveillance testing. (PRIO9ITY 1) 48.5.1.3 Rework the makeup pump and return to service.

(PRIORITY 1) 48.5.1.4 Complete the in-progress battery replacements (A, B, C, D, E, f). (PRIORITY 1) 48.5.1.5 Perform refueling interval surveillance of snubbers.

(PRIORITY 1) 48.5.1.6 Complete rework of terminations in the Bailey Cabinets in the Control Room (NNI/SFAS/RPS/ICS). (PRIORITY 1) 48.5.1.7 Perform biennial Diesel Generator Inspection and replace turbochargers. (PRIORITY 1) 0

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48.5.1.8 Define the critical items to be included in the PM

[] program. (This is to be an accelerated portion of the V planned PM Program Upgrade.) As a minimum, this will include the Manual Limitorque Operated Valves (105),

the Manual Non-Limitorque Operated Valves (135), other Manual Valves important to process flow control in Class 1 and steam generator heat removal applications (143), plant instrumentation required for surveillances, safety related HVAC and the Control Room normal HVAC system. (PRIORITY 1) 48.5.1.9 Complete Preventive Maintenance (PMs) on selected manual valves identified in 48.5.1.8 above.

(PRIORITY 1) s 48.5.2.1 Develop a departmental procedure hierarchy and writer's guide for Maintenance Procedures.

(PRIORITY 2) 48.5.2.2 Identify and prioritize maintenance procedures for generation and/or revision. (PRIORITY 2) l 48.5.2.3 Achieve authorized staffing levels within the PM organizations and activities. (PRIORITY 2) v 1

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48.5.2.4 Develop and/or revise the required programmatic procedures for the PM program to: assign responsibilitics, authority and accountabilities for the program; establish criteria and dafine the scope of the program; and define the interface with other work control processes. (PRIORITY 2) 48.5.2.5 Review existing PM tasks and frequency for critical equipment. Revise and augment as required by programmatic selection criteria. (PRIORIfY 2) 48.5.2.6 Perform Laboratory failure Analysis of the ICS S1/S2 switches and ICS Power Supply Monitor which were in place on December 26, 1985. (PRIORITY 2)

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l 48.6 HEALTH PHYSICS AND RADIOLOGICAL CONTROLS s

48.6.1.1 Relieve operators of special HP duties. (PRIORIlY 1) l 48.6.1.2 Prepare a procedure, for Health Physics Technicians to use for entry into unknown radiological conditions.

(PRIORITY 1) 48.6.1.3 Revise setpoints for plant gaseous effluent monitors to ensure unambiguous indications. (PRIORITY 1) 48.6.1.4 Issue a Radiological Event Directions Manual to provide more guidance for abnormal situations.

(PRIORITY 1) 48.6.1.5 Issue new manuals to separate event and instrument procedures from the Radiation Control Manual.

(PRIORITY 1) i

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48.7 10CfR50 APPENDIX I DISCHARGE GUIDELINES 48.7.1.1 Evaluate the current radioactive waste analysis methods and sensitivity relative to their ability to support operation needs and to provide confidence that discharges and the cumulative impact of discharges will satisfy the objectives of the environmental discharge requirements. (PRIORITY 1)~

48.7.1.2 Develop and implement the changes in Radiochemistry methods and controls necessary to provide confidence that discharges and the cumulative impact of discharges will satisfy the objectives of the environmental' discharge requirements. (PRIORITY 1)

O 48.7.1.3 Review and revise the off-site discharge calculation manual to incorporate changes in Radiochemistry methods and controls necessary per 48.7.1.2.

(PRIORITY 1) l 48.7.2.1 Evaluate the design of plant systems with the intent i

l to improve the ability to operate within Appendix I 1

Guidelines when operating with primary to secondary leakage. Implement plant improvements as appropriate.

l (PRIORI 1Y 2)

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48.8 ENERGENCY PREPARE 0 NESS O 48.8.1.1 Meet with NRC (Region V) to review / critique Emergency Action Plan. (PRIORITY 1)  ;

48.8.1.2 Update Emergency Plan and Implementing Procedures to address December 26 event Lessons Learned.

(PRIORITY 1) 48.8.1.3 Establish independent meteorological assessment capability. (PRIORITY 1) 48.8.1.4 Integrate EP commitments into commitment tracking program. (PRIORITY 1)

O 48.8.1.5 Evaluate notification / communication system and implement upgrades related to December 26 event lessons Learned. (PRIORITY 1) 48.8.1.6 Simplify Control Room dose calculation procedure.

(PRIORITY 1) 48.8.1.7 Implement new PASS procedures including core damage assessment. (PRIORITY 1) 48.8.1.8 Initiate " mini-drills" program for emergency preparedness. (PRIORITY 1)

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l 48.8.1.9 Coordination and support of the Training Group per 48.3 including: (PRIORITY 1) 48.8.1.9.a. Emergency Response Organization (ERO) identification 48.8.1.9.b. Plan / procedure update information 48.8.1.9.c. Facilities identification 48.8.1.9.d. Instruction materials upgrade

(

48.8.1.9.e. Scheduling 48.8.1.9.f. Tracking and documentation 48.8.1.9.a. Data base management 48.8.1.9.h. Management support of ERO participation l

l 48.8.2.1 Establish separate onsite and corporate plans for emergency preparedness. (PRIORITY 2) 48.8.2.2 Consolidate / cross index Emergency Preparedness (EP) 1 and Central files. (PRIORITY 2) l 48.8.2.3 Complete multi-parametric data base including EP records and schedules. (PRIORI 1Y 2) l 48.8.2.4 Provide positional analyses for ERO versus District staff. (PRIORITY 2) l l

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48.8.2.5 Define / implement public education program enhancements for emergency response. (PRIORITY 2)

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j 48.8.2.6 Integrate Emergency Plan Implementing Procedures and plant operating. procedures. (PRIORITY 2) 48.8.2.7 Complete installation of notification / communication system including verification, training and drills program. (PRIORITY 2) 48.8.2.8 Redefine Emergency Preparedness maintenance program.

(PRIORITY 2)

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._ . . . . . . . . . . _ . . _ - . . _ __ _ . - . . _ _ . _ _ _ . _ . . _ _ _ _ _ _ _ _ - _ _ . _ _ _ ~ , . _ _ . . _ . _ . . . . _ . _ . . _ . . . . . . . . _ _ - . . _ . . _ _ . _ . -

48.9 HUMAN FACTORS Implement modifications and procedure changes O

48.9.1.1 resulting from the post trip Human factors review including: (PRIORITY 1) 48.9.1.1.a. Provide operator training associated with AFW valves FV-20527 and FV-20528.

48.9.1.1.b. Provide accurate local hand jack position indication for AFW valves FV-20527 and FV 20528.

48.9.1.1.c. Improve interface between Security and Control Room personnel.

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48.9.1 1.d. Install long cord on red phone.

48.9.1.1.e. Relabel ICS power supply breakers S1/S2.

48.9.2.1 Modify Control Room access doors such that one door is used for exit and one for entrance. (PRIORITY 2) 48.9.2.2 Modify red phone power supply to eliminate spurious 1

ringing following power supply transfers. (PRIORITY 2)

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48.9.2.3 Assess capability to accelerate implementation of Control Room Design Review (CRDR) modifications (MOD 142) currently scheduled for Cycle 8 and Cycle 9 refueling outages. If possible accelerate implementation. (PRIORITY 2) 48.9.3.1 Participate in the INPO Human Performance Evaluation program. (PRIORITY 3) 48.9.3.2 Implement CRDR modifications (MOD 142). These modifications were identified in the District's submittal to the NRC in December, 1985, which documented the results of the Rancho Seco Control Room l

design review. (PRIORITY 3) '

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48.9.3.3 Review modifications incorporated following 12/26/85 event for impact and/or compliance with CRDR criteria / program. (PRIORITY 3) l l

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48.10 MANAGEMENT INFORMATION SYSTEM 48.10.1.1 Review implementation of NRC Generic l.etter 83-28 O

commitments and develop plan for program enhancement.

(PRIORIlY 1) 48,10.1.2 Provide site information system support for implementation and records management activities needed for restart. (PRIORITY 1) 48.10.2.1 Prepare a Nuclear Information Systems General Design, identifying and prioritizing improvement projects for implementation within three years. (PRIORITY 2) 48.10.2.2 Implement Vendor Data Program enhancements identified to achieve the program objectives. (PRIORITY 2) 4 8.10. 3.1. a . Complete Nuclear. Information Management System (NIMS) evaluation. (PRIORITY 3) 48.10. 3.1. b . Establish on-site facilities and organization to support NIMS hardware / software. (PRIORITY 3) 48.10. 3.1. c . Implement NIMS program / data. (PRIORITY 3) 9

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48.11 COMMIIMEN1 MANAGEMENT 48.11.1.1 Revise commitment management procedure to include tracking and compliance features. (PRIORITY 1) 48.11.1.2 Install a new commitment tracking system per 48.11.1.1. (PRIORITY 1) 48.11.1.3 Develop system / user documentation for the new commitment tracking system. (PRIORIfY 1) 48.11.1.4 Verify the commitment tracking ~ system database with respect to current known commitments. (PRIORITY 1) 48.11.1.5 Verify all commitments required prior to. restart are complete. (PRIORITY 1) 48.11.2.1 Verify the comprehensiveness of the commitment tracking system database. (PRIORITY 2) 48.11.2.2 Each Nuclear Department will establish specific milestones for reducing its backlog of open commitments. (PRIORIfY 2) 48.11.3.1 Integrate the commitment tracking system with the Nuclear Information Management System. (PRIORITY 3)

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48,12 CONFIGURA1 ION MANAGEMENT 48.12.1.1 Verify that Control Room drawings are current, in O

accordance with existing procedures. (PRIORITY 1) 48.12.1.2 Provide Nuclear Engineering support to plant operations to address and expedite configuration management issues. (PRIORITY 1) 48.12.2.1 Review and evaluate all temporary modifications and close out all existing abnormal tags that need to be converted to permanent plant modifications.

(PRIORITY 2) 48.12.2.2 Reduce the backlog of DCNs. (PRIORITY 21 48.12.2.3 Develop the System Design Basis documents for NEP manuals on 80P related systems. (PRIORITY 2) 48.12.3.1 Establish management direction for a Configuration Management program for Rancho Seco consisting of:

(PRIORITY 3)

I 48.12.3.1.a. Policy 48.12. 3.1. b . Specifications 4 8.12. 3.1. c . Computer hardware and software 48.12.3.1.d. Implementing procedures l 48.12.3.1.e. Training ,

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48.12.3.2. Establish or upgrade existing equipment and supporting r's

( ) documentation identification systems needed for total

plant configuration control. (PRIORITY 3) 48.12.3.3 Upgrade change control packages that control modification from change request through close out.

1 This includes identification of all affected documentation such as procedures, training plans, and simulator upgrades. (PRIORIfY 3) 48.12.3.4 Reorganize draf ting into a design /draf ting organization. (PRIORITY 3) 48.12.3.5 Train Nuclear Engineers to utilize the designers from the design / drafting organization identified in l 48.12.3.4 to reduce the engineering work load.

(PRIORITY 3) l

, 48.12.3.6 Develop new, or simplify existing procedures to clearly define the review process for drawings.

(PRIORITY 3) 1 48.12.3.7 Conduct a cost / schedule review to determine whether a CAD system can be justified for Rancho Seco.

(PRIORITY 3) 4 I

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i 48.12.3.8 Develop or upgrade existing systems to pro *;ide verification that configuration docume station _ reflects the true hardware configuration. (P<IORITY 3) 48.12.3.9 Develop or upgrade existing systems to provide the status of all documentatior. and equipment in a timely and accurate manner. (Pt.IORITY 3) 48,12.*s.10 Develop a work package system for all facility changes.

(PRIORITY 3)

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- 48.13 MATERIALS MANAGEMENT i

i 48.13.1.1 ' Conduct a review of the current Materials Management program. (PRIORITY 2) i 48.13.1.2 Develop and implement an action plan to improve the

performance of the Materials Management program.

(PRIORITY 2) i I

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4C Pl. ANT MODIFICATIONS AND MAINTENANCE IMPROVEMENTS 4C.1 INTEGRATED CON 1ROL SfS1EM (ICS) AND IN1ERFACING SYSTEMS 4C.1.a.1.a.1 Improve reliability of ICS power supplies.

(PRIORITY 1) 4C.1.a.1.a.1.a. Provide dedicated AC Inverters.

4C.1.a.1.a.1.b. Provide larger DC capacity.

4C.1.a.1.a.2 Implement Upgrades (PRIORITY 1) 4C.I.a.1.a.2.a. Reduce MFWP runback rate to 25%/ min.

4C.1.a.1.a.2.b. Substitute for RC flow the RCP status, i.e., number of RCP's.

4C.1.a.1.a.2.c. Remove SU FW flow correction to the MFW flow signal.

4C.1.a.1.a.2.d. Replace modules with improved Bailey revised modules.

4C.1.a.1.a.2.e. Remove FW temp correction from total FW demand signal.

4C.1.a.1.a.2.f. Revise RCP Runback rate to 25%/ min.

4C.1.a.1.a.2.0. Revise asymetric CR0 Runback rate to 3%/ min.

40.1.a.1.a.2.h. Improve post maintenance testing and surveillance.

40.1.a.1.a.2.1. Replace all S1 and S2 switches with  ;

i new switches.

1

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4C.1.a.1.a.2.3. Rewire TCS Power Supply Monitor to delete the " Daisy chain."

4C.1.a.1.a.2.k. Test / Inspect all cabinet wirewraps and lugs.

4C.1.a.1.a.3 Inspect electrical terminations within ICS.

(PRIORITY 1) 40.1.a.1.b.1 Upgrade ICS annunciation to remove ambiguities.

(PRIORITY 1) 4C.1.a.1.b.2 Provide computer inputs for ABT ( AC power supply) status. (PRIORITY 1)

) 4C.1.a.1.b.3 Add open/close status lights for ADVs, TBV's, MFW valves. (PRIORITY 1) 1 4C.1.a.1.b.4 Tag instruments with ICS processed signals.

(PRIORITY 1) 4C.1.a.1.b.5 Provide first-out MT." trip monitor. (PRIORITY 1) 4C.1.a.1.b.6 Provide adequate parameter trending independent of ICS/NNI. (PRIORITY 1) 4C.1.a.1.b.7 Provide clear labels for all S1 and S2 switches.

(PRIORITY 1)

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4C.1.a.1.b.8 Add all hot shutdown related parameters to SPDS, independent of ICS/NNI. (PRIORITY 1) 4C . I . a .1. c .1 Review procedures for adequacy in event of ICS failure. l l

(PRIORITY 1) 4C.1.a.1.c.2 Review / adjust minimum MFP speed. (PRIORITY 1) ,

4C.1.a.1.c.3 Trip ICS power on loss of NNI-X, Y, or Z power.

(PRIORITY 1) ,

4C.1.a.1.c.4 Trip MFP's on loss of ICS control. (PRIORITY 1) 4C.1.a.1.c.S Provide independent control-grade backup AFW automatic level control. (PRIORITY 1) 40.1. a .1. c . 6 Provide Class 1 bottled air supply for ADV's, AFW, MFW, and SUFW valves. (PRIORIfY 1) 4C.1.a.1.c.7 Modify Auxiliary Steam Reducing Station to fall at setpoint. (PRIORITY 1) 4C.1.a.1.d.1 Review ICS restoration procedures for adequacy, correct as necessary. (PRIORITY 1) 4C.1.a.1.d.2 Evaluate (and correct as necessary) final control element position on ICS power restore. (PRIORIlY 1) 9

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4C.1.a.2.1 Develop and implement those design changes or r'N enhancements that were not required to be implemented prior to restart, consistent with their assigned I

priority. (PRIORITY 2) 4C.1.a.3.1 Actively participate in the B&W Owners Group efforts to upgrade or replace the Integrated Control System.

(PRIORITY 3) 4C.1.b.1.1 Provide controls in the Control Room from which the operator can operate the TBV's, and which will cause these valves to remain closed on loss of ICS (DC) power. Note: ADV's will be controlled by EFIC.

(PRIORITY 1)

O 4C.1.c.1.1 Determine the contribution the ICS Power Supply Monitor (PSM) had in December 26, 1985 transient.

(PRIORITY 1) 4C.1.c.1.2 Evaluate the potential improvement to ICS reliability if redundant PSM's are installed. (PRIORITY 1) 4C.1.c.1.3 Determine the potential benefits to be obtained through the installation of independent ICS PSM's.

(PRIORITY 1) l 4C.1.c.1.4 Implement design improvements or document O

Q justification for not implementing ICS PSM modifications. (PRIORITY 1)

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4C.1.d.1.1.a Provide a window for the status indication of ICS Trouble (fan failure, power supply failure).

(PRIORIlY 1) 4C.1.d.1.1.b Provide a window for the status indication of ICS failure (loss of DC buss). (PRIORITY 1) 4C.1.e.l.a Participate in B&W Owners Group efforts to perform a generic failure / consequence evaluation of the Model 820 ICS. This effort was initiated August 1, 1986.

(PRIORITY 2) 4C.1.e.2.a Evaluate results and applicability of B&W Owners Group evaluation findings and recommendations to Rancho Seco.

(PRIORITY 3) 4C.1.e.2.b Conduct supplemental evaluations as required to achieve Rancho Seco specific information. (PRIORITY 3) 4C.1.e.2.c Develop and implement applicable modifications to Rancho Seco ICS. (PRIORITY 3) 4C.1.f.1.a Evaluate need for DC Bus battery backup based on reliability of power supplies, recent modifications, l etc. (PRIORITY 3) 1 1

! 4C.1.f.1.b Evaluate Hand / Auto station backup power. (PRIORITY 3) 1 0

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, 4C.1.F.1.c Participate with B&W Owners Group to enhance ICS and  ;

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f related power supplies. (PRIORIIY 3) i l

4C.1.f.1.d Develop and implement ICS modifications identified as j' necessary during evaluations. (PRIORITY 3) i i

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4C.2 NON-NUCLEAR INSTRUMENTATION (NNI) 4C.2.a.1.a.1 Improve reliability of NNI power supplies.

O (PRIORITY 2) 40.2.a.l.a.2 Upgrade NNI Modules to latest B&W models. (PRIORITY 2) 4C.2.a.1.a.3 Update NNI drawings to correct discrepancies identified in the deterministic failure analysis and the District's ongoing efforts to upgrade plant performance. (PRIORITY 2) 4C.2.a.1.b.1 Upgrade NNI annunciation. (PRIORITY 2) 4C.2.a.1.b.2 Provide computer inputs for NNI ABT status.

(PRIORITY 2) l 4C.2.a.1.b.3 Tag Control Room instruments with relevant power or signal channel identification. (PRIORITY 2) l l

1 4C.2.a.1.b.4 Provide adequate NNI parameter trending in IDADS (including recorder replacement). (PRIORITY 2) 40.2.a.1.c.1 Provide automatic trip of ICS power on loss of NNI-X, I

Y, or Z power. (PRIORITY 2) 4C.2.a.1.c.2 Revise procedures for loss of NNI. (PRIORIlY 2) 9

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4C.2.a.1.c.3 Review NNI restoration procedures for adequacy in

. event of NNI failure. (PRIORITY 2) 4C.2.a.2.1 Actively participate in the B&W Owners Group efforts to upgrade or replace the Non-Nuclear Instrumentation equipment. (PRIORITY 3) 4C.2.b.1.1 Evaluate the potential improvement to NNI reliability if redundant PSM's are installed. (PRIORITY 1) 4C.2.b.1.2 Determine the potential benefits to be obtained through the installation of independent NNI PSM's.

(PRIORITY 1) 4C.2.b.1.3 Implement design improvements or document justification for not implementing NNI PSM i modification. (PRIORITY 1) 4C.2.b.l.4 Inspect electrical terminations within NNI cabinets.

(PRIORITY 1) 4C.2.c.l.1 Tag NNI-affected indicator / recorder. (PRIORITY 1) 4C.2.c.1.2 Trip ICS power on loss of NNI-X, -Y, or -Z power.

(PRIORITY 1)

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4C.2.c.1.3 Make modifications to ICS to provide automatic control of ICS/NNI power. (PRIORITY 1) 4C.2.c.1.4 Provide separate annunciator windows to indicate:

(PRIORITY 1) 4C.2.c.1.4.a loss of NNI-X (DC) 4C.2.c.1.4.b Loss of NNI-Y (DC) 4C.2.c.1.4.c Loss of NNI-Z (DC, switching supply) 4C.2.c.1.4.d NNI trouble (fan failure, single power supply failure)

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4C.3 FEEDWATER AND STEAM SYSTEMS O 4C.3.a.1.1 Install EFIC and provide a Control Room panel as an extension to the H1SS panel. (PRIORITY 1) i 4C.3.a.1.2 ' Implement control grade modifications to close TBV's on loss of ICS power. (PRIORITY 1) 4C.3.b.1.a Develop and implement plant modifications to the auxiliary steam controls to assure valve control on loss of ICS power. (PRIORI 1Y 1)

. 40.3.c.1.1 Review operation of Main feedwater Pumps and status of incorporating lessons learned from October 2, 1985 event. (PRIORITY 1) 4C.3.c.1.2 Retune the ICS to reduce Main Feedwater System contributions to reactor trips. (PRIORITY 1)-

4C.3.c.1.3 Validate setpoints and proper initiation / interface l

with Auxiliary feedwater System. (PRIORITY 1) 4C.3.c.2.1 The recommendations contained in the B&WOG Availability Committee Report, 47-1159449-00, "MFW Pump Trip Reduction Program Final Report" will be evaluated for applicability to Rancho Seco and

, implemented as appropriate. (PRIORITY 2)

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4C.3.d.1.1 Walkdown the Main Steam Lines and verify that each service connection, greater than two inch diameter, is provided with capability to Isolate from the Control Room. (PRIORITY 1) 4C.3.d.1.2 Tag Control Room switches to clearly indicate valves associated with A- or B-Steam Generators. (PRIORITY 1) 4C.3.d.1.3 Evaluate Main Steam Line Supports for effects of flooding. (PRIORITY 1) 4C.3.d.1.4 Increase IDADS sample frequency for Main Steam / Main feedwater Pa,rameters. (PRIORITY 1)

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1 4C.4 EMERGENCY DIESEL GENERATOR RELIABILITY s

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4C.4.1.1 Replace Emergency Diesel Generator Turbochargers.

(PRIORITY 1) I

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g 4C.4.2.1 - Evaluate performance history of the Bruce-GM Emergency Diesel Generators and develop recommendations for

' re71 ability improvement. (PRIORITY 2) ~

j 1 4C.4.2.2 Determine and implement identified emergency diesel 3 r  ;

generator mod \fications. '(PRIORITY ?)

vh 4C.4.2.3 . Enhance the emergency diesel generator preventive nx s maintenance program. (PRIORITY 2) -

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4C.5 REACTOR COOLANT SYSTEM AND PRESSURIZER 4C.S.a.1.1 Issue ECNs for new pressurizer relief valve discharge O

pioing supports and support modifications.

(PRIORITY 1) 4C.S.a.1.2 Inspect, reanalyze, and redesign (as required) ring structure anchoring pressurizer relief valve supports to Pressurizer. (PRIORITY 1) 40.5.a.1.3 Construct new pressurizer relief valve supports and modify existing supports and ring structure (if required). (PRIORITY 1) 4C.S.a.1.4 Construct pressurizer support structure modifications and modifications to existing work platforms required to resist new pipe support loads. (PRIORITY 1) 1 0

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4C.6. ENHANCE THE POST ACCIDENT SAMPLING SYS1EM (PASS) OPERABILITY l

, 4C.6.1.1 Complete the SCAS panel rebuild. (PRIORITY 1) 4C.6.1.2 Complete associated peripheral PASS equipment upgrades.

(PRIORITY 1) 4C.6.1.3 Document the compensating equipment in the environmental lab. (PRIORITY 1) 4C.6.1.4- Complete work required to solve H2 mon toring heat tracing problems. (PRIORITY 1) 4C.6.1.5 Revise operating procedures and complete training on g revised system and conduct system functional test.

(PRIORITY 1) 4C.6.2.1 Replace Dionex program controller. (PRIORITY 2) 4C.6.2.2 Install R-15044 Sample Dryers. (PRIORITY 2) 40.6.3.1 Complete PASS decay heat valve replacement during the Cycle 8 outage. (PRIORITY 3) i

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r 4C.7 ACTIONS TO ENHANCE CON 1ROL ROOM /TSC AND NSEB HVAC - OPERABILITY AND RELIABILITY 4C.7.1.1.a. Prepare and implement a detailed action plan / testing program to identify excessive Control Room noise source (s) and propose modifications. (PRIORITY 1) 4C.7.1.1.b. Develop and implement those design and procedural changes needed to reduce noise levels to allowable limits. (PRIORITY 1) 40.7.2.1 Evaluate and implement design changes if necessary to improve balancing capabilities. (PRIORITY 2) 4C.7.2.2 Develop and implement the changes to install flow meters to facilitate surveillance testing of the Control Room /TSC HVAC filter units. (PRIORITY 2) 4C.7.2.3 Develop and implement the changes identified as necessary to facilitate maintenance of Control Room /TSC HVAC equipment (i.e., replace Air Handler Unit access doors, modify the lube manifold to condenser fans, etc.) (PRIORITY 2) 0

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4C.7.2.4.a. Develop and implement the changes necessary to add dampers through the TSC ceiling. ~(PRIORITY 2) 4C.7.2.4.b. Develop and implement the changes necessary to upgrade dampers in the wall between Control Room and TSC.

(PRIORITY 2) 4C.7.2.5.a. Investigate flow control through filter units and recommend improvements. (PRIORIfY 2) 4C.7.2.6.a. Investigate replacement of existing essential HVAC compressor motors. (PRIORITY 2) 4C.7.2.7.a. Develop improved methods to adjust NSEB essential air handler air flow. (PRIORITY 2) .

.4C.7.2.8 If required, develop and implement design change to add a sample manifold to filter banks in each of two units. (PRIORITY 2)

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l 4C.8 INSTRUMENT AIR SYSlEM RELIABILITY 4C.8.1.1 Complete IAS system review to identify hardware O

1 modifications required to improve system reliability.

(PRIORITY 1) l 4C . 8.1. 2 Replace leaking letdown filter valve operators.

l (PRIORITY 1) 4C.8.1.3 Add diesel-driven air compressor. (PRIORITY 1) l 4C.8.1.4 Provide bottled air backup to critical valves.

(PRIORITY 1) 4 C . 8.1. 5 Perform IAS walkdown to identify additional air leaks and any P&ID discrepancies. (PRIORITY 1) 4C.8.1.6 Develop and initiate Priority 1 modifications identified during system review. (PRIORITY 1) 4C.8.2.1 Develop and implement Priority 2 and 3 modifications identified in IAS review. (PRIORITY 2)

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4C.9 REACTOR BUILDING PURGE FLOW RATE MEASUREMENTS i

4C.9.1.1 Perform engineering evaluation of flow measuring 1

system deficiencies and identify appropriate ll modifications. (PRIORITY 2) l 4C.9.1.2 Install and test modifications identified during engineering evaluation. (PRIORITY 2) i a

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40.10 FIRE PROTECTION SYS1 EMS 4C.10.a.1.1 Develop and implement modifications to provide for O

manual operator override of a trip of the Auxiliary Building ventilation fans. (PRIORITY 2) 4C.10.a.1.2 Upgrade the control logic and schematic diagrams for the fire protection system. (PRIORITY 2) 4C.10.a.1.3 Develop and implement appropriate upgrades to prevent spurious signals of fire alarm systems on power transients. (PRIORITY 2) 4C.10.b.1.1 Develop and implement fire alarm and HVAC panel upgrades to separate the control circuitry and equipment for the Train A and Train B Dampers.

(PRIORITY 2) 4C.10.c.1.1 Identify vital areas of potential impact due to leakage through floors following actuation of fire protection systems. (PRIORITY 2) 4C.10.c.1.2 Evaluate effect of impact of potential leakage on safe shutdown equipment. (PRIORITY 2) 4C.10.c.1.3 Inspect all vital electrical equipment areas for potential leakage paths. (PRIORITY 2)

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4C.10.c.1.4' Evaluate the results of the leakage inspection and identify recommended corrective actions. (PRIORITY 2) 4C .10. c .1. 5 Develop and implement changes necessary to address recommended corrective actions from leakage inspection 2

evaluation. (PRIORITY 2) i 4C.10.c.1.6 Review and upgrade as necessary preventative f maintenance procedures to maintain drain lines clear i

of obstructions. (PRIORITY 2) i i .

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4C.11 MOTOR OPERATED VALVES 4C.11.1.1 Refurbish the 100 Safety Related M0V's. (pgroggjy j)

O 4C.11.2.1 Refurbish the 63 Non-Safety Related NOV's.

(PRIORITY 2)

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. 4C.12 CRITICAL PUMPS FAILURE ON LOSS OF SUCTION 4C.12.1.1 Evaluate procedures and provide training to prevent i recurrence of loss of suction pump failure.

(PRIORITY 1) 4 4C .12. 2.1 Engineering is to_ review design philosophy for suction' valve interlocks and alarms on critical pumps and

identify appropriate modifications. (PRIORITY 2) 1 J

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4C.13 MAINTENANCE PROGRAMS AND ACTIONS Inventory Calibrated Test Equipment (C1E) and O

4C.13.1.1 calibrate and/or control use to prevent use of uncalibrated CTE. (PRIORITY 1) 4C.13.1.2 Assure current calibration of all in-plant instrumentation used in the performance of surveillance testing. (PRIORITY 1) 4C.13.1.3 Rework the makeup pump and return to service.

(PRIORITY 1) 4C.13.1.4 Complete the in-progress battery replacements (A, B, C, 0, E, F). (PRIORITY 1) 4 C .13.1. 5 Perform refueling interval surveillance of snubbers.

(PRIORITY 1) 40.13.1.6 Complete rework of terminations in the Bailey Cabinets in the Control Room (NNI/SFAS/RPS/ICS). (PRIORITY 1) 4C.13.1.7 Perform biennial Diesel Generator Inspection and replace turbochargers. (PRIORITY 1)

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4C.13.1.8 Define the critical items to be included in the PM program. (This is considered to be an accelerated a

, portion of the planned PM Program Upgrade.) As a t

minimum, this will include the Manual Limitorque Operated Valves (105), the Manual Non-Limitorque Operated Valves (135), other Manual Valves important to process flow control in Class 1 and steam generator heat removal applications (143), plant instrumentation required for surveillances, safety related HVAC and the Control Room normal HVAC system. (PRIORITY 1)

) 4C.13.1.9 Complete Preventive Maintenance (PMs) on manual valves i

! selected due to their functional position, e.g.,

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isolation of active equipment such as pumps, control valves, heat exchangers, cross-ties. (PRIORITY 1) 4C.13.1.10 Repair valves investigated as a part of December 26, 1985 event troubleshooting. Includes FV-20527, l FV-20528, FWS-063, FWS-064. (PRIORITY 1) l 4C.13.2.1 Develop a departmental procedure hierarchy and

! writer's guide for Maintenance Procedures.

i (PRIORITY 2) 1 j 4C.13.2.2 Identify and prioritize maintenance procedures for 1

l generation and/or revision. (PRIORITY 2) f I

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4C.13.2.3 Achieve authorized staffing levels within the maintenance organizations. (PRIORITY 2) 4C.13.2.4 Develop and/or revise the required programmatic procedures for the PM program to: assign responsibilities, authority and accountabilities for the program; establish criteria and define the scope of the program; and define the interface with other work control processes. (PRIORITY 2) 4C.13.2.5 Review existing PM tasks and frequency for critical equipment. Revise and augment as required by programmatic selection criteria. (PRIORITY 2)

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4C.14 ONCE THROUGH STEAM GENERATORS (OTSG's) 4C.14.1.1 Complete Helium Leak Test of both OTSG's. (PRIOR!lY 1) l 4C .14.1. 2 Perform Eddy-Current inspection of OTSG's to recommend tubes for plugging. (PRIORITY 1) 4C .14.1. 3 Plug those tubes identified in 1 and 2 above. ,

(PRIORITY 1) 4C.14.1.4 Develop program and licensing documents necessary to sleeve tubes in lane region of OTSG's. (PRIORITY 1) 4C .14.1. 5 Install sleeves in selected tubes. (PRIORITY 1)

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I APPENDIX E i . SAMPLE PORTION OF i

ACTION PLAN 1 ,

ACTIVITY TRACKING REPORT l!

3 July 1986 1

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REEDMMEuGATIDN LIST Column Title Bescrintion talumn Title Descrintier.

Log No. RRR$ identifier defined as: Val /Inval defined as:

15.XXXX 5ection 15. 12/26/85 Transient IR - Invalid-Redundant Action List Recommendation

, 16.XXXX NRC Region V Recommended Action List 10 - Invalid for other reasons i 17.XXXX NUREG 0667 and B&W V - Valid 18.XXXX NUREG 1195 19.XXXX Selected Projects 20.XXXX Precurser review 21.XXXX Plant Staff Interviews Disposition Organization assigned by RRB to l

, 22.XXXX Deterministic Failure Consecuences Organization investigate and dispose of 21.XXXX BWOG Stop Trip Program proposed recommendation:

,' 24.XXXX RRRS Otservations CH - Chemistry (these items are not reconnendations but EM - Electrical Maintenance j , but may lead to recommen$ations.) proposed recommendation: ,

i 25.XXXX Dept. Managers Recommendations IC - I and C Maintenance LI - Licensing NE - Nuclear Engineering Recommendation Brief description of proposed reconnendations NO - Nuclear Operations RC - Regulatory Compliance RP = Radiation Protection Inttiator Person initiating recommendation TR - Training TS - Technical Support

• Disp Eng 01spositioning Engineer

$Y1 System 1 letter identifier of applicable Due Date Date dispositton is eue to RRRS plant system. See AP.3. Enclosure 6.4 from Disp. Eng.

! (additional desip ation for this report *** incidates disposition (if incluee a MPS Multiple Plant Systems: MGT - reeutred) completed.

Management NSA - No Systee App 11able)

\s,, feh Cat Defined as RRR8 recommendation for:

SU - Startup Cat Category - Area affected by recommendation N$ - Non Startup May inclues several areas and are eefined: unesterminced Status Results of RR8 review eefined as:

1 NA - NOT APPLICABLE PAG - Recommendation has been DC - PLANT M001FICATIONS forwarded to Performance MP - MAINTENANCE Analysis Group LL - LESSONS LEARNED REC - Recommendation has been TR - TRAINING returned for Clarification

! OP - OPERATIONS ANO PROCEDURES OSP - Recessendation forwareed to

EP - EMERGENCY PREPAREONES$ Dispositioning Dody

! 00 - DUALITY An0 OUALITY A$$URANCE ME - MANAGEMENT EFFECTIVENES$ Acted on Date Date recommendation reviewed by CM - COP 911TMENT MANAGEMENT Cassents RRRB C0 - CONFIGURATION MANAGEMENT RRR$ Comments HP - HEALTH PHYSICS RO - RECORDS ANO DATA BASE HM - MATERIAL MANAGEMENT MANAGEMENT E-1

Pete No. 1 07/81/06 MS MIDeGEATIlpt LIST 15500 E019I1951 WEUITM STRTIGI telIT M

• • • 4*Me* IECBEER 26,1905 TIUBEIDIT **********

lls WIL/ DISpmi1TIGl Ilm E meB SATIS S SYS DIT IIML OMMIZATISI pitIORITY STimE ETED WI 2.

15.0049 IEIJEL 31EMBE S1/52. IITDI2B, ICS E V DERIIEi 1 PAB 86/02/06 19501 FETGIB).

PIU/IE PppMpt! ATE SIZE PlWTECTIVE ICS DC V DERIlO 1 PAS 06/02/06 15.0108 FE FOR GE fLTDI0lTIIEi QNENT 10 FED LOAD CIIDlIT IUsilldi IETIEEN IllTEBlWITED CWmEL SYSTBI 0811ET H41CH2 AIS CGER.E H1R1.

DEUE PlWpER CMDIlslTIsl ETIEDI

.TE FUliE GIS UIETIGUI MMBL ILSO ~~

DEUE PIlopal C005115tT10ll IETIEDI DISTIldi FLES Ale LPSTIElWI PlWTECTIDIL (ETIst ITDI 3.F.8 CLDEUIE IEp0RT).

ICS DC V DSRIldi 1 PAS 86/03/06 15.0245 IERIFY M PUUIT TO EPUE M UISTIIEi 30A Its S1J IDi FEDER NEMER WITH A 40A 31EMER (DSIIEERIls DDE IETIE IH469).

(ETIBl ITDI 3.B.3, CLIBE IEPORT).

ICS V DEIRIIEi 1 Pali 06/13/06 15.0247-1 PEIFulWI IUll IEEllED BY EDI DC IH359. D5llEBtIIEi 05NE IETIE IH359 DREES M ELECTRIDL

!Ellillei SIMEES FDR M PGER SupFLY IENITORS IN M ICS 1881911 CA0!!ETS. (ACTI(M ITDI 3.F.6, CLoliUIE EPORT).

I 18.0045 ERUATE TE LEE (F A SIDEE PSER ICS DC V BERIIEi i Pali 06/16/06 ElpplY IOl! TOR F0lt ICS, IE M SIISI follTDit LElWES TE SYSTDI E-2

Pep No. 2 r//st a M MIDOGSITim LIST Nace EtO lansa eDeulTim Sw!W ImIT OE

• • • • • • • se IEnlER 5,1M5 TNESIEllT **e+eeeeee LE Vit/ DISptBITISI DE MIDeseRTISE SYS DIT IIML OMNIIZATIGl pilIORITY STATUS ACIED W is.

VIDENEZ TO SIIEE FAILUE.

OPERTISE BOLD IEBIFY ICS 19 V OPENITItBE 1 PAS 86/16/06 15.0857 Applupit!E panmusum Em 1891 IETIl31 ITEll 1.J les 3.L1, CURE EftRT. (10186-1271. M E-EleEIZATItBl STATUS F SPECIFIC ICS EElpleti IS lEED TD IERIFY M ICS IEDNEllY PIEB1K.

~ .

A DENLTY PIEDIE BGLD E ICS OP Y OpDulTISE 1 pilS E/27/06 15.015B IEYELDIED 10 PENIE RENIEZ TO M OPENITORS GI LIM F ICS IDEL THIS panmalir BGLD E IRITTEll i

TO DSNE LM (R L85, IE AppLICAILE, 151YE EBI CGIILEID TO J

M FOIllT Tigli IES plEBMS Its TDesumi1ES AE STMILIIED 150 LISER DDOSTRED OPElumR CGITIEL IEFOIE ICS TIRELISEDTIE IS (DeeCEB. M PEEDIE RR ICS IOER IESTORATIR 9GLD STIEMi (lOSINS S1/S2 TIRETER ICT EPANITE.Y. (IE/ stb IEETIIEi 2/10/86). IEEIFY LIB 815.01TT TO E PAlli F THIS IEIDOGSATISt.

1 (ACTION 1791 te, CLIERNE IEp0lli, j

ECTIGI Ell.

l ICS DC, V OleulTISEi 1 Pali 85/3B/86 15.0083 LIES F IllTHilulTED CSmEL SYSTEll l

PRIBINES SQLD DIIECT M (F l

(FElWITOR 70 PUllZ M ATIEERERIC Clflp VILVE IIS TLNDIE BYpIEi VILVE I

i E-3 i

- - _ _ - _ . - _ _ _ _ _ . _ _ - . _ . _ . . . _ _ - . - - - - - - ~ _ - _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

page No. 3 87/t1/86 M IEIDOEIEIRTIGl LIST NUDEI !E!!! IED. ERR ElEMTIls STRTIGI INIT GE eMeMeee IEEER 26,1985 TNUEIENT *eeeeeeeee VIEJ DISIMITICBI DATE UE STATLS ACTED (Il IEIDOEIGRTIGEi SYS Cili IIME. OMANIZATIGl MIORITY lEL _

AUTII1!E BIITDES IN M *!1DE' IEEE, fue M ELIARY FEED nEITER

• L Als N FLCW CtBITIEL (DITIELLEllS IN .

• 188UE.' MFDIE ATTEWi!Idi 70 IE-ESTIE.ISI IDER 70 M INTEBNITED CGITIEL SYSTBL GCE INTEBARTED CDITIEL SYSB IDElt IS STABLE, M VII.YES CIDI E EftNIED TD M INTEliRRTED (DfilllL SYSB BY GO!!Ei TO *l05ft." fue 'AUTD*

IESPECTIVELY.

W V OPERRTICIEi I PR6 06/02/86 EMIR M IWlEIPEILY TBDlilflTED ICS 15.0287 WIIE IN M INTEBUITED CDmEL SYSB CleMT. (EDIPLETD (ACTIGl ITEM 2A N CM EPORT). .

ICS W 0 OPEMTIGEi I PR6 86/0e/86 15.0298 IERLVE M CE F M DIFFUElEE IN M 'A6 FOSS' VII.LES F M TIIE DELAYS Of SI IIS 52 (ICS PGER SPP.Y IOIITGI SWITDES)

WITH M DES!91 Vit.1ES. (ACTIGi ITEM 2A lEXIT CitEE IEPGril.

l l

l V TMINIlii  ! PR6 85/3B/86 l ICS TR 5.0001 IIEll2E ' lolls E INTEBMTED CGITIEL SYSTBf' IN M TMINDEi PIEMllutt f

THIS TMINDti WILL AID M OPBUITDilS DURIIEi A 'pARTIll LIES F .

INBMs cDmoi. SYSTBr M Eu.

AS TUTIE. LDBS F INTIEBMTED CDfTIEL SYSTEM CABult. TIES. (ITEM IC, ACTIGI LIST).

E-4

pagello. 4 i 87/01/ 3 M MIDGEERTIGI LIST MIOC ED IIII.But EleulTIIS STRTItBI IIIIT (BE

• • • • • • • • e IEEBEER 5,1985 TWISIElff *****e***e O UE E. IEDOEIERTIGE SYS DIT Vill DISREITIGI IBML DEMIZATIGl pit!ORITY STimE DATE ACTED k 15.8898 EVII.lNITE M MILIMY STBul ICS, K V EletIIEi 1 PAS 06/02/06 REDUCIE VM.VE FISCTIM FW UBS F MC IllTERIED CGmEL SYSTBI E POIER fig eassinsur IggEILITY 10 (DITEL MILIARY STSUI PLEBE fue IIEUISIT COMECTM ETIGl 15 IECESSARY. (ACTIM ITBI 1.J IIS 3.L1 CLIBulE lEPORT).

15.0099.A plENIE A B CGITEL STATIM Ill ICS, E V BER! lei 1 PAS 86/02/86 M (DITEL EDI GI H1R110: ft.ull 15 famiglTIC CURIE F M .

ImEERERIC Dlas Vit.VES/TUBIIE BYpIWiVit.VES.

15.0099.B pleVIE A lel CGITEL STitTIGI Ill ICS, E V Battlei i plE e5/3B/86 M IDmEL EDI (Il H1R1 TO: lei p2 VIE nit 3L CURE F TIE IWYS/TNI.

15.II59.C plWV!E A IEW (DmEL STATIGI Ill ICS, K V EIERIE 1 PAS 86/02/86 TIE (DmEL MBI Gl H1R1 TO: 10 lei OVElutIE TIE EDIITIC CUENE F M fmEERetIC M5p VILVES/TimilE BYplEE VILES 188 ft.LIN M .

OpBumR TO Get M ATIEEplERIC Map Vit.YES Olt TIBBIIE BYplWi Vit.VES (2 PBCENT OpBI) lef!LE INTElWWITED CGmEL SYSTBI IOER IS LOST. OCTIGl ITBI 3.F.4 IIS 5 CUEUlk IEPORT, IEFUEEE BEillEBtIE DEUEZ IETIE IH357 A/B).

3.0009 IIGi STOP-TR!p plE E UWI TRIP ICS, 19 V IglIlff 1 //

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Page No. 5 57/01/ 3 M ECINEIORTIGl LIST 18800 ! ECD ltELElut IEIElulTIIS STRTIM LIET GE

• • • • • • • +e DEIBWt 26,1985 TIUISIEIR **********

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23.0008 IWO6 STOP-TRIP Pil0BNIl TRIP IERTICBI IECDOEleRTIWI TH06-ICS 1911

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23.0001 M E FLOW Il#UTS TO M ICS ICS, DC V EISRIIEi 2 //

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ICS, DC V DEIRIIEi 2 23.0007 506 STDP-TRIP PIEElWWI TRIP 1911 IEL'.TIGI IECGIGSATION T H 87-ICS. IEleWE ETU LIMITS FIDI ICS. //

ICS, DC V DEIRIIEi 2 23.0010 826 STIT-TRIP PIEBlWII TRIP IEELETI(31 IECSOBERTIGl 10l!

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APPENDII G CONTROL ~ ROOM / TECHNICAL SUPPORT CENTER ,

ESSEN AL AIR CONDITIONING SYSTEM

~

-SYSTEM ST6?'S REPORT l

, PREPARED BY: Bob Thomas M Date: 7-3-86 1-

' Revisi6n 0 ,

Et't'ective [tevision 'ODate i

INTERIM AP VALS:

l A2]f) S 94 "

Data Pert'ormancBgysis(3roup

d j fff/P4 -

Depu MahaieF, Nticlek, / Dad ,

Date TEs iwo APPROVAL.

SYSTEM ENGINEER l

Date TEST REVIEW GROUP Date __

i

• Date RESTART APPROVALS:

SYSTEM ENGINEER Date _

LEAD SYSTEM ENGINEER '

Date -

PERFORMANCE ANALYSIS GROUo Data ~ *

~

DEPUTY GENERAL MANAGER, NUCLEAR APPROVED AS REV. 0 ONLY j REV. 1 MATERIAL PROVIDED FOR INFORMATION ONLY G-1 l

_ ~ - - _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _

CONTROL ROOM / TECHNICAL SUPPORT CENTER ESSENTIAL AIR CONDITIONING SYSTEM SYSTEM STATUS REPORT O

G-2

CONTROL ROOM / TECHNICAL SUPPORT CENTER

^

ESSENTIAL AIR CONDITIONING SYSTEM SYSTEM STATUS REPORT PREPARED BY: Bob Thomas /Jeff Naleway Date: 08-04-86 Revision 1 08-07-86 Effective Revision 1 Date INTERIM APPROVALS:

Performance Analysis Group Date Deputy General Manager, Nuclear Date TESTING APPROVAL: Date SYSTEM ENGINEER Date TEST REVIEW GROUP Date RESTART APPROVALS: Date SYSTEM ENGINEER Date LEAD SYSTEM ENGINEER Date PERFORMANCE ANALYSIS GROUP Date DEPUTY GENERAL MANAGER, NUCLEAR G-3

...._. -. . . . -- - - - - - . . . . . . . . ~ . . . . _ . - . - . , . . . - . _ _ . - . - _ . . _ _ - . . - . . -

5 TABLE OF CONTENTS SUBJECT PAGE

1. EXECUTIVE

SUMMARY

1 -1

2. SYSTEM FUNCTIONAL DESCRIPTION 2-1 A. GENERAL DESCRIPTION -

8. FUNCTION / SOURCE CROSS REFERENCE C. SOURCES

3. TEST REVIEW 3-1

4. REVIEW OF RECOMMENDATIONS AND RESOLUTIONS 4-1 A. DISCUSSION B. INDEX OF PROBLEMS

5. REVIEW OF TEST RESULTS FROM DAVIS-BESSE 5-1

6. REVIEW OF MAINTENANCE HISTORY 6-1

7. SYSTEM INSPECTION 7-1 O

l i

l O

G-4 l

s EXECUTIVE

SUMMARY

l. GENERAL REPORT DESCRIPTION The Restart System Review and test program is explained in quality control instruction (QCI) 12. Attachment 11. QCI - 12 lists the selected systems for which System Status Reports are generated. The Control Room / Technical Support Center (CR/TSC) Essential Air Conditioning System is a selected system.

This report documents the information on the CR/TSC Essential Air Conditioning System to verify the System is suitable for return to power operation.

To expedite work, sections of the Status Report will be submitted to the Performance Analysis Group (PAG) for interim approval as the sections are completed or updated.

2. SYSTEM FUNCTION DESCRIPTION A. General Functional Description -

The Control Room / Technical Support Center (CR/TSC) Essential Air Conditioning System is designed to provide a suitable environment for equipment and station operator comfort and safety.

During certain abnormal events, as noted below, the CR/TSC Essential Air System is automatically actuated and started. The CR/TSC Essential-Air Conditioning System is comprised of two 100% capacity redundant trains. Each train is comprised of the following:

Essential Filtration Unit consisting of a moisture elimin3. tor, electric duct heater coil, two HEPA filter banks, two carbon filter banks, and a booster fan.

Essential Air eindler consisting of a medium efficiency filter bank, direct cxpansicae cooling coil, and circulation fan.

Essential Condrosing Unit consisting of a reciprocating refrigerant compressor, air cooled condensing coils, condensing fans, receiver,and associated ref tigarant piping and valves.

Fourteen dampers and associated ducts.

The CR/TSC Essential Air System performs the following functions:

1. Isolation of the CR/TSC from potentially radiologically contaminated air during a radiological event by automatically closing sixteen normal air conditioning system isolation dampers (eight isolation dampers are actuated by each train) and opening twelve essential air handler isolation dampers (six are actuated by each train).

G-5

(Continued)

2. Isolation of the CR/TSC from air potentially containing toxic gas during a toxic gas event by automatically closing eighteen isolation dampers (nine isolation dampers are actuated by each train) and opening ten essential air handler isolation dampers (five are actuated by each train).

3. Provide cooling for the CR/TSC during radiological, toxic gas, CR/TSC high temperature, or loss of of fsite power events by maintaining the CR/TSC temperature at 80*F or less.

4. Prevent infiltration of potentially radiologically contaminated air into the CR/TSC during radiological events by pressurizing the CR/TSC to 0.125 IWG relative to outside atmosphere.

5. Provide fresh, filtered, conditioned supply air for CR/TSC ventilation during radiological and loss of offsite power events by maintaining a flow of 3,200 cfm through the Essential Filtration Unit.

3. REVIEW PROBLEM / RESOLUTION

.1 General Discussion The problems identified for the CR/TSC Essential as Conditioning System 1 as of July 30, 1986 have been classified into four categories: l Modifications, Administrative (Training, Procedures, Drawings), Studies, '

and Maintenance. Figure 1, CR/TSC Essential Air Conditioning Report Scope / Sequence Chart, is a graphical representation of the problems.

.2 Approval Problems / Resolutions

.1 Reference time meters to facilitate carbon filter testing after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> elapsed operating time should be installed.

(Priority 2)

.2 Test Agent Injection Manifolds necessary for inplace filter leak testing must be designed, installed, and tested. (Priority 1)

.3 Modifications to~ reduce Control Room noise generated by Essential Air System operation must be designed and accomplished. (Priority 1)

.4 Vibration of components in the Condensing Units should be evaluated. Modifications may be necessary. (Priority 1)

.5 The Normal Air System low air flow switch should be relocated from the cold air supply duct to the return air duct. (Priority 2)

.6 Pressure taps to enable measurement of Technical Support Center pressure relative to the atmosphere should be designed and installed. (Priority 2)

.10 Controls for condenser fan cycling to facilitate system operation during cold weather must be designed and installed. (Priority 1)

O G-6

.~ -

.11 Modifications to the Refrigeration System to facilitate maintenance

, must be designed and installed. (Priority 1)

.12 Fire dampers should be installed between the Control Room and the l Technical Support Center. (Priority 2) j

.2 Maintenance (13 thru 23,25,26)

.13 Hi temperature actuation of the Essential Air System must be l investigated and resolved. (Priority 1)

.14 The Control Room Normal Air distribution system must be balanced to ensure uniform cooling. (Priority 1)

.15 Actuation of the Essential Air System by the Chlorine Detectors must

be investigated and resolved. (Priority 1)

.16 Infiltration thru the Essential Air Handler access door must be resolved. (Priority 1)

.17 Essential Air-System isolation dw;drs are not functioning properly. (Priority 1)

.18 The CR/TSC Toilet Area Exhaust Dawar operation must be investigated. (Priority 1)

.19 The Shif t Supervisor exhaust damper operation must be investigated.

(Priority 1)

)O

.20 The Control Room Pressurization Problem must be resolved.

(Priority 1)

.21 Operation of the CR/TSC Essential Air Refrigerant System is compromised by certain piping and control problems. (Priority 1)

.22 Th'e Control Room does'not appear to be uniformly cooled by operation of the Essential Air System. (Priority 1)

.23 Air flow thru the Essential Air Filtration Units may be excessive for several minutes after the systems are actuated. (Priority 1)

.25 The flow transmitter for Essential Air Train A must be calibrated.

(Priority 1)

.26 The flow transmitter for Essential Air Train B must be calibrated.

(Priority 1)

.3 Administration (24,27 thru 31)

.24 A Flow Transmitter Calibration Procedure must be prepared for calibration of Essential Filtration Unit flow transmitters.

(Priority 1)

O G-7 l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ _ - . .- - - -. . . . . - . _ - _ - . - - - - . - _ . _ _ .

.27 Operating Procedure A.14 must be revised to provide operators with more complete and detailed instructions for operating the CR/TSC Essential Air System. (Priority 1)

.28 Surveillance Procedures must be revised to assure that damper function problems are not masked. (Priority 1)

.29 Procedures for accomplishing routine and non-routine maintenance tasks should be prepared. (Priority 2) l

.30 Additional Training is necessary to enhance operator knowledge of l the CR/TSC Essential Air System. (Priority 1) l

.4 Studies (32 thru 39)

.32 Carbon filter tray mounting in the Essential Filtration Units should be evaluated for improvement to facilitate maintenance. (Priority 2)

.33 The affect of a freon leak in the Air Handler evaporator coil upon l CR/TSC occupants should be studied. (Priority 2) -

.34 The long term affects of diesel exhaust upon the Essential Filtration Unit carbon filter should be evaluated. (. Priority 2)

.35 The affect of CR/TSC Essential Filtration Unit initial flow rate on CR/TSC radiation exposure must be evaluated. (Priority 1)

.36 The affect of Essential Air System Safety Features Actuation upon Control Room operator work load should be evaluated. (Priority 2)

.37 Essential Air System operating status indication in the Control Room must be evaluated. (Priority 1)

.38 Essential System Compressor Compartment temperature curing hot weather operation must be evaluated. (Priority 1)

.39 Suitability of the Compressor Motors for the application s'toold be evaluated. (Priority 1)

.3 Problems / Resolutions for PAG Approval l

.1 Modifications (7,8,9) l l .7 The CR/TSC Radiation Monitors which actuate the Essential Air System l must be modified to prevent inadvertant actuation. (Priority 1)

.8 Power Supplies for the Essential Unit Controls and instruments should be revised. (Priority 2)

.9 Arrangement of the Refrigeration System components should be changed to facilitate maintenance. (Priority 2) i O

G-8 i

.2 Naintenance

. No additional problems for approval.

3

.3 Administration

( No additional problems for approval.

.4 Studies

^

No additional problems for approval. ,

~

, .4 PROGRESS (August 8, 1986) 7,

.1 -Nodifications -

g, -

< :s ri .

.1 Engineering Change Notice (ECN) A-5782 has been~. initiated to install reference time meters. (Problem 4-1) 4 s

. s . .

.2 ECN R-0788 has been initiated to install Test' Agent In.jection / j Nanifolds. Purchase Request 43079 is being processed tcipurchase the necessary hardware. (Problem 4-2) \ w; -

<2 s 1

.3 An evaluation of Control Room noise levels has been completed. A T reduction of Essential System air flow has been proposed to reduce *"

noise level. ' Preparations are being made to obtain sound measurements with the' Essential System flow reduced to 80% of the '

present flow. (Problem 4-3). ,,

.4 Both Trains of the Essential Air System have been operated to obtain.- -'

data for engineering evaluation of condensing Unit component 2/

vibration, Compressor Compartment temperatures, and uniformily of ,; 'a Control Room cooling. (Problems 4-4,4-13,4-22,4-38) i c

. N -( .

.5 ECN R-0314 has been initiated to relocate the Control Room Norrdi ..

Air System Flow Switch. (Problem 4-5) 3 3 s i

.6 ECN R-0182 and ECN R-0164 have been initiated to install ra'diatioh ~ .

monitor ground cables and change the radiation actuation of'the _ ,

l Essential System. (Problem 4-7) -

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.7 ECN R-0796 has been initiated to provide a newpcwer supply for , e  :

illuminating lights. (Problem 4-8) ,

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.8 ECN R-0789 has been initiated to provide isolation valves fod j D compressor instruments to facilitate calibration. (Problec 4-9) 1

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.9 ECN R-0769 has been initiated to provide controls for condenter' cycling. (Problem 4-10) 1 ,

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.10 ECN R-0764 and ECN R-0763 have been initiated to ads \ fire damperr in .

Essential System duct work and upgrade fire damper"pn Normal System duct work respectively. (Problem 4-12) iI /

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.11 Work Request Number (WRN) 115248 has a been initiated to obtain data which will be used for part of the evaluation of the Normal Air ~

Distribution System. (Problem 4-13 and 4-14)

.12 WRN 116218 has been initiated to investigate Chlorine Detector actuation problems. (Problem 4-15)

.13 WRN 116217 has been initiated to eliminate Essential Air Handler access door infiltration. (Problem 4-16)

.14 WRN 114662 and WRN 114663 have been written to investigate isolation damper problems. The majority of the dampers have been checked for proper function. Four dampers exhibiting functional discrepancies have been identified. (Problem 4-17,4-18,4-19)

.15 WRN 115773 has been initiated to inspect the CR/TSC pressure boundaries. (Problem 4-20)

.16 Special Test Procedure STP 1002 has been accomplished to obtain data -

for evaluating the cooling capability of the Essential Air System.

(Problem 4-22)

.17 A Maintenance Procedure for cleaning and inspection of CR/TSC P.ssential Air System Dampers has been prepared and will be implemented during the week of August 11. (Problem 4-17,4-18,4-19,4-29)

.18 A detailed Project II Schedule for bringing the system to " System

. Ready" status has been prepared.

.19 Control Room temperature profile data, based upon operation of the Normal Air Conditioning System, has been obtained. (Problem 4-13)

O G-10

SYSTEM FUNCTION DESCRIPTION g A. General Functional Description -

The Ccntrol Room / Technical Support Center (CR/TSC) Essential Air Conditioning System is designed to provide a suitable environment for equipment and station operator comfort and safety.

During certain abnormal events, as noted below, the CR/TSC Essential Air System is automatically actuated and started. The CR/TSC Essential Air Conditioning System is comprised of two 100% capacity redundant trains. Each train is comprised of the following:

Essential Filtration Unit consisting of a moisture eliminator, electric duct heater coil, two HEPA filter banks, two carbon filter banks, and a booster fan.

Essential Air Handler consisting of a medium efficiency filter bank, direct expansion cooling coil, and circulation fan.

Essential Condensing Unit consist.ng of a reciprocating refrigerant -

compressor, air cooled condensing coils, condensing fans, receiver,and associated refrigerant piping and valves.

Fourteen dampers and associated ducts.

The CR/TSC Essential Air System performs the following functions: i

1. Isolation of the CR/TSC from potentially radiologically contaminated air during a radiological event by automatically closing sixteen normal air

( conditioning system isolation dampers (eight isolation dampers are actuated by each train) and opening twelve essential air handler isolation dampers (six are actuated by each train). Each Essential Air System Train is actuated at 10,000 1000 CPM.

2. Isolation of the CR/TSC from air potentially containing toxic gas during a toxic gas event by automatically closing eighteen isolation dampers l

(nine isolation dampers are actuated by each train) and opening ten essential air handler isolation dampers (five are actuated by each train). Each Essential Air System is actuated by 1 PPM chlorine gas.

3. Provide cooling for the CR/TSC during radiological, toxic gas, CR/TSC high temperature, or loss of offsite power events by maintaining the CR/TSC temperature at at temperature of 80*F or le:s. Ex:h Essential Air System train is actuated at 80 1.2F Control Room air tenperature.

4. Prevent infiltration of potentially radiologically contaminated air into the CR/TSC during radiological events by pressurizing the CR/TSC to 0.125 IWG relative to outside atmosphere. 1600 CFM outside air is used to l

pressurize the Control Room /TSC.

5. Provide fresh, filtered, conditioned supply air for CR/TSC ventilation

, during radiological and loss of offsite power events by maintaining a flow of 3,200 cfm through the Essential Filtration Unit (Maximum cooling capacity is 1,188 MBH).

G-11

B. REFERENCE DOCUMENTS REVIEWED

1. Control Room / Technical Support Center Habitability Study

2. Rancho Seco Technical Specifications 3.13 and 4.10

3. Engineering Change Notice A-3920

4. Engineering Change Notice A-3660

5. Ellis and Watts CR/TSC Essential Air Conditioning System Operating and Maintenance Manuals, M13.15-103

6. Updated Safety Analysis Report Sections

7. Surveillance Procedure SP 211.01 A, SP 211.018, SP 211.01C, SP 211.010, SP 211.01E, SP 211.01F Records

8. Troubleshooting Action Plan 1.R.1 and 1.R.2 .

9. ANSI N 509

10. ANSI N 510

11. ERDA 76-21

12. American Air Filter Air Filtration Units TSC and Control Room; M13.16-IM01

13. US NP.C Reg Guide 1.52

14. US NRC Nureg 0737

15. Special Test Procedures STP-189, STP-194, STP-198

16. Non-Conformance Reports

17. Maintenance Information Management System

18. Plant Operating Procedure A.14 O

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RESULTS TESTING HISTORY REVIEW TEST I. FUNCTION AND TEST REVIEW FUNCTION Isolation of the CR/TSC during a STP-162 New Essential HVAC Systam for

- radiological event by automatically the Control Room and TSC (Acceptance closing suction Normal Air Test). This test demonstrated and Conditioning System isolation dampers recorded the status of control, (eight per train) and opening twelve indication,-and position of all essential air handler isolation equipment associated with each train dampers (see per train). of the Essential HVAC System upon actuation in the various modes.

Included is a position and position indication check of dampers upon actuation in the radiological mode.

Isolation of the CR/TSC from air STP-162 Described above. Included is potent 1 ally containing toxic gas a position and position indication m

. during a toxic gas event by check of dampers upon actuation in the automatically closing eighteen toxic gas mode.

isolation dampers (nine per train) and opening ten essential air handler

-isolation dampers (five per train).

Provide cooling for the CR/TSC during STP-175 Control Room Temperature, radiological, toxic gas, CR/TSC high Humidity, and Air Velocity Survey.

temperature, or loss of offsite power This test was performed to measure and record the temperature, humidity, and D) events by maintaining the CR/TSC temperature at 80*F or less. air velocity in the Control Room while operating the HVAC system under specified parameters.

Prevent infilt. ration of potentially STP 164 Air Balancing Test Procedure radiologically contaminated air into Aux. Bldg. Essential Air Handling the CR/TSC during radiological events Units. This test was performed to by pressurizing the CR/TSC to 0.125 balance air flow to design IWG relative to outside atmosphere. specifications. CR/TSC pressurization to specified levels was demonstrated.

(

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FUNCTION TEST Provide fresh, filtered, conditioned STF 140A, SF-A-7A airflow capacity, supply air for CR/TSC ventilation and STP 1408, SF-A-78 airflow during radiological and loss of capacity, verified that specified offsite power events by maintaining a volume toxic airflow rates thru the flow of 3200 CFM thru the Essential Essential Filtration Units could be Filtration Unit. achieved at the minimum and maximum filter differtial pressures.

STP-164 Air Balancing Test procedure was conducted to balance air flow in the Control Room and TSC.

STP-173 In Place Leak Test procedure for HEPA filter banks demonstrated the integrity of the HEPA filter bank.

STP-174 In Place Leak Test procedure adsorber stage demonstrated the -

integrity of the carbon adsorber stages.

O O

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II. FUNCTIONAL TEST PROCEDURE LISTING The following special test procedures have heen conducted on the new essentiar H.V.A.C. System for the Control Room and Technical Support center to demonstrate its functional capabilities.

STP-140A Rev. 4 SF-A-7A Airflow capacity.

This test was conducted on 5/23/85 and was to verify that the specified volume flowrate of air can be achieved with the fan as furnished, under actual field conditions at maximum and minimum filter pressure drops. Filter condition was 4 simulated by using blanks on the HEPA filters to achieve desired pressure drops. Flow rates were determined by the traverse plane method.

STP-1408 Rev. 4 SF-A-7B Airflow capacity.

~

This test was conducted on 5/20/85 and w s to verify that the specified volume flowrate of air can be achieved with the fan as furnished, under actual field

conditions at maximum and minimum filter pressure drops. Filter condition was simuTated by using blanks on the HEPA filters to achieve desired pressure -

drops. Flowrates were determined by the traverse plane method.

l STP-151A Rev. 4 Auxiliary Bldg., HVAC Initial Operation of condensing fans and -

compressor motor on air cooled condensing unit 545A.

l

! This te'st was conducted to ensure recording of important equipment initial operating data on unit 545A condensing Fans and Compressor Motor. This test

! verifies control schemes, bumps motors for rotation, checks for smooth operation and provides a run in period for all Fan Motors and Compressor Motor. This test provides a permanent record of Inrush and Running current, acceleration time, motor temperature rise, vibration and air flow.

I STP-1518 Rev. 5 Auxiliary Bldg.'HVAC Initial operation of condensing Fans and compressor motor on air cooled condensing unit 5458.

l i This test was conducted to ensure recording of important equipment initial l operating data on unit 5458 condensing fans and compressor motor. This test verifies control schemes, bumps motors for rotation, checks for. smooth operation and provides a run in period for all fan motors and for compressor motor. This test provides a permanent record of Inrush and Running current, acceleration time, motor temperature rise, vibration and air. flow.

STP-152A Rev. 3 Technical Support Center and Control Room HVAC Initial operation of Essential Air Handling Unit AH-A-545A.

This test was run to verify air flow, running current, vibration, voltage, and bearing temperature on the Essential air handling unit AH-A-545B.

STP-162 Rev. 3 New Essential HVAC System for the Control Room and TSC (Acceptance Test).

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L -__-. - -.-.- - _ - - - _ _ - _ _ _ _ _ - _ - _ _ _ _

The test was performed to demonstrate the operating functions of the new Essential HVAC System for the Control Room and Technical Support Center.

This test demonstrates and records status of control, indication, and position of all equipment associated with each train of the Essential HVAC when a start is initiated by the different modes of operation. The modes covered are manual, Radiological, Toxic, High Control Room temperature and SFAS actuation. The positon and indication of all Isolation Dampers is verified and cross checked with IDAD's indiction. All start and step switches and indicating lights are verified with each piece of equipment. This test covers both A+B units and verified all of the different isolation and interconnecting logic of both units.

STP-155A Rev.1 Technical Support Center and Control Room HVAC U-545A air cooled condensing unit acceptance testing.

This test was to verify that the air cooled condensing unit on train A of the essential HVAC system will perform satisfactorly. It first verified that all solenoids on the cooling coils will energize and deenergize at the proper tempef'ature and then leads into system evacuation and recharge of Freon-22. '

The expansion valves are adjusted, the system is run at the five different capacity modes and superheat temperatures are taken.

STP-155B Rev. 'l Technical Support Center and Control Room HVAC U-545B Air Cooled condensing unit acceptance testing.

This test was to verify that the air cooled condensing unit on train B of the essential HVAC system will satisfy the requirements of the test specification. It first verifies that all solenoids on the cooling coils will energize and deenergize at the proper temperatures and then accomplishes the system evacuation and recharge of Freon 22. This expansion valves are adjusted. The system is run at five different capacity modes and superheat temperatures are taken.

STP-164 Rev. 4 Air Balancing test procedure auxiliary Bldg. Essential Air Handling Units A and B.

This procedure was conducted to balance air flow throughout the C.R. and T.S.C. to design specifications. It measures air flow thru the different ducts by the. flow hood and traverse plane method and verifies Radiological and toxic damper positions. This test includes data of all the flows in different area's from both the A and B trains.

STP-171 Anco Air Aerosol mixing uniformity. _

This test was conducted to verify that tracer, dioctylphthalate (DOP) and refrigerant gas, injection and sample parts are located so as to provide proper mixing of the tracer in the air approaching the component stage (HEPA  !

filter bank or adsorber stage) to be tested, or the sample plane. The test is  ;

accomplished by injecting DOP or refrigerant upstream of the sample point at approximately four to five times the background dust concentration, and taking i concentration readings in a sample plane parallel to, and approximately one i foot upstream of the filters.

OI I G-16 hl l

9

4 STP 172 Anco Air Distribution Test Procedure HEPA and Carbon Filters.

1

) This test was used to verify that airflow distribution across each filter or ~ l adsorber in a system is reasonably uniform relative to the average flowrate in the total system. In this way, it can be verified that the HEPA filters will i load evenly, and the residence time of the adsorbers will meet design requirements. This test is applicable only for acceptance testing following

original installation and modification of the air cleaning system.

STP-173 In place leak test procedure for HEPA filter banks.

1 This test was performed to verify that the HEPA filter have not been damaged in any way, have been properly installed in their holding frames, no leaks are present in the holding frame or between the holding frame and the filter housing and no bypasses in the installed filter bank system that would compromise the function of the filters. The test consists of injecting DOP

. smoke upstream of the filter bank to mix with the air flowing through the filters. Concentrations obtained downstream are compared with concentrations upstream to determine leakage. ,

STP-174 In place leak test procedure, adsorber stage.

This test was performed to verify that the adsorber filters have not been

damaged in any way have been properly installed in their holding frames, have i no leaks in the structural frame or between the frame and the filter housing, that the installed adsorber bank does not have any voids or bypasses that would compromise the function of the adsorber system. The test is accomplished by injecting a tracer gas into the airstream upstream of the j

adsorber bank. Tracer concentrations are then determined upstream and

, downstream of the bank and penetre. tion (percent leakage) is calculated from

! the ratio of the concentrations.

STP-175 Control Room Temperature, Humidity and Air Velocity Survey.

This test was perforined to measure and record the temperature, humidity and air velocity of the Control Room while operating the HVAC system under specified parameters. This test accomplished a twenty-four hour run 'of each of the Essential trains and the Normal train while recording the temperature and humidity, also velocity readings were taken periodically.

STP-189 Control Room / Technical Support Center Pressure Test.

The purpose of this test is to measure and record the differential pressure of the Control Room / Technical Support Center.

The test is accomplished by using a monometer to measure the Control Room /T.S.C. pressure in relation to the outside pressure with first the A train running and second with the B train running. At least two readings should be taken in the Control Room and two readings in the Technical Support Center. The specification requires that 1/8 IWG (0.125) be maintained while in the Radiological mode.

G-17

The following Special Test Procedures (non-function) have been conducted on the new Essential H.V. A.C. System for the Control Room and Technical Support -

center to prove structural integrity.

STP- 138 Rev. 3 A Train Duct leakage test, Auxiliary Bldg. Control Room and TSC System.

This test was run to verity the integrity of the duct work after completion of initial construction and af ter any major system modification or repair (ANSI N510-1980 table 1). The test is accomplished by pressurizing the duct to a given pressure through a gas meter, once pressure is achieved it is held for a specified time, the meter reading at time of pressurization and completion of test are used to calculate the leakage. Total leakage cannot exceed 10 CFM for both trains.

STP-130 Rev. 3 B Train Duct work leakage test, auxiliary Bldg, Control Room and T.S.C. System.

This test was run to verify the integrity of the ductwork af ter completion of initi'al construction and after any major system modification or repair (ANSI N510-1980 Table 1). The test is accomplished by pressurizing the duct to a given pressure through a Sas meter, once pressure is achieved it is held for a specified time, the meter reading at time of pressurization and completion of the test are used to calculate the leakage. The total leakage is not to exceed 10 CFM for both trains.

STP-148A Rev. 3 SF-A-7A Filter housing leakage test auxiliary building, Control Room, and Technical Support Center system.

This test was conducted to verify the integrity of the filter housing after completion of initial construction. (ANSI N510-1980, Table 1, Section 5, Visual Inspection, and Section 6, Housing Leak Test and Section 7, Mounting Frame Pressure Leak Test). The test is accomplished by using a blower with a gas meter to pressurize the housing and then measure the flow required to hold the pressure.at this point for a specified period of time.

STP-1488 Rev. 3 SF-A-78 Filter Housing Leakage Test Auxiliary Building, Control Room, and Technical Support Center System.

This test was conducted to verify the integrity of filter housing af ter completion of initial construction. (ANSI N510-1980, Table 1, Section 5, Visual Inspection, and Section 6, Housing Leak Test and Section 7, Mounting Frame Pressure Leak Test) the test is accomplished by using a blower to pressurize the housing blowing through a gas meter to indicate the flow, when the specified pressure is reached the gas meter is read and recorded, the pressure is held for a specified time pericd and then the meter is read again, the final reading minus the starting reading divided by the time run is equal to the leakage in CFM.

STP-167A Rev. 2 Air Handling unit AHA-545A, Auxiliary Building, Control Room and T.S.C.

O G-18

This test was performed to verify the integrity of the air handler housing after completion of initial construction and after any major system modification or repair. This test is performed by blocking off all openings -

in the ductwork from the air handler housing assembly except the test air inlet port. The sealed off portion is pressurized to 8.1" IWG with a blower feeding through a gas meter which indicates flow. When the desired pressure j; is achieved the gas meter is read. Pressure is held for a specified time period and then the gas meter is read again. - The final reading minus the starting reading divided by the time indicates housing leakage in C.F.M.

STP-1678 Rev. 2 Air Handler unit AHA-5458, Auxiliary Building, Control Room and T.S.C.

This test was performed to verify the integrity of the air handler housing after completion of initial construction and after any major system modification or repair. The test is performed by blocking off all openings in

the ductwork from the air handler housing assembly except the the air inlet i part. The sealed off portion is pressurized to 8.1" IWG with a blo'.er feeding through a gas meter. When the unit reaches set pressure gas meter is read agairr. The gas pressure reading at conclusion minus the reading at the .

beginning divided by the time indicates leakage in C.F.M.

SP 211.01A: Control Room / Technical Support Center Emergency Ventilation System Loop A Monthly Surveillance Test.

< During performance of this procedure on June 20, 1986, the air handler fans were noted to be running but the compressor did not operate.

SP 211.01B; Control Room / Technical Support Center Emergency Ventilation System Loop B Monthly Surveillance Test.

During performance of this procedure on May 28, 1986, dampers HV-54706 and HV-54710 did not position properly.

III. REVIEW FINDINGS

, All functions of the CR/TSC Essential Air System have been demonstrated I by testing, however, based on the poor operating and extensive

modifications, nearly all testing must be repeated.

Surveillance procedures must be revised to provide a more thorough check of compressor and damper functions.

1 iO 4

G-19 vp----,.--,,-s---, , e ,w w ~.,w-,,-, -emem-,- me,--,,,,,m--,----v.-,,,+--eve, -

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IV. RECOMMENDATIONS

- 1. Proper function of the CR/TSC Essential Air System dampers operating in -

the radiological mode must be demonstrated upon completion of damper maintenance.

2. Proper function of the CR/TSC Essential Air System dampers operating in the toxic gas mode must be demonstrated upon completion of damper maintenance.

3. Proper function of the CR/TSC Essential Air Refrigeration System must be demonstrated upon completion of refrigerant system modifications.

4. Capability of each CR/TSC Essential Air System train to pressurize the CR/TSC must be demonstrated upon completion of damper maintenance and balancing of Essential Air Filtration Unit flows.

5. Capability of each CR/TSC Essential Air System train to provide fresh, filtered, conditioned supply air for CR/TSC ventilation must be tiemonstrated upon completion of refrigeration system modifications. -

O G-20 O

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4. REVIEW OF RECOMMENDATIONS AND RESOLUTIONS A. Discussion -

This section of the report contains a composite of all problems identified to date. The combined effect of the solutions being applied to the Control Room / Technical Support Center Essential Air Conditioning System wil1 be considered to determine if all known

deficiencies will be corrected.

B. Index of Problems The Index of Problems (Table 1) provides a summary of the problems associated with the Control Room / Technical Support Center Essential Air Conditioning System.

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[ TABLE 1 INDEX OF PROBLEMS Tracking O

Priority Problem Problem No. No.

2 Reference time meter addition 4-1 19.00118 1 Test Agent injection manifolds necessary for in-place filter leak testing have not been installed and tested. 4-2 26.0016 1 High Noise in the Control Room due to Essential HVAC 4-3 19.0010A 2 Excessive vibration of condensing units. 4-4 19.0010L 4-5 2 tontrol Room normal HVAC Flow Switch Relocation 26.0017 2 Pressurization measurement access for TSC 4-6 26.0001 1 The Control Room /TSC Essential HVAC System is subject to inadvertent actuations due to spurious signals from the radiation monitors. 4-7 26.0154 2 Powar supplies to Essential HVAC units do not follow normal engineering practices for providing reliable power to the controls and instruments. 4-8 26.0153 2 The Refrigeration System arrangement does not facilitate convenient or efficient maintenance of the system. 4-9 26.0152 1 Cooling capability of the CR/TSC Essential Air System is marginal during cold weather operation. 4-10 26.0167

.1 Control Room /TSC Essential HVAC System Refrigeration Problems. 4-11 26.0013 2 Work on fire dampers between the Control Room and TSC is not complete. 4-12 19.0011A 2 The CR/TSC Essential Air Conditioning System is subject to spurious actuation caused by high temperature in certain areas of the Control Room. 4-13 26.0018 2 Control Room normal HVAC air balance 4-14 26.0011 1 The CR/TSC Essential Air Conditioning System is subject to spurious actuations caused by the chlorine, gas detectors. 4-15 26.0025 O

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TABLE 1 INDEX OF PROBLEMS 1

Priority Problem Problem Tracking No. No.

1 Infiltration through the CR/TSC Essential Air Handler (AH-A-545A) access door is occurring. 4-16 26.0008 1 Isolation Dampers not operating properly 4-17 26.0004 1 CR/TSC Area Exhaust Fan Isolation Damper A(HV-54727) is not positioning properly when actuated and/or damper position displayed by IDADS is not proper. 4-18 26.0002 1 Shift Supervisor Office Exhaust Fan Isolation -

Damper B (HV-54732) is not positioning properly when actuated and/or damper position displayed by IDADS is not proper. 4-19 26.0003 i-1 Control Room pressure is not maintained at or b above 0.125 IWG relative to outside atmosphere by operation of either CR/TSC Essential Air j

System Train. 4-20 26.0006 [

j 1 Operation of the CR/TSC Essential Air Refrigerant system is compromised by certain lL l piping and control problems. 4-21 26.0014

[

i 1 The Control Room does not appear to be uniformly cooled by operation of the Essential Air System. 4-22 26.0012 1 Air flow through the Essential Filtration Unit may exceed the maximum rates specified in Technical Specification 4.10 for several minutes after the i

systems are actuated. 4-23 26.0010 1 There is no approved procedure for calibration l of the CR/TSC Essential Filtration Unit Transmitter. 4-24 26.0009

~

1 CR/TSC Essential Filtratilon Unit A Flow Transmitter j (FT-54701) and/or related IDADS components 4-25 26.0005 1 CR/TSC Essential Filtration Unit 8 Flow Transmitter (FT-54702) and/or related IDADS components. 4-26 26.0007 1 Operating Procedure A.14 does not include adequate instructions for operation of the CR/TSC Essential

{ Air System 4-27 19.0010D j G-23 i

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_ .~ - -..___ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ . . _ _ _ _ __ __ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ - _ -

l TABLE 1 INDEX OF PROBLEMS Priority Problem Problem Tracking No. No.

1 Surveillance Procedures SP.211.01A and SP.211.010 do not provide a check of all possible damper limit switch positions. 4-28 26.0023 2 Approved procedures for accomplishing non-routine complex refrigeration system maintenance tasks do not exist. 4-29 26.0024 1 Operation and control of the CR/TSC Essential Air System is not understood by operators 4-30 26.0021 1 The "Hardcast" tape used to seal the Essential Air Handling units against air infiltration has not been documented as being environmentally qualified. 4-31 26.0150 2 Carbon filter tray mounting in the CR/TSC Essential Air Filtration Unit must be evaluated for appropriate filter mounting capability'. 4-32 19.00110 2 Freon leak in Essential Air Handling Unit 4-33 26.0151 2 Diesel Exhaust into HVAC intakes 4-34 19.0010G 1 The affect of CR/TSC Essential Filtration unit initial air flow rates on CR/TSC radiation exposure is not known. 4-35 26.0020 2 Automatic initiation of CR/TSC Essential Air System operation increases Control Room operator work load during certain critical plant operating phases. 4-36 26.0019 1 Control Room /TSC Essential HVAC Controls and Operating status indication in the Control Room 4-37 26.0022 1 Temperature in the Essential Unit Compressor Compartments may become excessive during hot weather operation. 4-38 26.0015 2 Quality of the existing CR/TSC Essential Air System Refrigerant Compressor motor is questionable. 4-39 19.0011H O

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SYSTEM PR08LEM Control Room /TSC Essential HVAC System (Elapsed Time Meters) -

3 TRACKING NUMBER: 19.0011.8 PRIORITY: 2 l SOURCE: Interviews, Technical Specification 4.10; RRR819.00118 DESCRIPTION Technical Specification 4.10 requires sampling of the carbon filters in the essential filtration after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of unit operation. Presently, no automatic monitoring device has been supplied to monitor the total elapsed time; therefore, operating time is estimated based upon Operator Logs in the Control Room. Since these estimates are conservative, the carbon filters may be sampled more other than is necessary.

INVESTIGATIDN -

N/A RESOLUTION OF PROBLEM j A design change has been initiated to add elapsed time meters which will be i connected to the breakers for the filtration units to monitor total operating time j of each unit. This change is being. implemented under ECN A-5782. The surveillance

procedures will be revised to check these timers to verify elapsed operating time i

since previous carbon replacement.

TESTING Startup testing of the modification will be performed to verify proper operation of the elapsed time meters. Periodic surveillances will monitor the elapsed operating l time of the Control Room Essential Filtration Units by these meters.

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SYSTEM PROBLEM Test agent injection manifolds necessary for in-place filter leak testing have not been installed and tested.

TRACKING NUNBER: 26.0016 PRIORITY: 1 SOURCE: NCR S4761 DESCRIPTION HEPA and carbon filter banks in the Essential Filtration Units must be tested periodically for leaks to satisfy Technical Specification 4.10 requirements. This testing is accomplished by injecting a test agent into the airstream ahead of each filter bank. Injection manifolds are necessary to ensure all regions of the filter banks are uniformly challenged. The details of this are included in ANSI N 510, Testing of Nuclear Air-Cleaning Systems. Test manifolds have not been installed in -

the Essentf'al Air Filtration Units.

INVESTIGATION N/A RESOLUTION OF PROBLEM Test agent injection and sampling manifolds shall be installed between the two carbon filter banks on each filtration unit. This will be accomplished in accordance with ECN R-0788.

TESTING The test agent injection manifolds shall be tested in accordance with ANSI N510-1980 to demonstrate that the test agents are uniformly dispersed in the airstream entering the second bank of filters.

O G-26

SYSTEM PROBLEM Excessive noise in the Control Room caused by operation of the Essential HVAC System.

TRACKING NUMBER: 19.0010. A PRIORITY: -1 SOURCE: Meeting minutes of HVAC Upgrade Meeting held 12/04/85 (J. Kaleway/D.

Abbott to distribution, 12/09/85); Selected Projects; Interviews; 88N Laboratories Inc. Letters, 11/17/85 and 12/09/85 DESCRIPTION Operation of each train of essential HVAC equipment produces undesirable sound levels in the Control Room and impairs communication within the Control Room and while using the telephone. When both trains of equipment operate simultaneously, as occurs during automatic actuation of the essential system, the sound levels increase and have a severe impact on conumnication.

INVESTIGATIDN --

An evaluation of the noise and vibration levels of the essential HVAC System has

been performed to establish the cause(s) of the high noise levels. Based on the i results of this evaluation, immeaiate and long term resolutions will be pursued.

RESOLUTION OF PROBLEM The operating. procedure for the HVAC systems presently specifies that is both trains are initiated by automatic signal, one train should be secured.

Therefore, the following actions are recommended to be complete prior to restart:

1. Evaluate and replace, as necessary, air registers to reduce flow velocities through the registers.

2. Evaluate the total supply of conditioned air to determine if the air

. quantity being supplied to the Control Room and TSC may be reduced while still providing adequate cooling and pressurization. If so, then this

! change should be implemented.

If the previous items are not ef fective or practical, then the following long term actions are recommended:

1. Add additional balancing dampers to provide better air distribution control.

2. Replace ductwork from original system utilized with new distribution system with larger size duct to accommodate the high air flow.

TESTING Upon completion of any modifff: tion to the essential system for noise abatement, the

balance of the system shall be arified and corrected as required. The noise levels 4

shall then be measured to determine effectiveness of modification.

. s G-27

SYSTEM PROBLEM Control Room /TSC Essential HVAC Systems (High Vibrations on Condensing Units)

TRACKING NUMBER: 19.0010. L PRIORITY: 2 SOURCE: Interviews; Meeting minutes of HVAC Upgrade Meeting held 12/04/85, (J. Naleway/D. Abbott to distribution, 12/09/85); RRRB Action Plan 19.0010.L DESCRIPTION Operation of either of the Essential HVAC Systems causes a substantial vibration in the related condensing unit. This vibration has caused pipe straps to work loose

• and fall off and could therefore impact the integrity of refrigerant system if a pipe were to break.

INVESTIGATIDN Maintenance evaluated the vibration level of the CR/TSC Condensing Units in December 1985 and found that one of the compressors was missing a part which could affect the alignment. The compressor and motor for both units were realigned which resulted in reduced vibration.

Reconenended actions for short term:

1. Perform additional investigation on the units to determine source of vibration.

2. Perform preventive maintenance on the units to routinely check the condition of all supports internal to the equipment housing.

RESOLUTION OF PROBLEM Later TESTING Vibration levels will be checked following any modifications intended to reduce vibration.

O G-28

SYSTEM PROBLEM Control Room Normal HVAC (Flow Switch)

TRACKING NUMBER: 26.0017 PRIORITY: 2 SOURCE: Interviews; Meeting minutes of HVAC Upgrade Meeting held 12/04/85 (J. Naleway/D. Abbott to distribution, 12/09/85); SFR 280-85.

DESCRIPTION A paddle-type low flow switch was installed'in the cold air supply duct to shut off the normal air handler upon actuation of the Essential HVAC System due to closure of the isolation dampers. During the colder periods of the year there is a smaller demand for cooling in the Control Room, such that the air flow through the cold duct is decreased. This air reduction was enough to cause the flow switch to trip and shutdown the normal air handler and lose all normal cooling to the Control Room.

l Af ter a sho'rt time, the Control Room heats up and causes the Essential HVAC System -

! to automatically actuate upon Control Room high temperature.

INVESTIGATION N/A RESOLUTION OF PROBLEM

, A design change has been initiated to relocate the low flow switch from the cold air

, s duct to the return air duct where a steady air flow exists, independent of the heating / cooling load in the Control Room. .This change is being implemented under ECN R-0314.

TESTING Startup testing of this modification will be performed to verify proper operation of the flow switch upon completion of the modification.

l G-29

SYSTEM PROBLEM Control Room /TSC Essential HVAC (Pressurization Measurement Access for TSC)

TRACKING NUMBER 26.0001 PRIORITY: 2 SOURCE: Interviews; Technical Specification 4.10, STP-189 DESCRIPTION Technical Specification 4.10 requires that the Control Room /TSC Essential HVAC System be capable to pressurize the Control Room and TSC to 0.125 IWG (min.);

however, presently, no pressure tap exists to measure the pressure in the TSC.

Pressure readings are presently being taken by putting one tube under the door into the TSC.

INVESTIGATI,0N ,

N/A RESOLUTION OF PROBLEM ~

Install permanent pressure taps through the TSC wall such that future pressure measurements may be made from a constant measuring point. ~

It is recommended that one tap be placed through the north wall into the Auxiliary Building main corridor and that a second tap be placed through the south wall into the NSEB Bridge so that measurements may be taken against both outside reference points.

Permanent pressure taps will be installed in accordance with an ECN.

TESTING O

G-30

SYSTEN PR08LEN

! The Control Room /TSC Essential HVAC System is subject to inadvertent actuations due -

to spurious signals-from the radiation monitors.

TRACKING NUN 8ER: 26.0154 PRIORITY: 1 SOURCE: Interviews, Occurrence Description Report DESCRIPTION

-The radiation monitors were not grounded as recommended by the manufacturer of the equipment, such that output signal spikes were developed by the equipment causing the Essential HVAC System to start inadvertently.

INVESTIGATION RESOLUTION VF PROBLEM For restart, the radiation monitors shall be grounded per the manufacturer's recommendation to remove the spurious signal spikes and the resulting inadvertent actuations of the Essential HVAC System. This work is being performed by ECN R-0182.

In addition, the signal from the alert channel of the radiation monitors will be modified such that it no longer actuates the Essential HVAC System, but only l provides an alarm to alert the operators of increasing radiation levels. This

/'"'N change is being performed by ECN R-0164.

TESTING l

l Testing of these changes will be performed to verify the function of the changes is

! satisfied, i

i I

i G-31

SYSTEM PROBLEM Power supplies to Essential HVAC Units do not follow normal engineering practice for providing reliable power to the controls and instruments.

TRACKING NUMBER: 26.0153 PRIORITY: 2 SOURCE: Meeting Minutes of HVAC Upgrade Meeting; Project Proposal 86-018; Calc # Z-EDS-E0583 DESCRIPTION

1. The control power supply for the filtration units SF-A-7A and SF-A-78 (power supply units and Raymond actuators) are supplied from 120V distribution panels which could have steady state voltage availability below that which is recommended for operation of the equipment.

2. The C' lass 2 instruments of the filtration and air handling units share the -

same power supplies with the lighting and receptacles of the units. This does not follow established practices for providing power to instruments.

INVESTIGATION N/A RESOLUTION OF PROBLEM For short term improvement of the controls and instrumentation of the Essential HVAC Units, the following actions are recommended:

1. The control power supply for the filtration units shall be relocated from the 120V distribution panels S1A3 and SlB3 to the distribution panels S1A2-1 and S182-1 supplied by inverters SIA2 and SlB2 which have enough capacity to supply the equipment as recommended.

2. The illuminating lights for the essential air handlers and filtration units will be provided with a new control power supply such that switching transients in the lighting circuits do not interfere with the instruments.

This work is being accomplished by ECN R-0796.

l TESTING l

N/A l

l l

G-32 I

u __ _ _ _

SYSTEM PROBLEM

[ 'h The refrigeration system arrangement does not accommodate convenient or efficient

~

V maintenance of the system.

TRACKING NUMBER: 26.0152 PRIORITY: 2 SOURCE: Interviews with Maintenance personnel.

DESCRIPTION The arrangement of the refrigeration system components does not facilitate maintenance of the system due to:

1. insufficient isolation capability,

2. difficult access to components,

3. extreme difficulty in removing major components for repair or replacement.

INVESTIGATION N/A RESOLUTION OF PROBLEN Due to the short Tech Spec LCO for returning the essential equipment to service following actions are recommended prior to restart:

1. Provide procedure and equipment (monorails and hoists, etc.) for removing and reinstalling the compressor, compressor motor, and other major components.

This action will also require slight modifications to the condensing unit to acconnodate relocating obstructions of compressor removal.

2. Isolation valves shall be provided as required to facilitate isolation of the components for removal for repair and calibration purposes without releasing the refrigerant charge.

l The following actions are recommended for short term completion:

1. Relocate the liquid line sightglass so that it is more accessible for periodic monitoring.

2. Provide a refrigerant level gauge at the receiver tank to facilitate verification of Freon level during maintenance and troubleshooting of the system.

3. Reroute the lube manifold to the three condenser fans to ensure that each fan motor bearing receives appropriate lubrication.

G-33

SYSTEM PROBLEM (Continued)

The refrigeration system arrangement does not accommodate easy or efficient maintenance of the system.

TESTING Testing will be performed as necessary following completion of the modifications to verify proper function and integrity.

9

~

G G-34 O

SYSTEM PROBLEM

['~'N Cooling capability of the CR/TSC Essential Air System is marginal during cold weather operation.

TRACKING NUMBER: 26.0015 PRIORITY: 1 SOURCE: RRRB Action Item 19.0010C DESCRIPTION The controls of the refrigeration partion of the HVAC system do not permit reasonable flexibility of the system to respond to all variations in cooling loads with respect to ambient conditions. In particular, during low ambient conditions, the system fails to maintain adequate pressure in the suction line resulting in compressor trips. The consequence of'this condition is that the essential HVAC system cannot be relied upon to operate under all postulated accident modes. ,

INVESTIGATION RESOLUTION OF PROBLEM A modification was performed in September 1985 (Reference ECN R-0142) which opened a flow path for refrigerant under low load conditions by opening a solenoid valve which had erroneously ;been contra 11ar, to close at low load. Subsequent surveillance testing verified that the system was working satisfactorily, s

Revise the controls for the condenser fans to allow cycling of two of the three fans in response to ambient temperatures. This will provide improved responsiveness of the condensing capacity to cooling load with regard to ambient conditions. This change is being implemented by ECN R-0769 TESTING Testing will be performed upon completion of modifications to verify that the ref rigerant system functions properly.

G-35

SYSTEM PROBLEM Control Room /TSC Essential HVAC System (Refrigeration Problems)

TRACKING NUMBER: 26.0013 PRIORITY: 1 SOURCE: Meeting minutes of HVAC Upgrade Meeting held 12/04/85 (J. Naleway/

D. Abbott to distribution, 12/09/85); RRRB Action Item 19.0010C and 19.0010.J; NCR S-5028 DESCRIPTION Refrigeration System modification are necessary to facilitate efficient and effective system maintenance.

INVESTIGATION RESOLUTION'0F PROBLEM

1. Provide a system pumpdown capability to return refrigerant to the receiver when the system is shut down. This change will prevent damage to the compressor and enable more reliable startup of the system.

2. The hot gas bypass should be controlled through the evaporator to prevent gas through an inactive coil in conformance to good refrigeration practice.

3. A new P&ID should be developed to document the refrigeration system of the Control Room /TSC Essential HVAC System. This drawing will be in greater detail than the schematics provided by the vendor, and will facilitate maintenance and troubleshooting of the system.

Control Room /TSC Essential HVAC System (Refrigeration Problems)

4. Modification of the evaporator control sequence based on actual operating experience to provide closer incremental control is desirable. This will I require additional data to be taken during surveillances to monitor operating characteristics of the system.

TESTING Testing will be performed upon completion of the modifications to verify that the refrigerant systems function properly.

O G-36

\ > < , L SYSTEM PROBLEM A

Work on fire dampers between the Control Room and TSC;(' currently defer / red undeyMOD - s(

111) is not complete. , ,

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TRACKING NUMBER: 19.00ll.A '

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"\ w c PRIORITY: 2 'A '

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SOURCE: RRRB Action Item 19.0011.A t k lt n t

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DESCRIPTION 1 -

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Eight existing dampers are to be modified and two new dampers are to be insiaile k i

Long term operation with compensctory measures is riot des 1rable. The ability to ~ ,

complete this work prior to startup must be evaluated. '3

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• ( ,, , ,j INVESTIGATION i, g.

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The ability *to complete the work before startup will be evaluated. . If there is a '

reasonable possibility that the work can to completed in a timely Menner, tha work '4 s will be undertaken. 3p' RESOLUTION OF PR08LEM The eight fire dampers through the east wall of the control are being upgraded'fres -

1-1/2 hour-rated dampers to 3-hour UL-rated dampers by ECN R-0763. ; Tne s'upply a'nd -

return ducts for the Essential HVAC System through the TSC Roof will be equipped ,

with UL-rated fire dampers and smoke dampers to prevent smoke infiltration from the TSC to the Control Room through the Essential System ductwork. '

TESTING ., /

d

.t ,'}-

Verification of the proper operation of the fire and smoke dampers will be perforr,wo '

upon completion of,.the modifications. The fire dampers will then be tested *,

I periodically by Surveillance Procedure SP.201.03M. s -

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SYSTEM PROBLEM The CR/TSC Essential Air Conditioning System is subject to spurious actuations caused by high temperature in certain areas of the Control Room.

TRACKING NUMBER: 26.0018 PRIORITY:

SOURCE: Occurrence Description Report; September 7,1985 DESCRIPTION Temperature sensors which actuate the CR/TSC Essential Air System in the event of high Control Room temperature (80'F) are located approximately eight feet above tha floor in the vicinity of computer equipment. Temperature in this area is somewhat elevated relative to the average temperature in the Control Room. Spurious system actuation is caused by high temperatures in the vicinity of the sensors rather than excessive averags temperature in the Control Room.

INVESTIGATION The temperature at various point throughout the Control Room and Computer Room will be measured to determine the temperature gradient at these point with the normal HVAC System operating. The temperature at the sensors for the . Essential HVAC System will be compared to the overall distribution to determine if a trend exists that would cause the sensor to be subjected to temperatures higher than the average overall temperature resulting in spurious actuation.

RESOLUTION OF PROBLEM A change in the setpoint of the temperature controllers will be considered if it can be shown that a higher setpoint will still provide for a maximum temperature of 80*F in the general area as required by the design basis of the Essential HVAC System.

In addition, the Normal HVAC System will be rehalanced to ensure that the air distribution is adequate to nrovide cooling to all areas within the Control Room and Computer Room.

TESTING Balancing of the Normal HVAC will be performed by an STP followed by a determination of a Control Room temperature profile.

l O

G-38

SYSTEM PROBLEM

,p Control Room Normal HVAC System (Air Flow Balance)

TRACKING NUMBER: 26.0011 PRIORITY: 2 SOURCE: Interviews, Calculation 2-HVS-M0282, NCR S-5663

. DESCRIPTION The air balance of the Control Room Normal HVAC System has not been updated or verified since the original balancing was performed in 1973, despite several modifications which could affect the balance or the heat load. As a result, the HVAC equipment may not be satisfying the current loading in the Control Room.

INVESTIGATION Temperature' data will be obtained to enable evaluation of cooling throughout the .

Control Room area (Work Request Number 115248). .

RESOLUTION OF PROBLEM .

Short term recommended actions:

1. Initiate complete air flow balance of the Control Room HVAC System, including vibration testing of all rotating equipment.

l

2. Adjust and/or replace system components as required to support air balance (i.e., sheaves, belts, bearings, and thermal overload heaters). This work has l already begun.

3. Conduct a system review of the system design after the rebalance of the system to verify that system performance optimizes Control Room environmental conditions considerfng all the presently identified loads.

TESTING Testing will be performed as required to verify the balance of the system is adequate for cooling in the Control Room (i.e., temperature profiles will be taken at various locations in the Control Room and Computer Room).

! G-39 l

l l

SYSTEM PROBLEM The CR/TSC Essential Air Conditioning System is subject to spurious actuations ~

caused by the Chlorine Gas Detector.

TRACKING NUMBER: 26.0025 PRIORITY: 1 SOURCE: Occurrence Description Report, 06-30-86, WR# 116218, Interviews DESCRIPTION The chlorine gas sensor contains an electrolytic fluid which must be replenished periodically. Depletion of this fluid causes the detector to actuate.

Replenishment of this fluid should be accomplished in conjunction with other preventive maintenance items on a periodic basis.

INVESTIGATION RESOLUTION OF PROBLEM A procedure for performing periodic maintenance on the chlorine gas detectors will be written.

TESTING O

m e O

G-40

l l

SYSTEM PROBLEM Infiltration through the CR/TSC Essential Air Handler (AH-A-545A) access door is

~

( occurring.

TRACKING NUMBER: 26.0008 PRIORITY: 1 SOURCE: Occurrence Description Report. 06-30-86, Work Request Number 116217 DESCRIPTION Air leakage into the CR/TSC Essential Air Handler "A" with the system operating has been detected. This is a breach of the Control Room /TSC pressure boundary resulting in unaccounted infiltration of unfiltered air.

INVESTIGATION N/A RESOLUTION OF PROBLEM For restart, the access door gasket will be replaced and torquing of the retaining nuts will be verified.

For the long-term, the bolt-on access doors will be replaced with marine bulkhead type doors as recommended by ANSI N509-80 and ERDA 76-21. This will facilitate easy access into the units for maintenance as well as a leak tight joint.

. TESTING The new access doors will be tested for leakage in accordance with ANSI N510-80.

G-41

1 SYSTEM PROBLEM Isolation Dampers not operating properly.

TRACXING NUMBER: 26.0004 PRIORITY: 1 SOURCE: Interviews; NCR S-5611; STP-194; Troubleshooting Action Plan I.R.1 DESCRIPTION:

Control Room isolation dampers have experienced difficulty in reaching full-close position and may be leaking. Access doors were not installed near all dampers making maintenance of these dampers very difficult.

INVESTIGATION Investigation of proper damper position and proper position indication in actuated and unactuated dates is presently being performed in accordance with Work Request m Numbers 114662 ana 114663.

RESOLUTION OF PROBLEM

1. Initiate and implement a design change to add access doors where necessary for inspecting and maintaining the dampers.

2. Evaluate the dampers to verify that they are installed in accordance with

! the manufacturer's instructions. Contact supplier to verify that the installation is proper.

l

3. Review and update preventive maintenance requirements for the dampers and their components.

TESTING Proper function of the dampers and associated IDADS damper position indication will be verified by performance of an STP.

O G-42

SYSTEM PROBLEM

(g Control Room Toilet Area Exhaust Fan Isolation Damper A (HV-54727) is not positioning properly when actuated and/or damper position displayed by IDADS is not

( j proper.

TRACKING NUMBER: 26.0002 PAIORITY: 1 SOURCE: Selected Projects, STP-194 data, Troubleshooting Action Plan I.R.1 DESCRIPTION During performance of STP-194 (CR/TSC Essential Air Conditioning Safety Features Operation and Flow Detection Test), HV-54727 was indicated by IDADS to be in both open and closed positions simultaneously when the system was operated in the ventilation (high temperature / radiation) mode.

INVESTIGATEDN This problem was investigated under Work Request Numbers 111572 and 114655. The damper actuator was determined to be malfunctioning. Operation of the actuator was '

investigated in accordance with Work Request Number 115170. The actuator was fowtd to be functioning properly but did not produce sufficient torque to close the damper. NCR S-5611 was written.

RESOLUTION OF PROBLEM O

Investigation-of the discrepancy is to continue until a root cause is identified and the damper functions properly and consistently.

TESTING Proper function of the damper and the associated IDADS damper position indication will be verified by performance of an STP.

l

! Proper function of the damper will be periodically checked by performance of SP.211.01 A.

G-43

SYSTEM PROBLEM Shift Supervisor Of fice Exhaust Fan Isolation Damper B (HV-54732) is not positioning-properly when actuated and/or damper position displayed by IDADS is not proper.

TRACKING NUMBER: 26.0003 PRIORITY: 1 SOURCE: Selected Projects, STP-194 data, Troubleshooting Action Plan I.R.1; RRRB Action Item 19.0010.B DESCRIPTION:

During performance of STP-194 (CR/TSC Essential Air Conditioning Safety Features Operation and Flow Detection Test), HV-54732 was indicated by IDADS to be in both open and closed positions simultaneously when the system was operated in the ventilation (high temperature / radiation) mode.

INVESTIGATIDN This problem was investigated in accordance with Work Request Numbers 114656 and 114657. The damper was found to function properly. The damper position limit switches, which are the source for the IDADS damper position indication, were found to be operating intermittently. The limit switches were adjusted. IDADS indication of damper position was subsequently checked and found to be proper.

RESOLUTION OF PROBLEM Limit switches were adjusted. Damper function and damper position indication was checked and found to be proper.

TESTING Proper post maintenance function of this damper and the IDADS damper position indication will be verified by performance of STP.

Proper function of this damper will be periodically checked by performance of Surveillance Procedure SP.211.010.

i O,

G-44 l

l

- 1

SYSTEM PROBLEM Control Room pressure is not maintained at or above 0.125 IWG relative to outside -

7 O atmosphere by operation' of either CR/TSC Essential Air System train.

TRACKING NUMBER: 26.0006 PRIORITY: 1 SOURCE: Interviews with STP-189 Coordinator DESCRIPTION During performance of STP-189 (Control Room / Technical Support Center Pressure Test),

CR/TSC pressure could not be maintained above 0.125 IWG relative to outside atmosphere by operation of either CR/TSC Essential Air train.

INVESTIGATION The followihg checks of the CR/TSC Essential Air System dampers will be completed in -

accordance with Work Request Numbers 114662,114663,111572, and 114655:

1. Visual check of damper blade position in actuated and unactuated states

2. IDADS indication of actuated and unactuated damper positions.

An investigation of the Control Room /TSC boundary walls, floors and ceilings was performed under WR #115773 to identify any breached openings, however, none were

! located.

RESOLUTION OF PROBLEM Complete investigation of dampers and af fect repairs, as required.

TESTING Upon completion, perform STP-189 to verify the Control Room /TSC is pressurized to at least 0.125 IWG relative to outside atmosphere by operation of each CR/TSC Essential Air System train.

SP 211.01B and SP 211.010 will be performed to verify pressurization and flow are proper.

O G-45 i

SYSTEM PROBLEM Operation of the CR/TSC Essential Air Refrigerant Systems is compromised by certain piping and control problems.

TRACKING NUMBER: 26.0014 PRIORITY: 1 SOURCE: Interviews; System Engineer DESCRIPTION

1. A leak in the CR/TSC Essential Air Train A refrigerant piping is suspected.

2. A pressure regulator in the CR/TSC Essential Air Train A refrigerant system is subject to short cycling resulting in continuous surging.

3. Evaporator coil superheat should be checked and adjusted.

INVESTIGATION

-RESOLUTION OF PROBLEM Troubleshooting and maintanance as necessary will be performed to resolve the aforementioned discrepancies.

. A Maintenance Procedure will be initiation to describe the steps necessary to accomplish replacement of various refrigerant system components and to charge the system with refrigerant.

TESTING Proper function of. the refrigeration system will be verified following resolution by a post maintenance functional test.

O

, G-46 l

SYSTEM PROBLEM The Control Room does not appear to be uniformly cooled by operation Jof the O, ' Essential Air System.

TRACKING NUMBER: 26.0012 PRIORITY: 1 SOURCE: PAG l

i

! DESCRIPTION Air distribution in the Control Room with the CR/TSC Essential Air System operating is not believed to be adequate. This problem may be compounded by duct / diffuser modifications for noise reduction.

INVESTIGATION Data for ev'aluating the cooling capability was obtained by running each train of the -

system. Each train was operated for several hours; the B train on 07--19-86, the A train on 07-20-86. The Control Room was adequately cooled by the operation of each '

train; however, temperature fluctuations were experienced during compressor loading and unloading.

RESOLUTION OF PROBLEM The capability of each essential train to provide uniform cooling throughout the Control Room /TCS has been demonstrated. Temperature fluctuations encountered during

(N g'

evaluation runs will be reviewed and recommendations implemented. Cooling in the vicinity of the Bailey computer cabinet will also be improved.

TESTING The essential trains'will be operated to obtain data for evaluation of any changes and/or adjustments.

f e e G-47

SYSTEM PROBLEM Air flow through the Essential Air Filtration Units may exceed the maximum rates specified in Technical Specification 4.10 for several minutes af ter the systems are actuated.

TRACKING NUMBER: 26.0010 PRIORITY: 1 SOURCE: STP-198 data; Troubleshooting Action Plan I.R.1; RRRB Action Item 19.00

10.8 DESCRIPTION

Booster fan inlet vanes, which control flow through the Essential Filtration Units, are positioned in response to differential pressure developed across the carbon filter banks. When the units are actuated, there is no differential pressure; the inlet vanes are opened to the maximum position thereby allowing an excessive flow rate to be Dstablished. This condition persists for several minutes until a -

stabilized flow rate is established.

INVESTIGATION N/A RESOLUTION OF PROBLEM The Essential Filtration Unit Flow controls will be adjusted to reduce the time necessary to establish stabilized flow.

TESTING Upon completion of adjustments, time required to establish stabilized flow will be determined for each train.

O G-48

SYSTEM PROBLEM t

There is no approved procedure for calibration of the CR/TSC Essential Filtration -

Unit Flow Transmitter.

TRACKING NUMBER: 26.0009 PRIORITY: 1 SOURCE: RRRB Action Item 19.00

10.8 DESCRIPTION

Differential pressure across the carbon filter banks is the parameter measured to control flow thru the Essential Filtration Units. The flow characteristics are subject to change when tta carbon is changed, necessitating recalibration of the flow transmitter to assure the proper flow rate is maintained and proper indication of flow rate is indicated in IDADS.

INVESTI'3ATIDN N/A RESOLUTION OF PROBLEM A procedure for calibration of Essential Filtration Unit flow transmitters will be instituted; TESTING N/A I

l O. G-49

SYSTEM PROBLEM CR/TSC Essential Filtration Unit A Flow Transmitter (FT-54701) and/or related IDADS components:

1. Operate intermittently and/or provide intermittent flow indication on the IDADS monitor.

2. Control flow at a rate which is in excess of the Technical Specification 4.10 flow limit and/or indicate a flow rate which is in excess of the Technical Specification 4.10 flow limit.

TRACKING NUMBER: 26.0005 PRIORITY: 1 SOURCE: December 26, 1985 plant transient data, STP-194 (CR/TSC Essential Air Conditioning Safety Features Operation and Flow Detection Test);

Troubleshooting Action Plan I.R.1; RRR8 Action Item 19.0010.B. .

DESCRIPTION Flow control and indication of the filter unit is accomplished by a Foxboro Transmitter which monitors the pressure differential across the carbon filters and converts this into an electrical signal, which controls the booster fan inlet vanes in the unit and inputs to the IDADS computer. The computer then processes this signal in an equation and displays the flow rate in cfm.

During performance of STP-194, IDADS indication of flow through the Essential Filtration Unit A (SF-A-7A) was intermittent and indic' teda flow rate exceeded Technical Specification 4.10 maximum flow limit. Actual flow rate was determined to be 3010 CFM; within Tech Spec 4.10 Limitations.

INVESTIGATION This problem is be' ing investigated in accordance with STP-198 (CR/TSC Essential Air Filtration Unit Flow Test) and Work Request Numbers 114658, 115707.

Function of Flow Transmitter FT54701 was investigated in accordance with Work Request Number 114658 and I&C Periodic Maintenance Calibration Task No. 04767. A cause for intermittent IDADS flow indication was not identified. Investigation of intermittent IDADS indication is continuing in accordance with Work Request Number 115707.

O G-50

SYSTEM PROBLEM (Continued)

CR/TSC Essential Filtration Unit A Flow Transmitter (FT-54701) and/or related IDADS -

components: -

RESOLUTION OF PROBLEM Flow through the CR/TSC Essential Filtration Unit a (SF-A-7A), based upon measurements made by performing STP-198, is 3,010 cfm consisting of 1,507 cfm outside air and 1,507 cfm-return air. Flow through the filtration unit indicated by IDADS during the test was 3,540 cfm. IDADS indicated flow rate through SF-A-7A was 530 cfm greater than the actual flow rate.

The computer equation used to calculate the indicated flow rate will be revised to provide a more accurate IDADS indication of actual flow rate.

TESTING Continuous 'and accurate IDADS indication of air flow thru the CR/TSC Essential Air Train A will be verified by performance of an STP.

5 e

G-51

SYSTEM PROBLEM CR/TSC Essential Filtration Unit B Flow Transmitter (FT-54702) and/or related IDADS components control flow at a rate which is in excess of the Technical Specification 4.10 flow limit and/or indicate a flow rate which is in excess of the Technical Specification 4.10 flow limit.

TRACKING NUMBER: 26.0007 PRIORITY: 1 SOURCE: December 26, 1985 plant transient data, STP-194 (CR/TSC Essential Air Conditioning Safety Features Operation and Flow Detection Test);

Troubleshooting Action Plan I.R.1; RRRB Action Plan 19.0010.B.

DESCRIPTION Flow control and indication of the filter unit is accomplished by a Foxboro Transmitter which monitors the differential pressure across the carbon filter and converts this into an electrical signal, which controls the booster fan inlet vanes -

and inputs to the IDADS computer. The computer then processes this signal into an equation which indicates flow rate in cfm.

During performance o'f STP-194, IDADS indication of flow through the Essential Filtration Unit B (SF-A-78) exceeded Technical Specification 4.10 maximum flow limit. Actual flow rate was determined to be 3050 CFM; within Tech Spec 4.10 limits.

INVESTIGATION This problem is being investigated in accordance with STP-198 (CR/TSC Essential Air Filtration Unit Flow Test) and Work Request Number 115709.

RESOLUTION OF PROBLEM Flow through the CR/TSC Essential Filtration Unit B (SF-A-78), b'ased upon measurements made by performing STP-198, is 3,050 cfm consisting of 1,328 cfm return air and 1,725 cfm outside air. Flow indicated by IDADS during STP-198 was 3,793.

IDADS indicated flow rate through SF-A-78 is 748 cfm greater than actual flow rate.

Return air and outside air flow rates will be adjusted to 1600 CFM 10%. The computer equation used to calculate the indicated flow rate will be revised to ,

provide a more accurate IDADS indication of actual flow rate. l Balancing of outside and return air flow will be accomplished in accordance with

~

Work Request Number 114691.

e-TESTING ,

Proper air flow rate and IDADS indicated flow rate will be verified in accordance ,

with STP-198. l O'

G-52

SYSTEM PROBLEM Operating Procedure A.14 does not include adequate instructions for operation of the-O CR/TSC Essential Air System.

TRACKING NUMBER: 19.0010.0 PRIORITY: 1 i

SOURCE: Operator Interviews; Troubleshooting Action Plan I.R, RRRB Action Item 19.0010.D '

DESCRIPTION Operating Procedure A.14 is deficient in the following areas:

1. The difference between Train A and Train B in the present Interim Number 2 configuration is not clear.

2. The Timitations imposed on the system when in the isolation /stop mode are not -

delineated.

3. Criteria for stopping operation of the -individual trains after automatic actuation is not included.

4. Interaction with the Nuclear Service Electrical Bus Unloading Scheme is not discussed.

5. Instructions for placing. system in STANDBY are not complete.

INVESTIGATION N/A RESOLUTION OF PROBLEM Operating Procedure A.14 and related procedures will be revised.

TESTING N/A O G-53

3YSTEM PROBLEM Surveillance Procedures SP.211.01A and SP.211.01D do not provide a check of all possible damper limit switch position indications. As a result, damper position and/or position indication problems may be masked.

TRACKING NUMBER: 26.0023 PRIORITY: 1 SOURCE: Troubleshooting Action Plan I.R.1: STP-194 DESCRIPTION There are four possible damper position IDADS indications for each CR/TSC Essential Air System damper. However, only two of the four positions are recognized on the pertinent surveillance enclosures. As a consequence, conflicting damper position indications are not detected.

INVESTIGATTON N/A RESOLUTION OF PROBLEM Surveillance Procedures SP.211.01A and SP.211.010 will be revised to recognize the four possible position indications.

A Periodic Maintenance Procedure will be initiated to verify proper damper function once each refueling interval.

TESTING N/A l

1 l

e G-54 l

l

SYSTEM PROBLEM Approved procedures for accomplishing non-routine complex refrigeration system O maintenance tasks do not exist.

TRACKING NUMBER: 26.0024 PRIORITY: 2 SOURCE: Interviews with HVAC Maintenance personnel DESCRIPTION A Maintenance Procedure describing refrigerant system pumpdown, evacuation, recharging, and compressor change out is needed. The procedure must include a list of required special equipment, materials, etc.

RESOLUTION OF PROBLEM A Maintena6ce Procedure for accomplishing these tasks will be written.

INVESTIGATION N/A TESTING N/A ,

G-55

~

SYSTEM PROBLEM Operation and control of the CR/TSC Essential Air System is not understood by Operators.

TRACXING NUMBER: 26.0021 PRIORITY: 1 SOURCE: Operator interviews; Troubleshooting Action Plan I.R.2.

DESCRIPTION Additional training is necessary to enhance Operator knowledge of CR/TSC Essential Air Conditioning System operation. Clarification of backup power sources, various system operating modes, system availability in the various operating modes, and system interfaces is necessary.

INVESTIGATION N/A RESOLUTION OF PROBLEM Additional training will be provided to improve Operator knowledge of the CR/TSC Essential Air Conditioning System. Topics to be presented include:

1. Differences between CR/TSC Essential Air Train A and Train B in the Interim Number 2 configuration.

2. Clarification of backup power sources for the system.

3. Various system operating modes.

4. System availability in the various operating mode.

5. System interfaces.

TESTING N/A l

O G-56 l

l l - _ _

SYSTEM PROBLEM "N The "hardcast" tape used to seal the Essential Air Handling Units against air ~

infiltration has not been documented as being qualified for long term exposure to expected environmental conditions.

TRACKING NUMBER: 26.0150 PRIORITY: 1 SOURCE: System Design Engineer DESCRIPTION During the startup testing of the Essential HVAC System, it was revealed that the air handling units could not meet the leakage criteria of ANSI N509-1980 due to inadequate design and fabrication of the unit housing. "Hardcast" tape was used to seal all of the joints of the housing based on assurance of qualification from the manufacturer of the tape. Upon evaluation of the qualification information available,'it was determined that there was insufficient data available to establish a qualified life of the tape.

INVESTIGATION Bechtel has attempted to provide a qualification for the tape based on the limited available information from the manufacturer of the tape, however, has been unable to provide a qualified life of the material due to the proprietary nature of the components of the tape.

RESOLUTION OF PROBLEM Prior to restart, a procedure will be developed for periodic inspection of the tape application on the air handlers to verify that the integrity of the housing has been maintained. The inspection will be performed in conjunction with other testing to be performed on the units as a result of other modifications.

For the short term, the tape shall be submitted for environmental qualification to determine the expected life of the material.

TESTING A housing leak test which will. be performed following modifications to the air handler access doors (Ref: Tracking Number 26.0002) will also verify the integrity of the taped joints of the unit.

I G-57

SYSTEM PROBLEM Carbon filter tray mounting in the CR/TSC Essential Air Filtration Unit must be evaluated for appropriate filter mounting capability.

TRACKING NUMBER: 19.0011.D PRIORITY: 2 SOURCE: RRRB Action Item 19.00

11.0 DESCRIPTION

The carbon filter tray mounting frames may be improved to facilitate maintenance activities.

INVESTIGATION A study of the carbon filter tray mounting f rames will be performed to identify and -

evaluate improvements.

RESOLUTION OF PROBLEM Later TESTING In-place filter testing will continue to be performed by the existing Surveillance Procedures to verify that bypass leakage past the filter frames does not exceed Tech Spec limits.

O G-58 l

SYSTEM PROBLEM Freon leak in Essential- Air Handler. ~

C'- TRACKING NUMBER: 26.0151 PRIORITY: 2 SOURCE: Interviews with Human Factors Group DESCRIPTION INVESTIGATION An informal investigation of the effects of Freon 22 on habitability has indicated that the Freon in its gaseous form would act as an asphyxiant by displacing oxygen in the Control Room, however, a very high concentration of Freon is required to L endanger habitability. In addition, discussions with other plants having a similar type of HVAC system has revealed that were analyses were performed, it was found no l subsequent' actions (i.e., installation of Freon Detection Systems, etc.) were -

required.

RESOLUTION OF PROBLEM A further evaluation will be performed to document the effect of a Freon leak on Control Room habitability based on the quantity of Freon available in the system.

TESTING N/A O

G-59

-,-.,,m.,..,-,--e,. , , , , , - , . . . - , - -

. ,, - -_ _, ,-,,,,, --...__-,-.,-,a. - - -,,- - -, . -- -_.. ----

l SYSTEM PROBLEM Control Room /TSC Essential HVAC System (Diesel Exhaust into HVAC Intake)

TRACKING NUMBER: 19.0010.G PRIORITY: 2 SOURCE: Meeting minutes of HVAC Upgrade Meeting held 12/04/85 (J. Naleway/ l D. Abbott to distribution, 12/09/85); ODR No.85-154 l DESCRIPTION The exhaust of the existing diesel generators can be dispersed over the Auxiliary Building area, including where the Control Room /TSC Essential HVAC Unit outside air intakes are located. Concurrent operation of the diesel generators and the HVAC System could result in deterioration of the carbon absorbent material in the filter banks due to the exhaust fumes being drawn into the unit.

INVESTIGATTON An evaluation of the filtration unit outside are intake locations relati' e to the diesel exhaust stacks was performed and it was determined that there would be sufficient diffusion of the products of combustion in the exhaust stream with the surrounding air to avoid the problem of concentrated exhaust being drawn into the air intakes. In addition, a carbon sample was tested and the reduction in efficiency was not significant. Therefore, it was concluded that the diesel exhaust has a small impact on the efficiency of the carbon filters.

RESOLUTION OF PROBLEM Perform an evaluation to determine what possible long term effects th'e diesel exhaust may have on the carbon filters.

TESTING The refueling interval and 720 elapsed operating time surveillance procedures will continue to sample the carbon such that any degradation will be detected.

O G-60

SYSTEM PROBLEN  ;

I

N 1he affect of CR/TSC Essential Filtration Unit initial air flow rates on CR/TSC

.'adiation exposure levels is not known.

TRACKING NUMBER: 26.0020 PRIORITY: 1 SOURCE: USNRC concern DESCRIPTION Upon system actuation, air flow rates thru the CR/TSC Essential Air Filtration Units exceed the stabilized flow rates for several minutes. During this period, the concentration levels of airborne radioactive particulate and or gas is assumed to be at maximum levels. The com'iination of the flow rates and concentration levels may result in increased CR/TSC 9ccupant exposure levels.

INVESTIGATTON A study to evaluate the affect of CR/TSC Essential Filtration Unit hi flow rates and hi radiation concentration levels on CR/TSC occupant exposure levels will be completed.

RESOLUTION OF PROBLEM -

The flow controllers which actuate the valve actuators on the filter unit booster fan will be recalibrated to minimize the time required to stabilize the air flow through the unit.

TESTING h

G-61

SYSTEM PROBLEM Automatic initiation of CR/TSC Essential Air System operation increases Control Room Operator work load during certain critical plant operating phases.

TRACKING NUMBER: 26.0019 PRIORITY: 2 SOURCE: Troubleshooting Action Plan IR DESCRIPTION Both trains of the CR/TSC Essential Air Conditioning System are automatically started upon Safety Features initiation. The Control Room Operator must direct attention to this system, decide which train to shutdown, then take action. This is believed to be an unnecessary contribution to Control Room Operator workload during critical phases of plant operation.

INVESTIGATION A study will be performed to determine the necessity to start both trains of the CR/TSC Essential Air Conditioning System upon Safety Features initiation. The following will be considered.

1. Automatic start of only one CR/TSC Essential Air train in the event of a Safety Features initiation. Automatic start capability of the second train would be removed.

2. Staging of trains to automatically start only one CR/TSC Essential Air train upon Safety Features initiation. The second train would automatically start only if the first train failed to start or operate.

3. Removal of the CR/TSC Essential Air System automatic start capability from Safety Features. Neither train would be automatically started by Safety Features actuation.

The nature of modifications necessary to facilitate any of the aforementioned capabilities will also be considered. Modifications which add undue complexity to the controls will not be acceptable.

RESOLUTION OF PROBLEM l

Later TESTING Later l

J G-62 i

w

SYSTEM PROBLEM Control Room /TSC Essential HVAC (Controls and operating status indication in the -

O' Control Room.)

TRACKING NUMBER: 26.0022 l

PRIORITY: 1 SOURCE: Meeting minutes of HVAC Upgrade Meeting held 12/04/85 (J. Naleway/ ,

D. Abbott to distribution, 12/09/85); RRR8 Action Item 19.0010E and '

19.0011G DESCRIPTION The manual control and indication switch for the Control Room /TSC Essential HVAC System only provides limited information regarding the operation of the essential equipment. It does not have the capability to distinguish between SFAS, high radiation, or Control Room high temperature starts since all three are indicated by a single li'ght. No indication is provided for the status of the compressor. -

Therefore, to verify operating status of the essential equipment, the operator must either check the remote local indication or review the output of IDADS.

INVESTIGATION i

i N/A RESOLUTION OF PROBLEM For restart, the folicwing actions are recomended:

1. Revise the operating procedure to address how to distinguish between the actuating signals prior to isolating both units. Additional operator training should be provided to promote a good understanding of the characteristics of the system.

I For short term enhancement of information provided to the operators, the following actions are recommended:

i

1. The control switch for the Essential HVAC System should be modified to provide an indication of the signal which caused the system to actuate.

2. An indication for the operating status of the compressors should be added in the Control Room, either near the control switch for the system or in the 4

IDADS computer.

3. Evaluate the Essential HVAC System to determine what other system operation indications should be made available in the Control Room.

G-63

_ _ _ _ _ _ __ ____.___.i____.__._.______.__.-_____________.-_._ _ _ _ _

SYSTEM PROBLEM Temperature in the Essential Unit compressor compartments may become excessive during hot weather operation.

TRACKING NUMBER: 26.0015 PRIORITY: 1 SOURCE: RRRB Action Item 19.0010B DESCRIPTION Ventilation in Essential Unit compressor compartments is natural circulation.

During hot weather operation, temperature in the compartments are elevated when the system is in operation. The elevated temperatures may result in failure of the motor, compressor, or other component.

INVESTIGATION Temperatures in the Essential Unit compressor compartments will be monitored during hot weather operation. Data obtained will be evaluated to determine the necessity for installation of a compartment air ventilation fan.

The qualification documentation of the compressor, motor, and instruments located in the compressor compartment will be reviewed to determine if the elevated temperatures exceed the qualification temperatures of the components.

RESOLUTION OF PROBLEM Later TESTING Not required.

l .

O G-64 l

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SYSTEM PROBLEM

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Quality of the existing CR/TSC Essential hir Syste:n'refrigerar'tt ' "

compresscr moterdis~ ';

questionable.  ;

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a

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TRACKING NUMBER: 19.00ll.H 2 /  :

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PRIORITY: 2 '

. j#? ,. 9 it-  ;

SOURCE: RRRB 19.00ll.H )

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DESCRIPTION

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The compressor motors are ODP motors. Concern has been expressed about the, quality of the bearings, shaft keyway, and motor windings. gConsideration should he Civin to '

upgrading these motors to TEFC or Mill and Chemical motors,i, '

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' I'* ' ~ '

INVESTIGATION '

2 ',

, < \; -

y, . .i e An evaluation of the specifications for these motors will bre cantscted to'detennine '

the suitability of the existing motors for the intended farvi'ce and the; advisability ofupgradingreplacementmotorsformoresevereserviel.{ l j RESOLUTION OF PROBLEM t

Later TESTING D i Testing will not be necessary.

] s .

I

%/

G 65 L

i

l RESULTS OF REVIEW OF TEST RESULTS FROM DAVIS-BESSE I. Comparison of Testing Davis-Besse Tests Rancho Seco Tests TP 170.05 Control Room Heating. An approved formal procedure for Ventilation, and Air Conditioning testing the Control Room Normal Air Preoperational Test procedure checked conditioning system has not been system and component operability, identified.

control logic, and interlock functions prior to start up of the system but did not verify performance of the Control Room Emergency Ventilation System condenser or closure of the isolation dampers on hi chlorine signal.

ST 5076.01 is performed every 31 days SP 211.01A and SP 211.010 are to verify that the Control Room performed every 31 days. To verify Emergency Ventilation System is each respective train of the CR/TSC operable. This is done by starting Essential. Air Conditioning System is and operating each train for 15 operable, each train is operated for minutes, at least 15 minutes. .

ST 5076.02 is performed at least 'nce o SP 211.018 and SP 211.01E are every 18 months to verify that the performed at least once every 18 Control Room Emergency Ventilation months to verify that each train of System is operable. Operability is the Control Room Essential Air System verified by demonstrating: is operable. Operability is verified by demonstrating:

1. System air flow rate. 1. System air flow rate.

2. DOP removal efficiency of the 2. DOP removal efficiency of each HEPA filter bank. HEPA filter bank.

3. Halogenated hydrocarbon 3.. Halogenated hydrocarbon refrigerant test gas removal refrigerant test gas removal efficiency of the carbon filter efficiency of each carbon filter bank. bank.

O G-66 1

Davis Besse Tests- Rancho Seco Tests f'

\,

ST 5076.03 is performed every refueling outage to verify:

SP 211.01A and SP 211.01D are performed every al days to verify that

'- the Control Room / Technical Support

1. The Control Room Normal Air Center is isolated when each train of Conditioning System is isolated the Essential Air System is ' operated by a SFAS signal and a station in the radiological event mode. The vent radiation hi signal. capability of each train to start and operate upon SFAS was demonstrated by

2. The pressure drop across each performance of STP 194. SP 211.01B combined HEPA filter and charcoal and SP 211.01E are performed every absorber in the Control Room. _ refueling outage to verify that Emergency Ventilation System is differential pressure across the less than 4.4 I.W.G. while combined HEPA filter and Carbon operating at 3300 CFM 10% Absorbers is less than 6 I.W.G. while when supplying the Control Room operating at a flow rate of 3200 CFM with outside air. i 10%.

3. The air flow rate from the -

outside air makeup is 500 CFM i 10%.

ST 5076.04 is performed after every SP 211.01C and SP 211.01F are 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber performed after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation to verify that a charcoal adsorber operation to verify representative carbon sample is that a representative carbon sample is

, obtained and a laboratory analysis is obtained and that a laboratory l 1 completed within 31 days of sample aralysis is completed within 31 days removal. of sample removal to verify that

'] radioactive methyl iodide removal "

efficiency is at least 95%.

ST 5037.01 is performed at least every SP 211.01A and SP 211.010 are

18 months to demonstrate that the per?ormed every 31 days to verify that chlorine system is operable by tFe Control Room / Technical Support performance of a char.nel calibration. Center is isolated when each train of the Essentici Air System is operated in the toxic gas mode.

O G-61

II. Functional Test Procedures SP 211.01A (and SP 211.010); Control Room / Technical Support Center '

Emergency Ventilation System Loop A (and B) Monthly Surveillance Test.

SP 211.018 (and SP 211.01E); Control Room / Technical Support Center Emergency Filtering System Loop A (and B) Refueling Interval Surveillance Test.

SP 211.01C (and SP 211.01F); Control Room / Technical Support Center Emergency Ventilation System Carbon Absorber Filter 720 Hour Surveillance Test.

III. Review Findings The existing surveillance testing of the Control Room / Technical Support Center Essential Air Conditioning System, with the exception of isolation damper position indication, is adequate to meet the surveillance requirements of the Technical Specifications. However, -

surveillance procedures are not adequate to demonstrate cooling capability of each Essential Air Train.

IV. Recommendations A. Surveillance Procedures

1. SP 211.01A and SP 211.010 should be revised to include a check of all four of the possible position indications for each damper in the CR/TSC Essential Air System. This should be accomplished prior to restart.

2. SP 211.018 and SP211.01E should be revised to include measurements of flow thru the outside air intake and return air ducts to assure that a flow rate of 1600 CFM is maintained thru each flow path.

3. A new refueling interval surveillance procedure should be initiated to include an 8' hour operating period to check the cooling capability of the CR/TSC Essential Air System, a check of the Safety Feature actuation capability, a check of the radiological event actuation capability, and a check of the toxic gas vent actuation capability. This should be accomplished prior to restart.

4. SP 211.01A and SP 211.01D should be revised to include a check of refrigerant system operation.

O G-68

B. Periodic Maintenance Procedures

1. A new Periodic Maintenance Procedure should be initiated ~

to check the CR/TSC Normal Air Conditioning System. This will assure that the Essential Air System will not be unnecessarily challenged by failure of the Normal Air System.

2. A new Periodic Maintenance Procedure should be initiated to assure the Chlorine Detection System is adequately maintained. The procedure should include instructions for replenishment of material which has depleted or deteriorated during operation, adjustments and calibration, and functional checks.

3. A new Periodic Maintenance Procedure should be initiated to assure Essential Air System Dampers are properly maintained. The procedure should include instructions for inspecting, cleaning, and lubricating the dampers and -

damper actuators.

G-69

l l .

Results of Review of Recent Maintenance Activity 1 and Maintenance History Trends Complete Mechanical Work Recaest Listing W.R. 96861 Disassemble and inspect V-545A compressor with vendor.

W.R. 96B62 Disassemble and inspect V-5458 compressor with vendor.

W.R. 97237 Replace damaged carbon adsorber trays.

W.R. ,

97754 -

l Realign V Belts on AHD-A-545 A+B Air handlers.

W.R. 97755 l

SF-A-7B Filter unit replace flex connection on fan discharge.

[

W.R. 97752 AH-A-545A - Inspect and check vibration in fan and motor.

W.R. 97747 U-0545B - Inspect noise in Bearings on Compressor.

W.R. 97776 SF-A07A - Rebalance Fan W.R. 97779 l U-545A - Replace Gasket in Suction line.

W.R. 9779B AH-A-545A - Install kit in thermo expansion valve.

W.R. 97799 i

AH-A-5458 - Install kit in thermo expa:ision valve.

i 1

r--70 O

L Complete Mechanical Work Request Listing (Continued)

W.R. 98034 Install protective corners over Differential Switches on both A and B train.

W.R. 98035 SF-A-7A - Replace plexiglass sight glasses.

W.R. 98035 SF-A Replace plexiglass sight glasses.

W.R. 98244

.SF-A Replace plexiglass sight glasses. .

W.R. 98246 RS Repair small hole in carbon adsorber tray.

W.R. 98254 U-545A - Replace suction valve per N.C.R.

W.R. 98344 Train A - Repair leak in refrigerant copper pipe.

W.R. 98345 Train B. - Repair leak in refrigerant copper pipe.

W.R. 98345 l U-545 A and 8 - Refill compressor with oil.

W.R. 98538 U-545A - Remove motor end bells + inspect hearings.

W.R. 98752 U-545A - Replace notor bearings per N.C.R.

W.R. 94676 j

U-545 A and B - Various brazed joints need to be re-brazed.

G-71

Complete Mechanical Work Request Listing (Continued)

W.R. 94727 l

U-545 A and B - Remove compressor head gaskets and replace.

1 W.R. 110455 i

U-5018 Change filter on air conditioning unit.

l l W.R. 109120 Replace missing bottom door latch on U-5458 Essential a.c.

W.R.103962 Replace filter driers at evaporator coil in 545-A.

W.R. ,103962 _

Replace filter drier at evaporator coil in 545-B.

W.R. 103960 Replace filter driers on AH-U-28-NSEB condensing units.

l l W.R. 10128 l

Replace air filter on AHA-545A air handler.

W.R. 101096 Check the hot gas bypass valve on U-545-8 train.

l W.R. 101073 Replace gaskets on HEPA filters SFA-7A/B Essential HVAC.

Maintenance Review These two units have required considerable work for the type of service that is required of them. Part of this, of course, was the disassembly of the two compressor's for vendor inspection.

One common problem is noise and bearing failure in fans and compressors with several having been replaced in only one year of service. It is evident that some of these problems were with the units at the time they were put into service and also that some of the other requests were to repair damages incurred during construction.

G-72 O

Complete Mechanical Work Request Listing (Continued)

It is difficult to identify any maintenance trends from the limited service these units have had but it indicates the bearings in compressors and motors will require monitoring routinely.

The filter driers and Air flow filters will also require and servicing or changing on a routine basis.

l G-73 e -

Complete Damper Work Request Listing W.R.116355 Damper HV-54707 did not operate.

W.R. 114662 Damper HV-54705 did not indicate properly. Thirteen dampers were checked on this work request. Four were found to have indication problems.

W.R.111572 Damper HV054727 did not indicate closed.

W.R. 114656

, Damper HV-54732 Damper indicates both open and closed at same time. ,

W.R.113808 Damper HV-54732 Damper indicates both open and closed at same time.

W.R. 114655 Damper HV-54727 Damper indicates both open and closed at same time.

W.R. 103995 Noise in solenoid of actuator on HV-54732.

W.R. 103997 HV-54731 Noise in Solenoid when energized'.

W.R. 103990 HV-50135 Noise in Solenoid when energized.

W.R. 114864 HV-54715-M Adjust limit switch.

~~

W.R. 113609 HV-54710 Failed to reposition while conducting SP-211.010.

O G-74 1

= -

4

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1 I

Complete Damper Work Request Listing (Continued) qO W.R.115046 i V HV-54706 Failed to reposition ~while conducting SP-211.010.

W.R. 97527 HV-54733 Replace limit switch.

W.R. 97761 HV-547228 Replace pnewmatic opertor with same type.

W.R. 98250 Hv-54705 Replace motor or repair open winding.

W.R. 99494 FV-54702 Replace limit switch.

Damper Review The damper problems almost all fall into the category of incorrect indication and or incorrect close or open position. It appears that a routine maintenance program on the dampers which requires cleaning and relubricating

. on a periodic basis will remedy most of these problems. Some of these work requests reflect a lack of complete understanding as to where the damper should be when all conditions are considered, when running surveillance procedures.

l l One other common problem has been noise in the solenoid valves when they are l energized if this continues'to increase they will require replacement.

G-75 t

Complete Instrumentation Work Request Listing W.R. 114660 Investigate Radiation /H1 Temp Auto Start Circuit on Train B.

W.R. 114661 Investigate Relay CRSFB3/KAJ.

W.R. 114658 FT-54701 Flow indication intermittent when running.

W.R. 99967 Replace oil switch on V05038.

W.R. 38479 FT-54701 Transmitter has no output.

W.R. 98481 A Train flow indicating loop reads high.

W.R. 101524 PDISH-54701 Zero adjustment screw is jammed.

W.R. 101525 POISH-54702 Zero adjustment screw is jammed.

W.R. 101526 PDISH-54706 Zero adjustment screw is jammed.

W.R. 101544 POISH-54702 High alarm contacts are stuck closed.

W.R. 101559 PDI-54706 Indicator is sticking.

W.R. 88508 V-5458 Rework Fuse blocks in cabinet.

O G-76 J

Complete Instrumentation Work Request Listing (Continued)

( W.R. 96368 PDT-54702 Remove cross threaded instrument fitting.

W.R. 108836 Replace lens on power off lite on V-545A compressor.

I W.R. 101090 Calibrate oil pressure switch PDSL on N.S.E.B. Air Cond.

Instrumentation Review The thstrumentation work has been acceptable for this period of time after -

startup. Instrument failures are the greatest the first year and these 4

systems have not had that many. The,PDI's that were furnished with the units had some problems when delivered and were sent back for replacement. The new units have not experienced as many calibration problems.

A computer tracking system for instrument calibration is being developed and will facilitate problem identification.

i h

i l

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i G-77 L_ )

RESULTS OF SYSTEM MATERIAL CONDITION WALKDOWN A walkdown to determine the material condition of the CR/TSC Essential Air Conditioning System is currently in progress. The system has been in place for about one year; therefore, significant deterioration of the various components is not expected. A preliminary walkdown to evaluate the external condition of the system revealed that the system is in very good material condition.

Dampers are the first of the items to be examined in the walkdown. Access doors which allow visual examination of the dampers exist for 22 of the 28 system dampers. Six dampers do not have access doors- duct must be removed to gain access. Four of the 28 dampers have been determined to have functional problems.

O O

G-78

pu 4

+%

++ APPENDIX H SAMPLE TEST SPECIFICATIONS The attached Sample Test Specifications are representative of the test specifications which will be used in the test program.

Independent Control of AFH Control Valves Page H-1 O Independent Control of ADV's and TBV's Page H-3 Fire Protection Valve Drain Line Page H-5 Functional Test

+%

4+w O

r'x TEST SPECIFICATION

( ) Independent Control of AFW Control Valves Hevision 0 MODIFICATION NUC ENG ZANT ENGINEER NUC DESIGNATED ENGINEER

+

PURPOS To verify proper operation of the following:

1. Dual steam generator level indicator on H2PS.

2. Preset automatic positioning of the Auxiliary Feedwater (AFW) control valves on loss of ICS DC power.

3. Two manual control stations for the AFH control valves on H2PS.

REFERENCES:

'(~'N

• ECN R-0357A and associated drawings.

J General Calibration Procedures I-011.

'o'

• Calculation for preset AFH valve bias for Loss of ICS Power -

Z-FHS-IO105.

PREREQUISITES:

1. Construction complete and turned over to Startup with ny ptions duly noted and recorded.

2. Ensure simulated loss of ICS does not adversel e et the plant.

3. Auxiliary Feedwater lined up to permit strok of the AFW control valves.

TEST METHOD:

I&C Maintenance Department to veri e following:

1. Installation and operation of Solenoid Valves FY-20527C and Fy-20528C.

2. Installation and operation of Auxiliary Feedwater Valve Manual Control Panels HC-20527 and HC-20528 (located in H2PS).

l 3. Installation and operation of I/PS FY-205288.

4. Installation and operation of Steam Generator A and B Startup Level

! (/3,)

l

~./

Indicators LI-205038 and LI-20504B respectively.

1 H-1

The above will be verified by the following methods:

1. I&C Maintenance Departmen to verify installation and operation of s instrument loops by calib g instrument loops using loop form per General Calibration P du e I-Oll.

2. WithICSpoweravhable,andManualControlStationsinautomatic, verify that So no1Y Valves FY-20527C and FY-20528C are de-energized and verify t lowing:

a. Red cator lights above L&N Manual Controllers are not lit.

b. Control Valves receive pneumatic control signals from ICS via

/P's by using existing Bailey Hand / Auto staticns on Panel HlCO.

3. Simulate a Loss of ICS Power then verify that Solenoid Valves FY-20527C.and FY-20528C are energized and red indicator lights above L&N Controllers are lit. Then perform following:

a. With L&N Controllers (HC-20527 and HC-20528) in automatic, adjust the preset bias potentiometers to provide the required initial Auxiliary Feedwater Control Valve position for loss of ICS DC power. The valve position will be determined from the calculated AFW flow needed. Record the valve position at both AFW Valves L&N Manual Controllers (HC-20527 and HC-20528) indicator,

b. Place L&N Controllers in manual and calibrate such that 0 to 1007.

Indication on the Manual Control meter corresponds to O to 1007. -

Auxiliary Feedwater Control Valve stem travel. Verify "bumpless" transfer from Auto to Manual for the L&N Manual Controls.

4. Place the Manual Controllers in "AUT0" and remove simulated Loss of ICS Power and verify FY-20527C and FY-20528C deenergize a the red indicator lights go off.

a. Place L&N Controllers in " MANUAL" and verify u ontrol as in 3.b. Insure red indicator lights are lit.

b. Place L&N Controllers in "AUT0" and veri turn to normal ICS control by modulating AFW valves us th Bailey Hand / Auto Stations on HlRC.

5. With Manual Controllers in "AU no simulated Loss of ICS Power verify normal ICS Control as i for each of the following conditions.

a. Removal of AC power to the 24 VOC power supply for HC-20527 and HC-20528.

b. Removal of all AC power from SIN 1-1 to H2PS feeding tta new equipment installed by this modification.

ACCEPTANCE CRITERIA:

Equipment tested and results meet criteria under Test Methods.

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r TEST SPECIFICATION

_( N) wi Independent Control of ADV's and TBV's Revision 0 MODIFICATION .

NUC ENG ZANT ENGINEER -

NUC DESIGNATED ENGINEER RT-UP COORDINATOR PURPOSE To verify proper operation of the following:

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1. Selector switches for the Atmospheric Dump Valves (ADV's) and Turbine Bypass Valves (TBV's) on the HIRI panel.

2. Auxiliary relay contacts in the ICS.

REFERENCES:

ECN R-03578 and associated drawings.

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1 t PREREQUISITES: ,

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1. Construction complete and turned over to Startup with any exceptions duly noted and recorded.

2. Ensure simulated loss of ICS does not adversely effect th ant.

3. Plant in cold shutdown and secondary side of Steam Ge tors

depressurized or in a configuration to permit te g.

TEST METHOD:

i I&C Maintenance Department to verify the fo in :

1. Installation and operation of the h n TBV Override Switches, j HS-20562C and HS-205618 respec elyUTocated in H1RI.

2. Hiring to ICS relay 86/ICS-PSM c ntacts.

The above will be verified by the following methods:

ADVs

1. Hith both hand switches HS-20562A (located on Shutdown Panel H2SD) and HS-20562B (located on Atmospheric Dump Valves Manual Control Panel) in

" Auto" and HS-20562C (located on HIRI) in " Normal" verify:

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a. Solenoid Valves PY-205620, PY-20562E, PY-20562F, PY-205710, PY-20571E and PY-20571F are energized.

1. Simulate a loss o power and verify that the solenoid valves in 1.a y deenergize. Verify ADV's are closed.

2 Place the Override Switch HS-20562C in " Disable" and verify at he solenald valves in 1.a. above reenergize.

3. ceWie ADV Override Switch HS-20562C in " Normal" and remove simulated Loss of ICS Power Condition, verify that the lenoid valves in 1.a are energized.

Place the ADV Override Switch HS-20562C in "Close", verify that solenoid valves in 1.a. are deenergized. Verify ADV's are closed.

TBVs

2. With TBV hand switch HS-20561A (located on H2SD) " Auto" and HS-20561B (located on HlRI) in " Normal" verify:

a. Solenoid valves PY-20561A, PY-20563A, PY-20564A, PY-20566A are energized:

1. Simulate a loss of ICS Power and verify that the solenoid valves of 2.a. are deenergized. Verify TSV's are closed.

2. Place the TBV Override switch HS-205618 in " Disable" and verify that the solenoid valves of 2.a. above reenergize.

3. Place the TBV Override Switch HS-205618 in " Normal" and remove the simulated Loss of ICS Power Condition, verify that the solenoid valves in 2.a ara energized.

4. Place the TBV Override Switch HS-20561B in "C Ae", verify that solenoid valves in 2.a. are deenergi . Wrify TBV's are closed.

ACCEPTANCE CRITERIA:

1. Equipment tested and results meet cr ter nder Test Methods.

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O TEST SPECIFICATION (i 4 Fire Protection Valve Drain Line F tional Test Revision 0 MODIFICAT N NUC NG GE)GNIZANT ENGINEER b

NUC PS DESIGNATED ENGINEER TART-UP COORDINATOR PURPOSE:

This modification includes the addition of a drain line from the TSC fire protection valve, through the switchgear room and corridor, to a floor drain located in the ventilation equipment room. The drain line will have a seal trap which, when filled with water, will maintain the control room pressure boundary to preserve control room habitability as prescribed by Nureg 0737. This test will verify that the CR/TSC essential HVAC can maintain pressurization of the areas in accordance with tech spec limits following the completion of this modification.

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REFERENCES:

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\s/ 1. Tech Spec 4.10 " Control Room / Technical Support Center Emergency Filtration Unit"

2. NUREG 0737, Item III.D.3.4 - Control Room Habitability Requirements PRE-REQUISITES:

1. Grouting of penetration for drain line is comple

2. Water level in the seal trap is a minimum of s inches.

3. All other breached penetrations are fit wi h appropriate temporary or permanent closures.

TEST METHOD:

1. Independently operate the "A" an "B" CR/TSC Essential HVAC Systems in the radiological /h1 temp mode.

2. Measure the pressure differential between the Control Room and outside atmosphere using the 1/2" dia. stainless steel tube.

3. Measure the pressure differential between the technical support center and the corridor north of the TSC.

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r ACCEPTANCE CRITERIA:

1. The pressure of the Contr Room relative to the outside atmosphere shall equal or exceed 1/8 g.

l 2. The pressure of t Te ical Support Center relative to the corridor shall aqual or ex d 1/8" w.g.

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APPENDIX I QCI No.12, Eev. 1 July 11, lo c6 Attach:e te No. 3 Page 1 or 6 w

PIM'T STAFF INTERVIEW 1.0 Purpose The interview process shall identify any previously unidentified, but "known" problems which may impact the safe or reliable operation of systems or components of the plant. This procedure is to provide guidelines for the interview process.

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2.0 Scope Interviews will be conducted with, among others, plant operators, operations tech support engineers, plant engineers and technicians, training instructors, design engineers and quality department personnel. These personnel will be encouraged to identify system, component, or operational related problems, or concerns which they are aware of and any recommendations to. resolve them.

It is intended to interview personnel from each plant organizational group. Volunteers will be requested. All volunteers will be interviewed. If insufficient personnel volunteer, the interview program coordinator will perform a random selection of interviewees.

The minimum number of.personnuel to be interviewed are derived from QAIP

.Q 3, which is based on MIL-STD-105D. Other interviewees may be added at the discretion of the Interview Program Coordinator (Enclosure 1).

3.0 Res ponsibilities The responsibility for the conduct and docunentation of the interviews rests with the Interview Program Coordinator. The interview teams are responsible for conducting and documenting the interviews. The l Interview Program Coordinator is responsible for compiling the results of the interviews. The Interview Program Coordinator shall ensure that interviews are thorough and are completed in a timely canner. The Interview Program Coordinator is responsible to the Performance Analysis Group for the conduct of this program.

l Initially four interview teams will be utilized consisting of human I factors engineers and other knowledgeable plant representatives. These interviewers will be responsible for conducting the interviews and compiling the data. The interviewers are accountable to the Interview Program Coordinator for tabulation of interview recommendations.

Interviewers can be added or deleted as deemed necessary by the Interview Program Coordinator.

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QCI No.12,, Rev. 1 July 11, 1986 At tachment 3 Page 2 of 6 4.0 Procedure The number of interviews that will be required to ensure adequate coverage are listed 'in Enclosure 1. These are the minimum number required, but if the interviewers, with concurrence of the Interview Project Coordinator, determines more interviews are desirable, they may conduct them.

The interviews may be conducted with group of personnel to provide a synergistic approach, or on an individual basis, or a combination of both. The approach used will be determined by the interviewers with concurrence of the Interview Program Coordinator.

The interview will be accomplished by requesting the area supervisor to provide the name(s) of personnel that have volunteered to be interviewed. If insufficient personnel volunteer to meet the requirements of Enclosure 1, the Interview Program Coordinator shall select additiona1 personnel to be interviewed.

All nuclear managers and their superintendents shall also be approached for interviews.

All concerns, p oblems or recommendations identified during the interview shall be recorded on the Personnel Interview form, Enclosure

2. A copy of the questionnaire used for the interviews is included in the program plan.

All cor.cerns, problems and their respective recommendations (if any) from the Personnel Interview Form will be compiled into one list by the interview team. Redundant concerns, problems or recommenations will be coaso11 dated into one statement to reduce the number of concerns. If re.lundant concerns are consolidated, the total number of people with that concern will be noted by placing that number in parenthesis at the end of the consolidated concern.

The compiled list of concerns / recommendations will be forwarded to the Recommendation, Review and Resolution Board for Processing. The Interview Program Coordinator shall be considered the initiator on the Recommendation / Resolution forms.

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QC I No .12, ?,ev. 1 July 11, 1966 Attach:ent 3 Page 3 of 6 v

Af ter the concerns / recommendations have been dispositioned, it will be the responsibility of the Interview Program Coordinator to inform the interviewees of the disposition of their concerns. For group interviews, everyone in the group should be notified.

5.0 Budget and Plan The Interview Program Coordinator assesses the resources required to complete the tasks as prescribed and shall present the Performance Analysis Group with a plan that identifies the human and other ' resources required to execute the - tasks and the approxinate time required to complete all tasks. The PAG shall review the plan for completeness and for approach and shall direct approval to proceed or revisions. The PAG shall ensure that resources are available for the approved plan.

6.0 Interview Guidelines The interviewers shall ensure that interviews specifically address the following plant systems / areas: Reactor Coolant System, Reactor Auxiliary Systems, Secondary Systems, Plant Support Systees, and '

Instrumentation and Control Systems and electrical distribution, as they b apply to the interviewee's experience and knowledge.

The purpose of the interview should be kept obvious during the-interview: the identification of potential or kncwn problems within systems, components, or operational guidance that could impact reliable plant operation. Remember that the main point of the interview is to learn what others believe may be potential or actual problems. The interview team must stimulate the interviewee into describing total e periences.

The interviewers shall develop a questionnaire based upon the background of the interviewee. This questionnaire shall be used during the-interview to help the interviewee recall previous problems or concerns they or others have had.

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QCI No.12, Rev. 1 July 11, 1986 At tachnent 3 Enclosure 1 Page 4 of 6 MINIMUM NUMBER OF PERSONNEL TO BE INTERVIEWED Plant Operators Shift Supervisors - 3 Senior Control Roon Operators - 3 Reactor Operators - 5 Auxiliary Operators - 5 Equipnent Attendants and Power Plant Helpers - 8 Operations Tech Support Tech Support Engineers 'S Shif t Technical Advisors - 2 Scheduling - 2 Plant Maintenance I&C Technicians - 5 I&C Engineering - 2 Conputer Technicians - 3 Electrical Maintenance - 8 Electrical Engineering - 3 Electrical Techs - 3 Mechanical Maintenance - 8 Mechanical Engineers - 3 SMUD Security - 4 O

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', QC I No .12, F.ev . 1 July.11, 1986 Attachment 3 Enclosure 1 Page 5 of 6 MINIMUM NUMBER OF PERSONNEL TO BE INTERVIE*4ED NPS - 20 Licensing - 4 Training Instructors Operations Training - 5 Maintenance Training - 3 General Employee Training - 2 Nuclear Engineering Electrical - 8 I&C - 8

• b Mechanical - 6 Civil - 2 Project Engineering - 3 Quality Department QA Inspectors - 3 QC Inspectors - 3 QE Inspectors - 2 Corporate QA - 1 Rad Protection Senior Chem-Rad Assistant - 2 HP Tech - 5 Chemistry Senior Chem-Rad Assistant - 1 Chen-HP Tech - 5

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QCI No .12, Rev . 1 July 11, 1986 Attachment 3 Page 6 of 6 Enclosure 2 PERSONNEL INTERVIEW FORM No.

ITEM DESCRIPTION System Designation PLANT SAFETY / RELIABILITY - PROBLEM / CONCERN /CONSEQUFJCES RECOMMENDED SOLUTION (OPTIONAL)

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INTEEVIEWER'S INITIALS I -6 y