ML20046C551
| ML20046C551 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/02/1993 |
| From: | Crockett W, Dillard T, Giffin B, Miklush D, Ortore S, Vosburg D PACIFIC GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20046C542 | List: |
| References | |
| OLA-2, NUDOCS 9308110130 | |
| Download: ML20046C551 (494) | |
Text
D m m oCOnnEsFONDENCE August- 25.,[1993 D UNITED STATES OF AMERICA '93 AUS -4 P 3 :42 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
. .. g D
In the Matter of: ) Docket Nos. 50-275-OLA
) 50-323-OLA Pacific Gas and Electric Cor.pany )
) (Construction Period
'O (Diablo Canyon Nuclear Power ) Recovery)
Plant, Units 1 and 2) )
l
) l TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY
^
) ADDRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE I
() PART 1: Bryant W. Giffin, William G. Crockett,
( Steve R. Ortore, David A. Vosburg PART 2: Tedd A. Dillard O
PART 3: Bryant W. Giffin, David B. Miklush LO i
O e
9308110130 930802 O bDR ADOCK 05000275 PDR
)
TABLE OF CONTENTS
) Pace PART 1 - TESTIMONY OF GIFFIN/CROCKETT/ORTORE/VOSBURG . . . . . 1 I. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . 1
)
II. OVERVIEW . . . . . . . . . . . . . . . . . . . . . . . 4 III. DESCRIPTION OF DCPP MAINTENANCE AND SURVEILLANCE PROGRAMS . . . . . . . . . . . . . . . . . . . . . . . 10
) Technical Specification Surveillances . . . . . . . . 11 Other Surveillances . . . . . . . . . . . . . . . . . 12 ISI/IST Programs . . . . . . . . . . . . . . . . . . . 14 EQ Program . . . . . . . . . . . . . . . . . . . . . . 16 Maintenance Program . . . . . . . . . . . . . . . . . 18 IV. ELEMENTS OF AN EFFECTIVE AND COMPREHENSIVE
) MAINTENANCE PROGRAM . . . . . . . . . . . . . . . . . 22 Maintenance Organization ... . . . . . . . . . . . . . 25 Training and Qualifications . . . . . . . . . . . . . 29 Maintenance Facilities . . . . . . . . . . . . . . . . 34
) Preventive and Corrective Maintenance . . . . . . . . 38 Maintenance Procedures . . . . . . . . . . . . . . . . 42 Planning and Scheduling . . . . . . . . . . . . . . . 42 l
Post-Maintenance Testing . . . . . . . . . . . . . . . 51 Procurement of Parts . . . . . . . . . . . . . . . . . 53 Control of Measuring and Test Equipment . . . . . . . 58
) Root Cause Analysis Program . . . . . . . . . . . . . 58 Maintenance History / Failure Trending . . . . . . . . . 60
, V. MAINTENANCE AND EOUIPMENT AGING MANAGEMENT AT DCPP . . 62 Maintenance and Surveillance Programs and Activities . . . . . . . . . . . . . . . . . . . 62
/ 65 Plant / Equipment Improvements to Date . . . . . . . . .
Aging Management Program Directive . . . . . . . . . . 69 Steam Generator Tube Degradation . . . . . . . . . . . 72 Reactor Pressure Vessel Aging Management . . . . . . . 74 Erosion / Corrosion . . . . ... . . . . . . . . . . . . 77
.) EQ Equipment . . . . . . . . . . . . . . . . . . . . . 79 Structures . . . . . . . . . . . . . . . . . . . . . . 81 VI. ISSUES RAISED BY MOTHERS FOR PEACE . . . . . . . . . . 82 Containment Fan Cooler Units Backdraft Dampers . . . . 88
) Positive Displacement Pump Operating Procedures . . . 89 Steam Generator Feedwater Nozzle Indications . . . . . 91 i
)
)
Reactor Cavity Sump Level Indication . . . . . . . . . 93 Motor Operated Valves . . . . . . . . . . . . . . . . 96 Debris Found in Containment Building . . . . . . . . . 97
) Diesel Generator Turbo Charger Bellows Bolting . . . . 98 Diesel Generator Fuel Oil Piping Corrosion . . . . . . 99 Chemical and Volume Control System Leakage . . . . . . 100 Measuring and Test Equipment Control Deficiencies . . 102 Emergency Diesel Generator Surveillance Test Issues . 103 i Fuel Handling Building Ventilation Leakage . . . . . . 104
) RHR Recirculation Sump Screens . . . . . . . . . . . . 105 Turbine Governor and Stop Valve Malfunction . . . . . 106 4kV/12kV Cable Problems . . . . . . . . . . . . . . . 108 VII. CONCLUSION . . . . . . . . . . . . . . . . . . . . . . 112 i
)
PART 2 - TESTIMONY OF TEDD A. DILLARD . . . . . . . . . . . . 116
)
l PART 3 - TESTIMONY OF GIFFIN/MIKLUSH . . . . . . . . . . . . 160 l l
) I. INTRODUCTION , . . . . . . . . . . . . . . . . . . . . 160 II.
- PERFORMANCE EVALUATION OF DCPP'S MAINTENANCE AND SURVEILLANCE PROGRAMS . . . . . . . . . . . . . . . . 161 Introduction . . . . . . . . . . . . . . . . . . . . . 161 Plant Operating Performance . . . . . . . . . . . . . 163 Maintenance Goals and Objectives . . . . . . . . . . . 167
- 1. Industrial Safety . . . . . . . . . . . . . 168
- 2. Radiation Exposure . . . . . . . . . . . . 169
- 3. Personnel Contamination . . . . . . . . . . 170
- 4. Personnel Error Reduction . . . . . . . . . 170 171
- 5. Refueling Outages . . . . . . . . . . . . .
- 6. Corrective Maintenance Backlog . . . . . . 172
- 7. Overdue Preventive Maintenance Items . . . 172
- 8. Ratio of Preventive Maintenance to Total Maintenance . . . . . . . . . . . . . . . . 173
)
Qualitative Evaluations and Self-Assessments . . . . . 173 Maintenance Quality Assessment . . . . . . .. . . . . . 175 Continuous Program Improvement . . . . . . . . . . . . 177
- 1. Maintenance Process Improvement Project . . 178 r
3 2. Reliability Centered Maintenance . . . . . 179
- 3. Procurement Task Force . . . . . . . . . . 179 Regulatory Performance . . . . . . . . . . . . . . . . 180
- 1. NRC Programmatic Assessments . . . . . . . 180 y
ii
)
)'.
- a. NRC "Best Plants" List . . . . . . . 181
- b. SALP Ratings . . . . . . . . . . . . 182 h NRC Inspection and Enforcement 2.
Activities - . . . . . . . . . . . . . . 186 III. CONCLUSION . . . . . * - . . . . . . . . . . . . . . 188-
)
D l p ;
1 I
D D
D iii 3
)
GLOSSARY
)
ACRONYM DEFINITION AFR Audit Finding Report AFW Auxiliary feed water ALARA as low as reasonably achievable
) ANS American Nuclear Society ANSI American National Standards Institure AR Action Report ASME B&PV American Society of Mechanical Engineers Boiler and Pressure Vessel l BOP Balance-of-Plant
) CCPs Centrifugal Charging Pumps i CFCU Containment Fan Cooler Unit CHUG EPRI Erosion / corrosion users group DCPP Diablo Canyon Power Plant E/C Erosion / Corrosion l EDGs Emergency Diesel Generators
? EH electro-hydraulic EMF Engineering Manager's Forum EPRI Electric Power Research Institute EQ Environmental Qualification FHB Fuel Handling Building FME Foreign Material Exclusion TSAR Final Safety Analysis Report f GVs Governor Valves l
l HIT High Impact Team HPSI High pressure safety injection I&C Instrumentation and Controls IEEE Institute of Electrical and Electronics
) Engineers INPO Institute for Nuclear Power Operations IR Inspection Report l ISI Inservice Inspection IST Inservice Testing LCM Life Cycle Management
) M&TE Measurement and Test Equipment MFP San Luis Obispo Mothers for Peace l
MM Mechanical Maintenance l
MOV Motor Operated Valve MQAs Maintenance Quality Assessments MVT Modification Verification Tests
) NCRs Non Conformance Reports NOV Notice of Violation NPG '!uclear Power Generation Business Unit (PG&E)
NPRDS Nuclear Plant Reliability Data System NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System
' Nuclear Regulatory Reports NUREG OSRG Onsite Safety Review Group iv
)
i ACRONYM DEFINITION
) oVT operability verification Tests Positive Displacement Pumps PDPs j PG&E Pacific Gas and Electric Company i PIMS Plant Information Management System PM Preventive Maintenance PMT post-modification testing
) PSRC Plant Staff Review Committee PTS Pressurized Thermal Shock QA Quality Assurance QC Quality control QEs Quality Evaluations RCM reliability centered maintenance RHR Residual Heat Removal RMS Radiation Monitoring System j RPE replacement part evaluation
[ RTD resistance temperature detector RVRLIS Reactor Vessel Refueling Level Indication System h SALP Systematic A sessment of Licensee Performance SPDS Safety Parameter Display System SSCs structures, systems and components STP surveillance test procedures SVs stop valves TRG Technical Review Group
) Wo work order J
?
l D
)
i i
v ]
) 1
l ..
August 2, 1993 UNITED STATES OF AMERICA O NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 0 In the Matter of: ) Docket Nos. 50-275-OLA
) 50-323-OLA Pacific Gas and Electric Company )
) (Construction Period (Diablo Canyon Nuclear Power ) Recovery)
Plant, Units 1 and 2) )
.O )
TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY ADDRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE 10 PART 1: Bryant W. Giffin, William G. Crockett, Steve R. Ortore, David A. Vosburg O
O O
O O
I O
l l CL l 1 August 2, 1993 0 2 UNITED STATES OF AMERICA 3 NUCLEAR REGULATORY COMMISSION 4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD j C) 5 6 In the Matter of: ) Docket Nos. 50-275-OLA 7 ) 50-323-OLA 8 Pacific Gas and Electric Company )
9 ) (Construction Period 1 (Diablo Canyon Nuclear Power ) Recovery)
O 11 Plant, Units 1 and 2) )
12 )
13 TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY 14 ADDRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE
.O 15 I. INTRODUCTION
'O Please state your name, affiliation, qualifications 16 Q1 17 and current job responsibilities.
O 18 A1 (Giffin) My name is Bryant W. Giffin. I am the 19 Manager of Maintenance Services for Pacific Gas and 20 Electric Company ("PG&E") at the Diablo Canycn Power Plant O I am responsible for all maintenance and-outage 21 ("DCPP").
22 activities at DCPP. I have more than 25 years experience 23 working in the nuclear industry; 12 years with PG&E and
<O 24 over 13 years as an officer in the United States Navy's 25 nuclear power program. A summary of my professional 26 qualifications and experience is provided in Exhibit 1.
O (Crockett) My name is William G. Crockett. I am the 27
) 28 Manager of Technical and Support Services for PG&E at 29 DCPP. I am responsible for Engineering Support, including O
O 1 System Engineering, Plant Engineering, Training, Security, O 2 351 General Services. I oversee the development and 3 evaluation of the DCPP surveillance testing program. I am 4 tha primary site interface with the Design Engineering O 5 section of the Nuclear Engineering Services Department, l'
6 with PG&E's Computer Services Department, and with the )i 7 Nuclear Regulatory Services Department. I have 16 years O 8 of nuclear power experience in operations, maintenance, 9 and engineeriny, including 14 years at DCPP. I hold a 10 Senior Reactor Operators ("SRO") license (inactive) for O 11 both units at DCPP. A summary of my professional 12 qualifications and experience is provided in Exhibit 2.
13 (Vosburg) My name is David A. Vosburg. I am Director
- O 14 of the Work Planning Section in the Maintenance Services l
15 Department at DCPP. I am responsible for maintenance 16 planning activities and for scheduling of maintenance, I
O 17 post-maintenance testing, and plant modification 18 activities during normal and outage conditions. I have 16 l
j 19 years experience in the nuclear power industry, including i
3 20 13 years at DCPP. In my 13 years at DCPP, I have worked 21 in the Operations, Engineering, and Maintenance l
22 Departments. I maintained an SRO license for DCPP from f
'O 23 1982 to 1992. A summary of my professional qualifications 24 and experience is provided in Exhibit 3.
25 (Ortore) My name is Steven R. Ortore. I am the O 26 Director of the Electrical Maintenance Section in the 27 Maintenance Srrvices Department at DCPP. I am responsible t
28 for maintenance of all electrical equipment at the plant.
O 2
O r 1 I was formerly the Director of Material Services at DCPP, f
2 which included responsibility for procurement. I have 19 g
3 years of experience in the nuclear industry, of which 9 4 years have been at DCPP. A summary of my professional 5 qualifications and experience is provided in Exhibit 4.
a, 6 Q2 What contention will the panel address?
i O
7 A2 (All) We will address the San Luis Obispo Mothers for 8 Peace ("MFP") Contention I: Maintenance and Surveillance.
c, 9 MFP in the contention alleges that PG&E lacks a p
10 sufficiently effective and comprehensive maintenance and ,
I l
11 surveillance program to jcstify the operating license 4 l
, 12 amendment here at issue. That amendment would extend the l DCPP Unit 1 and Unit 2 operating licenses to the full 13 14 40-year term that was assumed in the original plant i
g 15 design. A copy of PG&E's license amendment application is 16 provided as E?.hibit 5.
l l
l 17 Q3 What is the purpose of your testimony?
,O l
18 A3 (All) The purpose of our testimony is to provide a i 1
19 response to the above contention. This testinony will.
) i 20 demonstrate that PG&E has implemented a comprehensive i l
21 maintenance and surveillance program at DCPP.
, 22 Maintenance and surveillance activities at DCPP address 23 degradation of equipment, whether due to equipment aging 24 or service wear, that has been experienced to date or that a
,)
0-1 is expected over a 4G-year plant operating lifetime. This O 2 specifically includes any pre-operational equipment 3 degradation. Contrary to MFP's claims, PG&E has 4 implemented its maintenance and surveillance programs in 5 an effe tive, and often outstanding, manner. PG&E O
6 recognizes the importance of maintenance and surveillance 7 to efficient, reliable, and safe operation. PG&E has 8 devoted considerable attention to this area and will 9
9 continue to do so for the full 40 years of operation to be I 10 authorized by the proposed license amendment.
O 11 II. OVERVIEW g 12 Q4 What is the definition of " maintenance" as the term is 13 used in your testimony?
14 A4 (All) Maintenance as used in this testimony includes:
O 15 those activities which are performed to assure that 16 structures, systems and components ("SSCs") will continue O- 17 to operate as designed, as well as those activities 18 necessary to repair or replace SSCs that are degraded or 19 cannot perform the intended function. Maintenance is Cy 20 considered to be the aggregate of actions at DCPP that (1) 21 minimizes the degradation or failure of SSCs, and (2) 22 promptly restores the intended function of SSCs if they o 23 experience operability or functional problens.
I 4 o-_
O 1 QS What is the definition of " surveillance" as used in 2 your testimony?
O 3 A5 (All) Surveillance as used in this testimony is the 4 aggregate of periodic tests and/or inspections that verify 5 that SSCs continue to function in accordance with 6 predetermined specifications or are in a state of 7 readiness to perform their particular safety functions.
O 8 Surveillance activities can trigger maintenance activities 9 based upon the results of the particular tests or 10 inspections.
O 11 Q6 What is the scope of issues raised in the contention 12 and addressed by your testimony?
l O
13 A6 (All) MFP in Contention I alleges that the maintenance 14 and surveillance pr' grams at DCPP are not comprehensive O .
15 and effective. Our testimony will demonstrate that the i 16 overall maintenance and surveillance programs established ,
17 at DCPP are comprehensive and effective in assuring the 18 safe and reliable operation of DCPP. The programs meet 19 all established requirements and are effective in 20 identifying and rectifying potential and actual sources of O
21 equipment degradation. Moreover, as will be discussed in
( 22 detail later in this testimony, the individual isolated 23 items cited by MFP as bases for Contention I indicate O
24 that, contrary to MFP's claims, our maintenance and I
s O
~ - _ _ _ _ _- _. _ _ __- -_ - ______ -____ _ _- _ - ______ - ____ _ -
D 1 surveillance programs are functioning well and as 2 intended, in a comprehensive and effective manner.
3 3 Q7 How do DCPP maintenance and surveillance programs 4 address " aging" of equipment 7 3
5 A7 (Giffin, Crockett) Equipment aging management is a l
6 concept that is integral to the overall philosophy of an 7 effective maintenance and surveillance program. PG&E 8 recognizes that SSCs can deteriorate as service life 9 increases due to " age-related degradation." " Age-related 3
10 degradation," as defined by the Electric Power Research l
11 Institute ("EPRI"), is the gradual deterioration in the 3 12 physical characteristics of an SSC due to aging mechanisms 13 and which occurs with time or use and could impair the 14 ability of the SSC to perform any of its design functions.
15 An " aging mechanism" is a process that gradually changes 3
16 the physical characteristics of an SSC with time or use.
17 Some examples of aging mechanisms are fatigue, erosion, 18 corrosion, radiation embrittlement, thermal embrittle'v:nt, 3
19 and service wear.
I 20 In the case of certain critical components which are l
21 subject to complex aging mechanisms (e.g., the reactor 3
22 vessel and steam generators), or components which may have 23 limited life, special programs have been developed and 7
24 implemented at DCPP to manage the aging process. These 25 programs are discussed in more detail later in this 26 testimony. For those components where special programs 6
3 L __
D 1 1 have not been developed, normal maintenance and 2 surveillance activities provide for and facilitate the 3 management of aging mechanisms. DCPP's maintenance and 4 surveillance programs are designed to (1) minimize l
L 5 deterioration through routine maintenance, (2) predict the J
l 6 rate of deterioration, and (3) detect the deterioration of i
7 equipment. Specific maintenance tasks respond to '
8 unacceptable SSC conditions either by equipment repair or j O
r I
1 l 9 replacement. !
1 10 DCPP, like other nuclear power plants, was designed to I
11 operate for at least 40 years. In order to assure safe I J
l 12 operation, a defense-in-depth philosophy was incorporated 13 into the plant design, the~ plant operating programs, and 14 the maintenance and surveillance programs. Specifically, U i 15 design features have been incorporated to provide the I 16 ability to test, inspect, and perform preventive 17 maintenance on SSCs. Furthermore, the plant was designed 18 with redundant trains of safety-related equipment to be l 19 able to accommodate equipment failures that are random in 20 time or location. The operators are also trained to J
21 recognize and respond to equipment problems, and the 22 naintenance programs are established to assure that 23 equipment problems, which are detected through routine
,J l 24 surveillance and operating experience, are corrected.
25 As the NRC has recognized in issuing numerous 40-year 26 operating licenses, and in issuing many license amendments 27 similar to that at issue here, effective and comprehensive 28 maintenance and surveillance programs assure that O
O 1 detectable aging effects are addressed prior to their
() 2 becoming a safety issue.
3 Q8 What is the relationship between the specific, O 4 individual events alleged by MFP in their bases advanced 5 in support of Contention I, and the effectiveness and 6 comprehensiveness of DCPP's maintenance and surveillance O 7 program?
8 AB (Giffin) Our testimony will show that the specific o 9 examples MFP uses to support its contention that the 10 maintenance and surveillance programs at DCPP are not 11 effective or comprehensive do not, in fact, indicate a 0 12 breakdown in the overall programs. None of the examples 13 resulted in a situation with safety significance. Some 14 are not maintenance-related and those that are do not o 15 prove an inadequacy in program scope or its overall 16 implementation. Some cf the examples are simply 17 situations in which the maintenance and surveillance O la program has worked precisely as intended. In all cases, 19 where specific equipment problems were identified or j 20 performance weaknesses were observed, corrective actions O 21 have been taken. PGLE believes that the examples cited by 22 MFP illustrate a maintenance program that is working and 23 reflect a strong commitment to continuous improvement.
O 24 MFP speculates that weaknesses in maintenance follow-25 from the DCPP rate case settlement and the financial 26 structure for DCPP rates. However, none of the examples O 8
D 1 cited by MFP in the basis for Contention I supports this D 2 speculation. None of the examples involved delay in 3 necessary corrective actions or repairs. In fact, 4 necessary repairs were made while the plant was operating.
}
) 5 Contrary to MFP, there are times when either unit of DCPP 6 is curtailed at PG&E's discretion in order for prudent l
7 repairs or inspections to be performed. These 8 curtailments evidence PG&E's commitment to safe operation
() 1 I
9 as the highest goal for DCPP.
O 10 Q9 Could you give us a quantitative measure of the i i 11 magnitude of the maintenance and surveillance program at {
t 12 DCPP?
O 13 A9 (Giffin, Vosburg) There are approximately 20,000 work 14 activities which require a work order performed each year
() 15 at DCPP. This includes 6,000 preventive and 7,000 l
l 16 corrective maintenance tasks. In addition, there are 17 approximately 10,000 surveillances performed each year, 1
C) 18 7,000 of which require a work order. Many of these work 19 activities consist of a large number of sequential steps 20 or elements. These are three year averages of the number
'O 21 of work orders performed. It is worth noting that tha 22 number of corrective maintenance tasks has decreased 23 steadily for the past three years. In 1990, there were 9 24 9,000; in 1992, 6,000; and we estimate fewer than 5,000 25 this year.
3 9 L_ _ __ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
1 O.
1 In the context of these numbers, it must be recognized i
2 that the design of the plant; the checks and balances !
O l 3 inherent in the maintenance and surveillance programs; 4 state of the art tools, test equipment, and facilities; 5 and the training and qualifications of the maintenance O
6 personnel all provide assurance that there will be a .
7 minimal number of errors made and that when errors are 8 made or equipment degrades, there will be no safety O
9 significance.
10 III. DESCRIPTION OF DCPP MAINTENANCE AND SURVEILLANCE PROGRAMS O
11 Q10 What programs at DCPP are included within the umbrella 12 of " maintenance and surveillance" programs? '
'O 13 A10 (Giffin, Crockett) At DCPP, PG&E has established 14 several programs or activities that addressing topic of
-O '
15 maintenance and surveillance of safety-related equipment.
16 These include:
17
- Technical Specification required surveillances;
- O 18
- Numerous equipment surveillances not required by the 19 operating license; 20
'O 21 Programs; ,
22
- The Environmental Qualification ("EQ") Program; and 23
- The Maintenance Program.
.O 24 These activities and programs collectively assure that any 25 significant degradation of plant equipment will be 10
'O
O 1 promptly identified and addressed throughout the proposed O 2 40-year operating license terms.
3 Technical Specification Surveillances
]
O 4 Q11 Please describe the role of Technical Specification 5 surveillances?
\
I O 6 All (Crockett, ciffin) DCPP Technical Specifications are 7 part of the plant NRC operating license and include 8 numerous requirements for testing and/or assessment of O 9 safety-related SSCs. PG&E has adopted a surveillance and lo testing program in accordance with the industry standard 11 ANSI N18.7-1976/ANS 3.2, " Administrative Controls and O 12 Quality Assurance for the operation Phase of Nuclear Power 13 Plants." In accordance with PG&E procedure NPAP-C3, 14 " Conduct of Plant and Equipment Tests," the DCPP o 15 surveillance and testing program administratively controls 16 the surveillance testing required by the Technical 17 Specifications. A computerized master schedule is used to l
'O 18 schedule and track the status of Technical Specification 19 surveillance tests and to ensure that these tests are 20 performed at the required intervals.
O. 21 over 10,000 Technical Specification-required 22 surveillance tests are performed at DCPP each year. The 23 actual number of SSCs tested is much higher because many O 24 surveillances actually test multiple components. If the 25 results of a surveillance test are not within specific 26 operability limits, the SSC is declared inoperable and the 0 11
O 1
'l Technical Specifications provisions are invoked. In these 2 cases, PG&E initiates problem investigations and necessary O
3 maintenance activities to restore the SSC to its design 4 condition in a timely fashion. Technical Specification 5 surveillance testing provides assurance that 6 safety-related equipment failures or substandard equipment 7 perfo"mance will not remain undetected and that the l-8 requirec reliability and state of readiness of SSCs to l' O
9 functicn in accordance with predetermined specifications 10 is maintained for the life of the plant.
O 11 Other Surveillances 12 Q12 Are there routine surv'eillances of SSCs at DCPP other 13 than Technical Specification surveillances?
f l
14 A12 (Crockett, Giffin) Yes. In a broad sense, there are 15 many plant activities in addition to Technical 16 Specification-required surveillance test procedures which 17 result in surveillance information about equipment 18 condition and performance. Examples of these activities 19 are:
20
- Routine operator plant equipment inspections as 21 required by Administrative Procedure DLAP O
22 OPl.DC3, " Auxiliary Operator Routine Plant i
23 Equipment Inspections;"
24 e Predictive maintenance program testing (e.g.,
25 vibration monitoring, oil analysis, and thermography) l 12 O
. 1 0
1 as specified in Administrative Procedure C-751, 2 " Predictive Maintenance Program;"
O 3
- Preventive maintenance program inspections (e.g.,
4 routine inspections of check valves) as required by 5 Administrative Procedures C-750, " Maintenance 6 Department Preventive Maintenance Program," and C-450, 7 " Instrument and Controls Preventive Maintenance;"
8
- Procedure functional checks (e.g., turbine trip q 9 testing during startup);
10
- Performance tests of important equipment not 11 controlled by the Technical Specifications (e.g.,
g 12 equipment controlled by Equipment Control Guidelines),
13 as described in Administrative Procedure DLAP ;
14 OPl.DC16, " Control of Plant Equipment Not Reg.uired by 15 the Technical Specifications;"
16
- Inspections performed as part of the erosion / corrosion l 17 m nitoring program required by Administrative 0
18 Procedure D-300, " Monitoring of Erosion / Corrosion .
19 Induced Pipe Wall Thinning;"
g 20
- Testing performed after maintenance to verify that 21 equipment performance has been restored to the 22 required level as specified in Administrative 23 Procedure C-6sa, " Post Maintenance Testing;" and O
24
- System engineer quarterly equipment walkdowns in 25 accordance with Administrative Procedure TS5.IDI, 26 " System Engineering Program."
O 27 All of these different tects and inspections provide I
28 detailed, specific, on-going information about the current ;
i 1
1 0
l l
)
1 condition of plant SSCs. These tests also provide 2 thousands of data points on the performance of plant 3 equipment, which are included and reviewed in the ,
l 4 preventive and predictive maintenance programs' Based on .
5 this information, maintenance is scheduled and performed 6 in a manner that maintains equipment performance at the
} 7 required level for the life of the plant.
[
l' 8 ISI/IST Programs 3 Q13 You listed in A10 above the ISI/IST Program. Please 10 briefly describe this category of surveillance testing.
11 A13 (Crockett) The ISI and IST Programs were initiated in 12 1985 for Unit 1 and in 1986 for Unit 2, corresponding to
) 13 the start of commercial operation of each unit. The ISI 14 and IST Programs comply with the requirements of 10 CFR 15 50. 55a (b) (2) and 50.55a(g), as well as the requirements in 16 the Technical Specifications. The ISI and IST Programs 17 include inspection, testing, and maintenance of pressure-18 retaining components (including their support structures) 19 as required by the American Society of Mechanical 20 Engineers Boiler and Pressure Vessel ("ASME B&PV") Code.
21 Components that are within the scope of the IST 22 Program are designated pumps and valves that are required j 23 to perform a specific function in shutting down the 24 reactor or mitigating the consequences of an accident. i 25 Periodic pump tests are performed in accordance with ASME I
I 26 Section XI, subsection IWP. Such tests measure
].
1 m __ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . .
O i
- 1 operational pump performance by observing, measuring and I
i 2 recording specific data such as pump / motor vibration, !
, 3 flow, and bearing temperatures. Also, periodic valve ,
4 4
4 tests are performed in accordance with subsection IWV of -
5 Section XI. These tests measure the performance of LO 6 power-operated valves, safety valves, and check valves, t
7 As applicable, depending on the valve type, these tests 8 check stroke, stroke time, seat leakage, and relief O 9 setpoints. For both pump and valve tests, data are 10 analyzed and compared to specific criteria for 11 operability.
12 The ISI Program specifically includes nondestructive 13 examinations such as visual, surface, and volumetric 14 examinations. The surface examinations are done with O 15 liquid penetrant or magnetic particle methods. The 1
16 volumetric examinations are done using ultra.'onic or ;
17 radiographic examination methods. The objective of these ,
O 18 examinations is to:
19 e Identify any unexpected service-induced component s
20 degradation, which would be evidenced by surface O 21 cracks, wear, corrosion, or erosion; 22
- Locate any evidence of component leakage during system 23 pressure or functional tests; and O 24
- Verify operability of components and integrity of 25 component supports. ,
26 10 CFR 50.55a(g) requires revision of the ISI and IST ,
O 27 Programs as necessary to comply (to the extent practical 28- within the limitations of design, geometry, and materials ,
15 O
i O
1 for construction of components) with the edition of the O. 2 ASME B&PV Code and Addenda in effect and adopted by the 3 NRC twelve months prior to the start of each 10-year 4 inspection interval. These programs ensure that pressure-O 5 retaining components will be adequately inspected, tested, 6 and maintained throughout the proposed 40-year operating 7 license terms.
O ;
8 EO Procram !
9 Q14 How does the EQ Program relate to maintaining plant -
0 10 material condition through the proposed 40-year operating 11 term?
9 12 A14 (Ortore) EQ is a rigorous program to confirm that 13 electrical equipment which would be relied on in the event 14 of an accident will be capable of performing its design 9 15 safety function to assure safe shutdown of the reactor, ,
16 despite exposure to the harsh environment postulated to 17 result from an accident. The process of environmental b
Q 18 qualification includes:
19 e determining which plant components are required to be 20 operable in a harsh environment; 21 e defining the environmental conditions which each O
22 component may be exposed to, and for how long; ,
23 e using appropriate testing and/or analysis to l
O 24 demonstrate the component will operate in the harsh ;
25 environment for the period of time; O- 16
)'
1 e determining a " qualified life" based on the expected 2 service conditions;
]
3 e identifying and implementing appropriate surveillance, 4 maintenance and procurement requirements to assure the 5 environmental qualification of the component is
]-
6 maintained; 7
- documenting all of the above.
] 8 The EQ Program for DCPP complies with the requirements 9 of 10 CFR 50.49. As applied to DCPP, 10 CFR 50.49 10 requires that electric equipment important to safety and 11 located in a harsh post-accident environment be
])
12 environmentally qualified, at a minimum, in accordance 13 with IEEE 323-1971 and the Category II positions in 14 NUREG-0588, dated December 1979. In accordance with 10 15 CFR 50.49(1), replacement equipment (for equipment that is 16 required to be environmentally qualified) is required to 17 be qualified in accordance with IEEE Standard 323-1974 and
{}
18 the Category I positions in NUREG-0588, unless there are 19 sound reasons to the contrary. The DCPP EQ program was 20 evaluated and found by the NRC to be in conformance with 21 applicable requirements in 1981 and 1985.
22 As described in detail later in this testimony, D- 23 maintenance of EQ equipment is a formal program at DCPP.
24 The master list of equipment required to be
-25 environmentally qualified is maintained as a controlled J- 26 engineering drawing and is revised as plant design changes 27 are implemented. Surveillance activities are performed-to 28 detect adverse trends in equipment performance or the J- 17 j
-O i 1 normal operating environment. Maintenance procedures
() 2 assure that the qualified configuration of equipment is 3 restored after maintenance and that appropriate 4 maintenance activities are conducted to preserve the O 5 qualified status. Equipment that is not qualified for the 6 entire 40-year operating license term is refurbished or 7 replaced prior to exceeding its " qualified life."
O 8 Maintenance Procram 9 Q15 The major element relevant to MFP's contention under Q 10 the maintenance and surveillance umbrella referenced in 11 A10 is the Maintenance Program. Please describe this 12 program generally.
O 13 A15 (Giffin) As noted above, maintenance is the integrated 14 means of maintaining the plant material condition 0 15 throughout the plant's operating life by managing the 16 effects of degradation and service wear on SSCs. The DCPP 17 Maintenance Program was developed with the basic O 18 philosophy that it is necessary to have the requisite 19 administrative and technical controls to ensure that 20 maintenance is performed in a timely, controlled, and safe O 21 manner, consistent with applicable requirements, the 22 license, and quality control criteria.
23 The DCPP Maintenance Program has been implemented O 24 through procedures which incorporate relevant information 25 from the Technical Specifications, design basis criteria, 26 industry standards, equipment vendor and manufacturer i
O is m- _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____
0 1 recommendations, NRC Safety Evaluation Reports and PG&E
- g. 2 maintenance experience at both nuclear and fossil plants.
3 They provide the means to monitor, inspect, maintain, and 4 test plant SSCs in a programmatic manner. The program 5 includes maintenance tasks on both safety and g
6 non-safety-related SSCs.
7 As will be described further in this testimony, 8 maintenance tasks are categorized as either preventive or
)
9 corrective maintenance. Preventive and corrective 10 maintenance tasks were initially identified and initiated 11 during the plant construction phase. Maintenance g
12 activities have been conducted throughout the construction.
13 phase, the system turnover to plant staff, the period from g 14 operational testing through the start of commercial 15 operation, and during commercial operation. This helps to 16 assure that SSCs were not adversely affected during the 17 period of plant construction. Additionally, Westinghouse O
18 was present onsite during the construction phase and has 19 provided preventive maintenance information and guidelines 20 for the Nuclear Steam Supply System ("NSSS") equipment it O
21 supplied. The Westinghouse guidelines also include 22 recommended chemistry controls and system layup 23 conditions.
O 24 In addition to procedural guidance on specific 25 maintenance activities, maintenance procedures also 26 provide for scheduling, implementing, and documenting O.
27 activities covered by the Maintenance Program. PG&E
\
28 installed a computer-based Plant Information Management 19
'O l
O 1 System ("PIMS") in 1985 to assist plant maintenance and 2 engineering personnel in these activities. PG&E also O
3 committed additional resources to develop state-of-the-art 4 machine shops, maintenance training facilities, spare 5 parts invent ries, and management systems. Staff O
6 resources and personnel training are provided to fully 7 implement and use these support resources. All of these 8 program elements are described in more detail later in O
9 this testimony.
1 Q16 Y u menti ned PIMS. What is PIMS?
O 11 A16 (Crockett) PIMS is a computer-based system that 12 utilizes a mainframe computer system at DCPP. PIMS runs O
13 on a local area network and is available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day to 14 over 3,000 users. It functions to improve information 15 fl w, w rk planning, and productivity. It is regarded as
'O 16 one of the most comprehensive information systems in the 17 nuclear industry.
g 18 PIMS specifically provides users with the ability to 19 manage and obtain up-to-date information directly related 20 to:
21
- Problem reporting and tracking O
)
22
- Regulatory commitment management 23
- Plant component information and history )
I l
1 24
- Maintenance task instructions and history f 0 i 25
- Measuring and test equipment calibration tracking )
26
- Materials purchasing and processing 20 g
3 1
- Inventory control g 2
- Radiation exposure tracking 3
- Personnel and training records 4
- Plant access management 5 A typical day will result in over 200,000 completed O
i i 6 PIMS electronic transactions.
7 Q17 What are the regulatory requirements for maintenance j 3
f I
8 and surveillance programs?
g 9 A17 (Giffin) The requirements for the implementation of 10 maintenance and surveillance programs are found or <
11 referenced in several places. The DCPP Technical g 12 Specifications provide surveillance requirements for {
13 safety-related equipment and specify the requirements for 14 the ISI/IST program as described above. The Code of 3 15 Federal Regulations, 10 CFR Part 50, also provides certain 16 specific requirements, such as the EQ requirements. These 17 provisions, however, contain no explicit requirements for 3 18 a maintenance program. PGLE has committed in the DCPP 19 Final Safety Analysis Report Update ("FSAR") to follow the 20 standards of ANSI 18.7-1976/ANS-3.2, " Administrative g 21 Controls and Quality Assurance for the Operational Phase l 22 of Nuclear Power Plants," as endorsed and modified by the 23 NRC, in Regulatory Guide 1.33, " Quality Assurance Program 24 Requirements (Operation) . " PG&E also has long been lg
, 25 actively involved with establishing maintenance guidance
)
26 for the nuclear industry. In 1971, a PG&E manager was the l
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i 3r i 1- chairman of the committee that authored ANSI 18.7/ANS 3.2.
sy 2 In 1976, one of our present executive officers coauthored I r
J 3 the 1976 version which we committed to follow. This ;
4 continuous involvement by PG&E demonstrates our management {
5 commitment to fostering comprehensive maintenance programs D
6 in the nuclear industry. These documents provide guidance 7 concerning.the maintenance and testing of safety-related 8 SSCs. Accordingly, the maintenance program PG&E has ;
- ) .
i 9 established at DCPP is much broader in scope than required 10 by any of these documents. For example, our program 7 11 includen many maintenance activities for nonsafety-related :
0 >
12 and balance-of-plant (" BOP") equipment.
13 The NRC is implementing a new maintenance rule ;
(10 CFR 50.65) which will become effective in 1996. Our
- l 14 '
O I 15 preliminary assessment is that our existing programs meet 16 the requirements of this new rule and that few changes to
- I 17 our programs will be required. ,
! 18 IV. ELEMENTS OF AN EFFECTIVE AND COMPREHENSIVE I
[,,
J 19 MAINTENANCE PROGRAM L
20 Q18 Focusing now on the DCPP Maintenance Program, is there L 21 any objective standard by which to judge the proper scope O
5 22 of such a program?
t 23 A18 (Giffin) Yes. Maintenance is a very complex and ;
24 interrelated program which includes (1) the identification t
25 of an equipment problem or potential equipment problem, {
22 ;
3- i
6 1 (2) the planning of the work activity intended to prevent g 2 a problem from occurring or to return the equipment to its 3 design condition, (3) the preparation of a detailed 4 procedure or work order defining the steps necessary to
- ) 5 perform the work, (4) scheduling the work activity, 6 (5) obtaining the parts which are necessary to complete 7 the task, (6) assigning qualified personnel to actually
[) 8 perform the work, (7) performing the work, and (8) 9 performing necessary testing at the completion of the work 10 to ensure that the equipment is in operable condition.
[) 11 With this scope in mind, the nuclear industry and the NRC 12 have generally agreed that the Institute of Nuclear Power 13 Operations ("INPO") document 90-008, " Maintenance Programs 14 in the Nuclear Power Industry," (Revision 1, March 1990),
[)
15 identifies the requisite elements for a comprehensive 16 maintenance program. INPO 90-008 was recognized by the
[) 17 NRC in its Statement of Considerations for the Maintenance 18 Rule, and the NRC expressly found that detailed 19 recommendttions for the conduct of maintenance, such as
() 20 INPO 90-008, should be developed by the licensee, not the 21 NRC. (See 56 Federal Register 31,313, July 10, 1991.)
22 These elements are:
) 23
- A well-staffed and qualified organization that is 24 provided with the tools and facilities necessary to 25 perform tasks effectively and efficiently; e 26
- A proper nix of corrective and preventive maintenance 27 to provide assurance that equipment degradation is 28 identified and corrected prior to failure; i
g 23
1:O '
- 1 1
1
- Accurate procedures or work instructions for craftsmen 2 so that the work activities can be performed in a i
)-
3 quality manner; a
4
- Maintenance planning and scheduling to assure that all 5 involved plant departments are aware of the activities
- 0 v
6 and interferences are minimized; ;
, 7
- Post-maintenance testing to verify that the 4 8 maintenance task was performed correctly and the 3 0 9 equipment is ready to be returned to service; 1 10
- Availability of correct and qualified parts and 11 material to support the repair and return to service 12 of the component; ,
i 13
- Control of measuring a'nd test equipment to ensure the 1 14 accurate performance of instrumentation and equipment O
, 15 used for calibrations, testing, and repairs; 16
- A detailed root cause analysis program to understand ;
17 the cause of equipment failures; and
- 0 1
- 18
- A maintenance history program to provide historical i 19 data for maintenance planning and to support trending 20 analyses of equipment performance.
O 1 21 Similar but less-detailed elements are identified in J 22 ANSI 18.7-1976/ANS-3.2. The elements of INPO 90-008 have
- O 23 been addressed in the DCPP Maintenance Program and are 24 explained in our Program Directive MA1 " Maintenance." The 25 Program Directive is included with this testimony as 1 1
26 Exhibit 6.
) I
- Q i
D 1 Maintenance Orcanization 2 Q19 The first element of a maintenance program is a well-3 3 staffed and qualified organization with necessary tools 4 and facilities. Let us break this element into three sub-5 topics and address each: the maintenance organization, 6 the training and qualifications of maintenance personnel, 7 and maintenance facilities. First, please describe the 8 maintenance organization for DCPP?
g 9 A19 (Giffin) The DCPP maintenance organization is clearly 10 identified and has been communicated to personnel so that g
11 responsibilities, lines of communication, performance 12 objectives, and mission are clearly understood. The DCPP 13 Maintenance Services Department organization chart is g
14 included as Exhibit 7 to this testimony.
15 About four years ago, the DCPP maintenance functions 16 were part of three different departments. A decision was
)
17 made to arrange the plant organization along functional 18 areas. Accordingly, the Maintenance Services Department 19 was established and the sections having maintenance
)
20 responsibilities were assigned to that department. The 21 Maintenance Services Department now consists of six 22 sections: Electrical Maintenance, Mechanical Maintenance, g
23 Instrumentation and Controls ("I&C"), Work Planning and ;
24 Scheduling, Materials, and Outage Management. There are g 25 approximately 598 employees in the Maintenance Services 26 Department, 300 union employees, 298 professionals. This ;
i i
g 25
L l
! l O '
l 1 organization is well-staffed and well-suited to its 2 purpose.
.O 3 Q20 Please describe each organization within the 4 Maintenance Services Department individually, with the
.O 5 section responsibilities and the staffing.
,O 6 A20 (Giffin) The Mechanical Maintenance Section is >
! 7 responsible for maintaining and servicing all mechanical l 8 components at DCPP. This includes the supervision and
! 9 programmatic controls needed to ensure that all components 10 10 are monitored and maintained in reliable working 11 condition. Engineering support within the section 12 provides procedural guidance, troubleshooting, and 13 techrical direction. The Mechanical Maintenance Section 14 consists of 157 employees.
15 The I&C Section is responsible for the maintenance and 16 periodic testing of all plant instrumentation. Duties 17 include the supervision and administration of the 18 preventive and corrective maintenance programs for that 19 equipment, and the implementation of the Technical
! 20 Specification instrumentation surveillance program.
! 21 Engineering support withi: the section is provided in the
- O (i 22 areas of problem troub'eshooting, test reviews and 23 direction, tracking and trending of component failures, 24 design change scoping and installation sponsorship, plant
[O 25 computer systems maintenance and programming expertise, 4
26 integrated communication interfaces, design engineering ,
26
- O
l l
CT l l
l r
! i support, and quality problem evaluations and reviews. The l
() 2 I&C Section consists of 123 employees.
3 The Electrical Maintenance Section is responsible for l i
4 maintenance of all electrical plant equipment, including d) 5 main generator and exciter, all motors, generators, i
6 switchgear, batteries, chargers, inverters and motor :
f 7 operated valves ("MOVs"). Engineering support is provided l O 8 within the section. The Electrical Maintenance Section 9 consists of 88 employees. ;
10 The Work Planning Center is comprised of two main ;
O 11 functional groups: Work Planning and Work Scheduling. The i
12 overall responsibility of the Work Planning Section is to y 13 provide the plant maintenance sections with detailed ;
! f O 14 working documents that reflect a safe, efficient work plan 15 that is in compliance with all plant procedures, programs 16 and regulatory requirements. The Work Scheduling Section O 17 is responsible for the coordination and scheduling of all 18 plant work activities in a manner that maximizes efficient <
19 utilization of plant resources and enhances the safe and j i
O 20 reliable operation of the plant. The Work Planning Center !
l t 21 consists of 105 full time personnel.
! 22 The Outage Management Section is responsible for the
'O 23 planning, scheduling and direction of all activities j j
24 associated with refueling outages and unscheduled outages .
l t i i 25 of sufficimit duration or complexity. The outages are (
hO 26 directed from the Outage Control Center, utilizing a
[ 27 representative from each of the major disciplines on a l
28 seven day a week, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day basis. Outage Management I O 27 l
t
() ;
l i ,
( 1 is also responsible for the Lessons Learned program which ?
(1 2 captures items that worked well or things that need to be 3 improved during outages. During non-outage periods, there
- 4 are 6 people assigned to the outage management team. ;
{} 5 These people coordinate the efforts of schedulers and 6 planners. During an outage, the maintenance services 7 organization is augmented by about 400 engineers and O 8 craftsmen.
9 The Material Services Section has about 118 full time 10 employees and is responsible for the procurement of O 11 material (equipment and parts), warehousing, issuance and l 12 repair of tools, and the calibration of measuring ar.d test l !
13 equipment. Procurement, within this section, also has
() 14 responsibility for defining the technical and quality 15 requirement of material requests.
i i
O 16 021 Are there examples of organizational methods utilized ;
l 17 at DCPP, beyond what you have just described, to foster l
18 continued improvement?
O. i 19 A21 (Giffin) Yes. As part of our commitment to continuous 20 improvement in the maintenance organization, we have l
'O 21 instituted the High Impact Tean (" HIT") concept. A HIT is i
( 22 a multi-discipline team with members from a vertical slice i
. 23 of the organization who are selected based upon the ,
- 5) 24 particular maintenance issue or project involved. HITS !
! 25 are formed for complex outage tasks, to facilitate 26 teamwork for new projects, or for other complicated multi- l l
,O 28 ;
1 .
l O i l
1 discipline tasks. HITS are given the resources, l i
O 2 abilities, and authority to implement actions and [
3 improvements so that tasks are accomplished in an !
4 excellent and improving manner. Industry and PG&E
'O 5 management have recognized the efficiency, quality, and j 6 productivity benefits of employee teams. The DCPP HIT ,
7 teams have proven to be a very successful example. By :
O 8 having a management team which supports continuous i 9 improvement and employee involvement, coupled with HITS to j 10 look for and implement improvements, a working environment O 11 exists at DCPP where change for the better is expected.
l 12 Trainina and Oualifications j f
- O 13 Q22 Please describe the qualifications of the personnel .
14 involved in maintenance activities.
l l
~O 15 A22 (Giffin, Crockett) PG&E has committed to the ;
T
- 16 requirements of ANSI /ANS 3.1-1976, "For Selection and l l
17 Training of Nuclear Power Plant Personnel." The !
O IB qualifications of the DCPP maintenance staff are quite -
l 19 high, and exceed the ANSI /ANS 3.1-1976 requirements. Five 20 of the seven section directors in the Maintenance Services' l O 21 Department have been either licensed or certified SROs, 22 three have advanced technical degrees, and they all have l
- . 23 over 15 years experience in the nuclear industry. The I
k) 1 24 maintenance foremen and general foremen have, on average,
- 25 15 years of nuclear experience. The qualification of i 26 maintenance craftsmen who perform specific maintenance O 29
- 3 ' l l
l 1 tasks are ensured by a rigorous training program. In the 2 NRC's last Systematic Assessment of Licensee Performance
,O T i 3 ("SALP") report on the Maintenance area at DCPP (February 4 12, 1993), the NRC listed as a noteworthy strength:
5 " Qualifications: The training and qualification ;
(O 6 program for Maintenance personnel was strong. !
7 Well maintained training facilities and a j 8 dedicated training staff were significant factors l 9 in good performance, as was the sense of ownership ,
l 10 shown by Maintenance personnel." -
0 11 Maintenance training programs at DCPP have been 12 accredited by INPO. These accredited training programs 13 include Mechanical Maintenance, Electrical Maintenance, O 14 I&C, and Maintenance Supervisory Training. INPO 15 accreditation is an ongoing process that is re-evaluated ,
16 every four years. The maintenance training programs at ,
O 17 DCPP were first accredited in 1988 and accreditation was
- l 18 renewed in 1992. As a result of obtaining and maintaining l 19 accreditation for all of the INPO identified programs, ;
O 20 DCPP has been designated as a full member of the " National '
l 21 Academy for Nuclear Training." .
22 In recognition of DCPP's commitment to training and O 23 the quality of its programs, DCPP's training' director was 24 selected to participate with INPO and the Department of ,
l 25 Energy in a project to assist the Russian nuclear O 26 organizations with their training programs.
1 I
^
27 Q23 How are maintenance personnel trained?
4 9
l 30 i O ;
G 1 A23 (Crockett, Giffin) DCPP training programs are designed 3 2 to provide journeymen and foremen who will perform 3 specific maintenance tasks with the skills and knowledge 4 needed to successfully and safely complete their work. The 9 5 training programs meet the requirements of ANSI /ANS 3.1, 6 the state-approved apprenticeship program, and the INPO 7 accreditation criteria.
3 8 The maintenance training programs are comprised of 9 distinct components:
10
- Apprenticeship 3 11
- Basic (or Fundamental) Qualification for Journeyman 12
- Advanced (or Select) Qualification for Journeyman 13
- Supervisor Training 3 14
- Continuing Training 15 The apprenticeship program is State of California 16 approved and is a negotiated contractual agreement with 3 17 the International Brotherhood of Electrical Workers. Once 18 in an apprenticeship, individuals are given two or three 19 years of training and guidance in their chosen field.
3 20 Examples are:
21
- Machinist (3 years) 22
- Electrician (3 years) 3 23
- Welder (3 years) 24
- Instrument Repairman (2 years) 25
- Control Technician (2 years - must have already 3 26 completed either electrical or instrument repairman 27 requirements).
3 31
1 1
l O- )
l 1
1- When job vacancies for craftsmen occur, they are ,
i 2 filled by graduates of the apprentice program. However, g
3 sometimes these vacancies are filled by journeymen from :
i 4 outside of the DCPP maintenance organization. These ,
- g. 5 journeymen may come with a variety of backgrounds.
6 Although they are considered to be journeymen, their 7 specific skills, knowledge, and experience may be ;
8 different than that specifically required for basic O
9 qualification at DCPP. Thus, all such personnel are 10 tested upon entry into the Maintenance Services l t
11 Department. The test results, along with interviews by O ,
12 supervision and review of previous work documentation, f 13 determine the initial training needed to meet the 14 requirements f r Basic (or Fundamental) Qualification for O !
15 Journeyman.
16 The topics of the Basic qualification training 17 Program, and the amount of training available, are shown O
18 in Exhibit 8. Once an incoming journeyman satisfies the i
19 Basic (or Fundamental) qualifications, further' advanced [
r 20 training on selected plant components is given based on O
21 need. Examples of advanced or plant-specific training, ;
i 22 and the time available, are shown in Exhibit 9. [
23 In addition to plant-specific training on components z) 24 and equipment, all maintenance workers attend continuing :
25 training seminars on a quarterly basis, except during_
g 26 outage periods. These seminars are used to address plant 27 and industry issues, changes recently made to the i 28- facilities or equipment, and recent " lessons learned." {
32 O- l
O i
1 These seminars also afford management an opportunity to r
O 2 stress areas of concern. ;
3 Q24 How do you utilize " hands-on" training at DCPP? ,
O ,
4 A24 (Crockett, Giffin) All aspects of DCPP maintenance 5 training programs stress " hands-on" training activities.
O 6 The majority of the classes are taught in a lab or shop 7 facility and take advantage of plant-specific training 8 aids. The classes consist of lectures with examples, !
O 9 coupled with actual hands-on experience on equipment 10 utilizing actual plant procedures and work packages 11 similar to those used to perform maintenance in the plant.
O 12 Examples of plant specific equipment available in the ;
13 training facilities are: !
14 e Reactor Coolant Pump (seal and motor alignment ;
O 15 portions) 16 o Steam Generator Channel Head and Tubesheet 17 e 10% Steam Dump Valve O 18
- Limitorque valves and operators 19
- Reactor Trip Breaker 20
- Digital Feedwater Controls O 21
- Nuclear Instrumentation and Protection Systems 22
- Diesel Engine 23 With the extensive training aids available in our O 24 training program, the majority of tasks requiring !
25 qualification can be fully accomplished in the training j
\
26 facilities. However, in soi.e instances additional "on the i I
O 33 l
l J
O 1 job" training in the plant is required prior to ,
O 2 qualification. An individual's qualifications are tracked 3 by a qualification matrix that identifies the tasks an 4 individual is qualified to perform. The foremen makes use O .5 of this matrix when assigning work to assure that only !
-i 6 qualified personnel are assigned to a job. When possible, ,
7 workers are also given the opportunity to train on new ;
O 8 equipment prior to its installation in the plant.
9 This extensive training program demonstrates a strong 10 commitment to the development of the skills and knowledge !
O 11 of maintenance personnel. In the 1993 SALP report, the j 12 NRC recognized the strength of PG&E's training in the 13 maintenance area noting: ,
O 14 " Plant Safety: Maintenance personnel were trained 15 and informed regarding overall plant safety system 16 availability and the significance of their 17 individually assigned work relative to its risk to 18 the plant."
O f 19 Maintenance Facilities 20 Q25 What are the facilities that support the maintenance ,
o, 21 programs at ocPP? ;
r t
22 A25 (Giffin, Crockett) PG&E has provided excellent O 23 facilities for the performance of maintenance activities.
24 State of the art training facilities, machine shops, l 25 calibration facilities, and warehouse storage areas are
'O 26 available to support the maintenance program and specific 27 maintenance tasks. These facilities include:
28
- Maintenance Shops Building O 34
3 1
- Warehouses 2
- Computer Center and Associated Equipment 3
3
- Machine Shops 4 o I&C Shop 5
- M&TE Lab 3
6 The Maintenance Shops Building is a $10 million j 7 facility, located adjacent to the plant protected area.
8 In this building are laboratories, shops, and classrooms 3
9 totaling 70,000 square feet of space devoted to the 10 training of plant personnel. Within the mechanical, g 11 electrical, and instrumentation shops and laboratories are 12 many of the same components and equipment installed and I 13 operating within the actual plant. These are used for the 1
14 troubleshooting and repair training described above. When 3
15 infrequent maintenance activities are planned, the 16 equipment in this facility is also used to rehearse the l 17 activity and refresh the skill of the craftsmen before the 3
18 activity is performed in the plant.
19 The main Warehouse is a $18 million facility with over l 3 20 100,000 square feet of storage area with a material 21 testing laboratory, a Quality Control inspection area, and 22 an environmentally controlled storage room. In addition l g 23 to the main Warehouse, there are several satellite 24 warehouses to store specialized equipment, e.g., the spare 25 generator storage building that houses one complete spare g 26 main generator, the spare rotor storage building that 27 contains three low pressure turbine rotors, Warehouse A 28 that is used to store corponents and materials that may be 6 35
l 0
1 radioactive, and Warehouse B that is used to store other. ,
C) 2 large equipment and material to be surplused. .
.3 The Computer Center and associated facilities are 4 located in the Administration Building at DCPP and occupy 0 5 25,000 square feet. The Center includes:
6
- A fully automated mainframe computer facility housing 7 an IBM 3090 400J series machine.
O 8
- A network operating center housing over 30 network 9 servers.
l 10
- A computer support and help desk facility. ,
f) 11
- An information systems prototype and testing facility. ,
i l 12
- The data communications hub internal to the plant and ,
4
! 13 the gateway to external PG&E facilities.
4 ,
lO 14
- Applications maintenance, development, and 15 administrative areas.
16
- A support staff of full time personnel O 17 The cost to staff and operate the facility is 18 approximately $12.5 million per year. This cost provides f 19 us with the capability to execute business transactions at O 20 the rate of 700-800 transactions per minute and to ;
21 maintain over 130 gigabytes of plant information available 22 for rapid on-line retrieval. Collectively, the staff and O 23 the facilities provide DCPP with state of the art 24 information technology services.
i 25 There are four machine shops at DCPP. The first is C) 26 the Cold Machine Shop Building located south of the Unit 2 27 Turbine Building. It is an $8 million facility consisting ,
28 of 35,700 square feet and is shared by the Electrical and O 36 ;
O 1 Mechanical Maintenance Sections. The second is O 2 strategically located between both units in the Turbine 3 Building. A third facility is located at the intake 4 structure and consists of 2760 square feet at a cost of Q 5 S720,000. These three " cold" machine shops are used to 6 work on equipment that is not radiologically contaminated.
7 Radiologically contaminated equipment is brought to a O 8 fourth facility, the " hot" machine shop, where special 9 controls are in place to clean the equipment, prevent the 10 spread of contamination, and protect the workers. All O 11 four machine shops have proper equipment, such as lathes 12 and drill presses, with appropriate environmental 13 conditions conducive to ensuring maintenance quality and O .14 work efficiency.
15 The I&C Shop is an $8 million facility with over 16 25,000 square feet of I&C repair shops (for repairs and O 17 instrument calibrations), telecommunication shops, a 18 medical facility, and office area. The I&C shop was 19 constructed to consolidate all I&C functions in one O 20 location to enhance communication and overall work 21 efficiency.
22 Part of the I&C Shop is the Measuring and Test O 23 Equipment ("M&TE") calibration lab. Here, high accuracy 24 calibration is performed on equipment that, in turn, is 25 used to calibrate plant equipment. Over 700 pieces of O 26 M&TE are calibrated monthly during non-outage periods, and 27 about 5,000 per month during outage periods.
O 37
! t V i 53 1 In addition to these facilities, DCPP has various I i
2 laydown areas, tool storage, repair and issue stations, O
3 and temporary facilities that are used during outages.
4 Together, all of these facilities ensure the effective and 5 erficient performance of maintenance activities.
O 6 Preventive and Corrective Maintenance 7 Q26 The next element is achieving a " proper mix" of O
8 preventive and corrective maintenance. How is this 9 element satisfied at DCPP?
i O
10 A26 (Ortore) Maintenance is generally defined as falling 11 into two categories, preventive and corrective.
12 Preventive maintenance tasks are periodic, planned, or O '
13 predictive actions taken to ensure that equipment 14 continues to maintain its design function. Preventive i
15 maintenance tasks are normally performed on a O :
16 predetermined scheduled basis prior to any equipment 17 failure or degradation. Annually we perform about 14,000 18 preventive maintenance tasks at DCPP. These activities 1 0
19 range from simple equipment inspections, replacement of 20 limited life items such as lubricants, filters, wear l 21 rings, bearings, and diaphragms, to major equipnent
- O 22 overhauls. Preventive maintenance tasks are selected so .
l 23 as to maintain equipment in a condition which will l
! 24 increase reliability and extend the life of equipment.
g 25 A Master Equipment List ("MEL") identifies and tracks
[ 26 the majority of plant equipment, both safety-related and l
l 38 O >
1 I
9 1 nonsafety-related. Starting with the MEL, the following 9 2 equipment is selected to be included in the preventive
)
3 maintenance program: (1) any installed equipment, either 4 NSSS or BOP equipment, needed for safe and reliable plant
,e 5 operation, (2) any equipment requiring preventive 6 maintenance based on PG&E commitments, (3) any equipment 7 whose malfunction can cause direct personnel injury, and
[] 8 (4) any equipment where the implementation of preventive 9 maintenance may cause a reduction in operating costs.
10 There are about 17,500 components included in the DCPP
[) 11 preventive maintenance program.
12 A program of predictive maintenance has also been 13 incorporated into our maintenance program at DCPP as a
[) 14 part of preventive maintenance. Predictive maintenance is 15 the continuous or periodic monitoring and diagnosis of 16 selected equipment parameters to provide early detection
[) 17 of equipment degradation prior to equipment failure. This 18 is accomplished by gathering and analyzing information, 19 predicting future degradation, and then taking action to
) 20 limit the degradation before partial or complete failure 21 occurs. For example, this could involve taking a 22 lubricating oil sample from a piece of rotating equipment
- ) 23 such as a pump or motor (which is a standard task),
24 analyzing the sample for metal wear products and other 25 particulates to determine the amount and type of wear
- 26 occurring, and trending these findings to determine the 27 degree of degradation that may have taken place. If 28 necessary, we will then recormend specific maintenance D 39
9 1 actions for equipment. Generally, DCPP utilizes a variety O 2 of nonintrusive equipment monitoring techniques including:
3
- Lubrication analysis (as just discussed) 4
- Vibration monitoring and diagnostics
- 5
- Air and motor operated valve diagnostic testing 6
- Acoustic analysis 7
- Infrared thermography C) 8
- Bearing temperature analysis 9
- Insulation resistance 10
- Non-destructive analysis C) 11
- Monitoring and trending of equipment data 12 The predictive maintenance program is described in 13 procedure, AP C-751, " Predictive Maintenance Program."
C) 14 Corrective maintenance consists of the repair and/or 15 restoration of equipment. As discussed above, this is a
, 16 normal and expected part of the maintenance program.
) 17 Corrective maintenance is performed when equipment is not 18 able to perform its intended function or is outside 19 operating limits. The maintenance action will return the C). 20 equipment to a specified performance level. For example, 21 if a valve packing is leaking and needs to be tightened, 22 that task would be listed as corrective maintenance. We
- 2) 23 perform on the average 7,000 corrective maintenance tasks 24 at DCPP each year.
O 25 How do we know that at DCPP we have achieved a " proper Q27 26 mix" of preventive and corrective maintenance?
- 2) 40
.-= . .. - -
i
'O i i
i 1- A27 (Giffin, Ortore) Within the nuclear industry, a high !
C) 2 ratio of preventive maintenance (including surveillances) l l
3 to total maintenance is indicative of an efficiently !
4 managed maintenance program which is effective in l l
0 5 maintaining the plant in a safe and reliable condition.
I 6 Obviously this is a very subjective assessment. It is in 7 our interest to minimize corrective maintenance, to the [
C) s extent practical, through preventive maintenance.
9 However, preventive maintenance cannot avoid all I
{
10 corrective maintenance, nor would such a goal be practical O 11 from an operating perspective. INPO formerly monitored 12 this parameter and had established 60 percent as the }
13 objective. !
O .14 At DCPP, we monitor preventive maintenance tasks as a -
15 percentage of all maintenance tasks and strive to maintain ;
1 16 the ratio above 60 percent. Over the past three years the C) 17 number of corrective maintenances performed have decreased ;
i 18 which can be attributed to the effectiveness of the !
I 19 preventive maintenance program. Based on-this performance O 20 and the experience gained since we began monitoring this 21 percentage, we believe that we are maintaining a proper !
22 mix. We believe also that the overall reliabilit'y of the 0 -23 plant to date, as measured by the plant availability and
! 24 plant capacity factors, is a good indicator of a proper }
Y l
! 25 mix. j 10 .
Y .
l !
41 j V !
9
) I l
i l
l 1 Maintenance Procedures 2 Q28 The next element of a maintenance program includes the 3 procedures and/or work instructions. Please describe how l 4 this works at DCPP.
1 <
l 1
[)
l 5 A28 (Vosburg) In order for maintenance activities to be )
6 performed in a consistently safe and efficient manner by !
7 the craftsmen, accurate procedures providing technical ]
8 guidance and direction have to be in place. We have ]
9 expended considerable effort at DCPP over the last several 10 years in the preparation of state of the art maintenance
, 11 procedures. There are currently about 6,000 written i l
l 12 procedures which are used to conduct specific maintenance i i 13 activities. Exhibit 10 to this testimony is a list of j 14 these maintenance and surveillance test procedures. These 15 procedures are periodically reviewed and updated, as 16 necessary. The procedures that the craftsmen use have
)
17 been prepared with input from the workers and provide the 18 necessary graphics and check lists. The relevant
(
19 procedures for specific tasks are incorporated, as
)
20 appropriate, into task-specific work packages described 21 below.
)
22 Plannina and Schedulinq 23 Q29 The next element concerns planning and scheduling of 24 maintenance tasks. How is this accomplished at DCPP7 L
42
O 1 A29 (Vosburg) With approximately 20,000 maintenance and i
o 2 surveillance work orders performed each year, it is 3 extremely important to plan and schedule the activities ;
4 correctly. This is one of the important functions of the O 5 Work Planning Section of the Maintenance Services 6 Department and is part of a detailed, comprehensive work 7 control process. The work control process provides the
~
9 8 integrated mechanism under which plant maintenance 9 activities and equipment problems (including both f i
10 preventive and corrective maintenance tasks) are o 11 identified, reviewed, prioritized, planned, scheduled, 12 performed, tested and closed out. The process ensures 13 that plant maintenance activities are planned and 0 14 performed in a safe, timely, efficient and controlled f
i 15 manner. The process is specifically designed to t
16 coordinate maintenance activities to minimize the time O 17 that safety-related equipment is out of service.
18 Q30 How are corrective maintenance activities identified !
O 19 for incorporation into the work control process?
i 20 A30 (Vosburg) A key objective in the overall maintenance l
j O 21 program is to ensure that actual or potential plant 22 equipment problems will be identified and documented in a 23 timely manner. One facet of the surveillance testing and o 24 preventive maintenance programs is directed towards early 25 identification of discrepant equipment conditions, with 26 the desired goal of repairing / replacing equipment before O 43
4 O' t 4
i i
i l 1 failure. Device calibrations also play a part in flagging ,
O 2 trends which aid in early detection and repair.
+
I 3 Maintenance tasks identified by these programs are 4 scheduled and tracked through the work control process. '
O 5 Any individual who discovers a problem in the plant is !
6 responsible for initiating on Action Request ("AR") or 7 reporting the problem to a supervisor who must then ,
O 8 initiate an AR in a timely manner. A significant factor ;
9 which contributes to the effectiveness of the problem f 10 reporting system at DCPP is that all personnel working at O 11 the plant nave access to the system, either directly or-At the plant site I 12 through their immediate superviscr.
13 alone, there are more than 1,800 computer terminals ,
O 14 available where ARs may be initiated and at least as many ,
15 plant personnel with direct access to the system. Since i 16 the AR system is used not only for documenting plant l t
O 17 equipment problems but also for documenting administrative ;
i 18 tasks, requesting design changes, requesting 19 interdepartmental support, etc., there is a high level of O 20 familiarity with the problem reporting process throughout j 21 the plant staff. ;
t O 22 031 once maintenance tasks are identified, what means 23 exist to track the scheduling, planning, and completion of i 24 those tasks?
O 25 A31 (Vosburg) The administrative controls for the work i 26 control process at DCPP are integrated into PIMS. Using -
O 44
I t
I l
1 this system, all data entered at any one computer terminal
() 2 during the process is immediately available at other i l 3 computer terminals throughout the plant. This attribute i I
4 makes the entire system highly responsive to plant
()- 5 maintenance needs. One advantage of PIMS is that the f 1
6 system itself assigns unique task numbers to new -
7 documents, thereby ensuring that none are inadvertently l
C) 8 misplaced. Once an AR is created, it cannot be ;
i 9 destroyed - even if it is taken to " rejected" status. 1 l
10 The ARs module within PIMS ensures that problems 0 11 receive appropriate levels of review and are tracked 3 12 through resolution. The AR is the key which {
13 electronically ties together all other modules which make i C) 14 up the maintenance process (equipment clearances, work i
15 orders, quality evaluations, post-maintenance test ,
16 requirements, etc). ;
C) 17 An overview of the DCPP work control process as it i 18 relates to a plant equipment problem is illustrated in 19 Exhibit 11. In the sequencing of a typical maintenance !
O 20 work request, the key elements of the process are- !
21 described below:
22
- Upon identification of a maintenance problem, the j
() 23 person who discovers it wil2 initiate an AR at a 24 convenient terminal. After filling in as much
! 25 information as is available at that time, the AR is
) 26 left in the " initiated" status and the individual's
! 27 supervisor is notified.
O 45 l
I.
O 1
- The supervisor reviews the AR for accuracy and assigns ;
2 a priority to the AR in accordance with the plant work 3 prioritization system. The supervisor is also 4 responsible to ensure that Operations is inmediately -
O 5 notified of any problem that is believed to have an ;
l 6 impact on plant safety or equipment operab'ility. This 7 notification is documented on the AR. The AR is then O This effectively #
8 taken to the " reviewed" status.
9 makes the AR available to the Work Planning Center.
10
- The Work Planning Center routes the AR to the !
O 11 appropriate planner, who is then able to begin 12 preparation of a work order, another electronic module 13 document which is linked to the AR. ,
O 14
- As part of the work order preparation, the planner 15 reviews the problem as documented on the AR and begins .
16 collecting information necessary to plan the O 17 maintenance activities. This may include (but is not -l 18 limited to): a walkdown of the jobsite; identification l 19 of relevant procedures, drawings and vendor manuals; 20 identification of special processes (such as welding);
21 interviews with foreman and craft; identification of 22 administrative limitations on the job secuence; and ,
O investigation of any salient information contained in 23 24 the PIMS Component or Maintenance History Databases. l 25
- The planner may use other specialized modules within ;
26 PIMS to electronically research and order spare parts, 27 flag quality-related aspects to the Quality Control .
28 group, request Radiation Work Permits for radiological l C) 46
l I
i f 1 work, request any clearance (equipment
)
2 isolation /tagout) which will be required in order to ,
3 commence work, and notify other disciplines of the 4 need for technical assistance.
)'
t 5
- When the planner has all necessary information ,
6 collected, he/she writes the work flow description 7 directly into the electronic work order. Quality J 8 related work orders are reviewed by the Quality 9 Control organization for the incorporation of needed 10 inspection hold points which are inserted into the ,
11 work order. The work package is then created which l 12 includes a copy of the work order, copies of 13 maintenance procedures referenced by the work order,
) 14 and other supporting documentation such as vendor 15 manuals and applicable plant drawings. The work 16 package is then assigned to the appropriate 17 maintenance foreman.
1B
- Another electronic link ties the work order to the 19 on-site scheduling software. The job is integrated by 3 20 the scheduling staff into the overall maintenance i
21 picture to ensure that it is performed in a time frame 22 commensurate with its priority. It must be merged
, 23 into the work stream along with other tasks to be 24 scheduled, such as corrective or preventive i 25 maintenance activities and surveillance _ testing.
26
- The maintenance task is performed by the appropriate 27 discipline, following which the work order is reviewed 2B by the foreman. The package is then returned to the
) 47
4 1 Work Planning Center for closure. The planner O 3 transcribes the results of the maintenance action into 3 the Component Maintenance History (another PIMS module 4 which provides a running history of maintenance to aid 9 5 in future planning efforts and/or research) and then 6 forwards the completed work order to Document Control 7 for archiving.
O 8 Q32 What considerations go into planning the scope of work 9 for specific maintenance casks?
O 10 A32 (Vosburg) Work pl'aning creates the integrated work 11 package based on the following considerations:
C) 12
- Consistency and conformance to standards and 13 requirements; 14
- Safety, reliability and effect on generation O 15 capability; 16
- The quality classification of the work, to assure that 17 the plant quality organizations are integrated into O 18 the work planning process; 19
- Integration with work scheduling to maximize the 20 efficiency of all work performed; and O 21
- The safety and radiation protection of plant workers.
22 Q33 Explain how maintenance tasks are scheduled.
G 23 A33 (Vosburg) The overall objective of the work scheduling 24 function is to coordinate and schedule plant work O 48
r s
i O
i l
1 activities in a manner that maximizes efficient O 2 utilization of plant resources and enhances the safe and 3 reliable operation of the-plant. To accomplish this we 4 developed a computerized scheduling program which ,
O 5 interfaces with PIMS. Periodic schedules identify when [
-6 equipment maintenance, testing or inspection is required. i 7 These computer schedules are routed to appropriate C) 8 sections for planning and performance of the work. The 9 scheduling program allows for integration of all work 10 activities into the schedule.
C> 11 The framework for the scheduling of plant maintenance i 12 and surveillance activities is built around a process ;
13 known as the Mode one Integrated Daily Schedule ("MOIDS"). ,
O 14 This process uses a 12-week rolling matrix which 15 identifies the required Technical Specification 16 surveillance tests for plant equipment and reserves blocks O 17 of time within the base schedule for the performance of 18 these tests. All other maintenance activities related to 19 the equipment are then identified and incorporated into O 20 these maintenance windows. Typical activities to be 21 included would be the routinely scheduled preventive 22 maintenance activities and any necessary corrective O 23 maintenance activities. Using this method, both the 24 number of times that equipment is removed from service and 25 the amount of time that the equipment is unavailable are O 26 minimized. Using the MOIDS concept, plant schedulers 27 along with senior licensed personnel from operations meet 28 twice a week to develop the schedule for the upcoming O 49
)
1 week. All activities to be performed on safety-related
) a equip:ent are reviewed by the group for plant safet) 3 impact. All preventive maintenance activities to be 4 included are reviewed to assute that work necessary to
) 5 enhance the reliability of the equipment is performed 6 within the normal maintenance window and is not 7 unnecessarily deferred.
) 8 An additional enhancement to the scheduling process 9 has been the assignment of a licensed Operations 10 representative to the scheduling group. This innovation
) 11 has proven instrumental in the improvement of 12 c immunication between the Planning, Scheduling, and 13 Operations organizations. The Operations representative
) 14 provides guidance to ensure that work is scheduled in a 15 way that optimizes use of Operations and Maintenance 16 manpower. He or she also provides input to the schedule 3 17 from a risk assessment standpoint and reviews scheduled 18 work to identify potential safety system interactions.
19 This aids in the early identification and resolution of
) 20 schedule conflicts and improves the overall efficiency and 21 safety of the final schedules.
) 22 Q34 How do you monitor work to control the backlog of 23 maintenance activities? I
) 24 A34 (Vosburg) The work control system at DCPP provides the 25 means for identifying, tracking, and controlling the 26 maintenance backlog. The backlog of corrective and I
) 50 l
O 1 preventive maintenance is tracked separately with specific 2 goals for each as established by management at the
)
3 beginning of each year, with the general intention of 4 reducing each as low as practicable. Since all 5 maintenance items at DCPP are individually tracked through 6 PIMS, status reports can be generated quite readily and 7 are available for management review.
O 8 Post-Maintenance Testing 9 Q35 Please detaribe the DCPP post-maintenance and post-10 modification testing program.
11 A35 (Crockett, Vosburg) Post-maintenance and 12 post-modification testing ("PMT") is a key component to 13 implementation of the plant maintenance program. The 14 primary objective of PMT is to ensure that all plant 15 equipment which Sas undergone maintenance or modification u
16 activities ha s be;" demonstrated to be fully functional or 17 operable prior o ~5ts :n to service.
18 Technical Specificatlens specifically require that
,O 19 safety-related equipment be properly tested and any l 20 associated problems resolved following maintenance or modification prior to declaring the equipment operable.
l,3 21 22 However, PMT at DCPP is not limited to Technical 23 Specification equipment. From an overall plant safety and 24 reliability standpoint, it is a good practice to perform 25 PMT following maintenance or modification of plant 26 equipment not required by the plant Technical 51 3
O 1 Specifications to ensure that the equipment will fulfill 2 its design function prior to returning the equipment to 3 service.
4 Generally, PMT requirements are specified in the work 5 order and in each case will be commensurate with the O
6 maintenance or modification work completed and the 7 importance of the equipment to plant safety and 8 reliability. At DCPP, PMT consists of two different types 9 of testing: Maintenance or Modification Verification ;
10 Tests ("MVT") and Operability Verification Tests ("OVT").
11 MVTs are typically those tests, inspections, or
.O 12 verifications which are performed by the implementing l 13 organization without actually operating the equipment.
14 For example, typical MVTs would include cleanliness
,.O 15 checks, electrical continuity and megger checks, and 16 instrument loop tests. OVTs are those tests specifically 17 designed to prove Technical Specification operability.
18 OVTs usually consist of performing the appropriate 19 Technical Specification surveillance test.
l O
l 20 Q36 How is PMT planned, controlled, and tracked?
f l
l (O '
- 21 A36 (Crockett, Vosburg) Identification and tracking of the 22 necessary testing begins with the work planner during the 1 23 development of the work package. Based on the scope of O
24 the work, the planner must decide prior to issuing the l 25 package to the field whether PMT is required and who is 52 i
'O.
O 1 responsible for identifying the required MVTs. Normally, O 2 to determine the MVTs, the planner routes the work package 3 to the appropriate review group to identify any necessary 4 in-process inspections, parts dedication tests, and static O 5 and dynamic tests specified in applicable industry codes, 6 standards and vendor ranuals. These tests and inspections 7 are normally included in the work order created on PIMS.
8 VTs are also identified and tracked by the PMT nodule O
9 within PIMS.
10 Documented completion of the required PMT is an 11 essential component of the equipment control process used O
12 by Operations when returning equiprent to service. Prior 13 to returning equipment to service, Operations will verify 14 that all the work is completed and that the PMT has been
()
15 successfully performed.
16 In total, the process for identification and tracking 17 PMT is tied directly to the work perforned and includes
()
18 rultiple cross-checks to ensure that adequate testing is 19 accomplished.
O 20 Procurement of Parts l
21 Q37 The next element requires that correct and qualified
'g 22 parts be readily available to support maintenance. How is 23 this achieved at DCPP?
g 24 A37 (Ortore) In order to assure high reliability of j 25 operation, equipment downtire must be mininized.
26 Accordingly, when a component is removed from service for ,
l 53 I 3
i l
1 maintenance, it is important that any parts necessary to j 2 restore or repair the component be readily available. The l 3 DCPp Warehouse and parts ordering system provide a l
l 4 comprehensive and reliable means for accomplishing this i
5 goal.
l 6 When a work order is produced, any parts that are l
l 7 required for the work are identified and that information 8 is electronically transmitted to the Warehouse. The parts 9 are staged for the craft foreman prior to the start of the i
10 work so they are verified and ready for installation 11 before removing a system from service. The computer j
12 system also orders any parts which have dropped below a 13 preset limit. The onsite warehouses together have more i
14 than 200,000 square feet of storage space and contain 15 about 59,600 different items (worth about $100 million) in 16 spare parts inventory.
f-h l 17 Q38 How do you assure that replacement components and 18 parts are procured in accordance with original design 19 requirements so as to maintain the design basis?
20 A38 (Ortore) To assure that replacement parts meet 21 original design requirements, a quality assurance program,
} 22 together with a careful engineering review of replacement 23 parts is used. Many American manufacturers and suppliers 24 of items to the nuclear industry have 10 CFR 50, Appendix
)
25 B, quality assurance programs that are audited by PG&E.
26 These manufacturers become our " qualified suppliers."
54
)
1 r
) !
I r
1 Safety-related components or parts purchased from j j 2 qualified suppliers must be procured: (1) as the identical l i
3 item, which is an exact duplicate of the original item 4 with identical quality assurance, quality control, y 5 technical and documentation requirements, and purchased 6 from the original supplier, or (2) a documented '
7 engineering evaluation (called a replacement part j- 8 evaluation) must be performed to ensure the replacement 9 part meets or exceeds the design requirements of the ;
10 original component. In total, PG&E devotes considerable l 11 resources -- both in engineering time and money -- to 12 assure that replacement components and parts are l
13 equivalent to or better than the original. [
) i 14 Q39 How do you assure the item you ordered is the one l l
L 15 received?
i 16 A39 (Ortore) Receipt inspection is performed on all items i
17 delivered to DCPP from the supplier. Receipt inspections :
l j 1B range from checking the model number and supplier ;
i :
19 documentation against the purchase order to a more l l 20 detailed inspection by the Quality Control organization l
j al for safety-related and graded quality items. Any item not r
22 fully meeting the criteria established in the purchase i i
[ 23 documents is placed on hold, thus-preventing its use in ,
24 the plant until the discrepancy is resolved.
L F
! i j- 55 f
1
J 1 To assist in receipt inspection, DCPP has a material 2 testing lab onsite that is used to test items. Equipment 3
3 in the lab includes:
4 e a portable X-ray fluorescent spectrometer for metallic
-s 5 material identification; a
6 e an optical emission spectrometer for metallic material 7 identification; 8 e an infrared spectrometer for elastomer identification
)
9
- Rockwell hardness tester for determining metallic 10 hardness tests; and 11
- Other equipment, including durometers, system g
12 scientific alloy analyzers, and normal lab facilities.
13 This lab is staffed with qualified quality control 14 inspectors.
g 15 Q40 How do you assure that the effects of age and
, 16 environment on stored items do not adversely affect the 17 quality of the item?
J.
18 A40 (Ortore) All items with shelf-life considerations have 19 the shelf-life determined based on criteria from the 20 manufacturer, industry standards [EPRI NP-6408, 21 " Guidelines for Establishing, Maintaining and Extending g
22 the Shelf Life Capability of Limited Life Items" 23 (NCIG-13)], or other sound engineering criteria. Material 24 that has an expired shelf life is segregated and not 25 released for issue without further documented evaluation.
26 When a new part or item is installed in the plant,
, 56 J
l b !
1 preventive maintenance is appropriately scheduled to take
[): _2 into account any impact of shelf life on the design life '!
i 3 of the part or item. ;
4 To reduce the effects of the environment on stored
[] 5 equipment, storage levels are established that correspond 6 to the specifications in ANSI N45.2.2, " Packaging, t 7 !
Shipping, Receiving, Storage and Handling of Items for i
.[) 8 Nuclear Power Plants." Four storage levels are ;
9 established, from level A through D, with the most j 10 stringent controls placed on Level A. For materials that ,
11 are highly sensitive to environmental conditions such as
-[] 1 12 electrical components, the material is stored in Level A !
13 storage. Level A storage provides temperature control 14 between 60 and 90 degrees and humidity control between 30 [
'[)
i 15 and 60 percent. The ventilation system is filtered to 16 provide an atmosphere free of dust. The main Warehouse at {
! i
[) 17 DCPP has an environmental room that meets or exceeds the J
.18 Level A storage requirements. The remainder of the l
19 Warehouse is storage level B, which is indoors with j
[) 20 temperature controls. (Storage level C is inside with no 21 temperature controls and level D is outdoors. For shelf l 22 life consideration, a minimum of storage level B is used.)
[] .23 To further reduce the effects of age and environment
)
24 on stored items, in-storage maintenance is performed.
25 In-storage maintenance is based on manufacturer
] 26 recommendations, past experience, and the significance of 27 the item to plant safety. In-storage maintenance can 57 h)
O 1 involve visual examinations, cleanings, rotation of g 2 shafts, and application of preservatives.
3 Control of Measurinc and Test Ecuipment 4 Q41 An ther aspect of maintenance is control of Measuring O
5 and Test Equipment (" METE"). How is this addressed at 6 DCPP?
O 7 A41 (Giffin) A program for the control and calibration of 8 M&TE has been implemented to ensure that equipment, such 9 as test meters and t rque wren hes, is functional and O
10 accurate to support maintenance activities. The DCPP M&TE 11 program is described in Administrative Procedure D-5, 12 " Control of Mechanical, Electrical and Instrument &
O 13 Controls Measurement, Test and Performance Monitoring 14 Equipment." The DCPP M&TE program has been enhanced 15 significantly over the past several years. We have g
16 invested in state of the art equipment and facilities for 17 the calibration of this special equipment. PIMS is used 18 to track and control M&TE. A highly trained and qualified g
19 staff is in place to manage and support the M&TE program.
20 These elements have improved our performance in the M&TE 21 area.
O 22 Root Cause Analysis Procram 23 g42 Another element of maintenance is a detailed root O
24 cause analysis program. How does PG&E address this 25 element especially in regard to equipment failure rates?
8 O
O 1 A42 (Giffin) One of the very important components of an
() 2 effective and comprehensive maintenance program is a 3 program which provides for a systematic analysis of 4 unplanned occurrences pertaining to maintenance. This
() 5 analysis is designed to identify the root cause of an 6 event. With a root cause, corrective actions can be 7 implemented so that occurrences of the same type can be
() 8 prevented. Root cause analysis at DCPP is controlled 9 by Procedure NPAP C-26 " Root Cause Analysis." This 10 procedure provides guidance in several analysis techniques
() 11 such as cause and effect analysis, event and causal 12 factors analysis, change analysis, barrier analysis, task 13 analysis, and human factors surveys. The procedure is C) 14 supplemented by training provided by both PG&E's Training 15 Department and outside industry experts. We have made 16 major investments in training our engineers and managers
() 17 in effective root cause analysis techniques as part of our 18 commitment to avoid repetitive problems.
() 19 Q43 How are root cause determinations made for DCPP?
20 A43 (Giffin) PG&E's root cause determinations have 21 different levels depending on the occurrence being C) 22 investigated. A root cause determination is required for 23 all quality problems. These'are classified as either g 24 Nonconformance Reports ("NCRs" - most significant) or 25 Quality Evaluation ("QEs" - less significant 59 C)
D 1 A Technical Review Group ("TRG") performs a root cause 3 2 analysis for each NCR. A TRG is a multi-disciplined group 3 established for each NCR and chaired by the DCPP 4 department with the most responsibility for the issue.
() 5 The TRG includes representatives from Quality Assurance, 6 Quality Control, and all involved departments. Agreement 7 on the root cause must be unanimous among the TRG members.
3 8 Effective corrective actions are developed by the TRG only 9 after the root cause of the event is determined and agreed 10 upon. The Plant Staff Review Committee ("PSRC") reviews 3 11 the root cause analysis for all NCRs. If the PSRC does 12 not agree, the TRG reconvenes to resolve the differences.
13 The PSRC also reviews the corrective actions recommended g 14 to ensure they are adequate to prevent recurrence.
15 A root cause determination is performed for all QEs by 16 the responsible department. Corrective actions are g 17 established after the root cause is determined. Root 18 cause determinations for all QEs are reviewed by the 19 Quality Control Department.
O 20 Maintenance History / Failure Trending 21 Q44 What does PG&E do in the area of component maintenance
- 7) 22 history and failure trending?
23 A44 (Crockett) A major component of root cause analysis 24 and component failure trending is the history of the 3
25 components. Currently there are approximately 187,000 26 individual components in the DCPP Component Data Base, 3 60
[) '
l :
1 each with its own maintenance history available in PIMS. l 1
[) 2 The information is readily available on PIMS when a i
3 problem occurs with a particular component. For example,
! 4 it is possible to determine the maintenance history of an [
i
[] 5 individual valve, or all valves with the same ;
6 model/ manufacturer, or all valves in a system. Component ,
l 7 experience is also available from an industry wide i
() 8 database, the Nuclear Plant Reliability Data System 9 ("NPRDS"), maintained by INPO.
t 10 PG&E uses component history data for two systematic
() 11 failure trending in several respects. First, all failures 12 of components necessary for accident mitigation or whose !
13 loss of function could initiate significant plant
() 14 transients are reported and included in NPRDS. The 15 failure report contains extensive detail, including
- 1 16 manufacturer, model number, serial number, supplier , j l
C) 17 application, failure mode, detection method, corrective
- 18 actions, symptoms of failure, systems affected, narrative !
19 of the failure, and cause.
l !
() 20 DCPP examines every component failure report in the 21 NPRDS database that results in a major transient. If the !
! 22 same component is used at DCPP in a similar application,
[) 23 PG&E investigates the failure to determine if our l l
24- application, maintenance practices, or surveillance ;
25 testing assure we will not have the same transient. When
- ) 26 corrective actions are required, the normal processes are 21 used. l
[) 61 1
I
l
)
i 1 Second, DCPP design and maintenance departments 2 routinely research failure rates in components as part of 3 the design or parts replacement process. Any DCPP 4 components failing at a rate higher rate than the industry 5 average are evaluated and an appropriate evaluation is
)
6 initiated.
7 Third, Nuclear Safety Engineering researches component 8 failure rates using the DCPP component history. Failure 9 rates are initially screened by computer using 10 continuously upgraded evaluation techniques. Components 11 that are identified in this process are then screened by 12 engineers. This methodical evaluation process will detect 13 failure trends occurring within systems by component type, 14 manufacturer /model, and by other groupings. This is an 15 approach designed to detect any aging failure mechanisms.
16 V. MAINTENANCE AND EOUIPEENT AGING MANAGEMENT AT'DCPP 17 Maintenance and Surveillance Programs and Activities 18 Q45 How does maintenance and surveillance at DCPP address 19 equipment aging? g 30 A45 (Giffin) As described above, equipment aging 21 management is inherent in maintenance and surveillance.
.l 22 PG&E has long recognized the benefits to be gained from ,
23 initiatives to identify and address the effects of 24 age-related degradation of SSCs. A commitment to aging ,
i 25 management has existed since construction at DCPP began.
62
) i i
3 l 1 As a result, PG&E has many nature maintenance and D 2 surveillance programs and practices in place at DCPP that 3 address equipment aging.
4 Some of the more significant programs and activities D 5 that assist in mitigating the effects of age-related 6 degradation include:
\
7
- The preventive maintenance program provides the D 8 necessary inspection, testing, and monitoring ,
9 activities and periodic equipment servicing and 10 refurbishment to maintain the reliability of the D 11 equipment.
12
- The predictive maintenance program seeks to forecast 13 the functional ability and necessary maintenance of D 14 SSCs.
15
- The corrective maintenance program addresses the 16 repair of plant SSCs. Corrective maintenance can l
[) 17 provide valuable input to the aging management program 18 regarding potential aging effects.
- 19
- Surveillance test programs, including ISI/IST, help to l
D 20 detect any degradation that might affect SSC 21 operability or reliability.
j 22
- Fatigue monitoring provides on-line fatigue cycle I) 23 monitoring of sensitive components.
24
- The EQ program defines qualified life and service 25 criteria for certain electrical equipment.
p 26
- A reactor vessel embrittlement management plan 27 outlines PG&E's long term strategy to assure the
) 63
l.
O P 1 integrity and operational life of DCPP's reactor 2 vessels. t 3
- A Motor Operated Valve ("MOV") testing and evaluation 4 program identifies and mitigates performance problems 5 with safety-related MOVs through design reviews and ,
6 testing activities.
7
- A steam generator Strategic Management Plan employs 8 inspection results, industry experience, and ,
9 engineering analyses to predict steam generator 10 degradation trends and recommend aging mitigation 11 strategies.
12
- Structural monitoring addresses the condition and 13 integrity of important plant structures.
14
- An erosion / corrosion program addresses the aging 15 effects of erosion / corrosion mechanisms in piping 16 systems.
17 Each of these programs or activities produce specific 18 results and corrective actions to maintain and/or restore 19 equipment to its required performance level. These 20 measures address aging effects regardless of whether the O
21 aging occurred prior to or during plant operation. In 42- addition, results from these programs are used to develop 23 new or enhanced strategies to improve maintenance and ;
24 monitoring activities, highlight strategic issues, and i l 25 increase understanding of aging mechanisms to insure that l !
26 necessary actions are taken for mitigation. ,
t
! 64 !
[) '
i
{
I I
f l 1 Plant /Ecuipment Improvements to Date
): 2 Q46 Since plant startup, has PGEE made or planned plant 3 equipment modifications or replacements at DCPP which 4 improve reliability or reduce the likelihood of future l
). 5 age-related degradation?
! 6 A46 (Giffin) Yes. A number of major plant modifications
) 7 designed to improve reliability or upgrade safety-related 8 equipment have been made during the eight years since DCPP
! 9 began operation. Several others are planned in the near 3 10 future. Some of the more significant modifications 11 include:
12 e Copper Removal - This project involved replacement of 3 13 all feedwater heaters and retubing of all moisture 14 separator reheaters. These changes resulted in the 15 removal of essentially all copper from the secondary
[F 16 side of the plant to increase the life of the steam
~ 17 generators.
18 e Condensate Polisher Addition - To increase the life of 3 19 the steam generators, a Condensate Polisher System was 20 added to process secondary water by ion-exchange.
l 21 e Ammonium Hydroxide Storage - To regenerate condensate
[1 22 polisher resin, a 6,000 gallon bulk storage tank for.
23 ammonium hydroxide'was added.
l
- 24 e SG Blowdown Rate Increase - To improve secondary water chemistry and thus increase the expected life of the
~
25
- 126 steam generators, the blowdown rate has been
[ '27 increased.
o 65 l-l
l
)
l 1
- Control Room Upgrade - A detailed control room design i
I) 2 review ("DCRDR") was performed in accordance with the 3 requirements specified in Supplement 1 to NUREG-0737.
l 4 Weaknesses in the man-machine interface between j) 5 control room operators and equipment were identified 6 in the DCRDR. Following review and approval by the ,
1 7 NRC, control room equipment upgrades have been j) 8 implemented.
i 9 e High Density Spent Fuel Pool Racks - The original fuel'
! 10 racks in each unit's spent fuel pool were replaced 11 with high density racks, increasing the capacity in -
3 12 each spent fuel pool from 270 to 1324 fuel assemblies. ,
13
- Improved Fuel Design - The reactor fuel used in each 14 unit is being replaced with an improved VANTAGE 5 3 ,
15 Westinghouse design.
16 e Baron Injection Tank (" BIT") Removal - In response to j 17 industry experience and NRC recommendations, the BITS ;
18 in both Diablo Canyon units were removed from service 19 to reduce the potential for boric acid crystallization ;
l j 20 in the Emergency Core Cooling System ("ECCS") piping 21 and valves which could potentially have degraded ,
22 safety-related equipment operability.
3 23
- Reduced Boric Acid Concentration - The boron i 24 concentration in the Boric Acid System has been i
25 reduced from 12 to 4 weight percent to reduce the 3 36 potential for boric acid crystallization in safety-27 related components.
t
J 1 e Digital Feedwater Control System - A digital feedwater 2 control system was installed to improve feedwater g
~
control performance and reliability. The enhanced 4 feedwater control features provided by this system
.. 5 reduces the likelihood of steam generator level-J 6 related reactor trips.
7
- Computer Replacement - The original plant process 8 computer was replaced with one having improved 3
9 man-machine interface, greater capacity, faster 10 response time, improved print and report capability, 11 improved retrieval of historical data, and complete J,
12 redundancy to prevent loss of information due to 13 single failure.
14 e ATWS Mitigation System Actuation Circuitry ("AMSAC") -
i 15 The AMSAC System was installed in both Diablo Canyon 16 units to ensure reactor protection during an 17 anticipated transient without scram ("ATWS") event l 18 that results in the loss of the secondary side heat 19 sink. AMSAC is designed to trip the main turbine, 20 initiate auxiliary feedwater flow, and isolate steam j 21 generator blowdown and sample lines during an ATWS l
22 with a low steam generator level condition.
i
% 23 e Chlorination System Modifications - Modifications to j
24 the Chlorination System include (1) the use of liquid 25 hypochlorite to control microbiofouling instead of 26 gaseous chlorine, (2) implementation of continuous g
27 chlorination of the auxiliary saltwater system to 28 control macrobiofouling (invertebrate marine life),
_, 67
.)
D !
1 1 and (3) possible use of intermittent injection of a 2 chlorine / bromine mixture to prevent macrofouling in i g
3 the Circulating Water System.
4
- Fatigue Monitoring - PG&E has installed an on-line 5 fatigue monitoring system at Diablo Canyon that will 6 continuously analyze plant operational data to track 7 fatigue usage of critical reactor coolant system B components.
9
- Additional Diesel Generator - Addition of a sixth 10 diesel generator will provide each unit with three 11 dedicated diesel generators. This will enhance 12 reliability of the onsite power distribution system by 13 eliminating dependence on a swing diesel generator and 14 the associated procedural complexities. Installation, 15 testing, and tie-in of the sixth diesel was completed l
16 in April 1993.
l 17
- Plant Process Protection System Upgrade - This project 18 will upgrade the Process Protection System by l
l 19 replacing the existing HAGEN 7100 equipment with a l 20 Westinghouse Eagle 21 system. New steamline break h.
l 21 logic and steam generator low-low level trip time f 22 delay options will be included in the upgrade. These l
23 changes will improve the reliability and availability 24 of the Process Protection System. The digital 25 microprocessor-based system with computer-enhanced 26 testing will also minimize the likelihood of personnel 27 error during surveillance testing. System 68
[) j 1 installation is scheduled for the spring of 1994 for l l
2 Unit 1 and the fall of 1994 for Unit 2.
O l 3
- RTD Bypass Elimination - This project will replace the l
4 resistance temperature detector ("RTD") bypass loop 5 piping with fast response RTDs installed in the hot ,
O
~
6 and cold legs of the Reactor Coolant System. Plant i 7 downtime and radiation exposures will be reduced and l 8 numerous snubbers can be eliminated. Installation is D 9 scheduled for the spring of 1994 for Unit 1 and the 10 fall of 1994 for Unit 2. ;
11 e Radiation Monitoring System ("RMS") Upgrade - This D 12 project will upgrade the present RMS to improve ,
13 performance and reliability, and reduce required 14 maintenance. Most of the work is scheduled to be O 15 completed by 1995.
16 Many of these major modifications, made subsequent to C 17 the issuance of the operating licenses for DCPP, 18 constitute upgrades in plant equipment or operating l 19 systems which will help minimize the effects of -
D 20 age-related degradation on the plant over the remainder of ,
21 the 40-year operating life contemplated by the license i
- 22 amendment request. *
[) l t
l 4 23 Aging Management Procram Directive '
l 24 Q47 Has PGEE formalized its aging management efforts?
t i
69 i a
i I
O 1 A47 (Giffin, Crockett) Yes. The overall aging management t O 2 program for DCPP was recently established pursuant to 3 Program Directive TS1, " Plant Aging Management." This 4 program encompasses and augments the many existing O 5 programs (listed in A45 above) that address age-related 6 degradation over a 40-year operating life. j 7 The DCPP aging management program collects data and O 8 utilizes input from many resources to insure that PG&E !
9 stays on the " cutting edge" of new developments and 10 technologies for detecting and responding to age-related :
O 11 degradation. New research findings, industry operating 12 experience, and information from the NRC, EPRI, 13 Westinghouse, i.nd other sources are considered for O 14 inclusion in appropriate programs. A process is also 15 defined to integrate PG&E and industry experience into the 16 existing DCPP maintenance and surveillance programs.
O 17 A major element of this enhanced aging management 18 program involves increased emphasis on aggressive 19 participation in activities that capture new technological O 20 advances and tools developed to improve nuclear plant life 21 cycle management / aging management capabilities. As a 22 result, during the first half of 1993 PG&E pursued this O 23 objective through the following initiatives:
24
- In January 1993, PG&E joined the EPRI Life Cycle 25 Management (" LCM") subcommittee. EPRI is a recognized O 26 leader in research and development activities 27 involving all aspects of the power industry. The LCM 28 subcommittee has been established as a forum to O 70
)
1 develop and/or assemble state of the art technical j 2 information , tools, and methodologies to assist 3 nuclear power plant licensees in optimizing plant 4 performance by identifying, evaluating, and mitigating
) 5 the effects of plant aging to support continued safe, 6 reliable long term nuclear plant operation.
7
- Acting on a PG&E initiative, the Region V Engineering
) 8 Manager's Forum (" EMF"), consisting of representatives 9 from Region V licensees, recently authorized the 10 establishment of an aging management subcommittee.
11 One of the primary objectives of this subcommittee is
)
12 to share, collaborate, and disseminate information .
13 regarding issues and developments involving aging
) 14 management.
15
- In February 1993, utility owners who are members of l 16 the Westinghouse Owners Group authorized an $8 17 million, five year project to manage nuclear plant
)
18 life cycle management / license renewal issues facing 19 the industry. One of the primary objectives of this program is to develop technical reports (based upon
) 20 21 ongoing research, technical evaluations and plant -
22 experience) and other tools that guide member l 1
23 utilities in their efforts at managing aging in
) '
24 important plant components. PGLE has taken a i 25 leadership role in this effort by serving on a core l
26 project management team.
I 71 l
O 1 The above initiatives enhance the existing, ongoing 2 PG&E efforts to identify, evaluate, and implement evolving 3
3 aging technologies, tools, and methodologies.
4 Q48 Are there other programs at DCPP to address O
5 aging / maintenance of specific equipment?
6 A48 (Giffin) Yes. As we stated at the outset of this O
7 testimony in A7, for certain critical components subject 8 to complex aging mechanisms, or for certain components I 9 with a limited life (e.g., some EQ equipment), special O
10 maintenance programs have been implemented at DCPP. Some 11 of these nature programs are also listed in our response 12 A45 listing existing programs that effectively address O
13 equipment aging. We will now describe a few in greater 14 detail.
O 15 Steam Generator Tube Degradation 16 Q49 How does PG&E address steam generator tube 17 degradation?
O 18 A49 (Giffin) At DCPP, steam generator tube degradation is ,
19 monitored and managed by careful chemistry control during g
20 operation and by an extensive cleaning and inspection 21 program during each refueling outage.
22 During plant operation, chemistry conditions in the O
23 condensate, feedwater, steam generators, makeup water, and 24 other systems are monitored and controlled in accordance l 72 b
i O i i
1 with the chemistry guidelines published by EPRI (PWR O 2 secondary Water Chemistry guidelines - Revision 3, May 3 1993). These guidelines reflect state-of-the-art industry
- 4 thinking with respect to measures for reducing corrosion i .
.O 5 and for enhancing steam generator reliability.
6 During refueling outages the following techniques are 7 used for tube cleaning and inspection:
O 8 . " sludge lancing" to remove sludge accumulated during 9 operation; 10 * " Pressure pulse cleaning" to aid in sludge removal in 0 11 the upper area of the tube bundle; and ;
12
- Eddy current inspection to assess the condition of (
13 tubing (PG&E is actually inspecting by this technique O 14 a much larger scope than required by applicable 15 guidelines to proactively assure detection of trends).
16 PG&E is also an active member of the EPRI Steam ,
i O 17 Generator Reliability Project. Participation in this and ;
18 other industry groups helps keep PG&E aware of new l 19 developments related to steam generators. As new i O 20 understanding of steam generator tube degradation emerges r l 21 and new processes and recommendations for mitigation of 22 age-related degradation are developed, PG&E has actively [
l ?
O 23 pursued prompt implementation of such initiatives. !
I 24 Examples of steam generator maint 2 nance initiatives at ,
25 DCPP are heat treatment of row I and 2 tube U-bends and ,
L() 26 "shotpoening" of-all hot leg tube ends.
I 27 PG&E has also prepared a comprehensive Steam Generator ,
28 Strategic Management Plan which compares the status of
.O 73 i t
l :
t C) i i
1 DCPP steam generator tube degradation with that of other O 2 steam generators throughout the industry and predicts ,
3 performance over time. Based on the current number of .
4 so-called " defective" tubes, DCPP compares well with the O 5 industry. Tubes are defined as " defective" with 6 degradation equal to or greater than 40 percent through 7 wall. Defective tubes are required to be repaired by 0 8 either plugging or sleeving. Unit 1, with 43 tubes 9 (0.3 percent of total tubes) plugged, is in the 20th 10 percentile of the industry (i.e., 80 percent of plants i O 11 have higher steam generator tube degradation rates); Unit i
12 2, with 76 tubes (0.6 percent) plugged, falls just above l 13 the 50th percentile. If this level of performance ;
r O 14 continues, we predict that less than 10 percent of the 15 tubes in each unit would require repair before the end of }
16 life based on license recapture (i.e., 2021 for Unit 1 and !
O 17 2025 for Unit 2). It is also estimated that most of these 18 tubes could be repaired by sleeving so that the overall i 19 effect on steam generation and plant performance would be >
l O. 20 minimal. :
i 21 Feactor Pressure vessel Aoino Manacement ,
() 22 Q50 How does PGEE address reactor pressure vessel l 23 embrittlement? !
l I
L() 24 A50 (Giffin) PG&E has and will continue to comply with NRC
( 25 regulations governing Reactor Pressure Vessel ("RPV")
26 surveillance and integrity codified at 10 CFR 50.61 and i O 74 1
O 1 Part 50, Appendix H. The DCPP Reactor Vessel Radiation O 2 Surveillance Program is designed to monitor changes in 3 material / mechanical properties of the DCPP RPVs over the 4 operating life of the plant to assure safe continued O 5 operation of the vessel. DCPP's surveillance program was 6 designed to meet the requirements of ASTM E-185, " Standard 7 Practice For Conducting Surveillance Tests For Light Water C) 8 Cooled Nuclear Power Reactor Vessels."
9 Reactor vessel integrity is also evaluated through 10 periodic 10-year in-service inspections which are required C) 11 by ASME Section XI. The in-service inspections require 12 volumetric inspection of all pressure retaining welds, and 13 all full penetration nozzle welds; volumetric and surface C) 14 inspection of all pressure retaining dissimilar metal 15 welds and visual inspection of 25 percent of the partial 16 penetration nozzle welds external surface. The first C) 17 in-service inspection was successfully conducted on both 18 units during the most recent outages, with no adverse I
i 19 findings.
I C 20 Alternative fuel management strategies are also l
21 continually being evaluated to determine the most 22 effective way to reduce reactor vessel exposure to neutron
) 23 irradiation. To date these strategies have resulted in two 24 fuel arrangement changes. DCPP implemented Phase 1 of the 25 fuel management strategy in the first reload of each unit O 26 by incorporating a " low leakage" fuel loading pattern 27 which yielded approximately a 30 percent flux reduction 28 compared to standard loading patterns. A low leakage h) 75
i O
1 loading pattern is achieved by loading relatively low O 2 enrichment fuel assemblies which have been previously 3 burned in the preceding cycle in the core periphery to 4 reduce the number of neutrons produced near the vessel l
g 5 wall. Phase 2 of the fuel management strategy was 6 incorporated into the cycle 3 design which targets flux 7 reduction measures in specific areas (i.e., the welds O 8 adjacent to the corner baffle locations).
9 PG&E has also recently implemented a comprehensive, 10 state of the art Embrittlement Management Program which is O 11 designed to manage all of the issues relating to reactor ,
12 vessel embrittlement, g 13 QS1 Has PGEE's regulatory compliance in this area been 14 documented?
i 25 Asl (ciffin, Crockett) Yes. Compliance with all NRC O ,
16 regulations governing vessel integrity has been documented 17 in PG&E's response to Generic Letter 92-01 (PG&E Letter 18 No. DCL-92-150, dated June 30, 1992). In addition, PG&E 9 ;
19 has calculated the Reference Temperature for Pressurized 20 Thermal Shock (" PTS") for each veld metal and base metal 21 in the DCPP beltline region for neutron fluences O
22 corresponding to 40 operating years. Since all materials 23 meet the screening criterion in 10 CFR 50.61, neither 9 24 additional flux reduction nor plant specific PTS analyses ;
i
, 25 are required to comply with the PTS rule. Details of the ;
76 ,
10 t
) )
1 PTS evaluation were submitted to the NRC in 1992 (PG&E 2 Letter No. DCL-92-056, dated March 6, 1992).
3 Erosion / Corrosion 4 QS2 How does PGEE address piping erosion / corrosion 5 concerns?
6 A52 (Crockett) The term Erosion / Corrosion ("E/C") in the 7 context of nuclear power plants refers to the process of 8 wall thinning in susceptible piping or other pressure 9 boundary components caused by the flow of water or wet 10 steam. E/C is a normal part of the plant aging process; 1 11 consequently, its management is an integral part of normal l 12 maintenance at DCPP. All DCPP secondary-side piping h 13 systems operating above 212*F, containing single or 14 two-phase water (but not dry steam), are susceptible to I
, 15 the effects of E/C and are therefore included in a DCPP D-16 E/C Monitoring Program. Notably, safety-related piping 17 which is fabricated of stainless steel is not susceptible 18 to E/C degradation.
19 The rudiments of the DCPP E/C Monitoring Program were 20 established in 1983, well before commercial operation was 21 achieved. Later, in early 1987, PG&E formed a 22 multi-discipline E/C Task Force. The Task Force is l
23 charged with developing and maintaining a broad-based 24 perspective on E/C monitoring at DCPP. Rather than 25 relying upon any single methodology for the prediction of 26 E/C-susceptible locations, the Task Force has maintained a 77 D-
l 1 defense-in-depth approach to address E/C at DCPP. PG&E's 2 defense-in-depth philosophy includes monitoring E/C 3 experience through a number or industry and NRC sources. ,
4 The DCPP E/C inspection scope incorporates ,
5 approximately 160 components per unit in the most recent ;
l 6 outage inspections. The most severe examples of E/C 7 degradation found at DCPP are highly localized, and i
8 typically occur downstream of control valves and at
- 9 reducing orifices in areas of extremely high fluid f 4
j 10 turbulence. Piping at all such locations has been or is 11 in the process of being replaced with E/C-resistant 12 material such as stainless steel or chrome-moly steel.
l L
13 These replacements are a permanent solution to the
)
14 degradation problem, as proven by continued monitoring. :
15 The next-fastest wearing piping occurs in the high l 16 pressure extraction steam piping exiting the high pressure 17 turbine. The majority of this piping has already been h
18 replaced at DCPP with stainless or chrome-moly steel as a-19 permanent solution, as proven by continued monitoring.
)
20 The replacement of the remainder of this piping at DCPP is j 21 scheduled to be completed within the next several fuel 22 cycles.
23 The DCPP E/C Monitoring Program was identified as an L
1 24 engineering strength in the 1992 NRC SALP Report. We 25 believe the E/C program is an integral part of the PG&E 26 corporate commitment to the safe, reliable operation of i
27 DCPP.
i
- 78 1
l
) l l
l 1 EO EauiDment
) 2 QS3 How does PGEE maintain the continuing qualification of 3 electrical equipment within the scope of the EQ program?
4 l
4 A53 (Ortore) As discussed earlier, each item of EQ 3
5 equipment has a calculated qualified life corresponding to 6 the time the equipment can operate under its normal, ;
7 installed operating conditions and still be considered 3
8 qualified for the postulated post-accident harsh
- 9 environment. For equipment whose qualified life is less
) 10 than 40 years, replacements are scheduled and made before l '. the end of the qualified life (in accordance with 12 procedure AP D-756, " Maintenance and Surveillance of
) 13 Electrical and I&C Environmentally Qualified (EQ) :
14 Equipment").
15 PG&E considers the anticipated effects of normal l 16 operating environmental conditions, including temperature, 3
17 corrosion, dynamic interactions and radiation exposure, in 18 the overall design of equipment utilized for i
~
19 EQ equipment in
) safety-related service at DCPP.
20 particular is specified, designed, and fabricated for the 21 anticipated service conditions. Required periodic 22
) inspections, tests, and surveillances provide assurance of 23 continued equipment performance within these operating 24 environments. To the extent EQ requirements dictate [
25 particular maintenance requirements (e.g., to preserve 26 qualified life under normal conditions, or to simply l i
27 preserve qualification such as by maintaining equipment
) 79
i h
i
) !
i t
i i seals), these requirements are incorporated into the j 2 maintenance practices for the specific equipment. This 3
i 3 process is also established by AP D-756. j 4 Equipment qualified lives are specifically calculated l 5 based on the anticipated service environment.
3 6 Environmental conditions which might have adverse effects [
i 7 on of these components are monitored by normal l i
8 surveillance test procedures ("STPs") and periodic
- 3 9 functional checks during routine operations and preventive !
10 maintenance. For example, DCPP procedures STP I-1A, 11 " Routine Shift Checks Required by Licenses" and STP I-1B, l 3
12 " Routine Daily Checks Required by Licenses," specifically ;
13 provide for the monitoring of outside and inside- l 14 containment temperatures, respectively. These temperature 3
15 measurements assure that Technical Specifications are met :
16 during plant operation for containment average air 17 temperature monitoring (Technical Specification 3/4.6.1.5) 3 18 and for important outside containment area temperature j 19 monitoring (Technical Specification 3/4.7.11). If the i 3, 20 temperature recorded during these surveillances exceeds 21 the limitations in the Technical Specifications, i 22 corrective measures are taken to restore the temperature j
) 23 within limits and an evaluation is performed by 24 engineering to identify any necessary reduction to the {
25 qualified life of EQ equipment.
t 26 PG&E also has established a program for temperature
) j 27 monitoring in connection with maintenance of EQ equipment.
28 Two maintenance procedures provide guidance: ;
80
). ;
l 1
)
1
- MP E-57.4, " Environmental Qualification Maintenance
) 2 and Survey of Containment Penetrations, Cable and 3 Splices": and 4
- MP E-57.8A, " Temperature Monitoring".
5 Procedure MP E-57.4 provides the necessary guidance for
)
6 the visual inspection and insulation resistance testing of [
7 cable penetratjens and splices in accordance with EQ :
8 requirements. MP E-57.8A provides guidance for the
) ,
9 methods of specific device temperature monitoring to l 10 obtain qualitative temperature information. In accordance ,
11 with these procedures, temperature indicating stickers are
) t
~
12 placed on various EQ devices to identify any equipment ,
13 that may be exposed to temperature extremes higher than 14 previously considered for qualified life purposes. These 15 stickers identify momentary peaks and are sometimes l 16 augmented by continuous temperature recording devices. .
17 This program allows engineering to again consider whether 18 local conditions cause any impact on the qualified life of j 19 equipment.
)
20 structures f l
l 21 QS4 How does PG&E maintain plant structures?
)
22 AS4 (Giffin) Maintenance of structures at DCPP has been i i i
- 23 based upon experience and proven practices that PG&E has f i
j 24 employed over decades of operation at its generating 25 facilities. Aging of passive, long-lived structural {
l 26 concrete and steel in particular is caused by processes !
I i
81 l
t
1 O )
l 1 that are well understood and readily detected. Conditions O 2 such as spalling or cracking of concrete, corrosive or 3 caustic attacks from leaks, spills or exposure to the 4 environment, mechanical damage, and rust are routinely g 5 identified and reported by plant personnel (e.g., ,
6 operators, firewatches, and security personnel) as they 7 move about the plant. They are investigated and l
O s appropriate action taken. In addition, as noted 9 previously, at DCPP system engineers conduct periodic .
I 10 walkdowns of their systems. During their inspections, ;
9 11 they can observe signs of structural degradation.
12 For safety-related structures, functional surveillance l
13 requirements are specified in the Technical !
l lO 14 Specifications. This periodic surveillance testing 15 verifies operability of these structures. For example, 16 functional integrity of the DCPP containment structures l 17 must be routinely tested and documented by local and !
9 18 integrated leak rate surveillance test procedures. .
19 Containment coatings are also inspected under a special !
O 20 program.
l 21 VI. ISSUES RAISED BY MOTHERS FOR PEACE
! i
!O i 22 Q55 Are you familiar with the Supplement to Petition to 23 Intervene filed by the MFP on October 26, 1992 alleging i
9 24 certain problems with PG&E's maintenance and surveillance 25 program for DCPP as specified in the bases for proposed i 26 Contentions I and IV? ;
- o a2 1
O (All) Yes.
1 A55 O
2 Q56 Are you also familiar with MFPs Supplemental Response 3 to First and Second Sets of Interrogatories and Request O 4 for Production of Documents Filed by PG&E which the MFP e
5 filed on June 21, 1993, and in particular Attachment C, ,
6 which was MFPs unfiled " Reply to NRC and PGEE Responses to i O 7 Petitioner's Supplement to Petition to Intervene dated g 8 December 10, 1992," relating to Contentions I and IV 9 regarding alleged deficiencies in PGEE's maintenance and O lo surveillance programs? ;
l 11 A56 (All) Yes.
O 12 Q57 In the original contention bases and the unfiled reply t
13 of December 10, 1992, MFP has identified a number of o 14 examples which they allege support their contention that 15 the DCPP maintenance and surveillance programs are not 16 sufficiently effective and comprehensive. Could you O 17 comment on these assertions?
18 A57 (All) MFP in these documents has identified a number O 19 of specific events at DCPP which they allege demonstrate ,
20 that DCPP's maintenance and surveillance programs are not ,
21 sufficiently effective and comprehensive. Many of these O 22 events have been reported and documented in our extensive 23 process of identifying problems, thoroughly investigating 24 them, taking corrective action, and improving our )
O 83 i w.
D- !
1 operations to prevent recurrence. Specific problems which 2 have been identified and discussed by MFP in support of 3 their contention which will be discussed in greater detail 4 later in this testimony, involve containment fan cooler i
5 unit backdraft dampers, the positive displacement pumps !
D 6 operating procedures, steam generator feedwater nozzle 7 indications, reactor cavity sump level indication, the l 8 motor operated valve program, debris found in the D
9 containment building, diesel generator turbo charger 10 bellows bolting, diesel generator fuel oil piping 11 corrosion, chemical volume and control system leakage, D
12 measuring and test equipment control deficiencies, f 13 emergency diesel generator surveillance test issues, fuel
, 14 handling building ventilation leakage, residual heat D i 15 removal ("RHR") recirculation sump screen deficiencies and 16 turbine governor and stop valve malfunction.
) 17 Q58 Can you characterize these examples in a general sense ,
18 and put in perspective their nature, significance and 19 relationship to the overall maintenance and surveillance >
20 program?
i I 21 A58 (All) The alleged problems are in some cases examples J
22 where something has fallen short of our expectations.
23 However, these identified problems have been relatively ,
24 minor, had no safety significance, have been thoroughly
)
25 investigated and pursued, and have resulted in 26 improvements to our maintenance and operations practices.
84 J
-s- - - - - -_ -_
I O !
t 1 In most cases, the problems are examples of how effective g 2 and thorough PG&E's maintenance program has been in f 3 addressing isolated errors or omissions and aggressively 4 following up to assure that we are continuously improving {
O 5 ur performance. l 6 It is also important to note at the outset that the !
7 number of problems identified by MFP is very small when O 8 c mpared to the very large number of tasks and individual 9 actions which comprise our overall maintenance and ;
i !
10 surveillance programs. Over the past three years, PG&E ,
O 11 estimates that there have been more than one million l 12 individual tasks and activities conducted during an -!
13 operating cycle per unit in our maintenance and j i
14 surveillance program. The relatively low rate of problem
- z l) 15 occurrence (e.g., reportable events, missed surveillances, l l
16 nonconformances) indicates a well-functioning program, !
17 with an active and aggressive problem identification and O l 18 resolution effort. Nonetheless, it is important that even f 19 a relatively small number of problems be properly ,
9 20 identified, documented, resolved, and used to improve the 21 overall effectiveness of the DCPP maintenance and l l
22 surveillance program. These identified problems have been ;
i i
- g 23 extensively reviewed, in accordance with DCPP's problem i
24 resolution program described earlier in this testimony.
l f 25 Root causes have been determined and measures taken to ,
!g 26 prevent recurrence. i l
i !
85
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i $
1 Q59 Can you draw any additional conclusions about the j 2 specific events discussed by HFP (in the documents 3 referenced in Q56) relative to any pervasive patterns or 4 recurring events?
O 5 A59 (All) As we review these specific events, as well as 6 the events characterized in the problem report documents 7 referenced by the MFP, we see no pervasive or systematic 3
8 problems that would indicate a programmatic weakness in 9 our maintenance and surveillance programs. Rather, we see g 10 random occurrences or omissions that are typical of human ,
11 shortcomings when engaged in a complex array of human 1 12 activities. In some cases, a procedure was not j 13 sufficiently explicit. Other occurrences involved .
14 individual inattention to detail; while still others i 15 involved the exercise of judgment by an individual that j 16 some activity is acceptable, when a later, more thorough, 17 review determines it to be unacceptable. We conclude that 18 these are random problems of the type that one might i
- j. 19 normally anticipate to occur in the implementation of a 20 nuclear plant maintenance and surveillance program. In 21 addition, a defense-in-depth philosophy was incorporated j 22 into the plant design, the plant operating program, and ;
23 the maintenance and surveillance programs at DCPP. This l
l 24 includes redundant trains of safety equipment, cperator 25 training to recognize and respond to problems, and j 3
26 surveillance and maintenance to identify and correct 27 problems and potential problems. This defense-in -depth 80 Q i l
l
O 1 assures that individual random equipment failures and 2 personnel errors will have no effect on safe operation of C
3 the plant.
4 On the other hand, as we have previously stated, it is 5 extremely important that we learn from these problems, D 6 investigate them and understand them fully, and 7 continuously improve our maintenance and surveillance 8 programs based on what we have learned. Each identified O
9 problem has its associated corrective action, which often 10 includes improvements to procedures, training, work 11 implementation, or work control. During this evaluation O
12 process, PG&E critically assesses how it conducts its 13 activities. Thus, it should not be surprising that we 14 identify mistakes and errors and use this information to O
15 formulate new and/or better ways to conduct our 16 maintenance and surveillance activities.
17 Additionally, PG&E engages in a variety of D
18 self-critical evaluations and assessments which are l 19 designed to provide insight into and improve maintenance l 20 and surveillance at DCPP. This constant search for ways lO l
21 to improve is a hallmark of PG&E's and has contributed to l
l 22 the NRC placing DCPP on the "best plants" list for the 23 past two years. Far from being an indication of D
24 programmatic weakness, PG&E's documented investigation and 25 self-critical evaluation of the items cited by MFP, 26 indicates that the overall maintenance and surveillance O
27 program is strong, effective and comprehensive.
87 D .
O .
1 Containment Fan Cooler Units Backdraft Dampers ,
2 Q60 For each of the specific events discussed by the MFP, 3 please provide a very brief description of what went wrong 4 and what was done about it. Let's begin with the 5 Containment Fan Cooler Units Backdraft Dampers.
6 A60 (Giffin) Each of the DCPP units has five Containment 7 Fan Cooler Units ("CFCUs") within the containment to ;
8 provide ventilation and cooling during normal operation as 9 well as postulated accident conditions. Each CFCU has a 10 backdraft damper downstream of the fan. The backdraft O
11 damper is designed to close on reverse air flow and 12 prevent the fans from rotating in a reverse direction. ,
13 The design basis accident reverse flow could be caused by 14 higher pressure in the discharge duct resulting from a t 15 postulated pipe rupture in the lower portion of the 16 containment structure.
O 17 The problems with the backdraft dampers discussed, 18 among other places, in NRC Inspection Report ("IR") 92-17, 19 involved loose counterweights on the dampers and the 20 incorrect assembly of certain linkages connecting the 21 damper vanes caused by inadequate maintenance practices.
22 PG&E identified and evaluated these problems and 23 determined that the CFCUs were operable in the as-found 24 condition. Accordingly, there was no safety significance 25 to the issue. The NRC found no programmatic deficiency or 26 breakdown in PG&E's maintenance program and did not take 27 any escalated enforcement action.
88 0 ,
i I O !
l 1 Nonetheless, because the inadequate maintenance l
O 2- practices fell short of our expectations PG&E formed a HIT 3 team to assure adequate planning and coordination of 4 maintenance on CFCUs. Our investigations revealed several O 5 areas for improvement in the maintenance and surveillance 6 of the CFCUs, including improved procedures for damper 7 linkage servicing and tightening of damper counterweights, O 8 improved procedures for post-maintenance inspection and l
9 testing, improved training, and additional emphasis on 10 attention to detail when working on the plant ventilation 0 11 systems.
12 In addition, we increased system engineering support ;
i 13 for this type of ventilation system and improved our D 14 efforts in performing thorough and probing investigations 15 of problems in ventilation systems. Later, as part of our 16 increased inspection efforts on this system, small cracks D 17 were found in some of the damper vanes. By the end of the 18 recent Unit 2 refueling outage, all of the damper vanes in 19 both units had been replaced with material which is less D 20 susceptible to initiation of fatigue cracks.
l l 21 Positive Displacement Pumo Operatino Procedures l 22 Q61 Please discuss the issues raised by MFP concerning the 1
23 positive displacement pump operating procedures in the l 24 1990-1992 period?
25 A61 (Giffin) Each of the Diablo Canyon Units has one 26 Positive Displacement (Charging) Pump (PDP) and two
) 89 1
i
O ,
1 full-capacity Centrifugal Charging Pumps ("CCPs") to g 2 maintain water level in the reactor coolant system 3 pressurizer during normal operation. The CCPs have higher 4 flow capacity and are also used to inject water into the 5 primary system under certain postulated accident ,
z) 6 conditions. During normal operation, it is preferable to ;
7 use the low flow PDP for charging. However, the industry 8 has experienced hydraulic / mechanical vibration problems O
9 with this type of high pressure, low flow reciprocating 10 pump for many years. At DCPP, the PDPs have required 11 relatively high maintenance, and vibrations under certain O
12 operating conditions have caused pipe cracking in some of i 13 the associated small diameter piping system. l g 14 In August 1990, the PDPs for both units were placed in ;
15 standby service status pending installation of in-line 16 vibration dampening devices. In the interim, the CCPs are g 17 being used for this normal charging function in accordance 18 with plant operating procedures. ,
19 In August 1992, PG&E received two notices of violation 20 concerning the lack of appropriate procedures for O
21 operation of the PDPs as a backup to the CCPs in the i
22 unlikely event of a fire in the centrifugal charging pump 23 room. our own investigation concluded that there was a lO 24 weakness in our procedures for using the PDPs as a backup
?
25 under this postulated fire situation. However, neither 26 the NOVs nor our own evaluation found any deficiency _in g
i 27 maintenance or surveillance of the PDPs. In fact, the [
, 28 problems related to a lack of clear procedural guidance ,
i 90 ;
!()
J l
1 provided by engineering and operations to the operating
) 2 staff on use of the PDPs in response to a postulated fire 3 involving both centrifugal charging pumps. Simply stated, 4 this deficiency was not a maintenance-related issue.
[) 5 Thr. operators' procedures and instructions were 6 clarified to better define the standby status of the PDPs 7 penaing the installation of the in-line vibration j 8 dampening equipment.
9 There was no safety significance due to placing the 10 PDP in the backup mode while the vibration problem was >
] 11 investigated because the combustible loading in the area 12 was limited, smoke detection and the wet-pipe sprinkler 13 system were available, and an hourly fire watch was in
] 14 place in the CCP rooms.
l 15 Steam Generator Feedwater Nozzle Indications ,
i 3 16 Q62 Could you address the MFP reference to indications of l 17 steam generator feedwater nozzle cracks in Unit 1 in 1992?
r
[) 18 A62 (Crockett) The steam generator feedwater nozzle is a 19 20-inch diameter piping connection through which feedwater l
20 flows into each steam generator. These nozzles and the
[) 21 immediate upstream piping are susceptible to interior 22 surface cracking as found at other nuclear power plants.
23 This surface cracking results from nozzle metal 24 temperature difference caused by certain relatively 25 infrequent operating flow conditions during which cold 26 water is flowing into the steam generator through a hot 91
[)
t l
. 1
i f
)
1 nozzle. During a visit to another nuclear plant, a PG&E
} 2 engineer observed a problem the plant had with feedwater !
3 nozzle cracking. When the engineer returned to DCPP, he 4 r commended that as a prudent measure, inspections for
] 5 similar cracking be performed at DCPP at the next 6 scheduled refueling outage. As a result, during the fifth 7 refueling outage for Unit 1, in September 1992, an 8 ultrasonic inspection was performed which indicated some 3
9 surface cracking indications in the steam generator 10 feedwater nozzle connection welds at DCPP.
j 11 Based on these findings, a short piping section and 12 the pipe-to-nozzle weld were replaced on all four steam 13 generators in DCPP Unit 1. Later metallurgical 3 14 investigations of the removed pipe and weld material, 15 using sophisticated detection techniques, determined that 16 the actual interior surface cracking was significantly 17 smaller than indicated by the original ultrasonic 3
18 evaluation. The ASME code allows such small surface 19 cracking and, thus, we could have actually continued 3, 20 operation without repair. Our crack growth projection ;
i 21 calculations demonstrated that crack growth would not have 22 exceeded code allowables for at least another full cycle 3
j 23 of operation.
24 Surface crack depth was determined to have been 25 acceptable during the previous cycle operation, and 26 repairs were not required for safety or code compliance
, 27 reasons. Thus, there was no safety significance to this 28 issue.
92
l l
) u 1 Nonetheless, to minimize future potential problems in 2 this area, a design change is being developed to install a 3
3 thermal sleeve device inside the pipe and nozzle l 4 connection to prevent contact between cold feedwater and i j 5 the hot nozzle connection at this location during the ,
6 relatively infrequent operating conditions of high .
7 temperature differential. In retrospect, this is an 8 excellent example of the proper functioning of the DCPP 3
9 maintenance and surveillance program, especially in 10 assimilating industry experience and proactively i
g 11 initiating repairs even where existing standards do not i
12 require such repairs. ;
i i
13 Reactor Cavity Sump Level Indication 3
14 Q63 MFP has raised a question regarding alleged inadequate 15 corrective action involving the Reactor Cavity Sump Pump j 16 Level Indication system. Could you address this question?
h 17 A63 (Vosburg, Crockett) On November 6, 1990, it was f g 18 discovered that both Wide Range Reactor Cavity Sump Level 19 instrumentation channels were inoperable. Investigative 20 actions subsequently determined that the instruments had l 21 been inoperable since August 21st of that year. The cause 3
22 for the failure of one of the indicators was due to a .
23 blown fuse. However, due to the intermittent nature of ;
24 the failure, our extensive investigation could not 7
, t 25 determine the exact cause for the failure of the other l i
93 l
I i
l l
\
1 channel. A suspect component was replaced and the ]
l )
[) 2 instrument was tested and returned to service. ;
l 3 The delay in recognizing these failures was due to the j
, 4 fact that the instrumentation normally indicates zero q D 5 percent sump level and the indicators had " failed low", l l i l 6 thus giving approximately the same indication. Sump level j I
7 is also displayed on the Safety Parameter Display System j D 8 ("SPDS") which provides the plant operators with certain ;
9 safety-related information. The SPDS provides an l l
10 indication when questionable input values are detected. !
i D 11 As a corrective action for this event, training was given l-l 12 to operations and maintenance personnel on interpreting 13 the SPDS displays with respect to failed channel D 14 indications.
15 On October 22, 1991, another intermittent failure l
! 16 occurred in one of the two sump level indicators and l
D 17 investigation determined that the indicator had been out 18 of service for approximately eight days. Again, the root 19 cause for the instrument failure could not be found. After
[) 20 replacing the instrument and much of the interconnecting 21 cable, the problem was resolved. The sump level i 22 instrumentation has operated normally since that time.
i
) 23 Since the failed instrument had been inoperable for 24 approximately eight days prior to discovery, the 25 effectiveness of the initial corrective action was
) 26 reexamined. Additional corrective measures were taken to 27 revise STP I-1B, " Routine Daily Checks Required By 28 Licenses", to include a step for a member of the
) 94
O 1 Operations staff to check the SPDS display each shift.
() 2 The STP now includes specific guidance with respect to 3 SPDS indications of a questionable instrument channel and 4 actions to be taken when channel problems are identified.
3 5 As further corrective action, PG&E has reviewed all SPDS 6 and control room indicators where a similar instrument
! 7 failure might be difficult to detect from the control room
() 8 and improved the procedure to cover these additional 9 control room indicators.
10 This event is not indicative of a programmatic 0 11 breakdown in the maintenance program. In each case, 12 extensive investigative effort was expended to determine 13 the specific cause of the instrument failures. However, 14 in some cases involving intermittent failures it can be 7) 15 extremely difficult to pinpoint the exact cause for an
! 16 equipment failure. The most likely causes of the problem 17 are addressed, the equipment tested and/or replaced, and 9
18 returned to service under increased surveillance. In this l
l 19 case, the initial actions taken to improve the capability 1
20 to identify a degraded channel were not effective. When f
9 l 21 this was recognized, more comprehensive actions were taken 22 to replace the faulty equipment and assure that any 3 23 possible failures in the future would be promptly 24 detected.
O I
I I
95 3
O 1 Motor Operated valves 2 Q64 MFP has identified DCPP's response to Generic Letter 3
3 89-10, regarding motor operated valves, as being 4 deficient. Could you discuss this issue and DCPP's g 5 response?
6 A64 (Ortore) An extensive MOV testing program is being g 7 implemented throughout the industry in response to NRC 8 Generic Letter 89-10. In their supplemental Petition, MFP 9 refers to NRC Inspection Report 91-39, dated January 24, 1 1992, which identified certain weaknesses in PG&E's MOV O
11 testing program involving calculation methodology, 12 selection criteria, verification parameters and trending.
,0 13 MFP's reading of the inspection report is out of 14 context. Notably, the NRC inspection report in fact 15 concluded that PG&E appeared to be developing an g
16 " aggressive" and comprehensive Generic Letter 89-10 1
17 program. No deviations or violations of NRC requirements 18 were found. The NRC inspection report comments regarding
)
19 " weaknesses" are typical of NRC inspection activities in 20 areas of development and improvement. The inspection 21 report also commented that our MOV program was proceeding 3
22 satisfactorily. Their comments and concerns were 23 evaluated by PGLE and our program has been adjusted in
, 24 specific areas to make further improvements.
25 PG&E believes that it has a good MOV program. PG&E is 26 monitoring the development of the industry's response 96 g
S ;
i j
l' through its participation on industry groups and
- g. 2 committees. Again, nothing in the NRC IR cited by MFP i 3 supports the conclusion that this aspect of DCPP's j 4 maintenance and surveillance program is deficient. $
ti G l 5 Debris Found in Containment Buildinc 6 Q65 MFP has stated that DCPP had a problem with debris in t
g 7 the containment building. Could you discuss this issue?
i 8 A65 (Crockett) The RHR recirculation sump in each 3 9 containment building provides a collection point for water {;
10 during a postulated accident, so that the water can be i 11 recirculated, cooled and returned to the system for g 12 accident mitigation. Housekeeping inside the containment.
13 building should assure that no loose debris (rags, papers, ;
i 14 clothing, plastics, etc.) can be transported to the ,
15 containment sump.
3 ;
16 In October 1991, near the end of the Unit 2 fourth 17 refueling outage, certain debris was found in the f 18 containment building during a PG&E Quality Control 3 ,
19 surveillance. The debris included a small plastic bag, 20 some wipealls, a tool bag, a_ water jug and a tool bin. An l
)< 21 engineering evaluation demonstrated that this small amount 22 of debris would not have impreted the operability of the- ,
23 safety systems and, therefore, it did not represent a j g 24 safety concern. Later, during the Unit 1 fifth refueling 25 outage, although there was overall improvement, some !
26 further problems with control of debris were noted. l 97 3
O' i
1 These problems with control of debris in the .;
() 2 containment building caused PG&E to recognize that the ;
3 program in effect to control material and housekeeping in ;
t 4 containment during outages needed to be reorganized and. [
O 5 strengthened to clarify responsibility for housekeeping 6 and control of loose debris. f I
7 Corrective action Las included the logging of i
() 8 all personnel in and out of containment at key times, f
9 logging and inspection of all jobs in progress every .
10 shift, periodic containment walkdowns, and a ;
+
C) 11 documented tailboard explaining the importance of 12 housekeeping rules for all workers allowed in containment. :
i 13 The containment coordinator is now responsible for ,
() 14 performing a housekeeping inspection of the accessible-15 work areas of the containment during each shift to verify-16 naintenance of housekeeping standards. PG&E believes that l
() 17 these corrective actions will prevent recurrence.
L 18 Diesel Generator Turbo Charcer Bellows Boltina f
() , 19 Q66 Could you discuss the problem with DCPP's expansion 20 bellows during pre-operational installation of the new 21 sixth emergency diesel generator? ,
O ;
22 A66 (Vosburg) An expansion bellows is used on the turbo l
23 charger for each emergency diesel generator to connect the
() 24 turbo charger to the exhaust piping. The issue cited by l i
25 MFP and addressed in NRC IR 92-14 (June 5, 1992) involved 1 26 insufficient written instructions for installing the lower O 98
O l' flange connecting the expansion bellows to the turbo 2 charger during pre-startup installation of the new sixth 3 diesel generator. The installation engineer verified the 4 correct bolting and gasket material to be used, and the 5 installation was completed. However, this action was not 6 documented in accordance with DCPP procedures. Because 7 the work was completed properly, the documentation 8 deficiency had no safety significance.
9 This issue was not a maintenance or surveillance 10 problem. It occurred during the installation of the new 11 sixth diesel generator during the construction phase. At 12 the time, the sixth diesel generator had not been turned 13 over to the plant, and was not yet part of the plant's 14 safety systems. The deficiency in documentation was 15 corrected and the level of attention to such documentation 16 was improved on the sixth diesel generator project.
O 17 Diesel Generator Fuel Oil Pipinc Corrosion 18 Q67 Could you discuss the corrosion problem discovered by 19 PG&E involving the diesel generator fuel piping?
20 A67 (Crockett) The diesel generator fuel oil system is 21 used to transfer fuel oil from underground storage tanks 22 to the emergency diesel generators. This small diameter 23 piping runs below ground in concrete trenches, on the 24 seaward side of the turbine building and within the 25 turbine building buttress area. The piping was installed 26 in the 1970s and has experienced corrosion on its external 99 0
=
)
i l
1 surfaces from the salt air environment. Over time, this ;
l h 2 environment corroded the piping's exterior surface in 3 areas where the external coating on the piping had not 4 been fully effective in preventing corrosion.
3 5 This problem, identified by PG&E in voluntary Licensee 6 Event Report 1-92-006, was discovered during a PG&E >
7 inspection. Certain local repairs were made in the highly
) 8 corroded areas. PG&E determined that the system's r 9 operability was not compromised even in the degraded s 10 condition. However, based on the general conditions of h) 11 the piping and the environment in the trench area, it has >
l 12 been decided to replace the entire length of piping and to '
i 13 use improved coating techniques in the new installation.
21 14 Accordingly, this issue - identified and corrected by 15 PG&E - does nothing to support MFP's contention. This l
16 type of problem and its resolution illustrates how the i
) 17 DCPP maintenance and surveillance program functioned 18 properly to find the deteriorating piping and then to 19 replace the piping with upgraded design materials or 3 20 construction techniques.
i t 21 Chemical and Volume control System Leakage 22 Q68 Describe the chemical and volume control system 23 leakage issue raised by MFP regarding valve diaphragm 24 degradation.
)
25 A68 (Giffin) During normal plant operation, the primary 26 function of the Chemical and volume Control System
) 100
i !
i l
0
~
1 ("CVCS") is to maintain reactor coolant system inventory. ;
[) 2 Parts of the system are heat-traced to keep concentrated f
i 3 boric acid in solution. During a postulated accident, ;
4 portions of the system are used to recirculate and supply 0 5 water to mitigate the accident. !
6 In June 1992, during a routine radiation survey, .
7 leakage was noted from a CVCS valve bonnet in a ,
O 8 heat-traced portion of the system. This leakage was 9 determined to be outside the design basis for CVCS 10 leakage. The leak was immediately stopped by tightening
() 11 the valve body-to-bonnet nuts. The issue was identified ,
12 by PG&E and reported in LER 1-92-009, Revision 1, dated 13 January 11, 1993.
() 14 Our investigation revealed that the cause of the 15 leakage was thermal degradation of the valve diaphragm due i 16 to a malfunctioning heat trace thermostat causing leakage I
O 17 through the body-to-bonnet joint. Corrective actions 18 included lowering the heat trace temperature for the 19 valve, reviewing every other valve in similar service in O 20 Units 1 and 2 to verify there were no other potential 21 problems, reviewing the entire heat trace system for 22 proper setpoints and installation, and improving our i
O 23 surveillance monitoring program.
24 A similar leak occurred in September 1991 due to a
! 25 different root cause. In that instance, valve ;
i
). 26 body-to-bonnet leakage was caused by an isolated case of 27 the valve vendor recommendations (concerning bonnet nut 28 torque and diaphragm replacement) not being adequately l
0- 101 i
I t !
O l 1 included in the preventive maintenance program. The
() .2 preventive maintenance program was reviewed and upgraded-3 to address this issue. No similar problems were found.
i l 4 The analysis of the as-found leakage in both events
() 5 determined these events were not safety significant or 1
i 6 programmatic.
7 Measurina and Test Ecuipment Control Deficiencies
()
l 8 Q69 Describe DCPP's response to measurement and test
- 9 equipment control deficiencies identified in 1991.
O 10 A69 (Giffin) An NRC inspection during February 1991 found 11 deficiencies in the control of Measuring and Test 3 12 Equipment ("M&TE") administrated by the Mechanical .
i 13 Maintenance ("MM") Department at DCPP. This inspection i
14 also found that PG&E's own Quality Control ("QC") and ,
g 15 Quality Assurance ("QA") Departments had previously 16 audited the MM M&TE program and found similar deficiencies 17 which had not been aggressively corrected. i g 18 The identified deficiencies were reviewed in detail to 19 det' ermine the potential impact on plant equipment. The 20 deficiencies involved calibration or use of torque 3 21 wrenches and nechanical gauging devices. The application l
l 22 of the torque wrenches and gauging devices was found to .
L
( 23 have no effect on the plant equipment. The overall f) c 24 maintenance program controls for safety-related work 25 (e.g., detailed work orders, post-maintenance operability
?
l 1
3 102 .
i l
i l
1 testing, etc.) were found to be adequate. Therefore, O 2 there was no safety significance to these deficiencies.
3 The corrective action for the deficiencies in the MM l
4 M&TE program included accelerating a previous manacement
.O 5 decision to transfer and consolidate all responsibility l
6 for MM M&TE equipment to the Instrument and 2cntrols (1&C) :
7 Department M&TE program -- a program which was, and is,
() 8 working effectively. Prior to the NRC inspection, PG&E 9 had initiated a QA audit to verify and supplement !
l l 10 information on the deficiencies found by PG&E's QC O 11 organization, but had not yet implemented improvement in a l
f 12 timely manner. Additional corrective action included i
13 improvement in the quality resolution process, better O 14 definition of management expectations concerning the 4
15 resolution of quality audits, and improved training on :
16 M&TE requirements.
O i 17 Emercency Diesel Generator Surveillance Test Issues i
18 Q70 In their reply to PGEE's response to the proposed 0 19 contention, MFP supplements the original basis of the {
20 proposed contention by citing " weaknesses" stated by the ;
j 21 NRC in IR 92-01. The MFP state that these " weaknesses" ;
r !
l() 22 challenge the maintenance and surveillance program. What
! 23 is your reaction to this issue. ,
e f
- O 24 A70 (crockett) During 1991, the NRC performed an extensive i 25 functional inspection on Diablo Canyon's electrical 26 distribution system. The NRC issued IR 92-01 as a result 1
'O 103 4
i Y
1 of this inspection. The specific " weaknesses" which were g 2 described on page 1 of this NRC report and discussed by 3 MFP were attributed by the NRC to the Engineering 4 Department. These weaknesses were not relevant to DCPP's 5 naintenance and surveillance programs. Two Severity Level O
6 IV NOVs were issued but they were separate and apart from 7 the NRC " weaknesses" discussed by MFP. Both NOVs related 8 to specific surveillance procedures which have been
)
9 corrected. The overall conclusion of the inspection was 10 that DCPP's electrical systems were acceptable, no 11 immediate safety or operability concerns were identified,
()
12 and no broad scope programmatic breakdowns were noted.
13 Fuel Handlinc Building Ventilation Leakage g
14 Q71 Describe the DCPP fuel handling building ventilation l
15 leakage issue raised by MFP's Contention IV (age-related x 16 degradation).
.J l
l l
17 A71 (Crockett) The Fuel Handling Building ("FHB")
3, 18 ventilation system for each unit is designed to maintain a 19 slight negative pressure inside the building. This design 20 is accomplished by removing more air (with the exhaust 21 fans) than what is supplied to the FHB (by the supply ;
c) 22 fan). Thus, air leakages will be inward and all potential 23 releases from the spent fuel pool are exhausted through l
)
, 24 filters. This functional capability is required to be j
(
25 verified by a surveillance test procedure every 18 months. 1 26 On September 1989, this surveillance was performed and 1
g 104 1
)
)
I fully met the specific requirements for negative O 2 pressures. The next required test was performed on 3 January 18, 1992. Although the measured pressure was 4 negative, it did not meet the specific operability 0 5 requirement. PG&E conservatively assumed that the FHB 6 ventilation system was inoperable at the time of fuel 7 movement during the Unit 1 third refueling outage in O 8 October 1989. Prompt action was initiated in accordance 9 with our problem resolution procedure to investigate the 10 situation, determine a root cause and implement corrective O 11 action.
12 Our investigation determined that the cause of system 13 inoperability was the existence of small leakage paths 0 14 into the building due to building and seal degradation.
15 Corrective action included sealing the leaks. Both FHBs 16 have also been re-sided. The new siding is the best b 17 available on tne market with an expected life of greater 18 than 25 years.
19 Analysis of the event showed that because of 20 conservatism in the ventilation system design, sufficient 21 negative pressure was maintained at the surface of the 22 spent fuel pool throughout the event and there was no O 23 safety significance.
24 RHR Recirculation Sump Screens O 25 Q72 MFP has stated that DCPP had a problem with gaps in 26 the containment building recirculation sump screens and 27 debris inside the sumps. Could you discuss these issues?
p 105
)
1 A72 (Crockett) Yes. RHR recirculation sumps are
) 2 surrounded by debris screens. In February 1990, the NRC 3 issued Enforcement Action 89-241 citing PG&E for three 4 violations related to the RHR recirculation sump screens.
) S One of the violations involved inadequate engineering and 6 construction completion, which resulted in unacceptable 7 gaps in the screen structure. The second involved
) 8 operational control that allowed an unattended open access 9 hatch in the screens. Neither of these two violations 10 bears any relation to maintenance or surveillance. Both
) 11 were corrected. The third violation included a reference 12 to poor performance of a visual surveillance of the sumps, 13 which resulted in unidentified debris remaining in the
) 14 sumps. This deficiency was an isolated personnel error --
15 an individual failed to implement an otherwise clear 16 procedure. This does not indicate a program problem. As
) 17 a corrective action, PG&E clarified requirements for 18 inspection of the sumps.
) 19 Turbine Governor and Stop Valve Malfunction 20 Q73 Describe the DCPP turbine governor and stop valve 21 malfunction cited by MFP.
J
)
22 A73 (Vosburg) The main turbine speed is controlled by four 23 governor valves ("GVs") during startup, operation and ;
1
) 24 shutdown. Four additional stop valves ("SVs"), located in 25 line with the GVs, are used to trip and isolate the 26 turbine during overspeed and other conditions. All cight
) 106
D 1 valves are controlled by an electro-hydraulic ("EH")
3 2 system.
3 In September 1992 during a Unit 1 shutdown, 4 simultaneous circumstances occurred which resulted in a 3 5 steam flow path being created through the SVs and GVs 6 after the turbine had been tripped, thereby causing the 7 turbine to accelerate. The operators controlling the l 3 8 evolution immediately tripped the turbine. No reactor or t
9 turbine limits were exceeded.
10 Our investigation into this event found the cause to ,
3 11 be a combination of factors, including a sticking pressure j 12 switch, low EH system pressure and steam leakage. ,
4 13 Corrective actions included replacing and upgrading the 3 14 pressure switch, correcting several conditions in the EH 15 system and revising our shutdown procedures to add 16 additional verification of EH system control parameters 3 17 during turbine shutdown. ;
18 Two other previous related events merit discussion. i 19 In August 1991, a problem with the Unit 2 turbine control 3 20 system resulted in a spurious GV opening. Rigorous j 21 investigative testing could not determine the root cause. {
22 The problem was intermittent and could not be reproduced. {
3 23 The most likely cause was thought to be a sticking ?
24 pressure switch, similar to the switch which contributed i 25 to the September 1992 Unit 1 event. The switch was
) 26 replaced. In March 1992, a spurious SV opening occurred !
i 27 on Unit 1 due to suspected steam leakage. The SVs were ;
I i
J 107 l j
)
1 operable and, therefore, no corrective action was D 2 necessary.
3 PG&E believes tPV its investigation and corrective 4 actions for these two events were comprehensive and D 5 thorough. In retrospect, operations personnel should have 6 implemented more positive interim compensatory action 7 (such as changes to the operating procedure or additional D 8 training) which may have prevented the third, September 9 1992, event. As part of our corrective action for the 10 third event we upgraded our use of interim compensatory D 11 measures.
12 No operating limits were exceeded during any of the 13 three events. This equipment is not safety-related. None D 14 of the events was safety significant.
15 4kV/12kV Cable Problems D 16 Q74 MFP has raised an issue regarding maintenance of the 17 4kV and 12kV underground cables between the turbino 18 building and the intake structure. Please discuss the D 19 recent problems experienced by DCPP involving these 20 cables.
D 21 A74 (Ortore) DCPP has experienced three failures of 4kV 22 cables -- two safety-related and one nonsafety-related.
23 Two of these failures were random in nature and time of 24 occurrence. One of the failures was detected by PG&E 3
25 during post-maintenance High-Potential ("Hi-Pot") testing.
26 Additionally, there were two area-specific failures of D 108
)
1 non-safety 12kV cables. Both types of cable are contained
}
2 in conduits located in underground duct banks routed 3 between the turbine building and the intake structure.
4 The safety-related and non-safety related 4kV cables i
5 supply electrical power to safety related cooling water i 3
E pumps and a non-safety related load center transformer, 7 respectively. The nonsafety-related 12kV cables supply 8 electrical power to the main circulating water pumps.
3 &
9 The failures which occurred in the 12kV 10 nonsafety-related cables were caused by exposure to a j 11 contaminant which was present in the underground conduits 12 in a localized area between the turbine building and the i
13 intake structure and which caused severe cable jacket and 14 insulation degradation. Both of these failures were 3
15 detected with the ground detection system, allowing ,
16 operator action to isolate the cable before complete 17 failure of the cable.
3 18 Neither of the two safety-related 4kV cable problems i 19 involved actual in-service failures. One involved a j 20 momentary alarm indication signaling a potential ground 21 fault. When this occurred, the redundant cooling water 22 pump was started and the cable with the potential fault 23 was isolated. The second case occurred during routine 3
24 "Hi-Pot" testing while the cable was out of service during 25 a refueling outage. One of the two safety-related 4kV 26 cables was in a location remote from the more recent 12kV ,
7 27 cable failures. Moreover, the 4kV cable jackets did not i 23 exhibit any physical degradation as did the 12kV cable
) 109 i__.____.___.______._ _ _ _ -_ _ _ .
O 1 jacket. These differences would suggest that there was no g 2 imminent common-mode cable failure mechanism involved with 3 these occurrences .
4 After failures occurred to the 4kV and 12kV cables, 5 the failed cable sections were replaced. Reviews of the 6 original design, installation, quality assurance and/or 7 quality control audits for the failed cables were 3
8 conducted. These reviews have concluded that the installed 9 cables are of acceptable quality and design for the'ir 10 specific applications and service conditions (wet or dry).
11 However, as a prudent measure, additional sections of 3
12 unfailed 4kV and 12kV cables have been replaced.
13 A contributory cause of the 12kV failures was believed 14 to be water carrying contaminants into the cable conduits.
3 15 While sump pumps are provided in the vaults, they were not 16 at that time part of the formal maintenance program and, 17 accordingly, were not maintained in an adequate manner.
g 18 After rains, the vaults would fill with water and flood 19 the conduits. The sump pumps have been repaired and are 20 now included in our preventive maintenance program.
3 21 PG&E believes that the maintenance activities with 22 regard to these underground cables are adequate, 23 particularly given the installations involved. To the g
24 extent our continuing analysis of these issues suggests 25 the need for further naintenance activities, they will be
, 26 included in our program.
110 g
)-
1 Q75 Are there other issues raised by the MFP in their 2 supplemental filing of December 10, 1992 regarding 3
3 Contention I, that were not maintenance and surveillance 4 issues? If yes, would you please comment on them.
D 5 A75 (Giffin) Yes, there are several issues raised by the 6 MFP in the document that are not maintenance issues. The
) 7 major issues are as follows:
8 e CCW, AFW, and Fuel Oil Transfer Pump Vault Drain (NRC 9 Enforcement Action 89-85, May 5, 1989) 3 10 -
These problems involved design or design basis 11 issues, not the ma'intenance and surveillance 12 programs. They were corrected by revisions to
)
13 operating procedures, reclassification of l 14 equipment, or upgrading design basis i
15 documentation.
)
16
- Feedwater Pump - Control System Power Supply Failure :
17 (Management Meeting of April 2, 1992).
18 This problem was a design issue involving a
)
19 particular electrical circuit design for redundant
?
20 power supplies, and it was not a maintenance or
- 21 surveillance issue. A complete reassessment of 3
22 this design to enhance the capability to withstand 23 local component failures was performed and 24 modifications to the electrical circuit were 25 implemented.
111
)
) !
l J
l 1 e Residual Heat Removal ("RHR") System Operational
) 2 Problems (NRC Enforcement Action 87-131, dated 3 August 7, 1987 and NRC NUREG 1269, " Loss of 4 Residual Heat Removal System, Diablo Canyon Unit i
D 5 2")
6 -
The loss of RHR occurred during mid-loop ,
7 operating conditions. The causes of this -
D 8 event were related to inadequate control of 9 system operation during this infrequent and 10 abnormal operating condition. The event 3 11 resulted in extensive investigative and 12 corrective action. The enforcement action 13 cited deficiencies in the operating procedure
- 3 14 control, the quality control inspection, and a 15 safety evaluation. These were not maintenance 16 or surveillance issues. Corrective actions 3 17 included several operating procedure changes, j t
18 upgrading of operator training, and 19 improvement of the temporary level indicating D 20 systems used during refueling outages.
r 21 In summary, none of these matters related to any 22 programmatic deficiency or breakdown in DCPP's maintenance :
O 23 or surveillance programs.
i
( k
' 24 VII. CONCLUSION
- 25 Q76 Do any of the problems cited by MFP and described 26 above indicate a pervasive concern relative to lack of ;
l l
D 112
)
l l
1 t
l l 1 timeliness in resolution of problems or a lack of proper
(
) 2 attention?
l 3 A76 (Giffin) No. In the vast majority of cases we p 4 consider our response to have been timely. In applying ,
i 5 resources to address issues, we consider safety
! 6 significance, operational necessity, and the time it takes 7 to perform a comprehensive assessment of the issue. For 8 the kinds of issues identified by MFP, we believe that I
9 thorough investigation and thoughtful corrective action j 10 are very important aspects. We also believe that we have ,
11 emphasized the need for thorough responses, and that the 12 extensiveness of our corrective actions reflects our ;
l j 13 desire to promote significant improvements to our 14 maintenance and surveillance program.
I j 15 Q77 Do these issues of events indicate a programmatic 16 deficiency or breakdown in PGEE's maintenance and 17 surveillance programs at DCPP? Why or why not?
\
)
- 18 A77 (Giffin) No. The issues or events described above do ,
19 not in any way represent a programmatic deficiency or j 20 programmatic breakdown in PG&E's maintenance and 21 surveillance programs at DCPP. These issues or events i
22 represent a relatively few random and isolated occurrences
) 23 that are inevitable in any large and complex program of 24 human activities. They are not concentrated in any one l 25 aspect of the program. Rather, they have involved
} 113
1 different equipment, systems, procedures, locations, or
} 2 operating conditions. Furthermore, they have occurred at 3 random times and in a random manner, without a particular 4 pattern relating to organization staffing or supervision.
). 5 Moreover, I want to re-emphasize that the 6 maintenance-related problems that have occurred at DCPP 7 represent only a small number of occurrences relative to
) 8 the very large number of maintenance activities, 9 surveillances, and individua' steps in maintenance work ;
10 orders. We perform more than a million individual ,
)
11 maintenance and surveillance tasks and activities at each 12 unit during each refueling cycle.
These problems and 13 issues have been aggressively handled and thoroughly
) 14 documented in a clear, self-critical manner, e
l 15 Q78 Are you aware of events or documents, other than those ,
) 16 cited by MFP as a basis for Contention I and discussed 1 17 above, that relate to maintenance and surveillance at 18 DCPP? j
) .
19 A78 (Giffin) Yes. Based on information provided to MFP by ;
I 20 PG&E during discovery in this case, MFP has identified l i
)
l 21 other documents which it intends to rely upon in this j t
22 proceeding. Some of these relate to plant maintenance and j l
23 surveillance activities.
r 24 Q79 How would you characterize these documents? l 4
l ll
O 1 A79 (Giffin) These documents generally are part of our normal 2 processes. As we have discussed elsewhere in this 3 testimony, when issues or problems are identified at the 4 plant, they are documented and thoroughly investigated to 5 prevent recurrence. A process is in place to track these 6 matters to resolution. We believe that a program of 7 aggressively finding and documenting problems, thoroughly 8 investigating them, and continually improving our 9 operations is essential to a complete and comprehensive 10 approach to plant maintenance.
11 In addition, as a means to foster continuous 0
12 improvement, we do perform from time to time candid self 13 assessments of our programs. In this context, the 14 existence of deficiency reports, critical
.O 15 self-assessments, and the like are not evidence of a 16 faulty maintenance program. Rather, they evidence a 17 properly functioning program.
O 18 Q80 Does this conclude your testimony?
O 19 A80 (Crockett, Ortore, Vosburg) Yes.
O O
115 O
O August 2, 1993 O UNITED STATES OF AMERICA '
NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD ,
i O
In the Matter of: ) Docket Nos. .50-275-OLA !
) 50-323-OLA Pacific Gas and Electric Company )
) (Construction Period (Diablo Canyon Nuclear Power ) Recovery)
.O Plant, Units 1 and 2) )
)
i TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY ADDRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE i
I
-O pART 2: Tedd A. Dillard ;
i e
- O 1
1 F
a N
l f
i .
t
, 4 5 .
!O -
I t
i O
l
lO l
1 August 2, 1993 ;
2 UNITED STATES OF AMERICA I
3 NUCLEAR REGULATORY COMMISSION '
I 4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD i
() 5 6 In the Matter of: } Docket Nos. 50-275-OLA 7 ) 50-323-OLA ,
8 Pacific Gas and Electric Company )
9 ) (Construction Period
'O 1 (Diabl Canyon Nuclear Power ) Recovery) 11 Plant, Units 1 and 2) .
)
12 )
13 TESTIMONY OF TEDD A. DILLARD ON BEHALF OF
'g 14 PACIFIC GAS AND ELECTRIC COMPANY 15 ADDRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE 16 Q1 Please state your name, affiliation, and current g 17 job responsibilities.
18 Al My name is Tedd Dillard. I am the Supervisor of 19 Component Programs for the Nuclear Division of Florida O
20 Power & Light Company ("FPL"). I was the Manager of 21 Maintenance at the St. Lucie Nuclear Power Plant from 22 May 1983 until November 1988. Prior to that, I was the
-() ,
23 Mechanical Maintenance department head at St. Lucie for j 24 eight years. My current responsibilities include ;
-g 25 supervising the efforts of a group of engineers i
26 responsible for the FPL Nuclear Division's technical r
27 expertise for major plant components at both FPL
- g 28 nuclear facilities. I am also responsible for 29 developing FPL's implementation plan for the l l
l 30 Maintenance Rule, 10 CFR 50.65. !
116 O
O 1 Q2 What other qualifications do you have in the area O 2 of nuclear power plant maintenance?
3 A2 In addition to my maintenance experience at FPL, I O 4 have had an opportunity during the past six years to 5 visit a number of nuclear power plants and participate 6 in several industry maintenance assessment efforts. I O 7 was corporate evaluator on an Institute of Nuclear 8 Power Operations ("INPO") maintenance assistance review 9 team evaluation of our Turkey Point Plant. I was a 0 10 peer evaluator on an INPO maintenance assistance review 11 team at the Ft. Calhoun Nuclear Plant. I was a member 12 of one of several teams who visited the Mihama Nuclear O 13 Plant of Kansai Electric Company in Japan in 14 conjunction with our quality improvement efforts. I 15 served on the Nuclear Management and Resources Council O 16 ("NUMARC") Ad Hoc Committee that developed the industry 17 response to the proposed Nuclear Regulatory Commission 18 ("NRC") Maintenance Rule. I also served on one of four O 19 committees that developed the implementation guidance 20 document for NUMARC that the NRC has endorsed with the 21 regulatory guide for the NRC Maintenance Rule.
O 22 I have more than 23 years of experience in United 23 States commercial nuclear power plants, the last 20 of 24 which has been with FPL.
O 25 A copy of my professional qualifications is 26 provided in Exhibit 12.
O 117
9 1 Q3 What contention will you address?
O 2 A3 I will address San Luis Obispo Mothers for Peace 3 ("MFP") Contention I, which is:
g 4 "The San Luis Obispo Mothers for Peace contends 5 that Pacific Gas & Electric Company's (PG&E) 6 proposal to extend the life of the Diablo Canyon 7 Nuclear Power Plant for more than 13 years (Unit 8 1) and almost 15 years (Unit 2) should be denied 9 because PG&E lacks a sufficiently effective and g 10 comprehensive surveillance and maintenance 11 program."
12 Q4 What is the purpose of your testimony?
O 13 A4 The purpose of my testimony is to provide my 14 opinion as to whether PG&E has implemented a C 15 comprehensive and effective maintenance program at the 16 Diablo Canyon Nuclear Power Plant ("Diablo Canyon").
17 PG&E specifically requested that I visit Diablo Canyon O
18 and review the maintenance program, procedures, and 19 program implementation at the plant in light of my 20 experience with maintenance at St. Lucie and in O 21 observing and evaluating other nuclear power plants.
22 It is my understanding that PG&E solicited my views, in 23 part, based on the fact that St. Lucie has been given O
24 Category 1 ratings by the NRC in the maintenance area 25 in Systematic Assessment of Licensee Performance 26 ("SALP") evaluations. St. Lucie is also one of the O
27 plants singled out by the NRC for distinction on its 28 "Best Plants" list.
9 118 i
D 1 Q5 What were your overall impressions of the Diablo 3 2 Canyon maintenance program?
3 AS Based on my review of the program and procedures, 3 4 my interviews with maintenance personnel, and my visit 5 to the plant, it appears to me that PG&E has 6 implemented a comprehensive and effective program. I 7 observed no significant weaknesses. In the time 3
8 available to me, I did not, of course, review all 9 details of the program. However, I should add also --
10 and I will spell this out in more detail below -- that 3
11 based on my experience, I believe I can spot quite 12 readily a well-maintained plant or a poorly-maintained 13 plant based on some important indicators. Diablo 3
14 Canyon appeared to me to be very well maintained by a 15 very professional and knowledgeable staff.
D 16 Q6 In general, explain the criteria you applied in 17 evaluating Diablo Canyon.
D 18 A6 I will use as a guide the elements contained in 19 INPO Document 90-008, Maintenance Proarams in the
'20 Nuclear Power Industry. This document is a compilation
]
21 of the performance objectives and criteria used by INPO 22 in evaluating nuclear power plants. I believe that the 7 23 objectives and criteria are comprehensive and that if 24 an organization effectively implements them the 25 resulting performance will be good, and more 119 3-
O ,
}
1 importantly, will continue to improve over time. Like O 2 many plants, Diablo Canyon has a maintenance program 3 that meets the elements contained in INPO 90-008. I ,
4 would like to point out that INPO 90-008 is not an O 5 organization chart or structure, but is a list, in no 6 particular order, of elements or functions of F 7 maintenance that have been shown to be important to the ,
i
-O 8 effective operation of a nuclear power plant. :
9 I note also that in one of the prehearing 10 conference orders in this case, the Licensing Board t
O 11 observed that the existence of adequate programs on ;
t 12 paper does not constitute proof of effectiveness, but ;
13 that "the implementation of these programs is the only O 14 real gauge of their effectiveness." Given my operating 15 background, I could not agree more. When I look at a 16 plant, Diablo Canyon included, I specifically look at -i O 17 how paper programs are implemented. r I
18 In addition to an overview of the programs in i
19 place at Diablo Canyon, I have, therefore, tried to O_ 20 make specific personal observations at the plant. It 21 is not the purpose of my testimony to respond in detail f 1
22 to any of the specific examples cited by MFP in the j l
O 23 bases for their contention. However, I have tried to l l
24 use observations closely related to their examples. ;
25 The):e are thirty INPO elements, but I am addressing the Cf 26 eignteen that I believe are most relevant to the l
27 cortention.
O 120
) 1 l
l 1 Q7 Please address your criteria one-by-one and [
2 provide your observations regarding Diablo Canyon. The ,
3 first concerns Management Assessment.
l 4 A7
- Manacement Assessnent 5 Performance Objective: Management and 6 supervisory personnel should monitor and 7 assess station activities to improve all 8 aspects of station performance.
9 Criteria: Line managers and supervisors are 10 responsible for and personally take part in 11 monitoring and assessing station activities.
12 Assessments by other independent groups, such 13 as Quality Assurance, are used by line
'. 14 managers and supervisors as a management tool
, 15 to assist them in assessing station 16 performance.
17 It is my observati~on that this element of the l
18 program is in place. Specific examples are:
I 19
- The housekeeping throughout the plant was i
I f 20 excellent.
j 21
- There were only about twenty action tags hanging 22 on equipment in the secondary part of the plant.
23
- The condition of both operating'and standby l
i 24 equipment was very good. ,
25
- Compliance with industrial safety work practices 26 was excellent. Everywhere I went people wore 27 their hard hats, eye glasses, and hearing
! 28 protectors. ,
29
- People were careful to ensure that fire and l
30 security doors locked properly behind them.
31
- Supervisors and managers that I talked to were ;
32 readily able to discuss the condition of 121
)
J h
1 equipment, and were very responsive to questions O 2 about the trends in performance and condition in i 3 their area.
4 It is my conclusion that the conditions that I O 5 have listed (and there are many more) would not be true 6 unless the managers and supervisors are doing an 7 excellent job in assessing plant conditions.
p l
8 QB The next criterion is Quality Programs. Please l 9 describe your observations in this area.
O 1
10 A8
- Ouality Proarams 11 Performance objective: Quality programs
! 12 should effectively monitor activities that O 13 affect safe and reliable plant operation, 14 provide feedback to line management on 15 quality of performance, and contribute to 16 improved performance. ,
l 17 Criteria: Quality programs reinforce and ,
O 18 support the line functions of managers and 19 supervisors. Line managers and supervisors .
20 are responsible for and held accountable for i 21 the quality of work performed within their 22 area of responsibility.
O. 23 There were a number of indications that quality
, 24 programs are healthy at Diablo Canyon. Some examples l
l 25 follow: :
i
- 26
- One of the best examples that the attitude toward i 27 quality programs is good is the comprehensive-28 plant aging management document, " Aging Management
) 29 Program," dated June 30, 1992, generated at a very 30 high level in the organization. This shows a 31 strong commitment to the issue. This is being i D 122 l l
t
i .
'O l
i l
l 1 addressed as a high level business objective. It O 2 is a very comprehensive document and includes all i
3 aspects of plant aging. It addresses how to 4 define terms and how to apply the programs. It O 5 discusses different age-related degradations and '
l 6 mechanisms. But most importantly, it outlines the l
7 relationship of the many activities, ,
f) 8 organizations, and outside factors that bear on 9 the issue. It also assigns responsibility for the i
10 implementation. I believe this document indicates O 11 that PG&E is clearly aware of the importance of 12 managing this issue and has taken action to do so.
13
- Examples of quality programs in place that I O 14 observed are:
15 -
Hot work permits 16 -
Equipment clearance tagging I) 17 -
Scaffold tagging 18 These examples demonstrate that quality programs 19 are in place and are working. ,
<O 20 Q9 The next criterion is Maintenance Organization and q 21 Administration. Please address this.
b 22 A9 e Maintenance Orcanization and Administration I 23 Performance Objective: The maintenance 24 organization and administration should ensure
<) 25 effective implementation and control of 26 maintenance activities.
[ 27 Criteria: Administrative controls are 28 employed in the conduct of maintenance O 123
I O
l 1 activities that affect safe and reliable 2 plant operation. Examples of such activities O 3 include scheduling of preventive maintenance, 4 use of special tools and lifting equipment, 5 and use of measuring and test equipment.
6 In my tour of the plant and discussions with plant )
O 7 staff, it was clear that the maintenance organization 8 and administration are being effectively implemented.
9 Observations are: !
O 10 . In all discussions with plant staff, they had a 11 ready answer for who was responsible for various 12 aspects of what was being discussed. For example, ;
O 13 when talking to technical support people about the I 14 Inservice Testing ("IST") program, they had a 15 clear understanding of how the check valve O 16 inspection program in maintenance fits in.
17
- The Work Planning Department plans and coordinates -
18 the work of all three maintenance disciplines with O 19 the Operations Department. This is a free-20 standing organization that must effectively -
21 coordinate the activities of many different O 22 departments. They are responsible for daily work !
23 and outage activities, too. The efficient outages l 24 that Diablo Canyon has had show how effectir' PG&E O 25 has been, as this is a major effort coordinating 7 26 several thousand activities in a sixty-day period.
j 27
- Significant work has been done at the intake ,
28 structure over the past couple years. The scope l 29 and volume of this work was substantial and 30 required a coordinated effort from many groups to 124
O i i I
1 be successful. The success of the improvements in i c
O 2 the intake structure is an excellent example of t 3 organizational effectiveness.
4
- An interdepartmental administrative procedure was
=
O 5 developed to improve the implementation of the '
i' 6 plant's response to Generic Letter ("GL") 89-10 i 7 regarding motor-operated valves ("MOVs"). This 50 8 document outlines the actions and accountabilities 9 of eight different plant organizations that play a 10 role in effectively responding to the requirements ,
O 11 of GL 89-10. The NRC noted in their inspection of 12 this area that "the inspection findings indicated t
13 that you appear to be developing an aggressive, o
0 14 well-integrated program for assuring MOV 15 reliability. Program strengths were found in the 16 area of program scope and your high impact team t
C) (HIT) approach to integrating the 17 18 multi-disciplined activities required for the 19 89-10 program." f O 20 These examples show, in my opinion, that the l 21 maintenance organization is properly defined and that i
22 proper administrative controls are in place. From what ;
O 23 I have seen, the organization appears to be effective. ,
24 Q10 Next, please address the Plant Material Condition 25 at Diablo Canyon.
26 A10 e Plant Material Condition 0 125 1
1 Performance Objective: The material G 2 condition of the plant is maintained to 3 support safe and reliable plant operation.
4 Criteria:
5 -
Systems and equipment are in good 9 6 working order; examples of this include 7 the following:
8 a. Fluid system leaks are minimized.
9 b. Equipment is appropriately 10 protected from adverse
- 11 environmental conditions.
12 c. Instruments, controls, and 13 associated indicators are 14 calibrated, as required.
15 d. Good lubrication practices are 16 evident.
O 17 e. Fasteners and supports are properly 18 installed.
19 f. Equipment, structures, and systems 20 are properly preserved and 21 insulated.
O 22 -
Material deficiencies are identified and 23 are in the work control system.
24 -
Temporary repairs are minimized and 25 permanent requires are made when 26 conditions permit.
27 -
Temporary environmental protection D 28 (e.g., dust, humidity, freeze, shock) is 29 provided for plant equipment when needed 30 to support construction, outage, or 31 maintenance activities.
32 -
Newly installed or modified 33 systems / equipment are verified to be in O 34 good working order prior to operational 35 acceptance by the plant staff.
36 Plant material condition is one of most visible 37 examples of maintenance effectiveness. The condition 38 of the equipment is the end to which all programs and 39 efforts point. While visual appearance alone is not a 40 guarantee of reliability, it is a very good indicator 41 of equipment condition. Diablo Canyon has an excellent i
3 126 i
lO I plant material condition. Some of the many examples g 2 are noted below:
3 e I toured two of the six emergency diesel generator l 4 rooms. The rooms were very clean with no dust or 5 trash anywhere. There were no equipment or tools l0 6 left anywhere in either of the rooms. There were -
I 7 only two action request tags hanging in both 8 spa es. The air start air supply system was very
.O ,
9 leak-tight as the compressors never cycled in the 10 ten minutes we were in the rooms.
11
- The 4kV switch gear rooms were in excellent
)
12 condition. There were no cross-under pipe manway 13 cover leaks. It h'as been my experience that they ,
14 are next to impossible to consistently make up g
15 leak tight.
16
18 condition. The valves, pipe hangers and pipe 19 snubbers were all in good condition. f 20
- The feedwater heater level control stations were O i 21 in excellent shape. They were_ leak-free, clean, 22 the lights behind the level site glasses were lit 23 and were in about as good a condition as any I g t 24 have ever seen. It has been my experience that the 25 level site glasses are difficult to prevent from l
26 leaking. The insulation on the feedwater heaters
)
27 was very clean and neat.
f ,
127 g
l 0 :
I i
1
- The main feedwater pump and drive turbines were in O 2 very good shape. These types of turbines have a [
3 low pressure control oil system and usually have a ;
4 number of small leaks on the reservoir and bearing l i
O 5 pedestals, in my experience. Tnese particular ;
6 ones were very clean and dry. The turbine l 7 platforms were clean and free of any debris. I O 8
- The condensate pumps, the condensate booster !
9 pumps, and the heater drain pumps were all clean l 10 and leak-free.
O 11
- oiablo canyon uses a large vacuum pump to draw 12 vacuum and a stream jet to maintain vacuum, so the 13 pump is in standby mode most of the time. It was ;
F O 14 in excellent condition.
15
- The condenser wells were in very good condition l f
16 with no evidence of leaks down in the well area.
O 17 The wells were very clean. There were no ;
18 materials or tools stored down in the wells. !
19
- The intake area, both top-side where the screen j
)
O, 20 drives are and below where the pumps and motors 21 are, was in excellent condition. This area has !
I 22 had a large effort directed at improving its j O 23 condition in the last year. Large areas of 24 concrete have been chipped out and replaced. They i
25 are about complete with this effort and the ,
O 26 results are outstanding. !
The intake screens are being replaced with new i 27
- 28 stainless steel all welded units.
O 128
D ,
1 There are many other specific observations I could O 2 make but I believe the examples provided demonstrate 3 that the equipment at Diablo Canyon is in very good 4 condition. This is one of the very best plants that I O 5 have seen. There is no doubt is my mind that a plant 6 could not be kept in this condition and not have an 7 effective maintenance program.
O 8 Q11 What were your observations regarding the Work !
9 Control System? l 1
O i 10 All
- Work Control System 11 Performance Objective: The control of 12 maintenance work should support the O 13 completion of tasks in a safe, timely, and i 14 efficient manner such that safe and reliable 15 plant operation is optimized.
16 Criteria: ,
C) 17 -
The work control system provides 18 management with an accurate status of 19 maintenance planning and outstanding 20 maintenance work.
21 -
Work planning includes considerations a C) 22 such as material, tool, and manpower ,
23 requirements; interdepartmental 24 coordination; safety considerations; f l
25 radiological protection requirements; 26 and quality control requirements. i i 27 Maintenance history records and NPRDS j C) 28 information are considered where 29 appropriate.
30 -
Advance planning is performed and -
31 routinely updated for scheduled outages.
32 Considerations such as work priority,
) 33 work procedures and instructions, 34 plant / system conditions, length of 35 outage required, prestaging of documents 36 and material, and coordination of 37 support activities are included.
129 l
O 1 Some of my observations follow:
O 2
- Diablo Canyon has a work planning department that 3 has the responsibility of planning all of the work 4 done by the Maintenance Department and scheduling O 5 all of the work performed on site. They do this 6 on a daily basis as well as for outages. Every 7 morning they meet with the representatives of all O 8 operating departments to go over the schedule for 9 the next few days. They meet on Friday afternoon 10 to review the plan for the weekend and again on 0 11 Tuesday to factor in what happened over the 12 weekend. This process is very tightly controlled 13 as a great effort is given to coordinate all the 0 14 activities necessary to accomplish a given job.
15 Jobs brought in late to this process must be 16 important in order to be included. The effort put
() 17 into the efficient planning and ccheduling pays 18 off in reduced manhours and equipment out of 19 service time. To do it well is the mark of a 0 20 strong organization, in my opinion, as it is 71 difficult to bring all different groups and 22 activities together on the dozens of jobs that are O 23 done on any given day.
24
- The maintenance action requests ("ARs") are part 25 of the maintenance equipment history, and the
,0 26 number and age of action requests are tracked by 27 department. The division directors are O 130
4:
1 1 responsible for the control of the backlog and age
) 2 of the action request in their areas.
3
- The excellent overall housekeeping condition of 4 the plant is an indicator of an effective work '
5 control system.
6
- The posting of hot work permits, equipment ,
7 clearances, and scaffold tags are also indications 8 of an effective work control system.
9 I believe that these examples show that the work 10 control system at Diablo Canyon is well conceived.
?
11 Q12 Please address the Conduct of Maintenance at 12 Diablo Canyon.
)
13 A12
- Conduct of Maintenance 1 r 14 Performance Objective: Maintenance should be j 15 conducted in a safe and efficient manner to 16 support plant operation. ,
17 Criteria: i 18 -
Personnel exhibit professionalism and j) 19 competency in performing. assigned tasks that 20 results in quality workmanship.
21 -
Maintenance personnel are attentive.to '
22 identifying and are responsive to correcting l 23 plant deficiencies with a goal of maintaining
[) 24 equipment / systems in an optimum material '
25 condition. t 26 -
Managers and supervisors routinely observe 27 maintenance activities to identify and 28 correct problems and-to ensure adherence to ;
29 station policies and procedures including 3 30 industrial safety and radiation protection.
t 31 This is an important aspect of maintenance 32 performance. The conduct of. maintenance is not as 3 131
(
F i
I 3 )-
l l
1 visible as the plant material condition, but is an j 3 2 important indicator of the way that the results are l 3 obtained. Some examples of my observations on the 4 conduct of maintenance at Diablo Canyon follow:
D 5 .
- Everyone that I observed out in the plant was very 6 conscious of industrial safety work practices. i 7 They all wore their safety equipment and were O 8 careful that all security and fire doors were 9 secured behind them. This showed me that the 10 habit of doing things right is a part of their 3 11 daily work. It has been my experience that it is 12 very difficult to get hundreds of people to do all 13 these small daily activities consistently well.
3 14
- During my tours of the plant, I noted many times 15 that supervisors were out in the field. In two 16 cases, I saw the mechanical department head 3 17 inspecting equipment in the plant at different 18 locations and times. In my discussions with him, 19 he displayed a good knowledge of the plant and the 3 20 equipment in it.
21
- During my tour of the intake area, I was impressed 22 with the area foreman's knowledge of and 3 23 commitment to the intake area. He was obviously 24 very enthusiastic about his work. He readily i
25 answered all of my questions, and in many cases 26 anticipated where I was leading to and volunteered
! 27 much more information. An example of this was on l 28 a question I had about the intake crane. He not i
3- 132
D-1 only answered my questions, but told me what the 2 plans were for future work on the crane. He 3
3 discussed experience they had on the cables, the !
4 plans to replace the cables with stainless, and
) 5 the schedule for non-destructive examination of 6 the hook. The crane had recently been included in 7 a plant-wide crane inspection program. Such
- ) 8 knowledge and enthusiasm is a clear example, in my 9 mind, of a very healthy commitment to the ,
10 effective conduct of maintenance. Such i
) 11 outstanding examples would not exist, I believe, i
12 unless a strong culture exists at the plant to 13 support it. ,
During my tour of the high voltage switchgear
) 14
- l 15 rooms, I noted a significant work activity where !
l' 16 large steel beams were being added to stiffen the l
) 17 walls. This was in response to a seismic upgrade. !
i 18 This work involved cutting, welding, grinding, 19 painting, and chipping in the relatively small l space between the switchgear cabinets and the
) 20 1
21 walls. The area had been very effectively -
22 controlled by the manner in which the scaffolding i had been erected, and clear, yellow plastic was
) 23 24 draped to prevent arc flashes from harming people 25 and dust from escaping fron the work area. This i 1
26 is a good example of effective conduct of
)
27 maintenance.
133
).
W - .- L .- y . v. w.
b i i
I believe these are examples wherein the 1
I j 2 maintenance program is being implemented very well at 3 Diablo Canyon. l 1
g-4 Q13 Did you observe a Preventive Maintenance Program 5 at Diablo Canyon?
t 6 A13
- Preventive Maintenance 7 Performance Objective: Preventive maintenance ,
8 should contribute to optimum performance and 9 reliability of plant systems and equipment.
10 criteria:
11 -
A preventive maintenance program is 12 effectively implemented and includes systems 13 and equipment that affect safe and reliable
[ 14 plant operation.
) 15 - Preventive maintenance, includinq predictive 16 maintenance activities, are performed at 17 appropriate intervals. These intervals 18 maximize equipment availability.
l 19 Considerations such as operational 20 experience, vendor recommendations, 21 engineering analysis, and cost / benefit 22 analysis are used as a basis to establish ,
23 preventive maintenance tasks and intervals.
24 -
Preventive and maintenance activities are ;
25 scheduled and performed within established 26 intervals. Preventive maintenance is waived l
27 or deferred only with management approval.
i l
28 Yes. Some observations related to preventive i
29 maintenance made during my plant tour follow:
30
- The implementation of activities for preventive 31 maintenance are done by the respective maintenance 32 organizations. The development and coordination
) '
33 of specific preventive maintenance activities is 34 the responsibility of the Preventive Maintenance ;
134
)
. . ~ ,_ _ ..
i i
1 Engineering Group. They are responsible for the r
) 2 technical aspect of preventive maintenance 3 activities and for determining the proper 4 frequency of the activities. The Group includes
) 5 representatives from the maintenance disciplines.
6 The Group reviews vendor recommendations, i
7 technical manuals, data provided by Reliability
) 8 Engineering, and equipment history files which .
9 include all action requests for that piece of 10 equipment. This appeared to be a comprehensive 3 11 and effective process and the overall plant 12 equipment condition reinforces this view.
13
- Diablo Canyon is implementing a reliability 3 14 centered maintenance ("RCM") project. This is an ,
15 ongoing effort, the results of which are available 16 plant-wide on their computer-based information
) 17 management system (known as "PIMS"). The basis of 18 why an activity is or is not recommended is 19 documented and can be referred to in the future to 20 support why a certain activity is recommended.
21 Factors included are:
22 -
why it is a critical component 3 23 -
what the RCM analysis results are !
24 -
what the system engineer analysis was ,
I i 25 -
what the vendor recommendations are 26 This RCM project is an effort by the maintenance l 27 organization to improve the performance of the 28 plant equipment. This project appears to receive .
135
9 1 high levels of commitment from management as it G 2 was mentioned by several management-level people 3 in discussions with them. This is an example, in 4 my view, of an organization that is looking for 9 5 ways to constantly improve its performance.
6
- The preventive maintenance program also ties in 7 with the Inservice Inspection / Inservice Testing 3 8 and performance monitoring activities at the 9 plant.
10 It is my conclusion that the preventive 3 11 maintenance program at Diablo Canyon is in place to 12 improve equipment performance.
3 13 Q14 Please address your next criterion, Maintenance 14 Procedures and Documentation.
3 15 A14
- Maintenance Procedures and Docurentation 16 Performance Objective: Maintenance 17 procedures and other work-related documents 18 should provide appropriate directions for 19 work and should be used to ensure that 9 20 maintenance is performed safety and 21 efficiently.
22 Criteria:
23 -
The preparation, review, approval, and 9 24 revision of procedures and other i 25 work-related documents are properly ;
26 controlled. :
I 27 -
Documents used in lieu of procedures 28 (such as excerpts from vendor manuals) 6 29 receive the same review and approval as 30 procedures.
I 31 -
Procedures and other work-related 32 documents such as vendor manuals, ,
I 3 136 ,
t :
l
) l l
I
! 1 I .1 drawings, reference materials, and j
! 2 posted job performance aids used in l
)
i 3
4 support of maintenance are technically accurate and up-to-date.
j l 5 Specific examples I observed wherein this element !
l l 6 is being effectively implemented at Diablo Canyon are:
l 7
- My review of selected procedures satisfied me that l
l 8 procedures are in place for effective control of l
9 maintenance activities. I did not review all '
b However, the procedures I did review i 10 procedures.
l l 11 were comprehensive, clear, and in general what
! 12 would be expected of effective procedures.
)
l 13
- Specific procedures examined include:
4 l
l 14 -
MA1, Maintenance (which is a program overview l
I 15 document and shows how all of the maintenance '
) 16 elements within the plant work together) ,
17 -
OM7, Problem Resolution I
i f
l 18 -
OM7 ID1, Problem Identification and 19 Resolution Action Request 20 - AP C-250, Preventive Maintenance Program ;
! 21 (electrical)
) 22 -
AP C-450, Preventive Maintenance Program 23 (I&C) 24 -
NPAP C-3, Conduct of Plant and Equipment l 25 Tests
.I 26 - AP C-62, Preventive Maintenance Living .
27 Program
)
AP C-352, Surveillance Testing and Inspection 28 -
k b 137 :
D .
1 Q15 Does PGEE compile and evaluate Maintenance 3- 2 Histories? -
3 A15
- Maintenance History O 4 Performance Objective: Maintenance history should 5 be used to support maintenance activities, upgrade i 6 maintenance programs, optimize equipment 7 performance, and improve equipment reliability.
8 Criteria: Maintenance history records are 1 O 9 maintained for systems, equipment, and components l 10 that affect safe and reliable plant operations.
11 Yes. Based on my discussion with maintenance i
12 personnel, reliability engineering personnel, O
13 preventive maintenance engineering, as well as some !
14 system engineering personnel, and upon my procedure 15 review, I believe it is reasonable to conclude that O
16 Diablo Canyon has an effective maintenance history 17 program. Some of my observations are:
18
- All of the action requests for a piece of C
19 equipment are permanently added to the 20 computerized history file. This aids in the 21 determination of equipment performance and i O
evaluation of the effectiveness of preventive l 22 23 maintenance activities.
24
- The reliability centered maintenance project l b information is entered into equipment history.
l 25 l
26
- The trending of equipment performance by the 27 reliability engineering group is an effective use 28 of equipment history.
t l
O 138 )
D 1
- Outage nanagement utilizes historical performance O 2 as a base for current outage planning.
3 Q16 Please address your next criterion, Maintenance O 4 Facilities and Equipment.
5 A16
- Maintenance Facilities and Ecuipment.
O 6 Performance Objective: Facilities and 7 equipment should effectively support the 8 performance of maintenance activities.
9 Criteria:
O 10 -
Maintenance facilities size and 11 arrangement promote the safe and 12 effective completion of work.
13 -
Measuring and test equipment (M&TE) is 14 calibrated and controlled to provide g 15 accuracy and traceability. Out-of-16 tolerance test equipment is removed from 17 service. Plant equipment calibrated 18 with out-of-tolerance test equipment is 19 evaluated in a timely manner for 20 operability, and is recalibrated as e 21 necessary.
22 Diablo Canyon has excellent facilities. The plant 23 as a whole has excellent facilities and the maintenance g 24 facilities are outstanding. During my tour of the 25 plant, I had an opportunity to see most of the 26 maintenance facilities, as well as the training e 27 facilities. Some specific examples follow:
2B
- There is a very good light machine shop and 29 mechanical facility between the units in the e 30 turbine building. This shop was very spacious and 31 very well equipped. There were four lathes, two 1
32 milling machines, and several smaller power tools
@ 139
-)
1 such as drill presses and grinders. It was also O 2 equipped with a nice overhead crane.
3
- The intake structure has its own free-standing 4 workshop area that is almost as large and well-O 5 equipped as the maintenance shop described above.
6 They also have an office and break / lunch building.
7 They have many special tools used in the repair of O 8 intake equipment.
9
- The I&C shop was very spacious and well-equipped 10 for a large number of I&C workers (I think I O 11 counted about 80 benches).
12 o The M&TE calibration shop was the best I have ever 13 seen. PG&E has the capability at Diablo Canyon to O 14 calibrate most equipment used in the plant; the 15 shop acts as a common calibration shop for all 16 M&TE equipment on site including health physics.
O 17 The calibration equipment was all in excellent 18 shape and the storage of calibrated equipment was 19 also excellent. PG&E separates the uncalibrated O. 20 instruments from the calibrated instruments, and 21 the history of individual calibrations is kept in 22 a computer history program. All jobs that an O 23 instrument was used on are kept in this history, 24 so if an instrument is found out of calibration 25 the work that was done with it is immediately ,
8 26 known. The individual that took me through the l
27 shop was quick to point out the many strengths in l l
28 this shop and demonstrated them to me. This is m
'> 140
i i
1 another example of a very knowledgeable person D 2 readily able to demonstrate the effectiveness of 3 the work they do. Highlights pointed out to me l 4 included the optical scanner used to identify both j 5 the M&TE equipment as well as the individual who 6 checks it out. ,
7
- The training facilities were also very good. I ;
D 8 will discuss them more in the element on training. I i
9 In summary, I believe that Diablo Canyon's r
10 facilities would compare favorably to any nuclear D 11 facility in the United States. ,
i i
12 Q17 Please address Maintenance Personnel Knowledge and -
D 13 Performance. !
14 A17
- Maintenance Personnel Knowledae and Performance i 3 15 Performance Objective: Maintenance personnel 16 knowledge and performance should support safe ,t 17 and reliable plant operations.
18 Criteria:
D 19 - Maintenance is performed by or under the ;
20 direct supervision of personnel who have r 21 completed applicable formal 22 qualification associated with the tasks 23 to be performed. ;
) 24 - Maintenance personnel knowledge is 25 evidenced by an appropriate 26 understanding of areas such as the 27 following:
maintenance policies and procedures
) 28 29 a.
- b. general plant layout >
30 c. purpose and importance of 31 plant / systems and equipment ,
32 d. effect of work on plant systems i
141 l
i 9 !
I e. industrial safety, including i 2 hazards associated with work on g 3 specific equipment / systems 4 f. radiological protection and as low ;
5 as reasonably achievable ("ALARA") ;
6 principles 7 g. job-specific work practice 8 h. cleanliness and housekeeping ;
g 9 practices 10 During my inspection of the plant, and in 11 discussions with plant personnel as part of my review 12 of the maintenance program, I had many opportunities to 3
13 observe the knowledge and performance of maintenance 14 personnel. Some of my observations are as follows:
3 15
- My discussion and tour with the Mechanical 16 Department head revealed that he was very 17 knowledgeable about the plant equipment and the
- g 18 programs and processes used to maintain the 19 equipment.
20
- The Electrical Department supervisor, who g 21 discussed Generic Letter 89-10 activities with me, 22 was very knowledgeable about the program and 23 equipment related to motor operated valves.
- p 24
- As I indicated in my comments on the M&TE shop, 25 the people who were associated with M&TE were 26 extremely knowledgeable of that program and how it 25 fits in with the rest of the plant activities.
3 28
- I will discuss this more later, but the training 29 supervisor for Maintenance demonstrated a strong
, 30 knowledge of the training aspect of maintenance, 31 as well as knowledge of how some of the training 142 3
[) I i
1 facilities are used to do maintenance on equipment 3 2 from the plant.
3
- The area foreman for the intake area was extremely 4 knowledgeable about this area. I covered this in 3 5 the element on plant material condition, but it is 6 a very good example of personnel knowledge and 7 bears repeating.
7) 8
- I observed discussions between the foreman and 9 some of the crew members related to work at the 10 intake structure. Based on the comments and 11 d tail f the discussions, I would say that the O
12 journeymen were very knowledgeable about the work 13 being performed.
14 I conclude that this element is being effectively
[)
15 implemented at Diablo Canyon.
16 Q18 Please describe what you observed regarding the 7) 17 Technical Support Organization and Administration 18 Performance Objective, p
19 A18
- Technical Succort Oroanization and Administration 20 Performance Objective: Technical support
[ 21 organization and administration should ensure 9 22 effective implementation and control of j 23 technical support.
24 Criteria:
25 -
The organizational structure is clearly g 26 defined.
27 -
Staffing and resources are sufficient to 28 accomplish assigned tasks.
29 -
Responsibilities and authority for each 30 management, supervisory, and 143 3
l-l 1 professional position are clearly 2 defined and understood.
O 3 -
Interfaces with support groups, I 4 including corporate groups and contract 5 services, are clearly defined and 6 understood.
7 -
Technical support personnel have 8 sufficient expertise regarding plant O 9 systems, components, and operations to 10 effectively investigate and resolve 11 plant problems.
12 My discussions with plant staff and my review of O 13 plant and corporate procedures show that the technical 14 support function is well established at Diablo Canyon.
15 Some specific observations follow:
O 16 . The data trending of equipment performance done by 17 the Reliability Engineering group is analyzed and 18 the results forwarded to preventive maintenance O 19 engineering. This is a good example of technical 20 support.
21
- The tie between the several department O 22 organizations needed to implement IsI, IsT, and 23 equipment qualification is generally clear and 24 well understood by the individuals involved.
O 25 . The motor operated valve activities are covered by 26 a program plan document that formalizes the plant
- 27 efforts to comply with GL 89-10.
l C- 28
- The corporate level aging document mentioned 29 earlier is also an example of effective technical
=
30 support.
31 I believe these examples show that this element is 32 clearly in place and is being implemented well.
[) 144 l
O 1 Q19 Please address the Surveillance Testing Programs ,
O 2 A19 e Surveillance Testina Procrams 3 Performance Objective: Surveillance O 4 inspection and testing activities should 5 provide assurance that equipment needed for 6 safe and reliable plant operation will 7 perform within required limits. ,
8 Criteria:
O 9 - Administrative systems and controls 10 ensure timely completion and review of 11 required surveillances.
12 -
Surveillance testing programs result in 13 a high degree of reliability of 14 equipment needed for safe and reliable 9 15 plant operations.
16 - Procedures used for surveillance testing .
17 contain. sufficient detail to ensure safe 18 plant operation during testing and 19 provide for consistent test performance O 20 and accurate results. Procedures ,
21 simulate, as near as practical, the 22 actual conditions under which the system 23 must operate on demand.
24 In assessing the effectiveness of this element, I O interviewed a number of plant personnel and made 25 26 specific observations of equipment and work being 27 performed:
O The 28
- The IST program appears to be well-defined.
29 supervisor of this program was very knowledgeable 30 about the requirements of IST. He pointed out to O me that they have a program document that clearly 31 32 states the relationships of the many activities 33 involved in completing the IST program 9 34 requirements. We discussed in detail how this ;
35 program relates to check valve inspections ,
36 performed by Mechanical Maintenance. This area is
.O 145
l h-
~1 often one that is not well integrated and leaves D 2 room for some problems. I believe that Diablo 3 Canyon has done an adequate job of ensuring good P
4 compliance of these two programs.
O 5
- The implementation of the ISI program appears to 6 be very good.
7
- The implementation of environmental qualification O 8 by the different maintenance departments is also 9 very good.
10
- The implementation of the motor-operated valve O 11 program is very strong. As I mentioned earlier, ,
I 12 PG&E has a program plan document that shows the 13 relationship of the different organizations that O 14 must coordinate well in the compliance to 15 GL 89-10. ,
O 16 Q20 Does PGEE address Performance Monitoring?
I 17 A20
- Plant Performance Monitorina
- b. 18 Performance Objective: Performance f 19 monitoring activities should optimize 20 plant reliability and efficiency.
l 21 Criteria: >
0 22 -
Programs are implemented to 23 routinely monitor, collect, trend, 24 and analyze performance data 25 (including thermal, hydraulic, 26 electrical, acoustical, and 27 mechanical data) for equipment, 28 systems and components important to 29 plant reliability and efficiency.
- 30 -
Approved procedures or guidelines 31 and knowledgeable personnel are 32 used to conduct performance -
146 l
I D
1 monitoring functions. Tests are l 2 conducted consistently to aid in 3 analyzing results.
3 4 Discussions with plant staff and inspection of !
5 equipment provided me with the following examples for g 6 this element: ,
7
- The gathering, trending, and analyzing of data i
8 done by Reliability Engineering is a good example j t
9 of this element.
9 10
- The RCM/ preventive maintenance activities also support plant performance monitoring. !
11 12
3 13 performance monitoring very well. !
14
- The actual valve testing of motor operated valves 1
15 is a part of performance monitoring.
3 16
- I believe that NRC SALP assessments, while not a 17 direct measure of plant performance, are a very t
18 good indication of the plant performance. A 3
19 review of the most recent SALP for Diablo Canyon .
20 satisfied me that PG&E is performing its licensed 3 21 responsibilities very well.
22 Q21 What were your observations regarding Maintenance 23 Personnel Training and Qualification?
3 l
l 24 A21
- Maintenance Personnel Trainina and Oualification Performance Objective: The maintenance
) 25 26 personnel training and qualification program 27 should develop and improve the knowledge and 28 skills necessary to perform assigned job 29 functions.
147 3
D 1 1 Criteria: Programs are established and 2 implemented for initial and continuing
) 3 training.
4 A significant contributor to maintenance strength 5 is training. A key indicator to the health of this
) 6 relationship is the attitude that maintenance has for 7 training and that training has for maintenance. I 8 observed a number of things that show me Diablo Canyon
- 3) 9 has a very healthy relationship: l 10
- I got very clear feedback from the maintenance l
11 manager that he is a strong supporter of training.
12 I also heard from at least two of his division 13 directors that he is a strong supporter of 14 training.
) 15
- The maintenance training supervisor attends the r
16 Maintenance Department's morning meeting each day 17 instead of his own training division morning 18 meeting.
19 e on a tour of the maintenance training facilities, 20 it was clear that training was well supported with 3 21 equipment and mock-ups. Some specific examples 22 are:
23 - The reactor coolant pump / seal mock-up was 3 24 about as realistic as it could be. It is 25 full-scale and is made of the actual 26 materials. This provides for a very ]l i
) 27 realistic training aid. Reactor coolant pump l
28 seal performance and change out of seals is a i
) 148
l 1 very important part of plant reliability, !
1
[ 2 outage activities, and radiation exposure.
l 3 -
They have a full-sized, real emergency diesel i
4 engine in training.
) 5 -
The electrical training area has a wide ,
l 6 variety of motor operated valve operators. .;
7 This would provide very important training in b 8 the proper maintenance of MOVs. .j 9 -
The I&C Lab was exceptional in my experience.
l 10 They have one channel of the reactor 11 protection system with all of the components, 12 including actual control rod drive motors and-13 drive shafts and rod position indication. l j !
le This facility is so realistic that they bring l t i
15 electronic circuit boards from the plant and ;
16 troubleshoot them on the training aid. [
t r b' 17 -
There were classes going on in all three ;
18 disciplines during my tour.
19 - About one-third of the training staff go to !
?
t 20 other plant departments during-refuelings to
[ ,
21 perform outage-related work. This is an 22 excellent way to keep the training staff'up-
[
23 to-date on plant activities and to strengthen j I
24 the personal ties.
i 25 Q22 Does PGEE effectively use industry reliability
?
26 data? ;
L t
) 149
l- i t
t 1 A22
- Nuclear Plant Reliability Data System ("NPRDS")
) 2 Performance Objective: The NPRDS should be l 3 effectively used to improve equipment 4 reliability and to report component failure !
5 information to the industry. ;
6 Criteria: NPRDS engineering data is
) 7 maintained up-to-date and in accordance with 8 program guidance. i 9 Discussions with plant personnel indicated that 10 the use of the NRPDS is effective. The RCM project 11 utilizes this information. The system engineers seem to .
12 be using it as well. The Preventive Maintenance 13 Engineering group utilizes NPRDS data in determining 14 the recommendations en preventive maintenance 15 activities.
)
16 Q23 How does PGEE address the element of maintenance 17 related to operating experience reviews?
)
18 A23
- In-House Operatina Experience Review 19 Performance Objective: In-house operating 20 experience should be evaluated, and 21 appropriate actions should be undertaken to 3 22 improve safety and reliability. ;
23 Criteria:
24 -
In-house events are screened for 25 significance and prioritized for 26 evaluation.
) 27 -
Rigorous investigation is performed on 28 significant in-house events to determine i 29 root causes, generic implications, and 30 necessary corrective actions to prevent !
31 recurrence.
32 Several examples of effective use of in-house 33 operating experience were evident to me:
i
)
t.
150 i l
i lO 1
- The plant-wide use of computerized action request O 2 ("AR") forms seems to strengthen the availability 3 of in-house information. All AR forms are 4 trackable via computer, so any issue can be I) 5 tracked to see if it is being responded to 6 properly.
7
- The AR is a permanent part of equipment -
() 8 maintenance history and figures in on the schedule 9 of work to be done and the tracking and trending 10 of repetitive failures.
O 11
- An interesting aspect to ARs for me is that they 12 can be for any type of problem. They do not have 13 to be for equipment deficiencies, but can be for O 14 program and process problems. This allows for a 15 very wide range of issues to be identified. This 16 seems to me to be an indication of an organization O 17 that is not afraid of criticism and change for the 18 better.
19
- While it is not clear who has accountability for O 20 root cause analysis (a situation also true at 21 other facilities), it seems to be because 22 everybody has accountability for it. Every time I
.O 23 asked, "Who had accountability for root cause?,"
24 the answer was, "I do," or "we do," or "here's how 25 it works," with a ready explanation. Significant
- t O 26 problems are addressed with an officially formed 27 team dedicated to whatever the problem is.
l I
lO 151 )
i
-j O
1 I believe that Diablo Canyon has a very effective 2 use of in-house experience.
O i
3 Q24 Did you observe any system for Industry Operating 4 Experience review?
O ,
5 A24
- Industry Operating Experience Review 6 Performance Objective: Significant industry 0 7 operating experiences should be evaluated, 8 and appropriate actions should be undertaken ,
9 to improve safety and reliability.
10 Criteria:
O 11 -
A comprehensive evaluation is performed 12 on applicable, significant industry r 13 operating experience, and appropriate ,
14 corrective action is completed in a 15 timely manner.
16 -
Sources of significant industry O 17 experience information reviewed for 18 applicability include the following:
19 a. INPO Significant Operating 20 Experiences Reports ("SOERs")
21 b. Significant Event Reports ("SERs")
,O 22 c. Significant by Others ("SO")
23 notifications 24 d. INPO Significant Event 25 Notifications ("SENs")
26 e. NRC letters, bulletins, and 27 information notices O. 28 f. Supplier and architect / engineer 29 reports 30 Most of the examples I provided for in-house 31 operating experience also apply to industry experience.
g 32 However, several additional observations are warranted:
33
- The success that Diablo Canyon has had with ,
O 34 respe t t the MOV/GL 89-10 issue is an indication 35 that they are effectively in touch with the rest ,
36 of the industry.
152 O ,
O 1
- Changes to the IST program reflect a strong tie O 2 with industry issues.
3 e The check valve program is one that has 4 effectively responded to industry activities.
5 Diablo Canyon performs a self-evaluation prior to
()
- 6 their INPO evaluation. They utilize the help of 7 peer evaluators from other plants to do this.
() 8 This evaluation appears to be more rigorous to me 9 than the actual INPO evaluation. They have about 10 28 people working for three weeks looking at all
() 11 areas. From the number of findings and the detail 12 involved, it is clear to me that the intent is to 13 find the weaknesses themselves and to strengthen
() 14 the performance from self-identified issues. This 15 is an indication of a very healthy organization.
) 16 Q25 What do you conclude overall from your review?
l l 17 A25 Based upon my review, I am satisfied that Diablo
() 18 Canyon does in fact have a comprehensive and effective 19 maintenance program. While there may be individual
! 20 deficiencies in the implementation of the programs, the t) 21 overall strength of the programs provides for 22 corrective action and continued improvement.
l 9
1 3 153
D 1 Q26 In raising Contention I, MFP cited a number of 3 2 incidents and reports that they believe show the Diablo 3 Canyon maintenance program to be less than adequate, or 4 at least less than adequately implemented. Do these 5 citations affect your view of the Diablo Canyon 6 program?
) 7 A26 I have not reviewed all of these incidents and 8 citations in detail. However, I am familiar with the 9 types of incidents and reports referenced. These do 10 not change my overall conclusions.
11 From my experience., it is unreasonable to expect, 12 given the complexity of a commercial nuclear power
) 13 plant, that there will not occasionally arise a 14 confluence of events that creates an unsatisfactory 15 condition. This is not to minimize the role that O 16 nuclear safety must play in operating a nuclear power j 17 plant. Nuclear safety nust be paramount. Eliminating 18 safety problems is of the utmost importance. It must 19 be the first consideration given to any situation. It 20 is extremely important that great initiative is given 21 to anticipating and preventing safety problems. In my 3 22 opinion, a healthy organization is one that is self-23 examining by nature, does not obscure the facts of a 24 situation, is ready to take responsibility for
) 25 problems, and initiates quick, effective corrective 26 action. During my observations at Diablo Canyon, I 27 found many instances to show that Diablo Canyon is such 3 154
r i O r r
1 a healthy organization and found no condition, word, or O 2 deed to suggest otherwise. !
3 Being alert to potential problems, ensuring that 4 problems are identified, making sure that problems are i O 5 entered on the proper tracking list, and taking timely, 6 effective corrective action on those problems are all i
i 7 signs of a healthy environment. Indeed, to suggest
() 8 that the identification of a problem was a weakness t
9 would reduce the probability of correcting it.
l
- O 10 Q27 MFP also speculates in their contention that PGEE ,
11 is guilty of putting off necessary maintenance to keep 12 the plant running and to maximize profits. Did you see
() 13 any evidence of this occurring at Diablo Canyon?
p O 14 A27 No. Furthermore, the speculation seems to me to c ,
15 be completely unrealistic. It assumes that you can run 16 a nuclear power plant just because you want to. If all f l '
l O 17 it took was a high level manager saying "you are going ,
\
18 to run," all nuclear power plants would have 100 i 19 percent capacity factors. It is not that easy! It is i
i O 20 very difficult to keep the equipment running reliably. ]
l 21 It takes a strong commitment at a high level, as well l
22 as a great deal of effort and experience. It has been e
O 23 my observation that only a very strong organization 24 with a resolute commitment to success can keep a 0 155 L _- _
i l i l 1 nuclear power plant running at a high capacity factor 2 over a long time.
3 Being able to maintain a high capacity factor l .
4 demonstrates in my mind the existence of strong D 5 programs. Once you have an organization that does 6 things well in one area, they tend to try to do things 7 well in all areas. I believe that the existence of a 8 consistently high capacity factor is the mark of a 9 well-run and well-maintained plant and, therefore, the 10 mark of a safely-run plant. ,
D 11 Q28 Do you have any additional observations?
D 12 A28 Let me make some final general observations. One l
13 of the examples that I have referred to in my testimony i
14 is the scaffold tagging program. It is my direct D 15 observation that this is an effective program at Diablo 16 Canyon. I believe that this observation has more i
17 significance than just a good scaffold program.
D 18 Some factors relating to scaffold tagging follow: ,
19
- It is relatively low on the priority list of 20 important things done at a nuclear power plant.
D- If you were going to cut corners, it would be a 21 *
- 22 place to do it.
L
! 23
- It takes a while, in my experience, for an I
24 organization to recognize that controlling l 25 scaffold tagging is necessary. ,
l
)
l 156 l^ t I i
D 1
- It is a fairly cumbersome process, as there are at 2 least five areas that are important and need to be 3 addressed in the process:
4 -
Fire (as a transient fire load)
C 5 -
Chemical (as a hydrogen release agent in 6 containment) 7 -
Seismic (as a missile or attachment to C 8 adjacent equipment) 9 -
Ventilation (flow blockage to equipment) 10 -
Industrial safety (slips, falls, dropped C 11 objects, and blocked exits) 12 e It is difficult to control as many different 13 people build and use scaffolds in many areas of C 14 the plant. In most cases, there is nothing to 15 physically prevent them from building and using a 16 scaffold.
17
- It takes a while for an organization to figure out 18 how to effectively control scaffold use.
19 e It takes a while longer, and some considerable 20 effort, to make it work.
21 There are several conclusions that I believe can l
22 rightfully be drawn from having an effective scaffold h 23 control program:
24
- You are not cutting corners. If you are doing 25 something like this well, you are probably also O 26 doing more important work well too.
27
- You have been paying attention to problems in your 28 organization.
O 157
I I i
I l
1
- You can develop and administer a relatively b 2 cumbersome process.
3 e Your people can tell that you think it is 4 important and therefore other things are important Y 5 too.
6 In this light, my observations of the scaffolding 7 program at Diablo Canyon increase my confidence b 8 regarding the overall effectiveness of the maintenance 9 program and the organization. !
10 I believe that the same general conclusion can be 0 11 drawn from other programs by my direct observation:
12
- Equipment clearances 13
- Equipment tagging O 14 . not work permits 15
- Attention to personnel safety 16
- Attention to fire and security doors 17 From my observations, PG&E appears to be effective 18 in these areas as well.
19 Finally, I want to reiterate that the plant is in
) 20 excellent overall condition. The operating performance 21 at Diablo Canyon has also been among the best in the l 22 nation. The SALP ratings are just about as good as 23 they can be. Maintenance / Surveillance was a SALP 1 for l 24 the most recent period, and I know that this is not an
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25 easy thing to do. This plant is also on the NRC's 26 "Best Plants" list, which also is a very difficult 27 achievement. I believe these general assessments alone l i
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1 demonstrate that the maintenance programs are in place 2 and that they are being implemented very well. ,
i 3 Q29 Does this conclude your testimony? {
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4 A29 Yes.
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l August 2, 1993 b
UNITED STATES OF AMERICA j NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD ,
d) '
4 I
In the Matter of: ) Docket Nos. 50-275-OLA
) - 5 0 -3 2 3-OLA Pacific Gas and Electric Company ) -
O ) (Construction Period (Diablo Canyon Nuclear Power ) Recovery)
Plant, Units 1 and 2) )
l
) l TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY O ARPRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE !
O PART 3: Bryant W. Giffin, David B. Miklush IO
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1 August 2, 1993
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2 UNITED STATES OF AMERICA ;
3 NUCLEAR REGULATORY COMMISSION j t
4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
5 i 6 In the Matter of: ) Docket Nos. 50-275-OLA ;
7 ) 50-323-OLA 8 Pacific Gas and Electric Company ) .
9 ) (Construction Period :
) 10 (Diablo Canyon Nuclear Power ) Recovery) 11 Plant, Units 1 and 2) )
12 )
13 TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY
) 14 ADDRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE 15 I. INTRODUCTION
)
16 Q1 Please state your name, affiliation, qualifications 17 and current job responsibilities.
)
18 A1 (Giffin) My name is Bryant W. Giffin. I am the 19 Manager of Maintenance Services for Pacific Gas and 20 Electric Company ("PG&E") at the Diablo Canyon Power Plant 21 ("DCPP"). I am responsible for all maintenance and outage 22 activities at DCPP. I have more than 25 years experience 23 working in the nuclear industry; 12 years with PG&E and
)
24 over 13 years as an officer in the United States Navy's 25 nuclear power program. A summary of my professional 26 qualifications and experience is provided in Exhibit 1.
27 (Miklush) My name is David B. Miklush. I am the 28 Manager of Operational Services for PG&E at DCPP. I am
) 160
O 1 responsible for Operations, Chemistry, and Environmental J 2 Engineering. I have more than 20 years experience in the 3 nuclear industry, and have been working at DCPP for more 4 than 15 years. I maintained an SRO license at DCPP from J 5 1982 to 1988. A sutrary of my qualifications and 6 experience is provided in Exhibit 13.
J 7 II. PERFORMANCE EVALUATION OF DCPP'S MAINTENANCE AND 8 SURVEILLANCE PROGRAMS 3 9 Introduction 10 Q2 What performance measures does PG&E use to determine 11 the effec'.iveness of the DCPP maintenance and surveillance 3 12 programs? How have the DCPP maintenance and surveillance 13 programs performed under each of these performance 14 measures over the last two SALP periods?
O 15 A2 (All) One of PG&E's most important corporate goals is 16 to " Operate the Diablo Canyon Nuclear Power Plant at the D 17 highest level of safety, reliability, and performance."
13 In order to meet this goal, the Nuclear Power Generation 19 Business Unit ("NPG"), which includes DCPP, has defined O 20 its goals and objectives in Program Directive OM2, 21 " Management Goals and Objectives." NPG's goals are 22 prepared annually and are formulated in the following e 23 categories:
24 e Safety and Quality of Operations 25
- Energy Production 0 161
O 1
- Cost Management O 2 . continuous Improvement 3 Maintenance activities have a large impact on how NPG 4 performs in meeting its goals. Maintenance effectiveness O 5 is not merely a function of day-to-day, individual 6 preventive and corrective maintenance tasks. A well 7 maintained plant exhibits long, uneventful runs between O 8 refueling outages with plant generation at a high percentage of generating capacity. Uneventful runs r
9 mean 10 minimal plant transients. High operating capacity factors 0 11 (percentage of maximum generation capability actually 12 delivered between outages) result in efficient delivery of .
13 energy to customers. Thus, in a programmatic sense, the O 14 effectiveness of maintenance is demonstrated by the 15 overall safe and reliable operation of the plant over 16 time.
O 17 With this in mind, NPG uses three broad performance 18 measures to evaluate the comprehensiveness and 2 19 effectiveness of the DCPP maintenance program in meeting ;
O, 20 PG&E's corporate goal for DCPP. These performance 21 measures include:
22 o Plant Operating Performance, ,
O 23 e Maintenance Services Department Goals and Objectives, !
24 and i
25
- Regulatory Performance !
O 26 Each of these performance measures will be discussed 27 below.
O 162
9 1 Plant Operatinc Performance 9 2 Q3 How does plant operating performance demonstrate the 3 effectiveness of DCPP's maintenance and surveillance 4 Programs?
O t
5 A3 (All) Experience has demonstrated that nuclear plants 6 with consistently high capacity factors, long continuous
- 7 operation, and short refueling outages are also the best 8 maintained plants. This is because reliable and 9 continuous operation at high capacity factors between S 10 scheduled refueling outages and across many operating 11 cycles simply cannot be sustained unless plant equipment 12 and facilities are maintained in a superb condition. In O 13 turn, reliable, event-free operation results in fewer 14 challenges to plant safety systems and less wear and tear 15 on safety equipment. Thus, good maintenance practices are O 16 not only evidenced by reliable plant operation, but 17 reliable plant operation itself reduces the wear and tear 18 that can lead to increased maintenance.
O 19 Diablo Canyon's operating history provides ample 20 evidence of the effectiveness of the DCPP maintenance 21 program. The combined lifetime capacity factor for both O 22 Diablo Canyon units since Unit 1 began commercial 23 generation in 1985 through July 1993 is 78 percent 24 including refuelings. In this respect, the Diablo Canyon
- 25 units are among the best operating nuclear units in the 26 nation. DCPP's capacity factor has steadily improved from 27 year to year over its operating history. Over the last 163
O 1 three years (1990-1992), the combined capacity factor for O 2 both Diablo units has been 85 percent. See Exhibit 14.
3 Among all nuclear plants worldwide, Diablo's performance !
4 consistently ranks among the best.
O 5 Q4 What measures are there of operating performance?
Q 6 A4 (All) A plant's overall operating performance is 7 defined by two components: (1) capacity factor between 8 refuelings, and (2) duration of its refueling outages. A g 9 commonly used measure of performance for the first is 10 " operating capacity factor," the percent of maximum 11 generation that is achieved from the time the breakers are g 12 closed at the end of a refueling outage, to the time power 13 production ends when the breakers are opened again at the 14 next refueling outage. !
O 15 QS Please address DCPP's performance in terms of capacity 16 factor.
O 17 AS (All) For a plant like Diablo Canyon with constant ,
18 tempe.rature ocean cooling, a 100 percent operating O 19 capacity factor is not attainable since the plant goes 20 through a period of power ascension testing coming out of [
21 an outage. Moreover, ocean-cooled plants like DCPP need 22 to conduct periodic scraping of the intake tunnels at i
.O 23 reduced power to remove barnacles and other sea life. ;
I O 164 ,
1
O 1 Diablo Canyon's operating capacity factor has averaged 3 2 95 percent over the last three years, has been over 92 3 percent each year over the last five years including 1993 4 year-to-date, and reached over 96 percent in 1992. See O 5 Exhibit 15. This is the best among similar plants in the 6 United States, and consistently among the very best of all 7 nuclear plants worldwide. In 1991, Unit 2 set a world D 8 record for length of continuous run, 481 days at power, 9 breaker-to-breaker between refueling outages. World class 10 performance like this is not achieved without a superb O 11 maintenance program. This is a further indication that 12 PG&E's preventive maintenance program has been effective 13 in precluding equipment failure and forced outages across O 14 multiple operating cycles.
15 Another, and perhaps more subtle, reason that our 16 excellent plant operating capacity factor is indicative of D 17 an effective maintenance program has to do with the 18 numerous Technical Specification surveillance tests which 19 are required to be performed during operation on the O 20 thousands of SSCs important to safety. All of these 21 surveillances have Technical Specification mandated 22 " ACTION STATEMENTS" which are required to be initiated if 9 23 an SSC is not operable because the surveillance result is 24 not satisfactory in every respect. This includes all )
25 supporting equipment, administrative requirements and test 8 26 results. The great majority of these " ACTION STATEMENTS" 27 require the plant to begin reducing power toward shutdown 28 if the problem is not fully resolved within a limited 8 165
O 1 period of time. To achieve a consistently high plant O operating capacity factor, the plant's maintenance program 2
3 must maintain and test SSCs in a well coordinated manner 4 at a very high level of performance to avoid " ACTION O 5 STATEMENT" power reduction or unit shut-down.
6 Q6 How is DCPP's performance of maintenance reflected in O 7 the duration of planned refueling outages?
8 A6 (Giffin) Certain periodic maintenance activities are O 9 performed every refueling outage, and the duration of the 10 outage (along with event-free post-outage operation) is an 11 indicator of how efficiently outage maintenance activities 12 are planned, scheduled and performed by a skilled, 13 highly-trained workforce. Since DCPP began operation in 14 1985, refueling outage times have been cut in half. See 15 Exhibit 16. DCPP refueling cutages are now consistently 16 among the shortest in the industry. In 1992, even with 17 the additional scope of DCPP's ten-year ISI anu reactor O 18 vessel steam generator shotpeening work, the Unit 1 fifth 19 refueling outage took only 59 days, a record for the unit.
20 In 1993, Unit 2 completed a similar scope refueling outage O This performance set a United States record 21 in 57 days.
22 for Westinghouse and Combustion Engineering plants 23 undergoing a 10-year ISI refueling outage. These short 4 24 outage durations reflect the comprehensive training of 25 DCPP maintenance personnel and thorough planning of 26 maintenance activities necessary to perform outage 9 166
) ,
1 maintenance tasks. The quality of the work performed has 3 2 a been demonstrated by the reliable, event-free operation ,
3 of both units since the outages were completed. Once 4 again, DCPP's consistency and quality are the hallmarks of J 5 a top-flight maintenance program. '
6 Maintenance Goals and Obiectives
) 7 Q7 How do DCPP's internal goals and objectives measure ,
8 the effectiveness of the DCPP maintenance program?
J 9 A7 (Giffin) As discussed previously, PG&E uses a 10 formalized set of goals and objectives for managing DCPP.
11 The objectives established for the Maintenance Services
) 12 Department are a subset of DCPP's overall goals, and 13 provide constant feedback to the workforce and to senior 14 management concerning overall effectiveness in the key
) 15 areas of maintenance activities. Each of these internal 16 indicators provides a different view of the process to a 17 specific organizational level or group. Some are i J 18 quantitative reports generated from PIMS records and some 19 are based on qualitative assessments. A comprehensive ,
20 analysis of maintenance performance is dependent on a D 21 review of all available data at any given point in time. !
22 The primary Maintenance Services Department objectives 23 include:
) 24
- Industrial Safety 25
- Radiation Exposure 26
- Personnel Contamination
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- Personnel Error Reduction O 2
- Refueling outages 3
- Corrective Maintenance Backlog ,
4
- Overdue Preventive Maintenance Items O 5
- Ratio of Preventive Maintenance to Total Maintenance 6 DCPP's performance in each area is discussed below.
f Q 7 1. Industrial Safety 8 Q8 How has DCPP performed relative to its industrial 9 safety goal in the maintenance area?
O 10 A8 (All) DCPP's industrial safety accident rate overall 11 has been 0.22 per 200,000 man-hours, substantially better O 12 than the U.S. nuclear industry median of 0.77 and the U.S.
13 best quartile of 0.27 per 200,000 man-hours. A culture i 14 which stresses attention to detail to produce safe working
,O 15 conditions also promotes professionalism and quality 16 overall.
17 The Maintenance Services Department experienced four ,
O. 18 industrial injuries in 1990, zero in 1991, and one in s 19 1992. So far in 1993, there has been only one injury. [
l 20 Even though Maintenance Services has not met its stated O 21 objective of reducing industrial accidents to zero, the 22 Department is performing at a rate better than NPG's goal t
23 of 0.5 per 200,000 manhours. DCPP's performance in this 1 O 24 area has usually been the best safety record in PG&E.
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J 1 2. Radiation Exposure 3 2 Q9 How has DCPP performed relative to its goal for 3 minimizing radiation exposure in the maintenance area?
3 4 A9 (All) PG&E establishes a goal each year designed to 5 minimize the radiation exposure that employees receive.
6 It has always been PG&E's philosophy to keep the dose 3 7 as-low-as-reasonably-achievable ("ALARA").
8 DCPP maintenance personnel achieved a collective dose 9 of 95 manrem in 1992 compared to our internal objective of D 10 96 manrem. This year our objective for similar scope of 11 work is 91 manrem and we are projecting a dose of about 60 12 manrem. During 1992, the majority of the dose, as in any 3 13 year, was received during the refueling outage. The dose 14 received in the 1992 Unit 1 outage was higher than we had 15 targeted. Consequently, in order to reduce the dose for D 16 1993, we formed a High Impact Team (" HIT") with the 17 responsibility to understand the reasons for the increased 18 dose and to implement improvements.
3 19 Due in a large part to these HIT team efforts, we were 20 able to reduce the dose received in the 1993 Unit 2 fifth 21 refueling outage by about 30 percent. This reduction is 3 22 also being seen in Maintenance Services personnel dose.
23 The improvement was an outstanding effort. We did as much 24 work in the 1993 refueling outage with fewer people and 4 25 received less exposure.
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i 1 3. Personnel Contamination
) 2 Q10 How has DCPP performed relative to its goal for 3 minimizing personnel contamination in the maintenance 4 area? ,
D 5 A10 (All) In 1991 and 1992, Maintenance Services i
6 Department personnel incurred 107 and 87 personnel
) 7 radioactive material contaminations, respectively. This 8 year, we estimate that the number will be about 55.
9 Again, the improvements that we have experienced in this
) 10 area have been due to the efforts of the plant staff. By 11 carefully planning work, taking time and being observant, j
12 it is possible to reduce the number of instances of
) 13 contaminations to almost zero. PG&E has expended 14 substantial effort in developing and providing a training e
15 program to provide workers with the knowledge necessary to
) 16 reduce the potential for contamination.
17 4. Personnel Error Reduction
) 18 Q11 How has DCPP performed on its personnel error 19 reduction goal in the maintenance area? I
) 20 All (All) We have established goals to reduce the number
- 21 of personnel errors which cause a plant transient or a 22 formal report to a regulatory agency. The goal for 1993
) 23 is 3 for each 1,000,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> worked at the plant. So far 24 this year there have been two instances of personnel
) 170
!O 1 errors which fall into this category (one in maintenance) !
O 2 which equals approximately 1.7 per 1,000,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> worked.
3 We have seen a continuous decrease in the number of 4 personnel errors since 1989. This is based on a reduction 5 in such events identified in LERs, Quality Evaluations,
-O 6 and Nonconformance Reports. We will continue to devote 7 attention to this area with the ultimate goal of reducing 8 personnel errors to zero.
O .
9 5. Refuelino Outaces g 10 Q12 How has DCPP performed on its goals for refueling ;
11 outages in the maintenance area? .
- g 12 A12 (Giffin) During a refueling outage, over a million man ;
l 13 hours of work and several thousand work activities are scheduled. Our performance in outage duration, radiation 14 g 15 exposure, and accident rate during refueling outages has I 16 continued to improve. As discussed above, a key indicator 17 related to outage performance is the low level of I
g 18 equipment problems experienced during subsequent 19 operation. The duration of outages at DCPP has been i i
20 reduced by almost 50 percent in the past five years. In-addition, the fact that we were able to reduce the ;
21 0
22 radiation dose by 30 percent in the recent Unit 2 fifth ,
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23 refueling outage is outstanding
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l l 1 6. Corrective Maintenance Backloc
) 2 Q13 How has DCPP performed on its goal for corrective 3 maintenance backlog?
) 4 A13 (Giffin) PG&E tracks the number of outstanding 5 corrective maintenance work requests at DCPP. The 6 objective of this goal is to provide an overall indication 3 7 of the condition of the maintenance program.
8 DCPP has consistently improved its performance on 9 corrective maintenance backlog over the last five years,
) 10 even while DCPP management has continuously raised the 11 " stretch" goal set for this indicator. As Exhibit 17 12 indicates, the corrective maintenance backlog has been
) 13 reduced from about 1,200 items prior to 1989 to about 600 14 items as of the end of July, 1993. Since DCPP performs 15 over 7,000 corrective maintenance work orders on an annual
) 16 basis, this backlog indicator demonstrates an excellent 17 and continuously improving trend in overall backlog I 18 management.
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l 19 7. Overdue Preventive Maintenance Items !
20 Q14 How has DCPP performed relative to its goal for
) 21 minimizing overdue preventive maintenance items? l
> 22 A14 (Giffin) This indicator provides an overall measure 23 of the timeliness of DCPP's Preventive Maintenance ("PM")
24 program. We monitor the number of PM tasks overdue. This 25 number has continued to be reduced because of an 172 1
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1 aggressive effort by the maintenance engineers and O 2 craftsmen. See Exhibit 18. What is important is the 3 overall trend, which at DCPP has been positive, i.e.,
4 decreasing. Again, the curve has a consistently O 5 decreasing number of overdue PM tasks for the past five 6 years. At the beginning of 1992 there were 137 overdue PM 7 tasks; this was reduced to 56 by the end of the year, and O 8 PG&E's goal is to reduce that number to 35 by the end of 9 this year. >
l O 10 8. Ratio of Preventive Maintenance to Total 11 Maintenance 12 Q15 How has DCPP performed relative to its goal for the i
O 13 ratio of preventive maintenance to total maintenance?
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14 A15 (Giffin) The purpose of the ratio of preventive to O 15 total maintenance indicator is to monitor progress in 16 achieving and maintaining a proactive maintenance program, i
17 DCPP's trend has been steady and positive, and we have 1 O 18 achieved the long-term objective of a 60 percent ratio of l
19 preventive to corrective maintenance over the past three !
1 j
20- years. This year we have averaged about 68 percent. j
- O 21 Oualitative Evaluations and Self-Asses, ents l
22 Q16 Please describe the self-evaluations performed by DCPP O 23 prior to INPO evaluations.
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1 A16 (Giffin) DCPP conducts a performance-based review of
) 2 the maintenance program once every INPO evaluation period. l 3 This self-evaluation provides for a qualitative assessment 4 of maintenance effectiveness, rather than a quantitative 5 5 score as is generated from the INPO evaluation itself. A 6 final report (Summary of Findings) is produced at the 7 conclusion of each assessment period. This report lists
) 8 the assessment group's findings (both strengths and 9 weaknesses) and recommendations for improvement. This 10 program is viewed as one means of assessing the
) 11 administrative structure, programmatic controls and 12 working environment in the maintenance organization. PG&E :
13 uses the report as a tool for continuously identifying .
) 14 areas for further improvement in our maintenance and other 15 programs.
16 The most recent two self-evaluations were conducted in -
l
) 17 1990 and 1993. In the maintenance and surveillance areas, t 18 as would be expected in these types of self-critical 19 assessments, the evaluations concluded that there were
) .-20 areas which could be improved. Action plans were prepared j 21 and implemented for the 1990 evaluation, and they are !
22 being prepared and implemented for the draft evaluation )
23 which was distributed in preliminary form in July 1993.
24 (The 1993 evaluation will be finalized prior to the next j 25 INPO evaluation in October 1993.) PG&E management 26 requires that each finding in these self-evaluations be
! 27 formally tracked, responded to, and corrected. In the 1
j 28 case of the specific findings of the 1990 and 1993 self-b 174
9 1 evaluations, the overall corrective actions that need to 3 2 be addressed are to improve maintenance personnel's 3 attention to detail on routine plant evolutions. The 4 Maintenance Services Department is responding to these J 5 issues by using continuous improvement techniques with 6 employee involvement. Even though areas for improvement 7 were identified in these self-assessments, there was no J Ei indication of any programmatic breakdown in maintenance.
9 Maintenance Ouality Assessment 3 10 Q17 How has PG&E's Quality Assurance program assessed the 11 DCPP maintenance program?
3 12 A17 (Giffin) Maintenance Quality Assessments ("MQA"),
13 conducted under the DCPP Quality Assurance Program, audit 14 multiple areas of outage related maintenance and evaluate 8 15 the operational readiness of selected plant systems.
16 The most recent MQA issued May, 1993, included 17 maintenance activities during the Unit 2 fifth refueling
- 18 outage. The assessment concluded that each of the audited !
19 activities had been effectively implemented. The report 20 also stated that the overall quality of maintenance and O 21 technical support demonstrated by each organization 22 provides confidence in the operational readiness of DCPP's 23 systems. Specifically, the audit nated that the steam
- 24 generator shctpeening had exceeded all expectations. The 25 planning and coordination led to a well-executed evolution 26 and the work was performed on time and with significantly 9 175
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1 less radiation exposure than anticipated. Also especially
) '2 effective, according to the audit, was the Generic Letter 3 89-10 MOV Testing. Field support by the foremen and 4 Electrical Maintenance engineers facilitated problem D 5 solving and helped keep the work moving smoothly. The 6 audit found some areas for improvement, and these areas i 7 have been documented on quality problem reports.
D 8 There were four Audit Finding Reports issued, three of ,
9 which were not maintenance-related but addressed design 10 documentation problems (associated with four air operated 3 11 valves) which were discovered in the course of the ,
12 assessment. The fourth finding was for incorrect torquing 13 of feedwater regulating valve capscrews to 16 ft-lbs
) 14 rather than the required 19 ft-lbs. These deficiencies did 15 not indicate a lack of programmatic control or overall 16 effectiveness. They are currently being resolved and 3 17 responded to in accordance with DCPP's required 18 procedures.
19 In December 1992, an MQA was issued for the Unit One D 20 fifth refueling outage. This audit also concluded that 21 the maintenance program, technical support of maintenance ,
22 and technical specification surveillance had been l 3 23 effectively implemented. Specifically cited as-effective' 24 were diesel generator maintenance and installation, and 25 calibration and placement into service of the Reactor i
) 26 Vessel Refueling Level Indication System ("RVRLIS"). )
27 Some deficiencies were noted and resulted in six Audit I 28 Finding Reports relating to specific issues, such as an ]
)
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1 incorrect radiographic test; procedural noncompliance
) 2 regarding reactor vessel water level during reactor head i 3 removal maintenance; incorporation of design criteria 4 memoranda requirements into certain surveillance test
)- 5 procedures; procedural noncompliance in control of some 6 M&TE; airborne radioactive contamination during steam 7 generator shotpeening; and a replacement parts evaluation
) 8 deficiency for a diesel engine fuel oil switch. DCPP has 9 responded with root cause analysis and corrective actions 10 identified for five of the six findings, and corrective
) 11 actions for three of these already have been completed.
12 The sixth, an incorrect radiographic test, is still under 13 investigation to determine if deficiencies exist. l
) 14 Again, these findings did not indicate any 15 programmatic problems, but did provide important feedback 16 to the DCPP maintenance program.
)
17 continuous Procram Improvement 18 Q19 What are some examples of " continuous improvement"
) 19 initiatives involving the DCPP maintenance and 20 surveillance programs?
) 21 A19 (Giffin) PG&E's operating and maintenance strategy 22 for DCPP is focused on continuous improvement. Over the 23 past few years, PG&E has conducted a number of internal
) '24 reviews of the maintenance program as part of its 25 continuous improvement strategy. Although these reviews 26 are specifically targeted at improving the efficiency of
) 177
1 the maintenance process, they provide a mechanism for O evaluating the quality of the maintenance organization in 2
3 terms of its internal processes. Examples of these 4 continuous improvement initiatives include:
O Maintenance Process Improvement Proiect 5 1.
6 In February 1992, DCPP initiated the Maintenance 7 Process Improvement Project. The Project consisted of a O
8 task force made up of employees from all levels of the 9 organization representing the DCFP maintenance process and 10 other related departments and sections. This task force 0 used continuous improvement techniques to identify 11 12 opportunities for improved efficiency in the maintenance 13 process. The Project was completed in December 1992, and O identified four specific recommendations with action plans 14 15 and eleven other recommendations worthy of further
~t 16 evaluation. The four specific recommendations are:
D Establish integrated maintenance teams aligned by 17
- 18 plant " systems";
19
- Establish a combined consumables, tools and O
20 equipment group; 21
- Modify the work priority system to provide an 22 integrated priority system for DCPP maintenance, 23 engineering and procurement personnel; 24
- Modify the minor maintenance program to improve 25 its utilization.
26 Each of these recommended improvements, along with j 27 necessary computer enhancements, is being implemented and l
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O 1 will add substantial efficiencies to DCPP's overall l
C 2 maintenanca process. )
3 2. Reliability Centered Maintenance 4 PG&E has an active process in place at DCPP to O 5 incorporate reliability centered maintenance ("RCM") '
6 improvements into DCPP's preventive maintenance programs.
7 RCM is a reliability-based methodology for optimizing O 8 preventive maintenance by analyzing the failure modes and 9 safety significance of equipment on a system-by-system 10 basis, and then improving condition-monitoring programs 9 11 for those systems to reflect the insights gained from the 12 analysis. PG&E has completed the first phase of its RCM 13 Program by completing a rigorous system analysis of the O 14 DCPP feedwater system and developing preventive i 15 maintenance program administrative procedures to reflect 16 the RCM process. The second phase of the program is well 0 17 underway at the plant, and involves analysis of 12 18 additional safety-related systems. Currently, the project ;
19 is nearing completion of the fifth and sixth systems (CCW O 20 and 4kV). If the second phase produces positive results, 21 a third phase will perform generic analyses of additional i
22 systems and components, and incorporate the insights into l O 23 DCPP's "Living Preventive Maintenance Program."
24 3. Procurement Task Force 25 In 1991, PG&E completed a multi-disciplinary review of O 26 the procurement process at DCPP. The purpose of this 27 review was to reduce overlap and duplication among various 28 departments, and to improve the reliability and timeliness 0 179 I
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0 1 of procurement of parts and equipment to support DCPP 2 maintenance and operations. The recommendations of the 3
Task Force have been implemented, and the DCPP procurement 4 organization has achieved a higher level of efficiency and ,
5 timeliness in meeting maintenance needs.
6 Reculatory Performance !
7 1. NRC Programmatic Assessments i 8 Q20 Please discuss examples of NRC assessments of the DCPP 9 maintenance and surveillance programs. .
'O 10 A20 (All)
As discussed above, the overall operating 11 performance of DCPP, as well as the numerous self-critical ,
'O 12 evaluations conducted by PGLE, are indicators which PG&E !
13 uses to monitor, arsess and continuously improve the i 14 performance of the DCPP maintenance and surveillance 15 programs. These measures demonstrate that the DCPP maintenance and surveillance programs are comprehensive, \
16 effective and superior. In addition, inspection, 17 J
'O
. 18 enforcement and assessment activities by regulators such as the NRC also provide PG&E with important information l 19 20 concerning the overall performance of all its programs, .
'O including maintenance and surveillance. These NRC 21 i I i 22 activities are discussed below. .
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O 1 a. NRC "Best Plants" List ,
O DCPP is on the NRC "best plants" list. How does this 2 Q21 3 relate to effectiveness of maintenance and surveillance at 4 DCPP?
O 5 A21 PG&E's excellent overall safety performance in 6 operating and maintaining DCPP has been recognized by the O NRC numerous times since issuance of the initial operating 7
8 licenses for the plant. Most importantly, four +
9 consecutive times over the last two years, the NRC has O commended PG&E's outstanding safety record as part of the 10 11 Commission's semi-annual review of licensee perfcrmance -
12 in February and June 1992, and again in February and June O 13 1993. These commendations, typically referred to as the 14 NRC's "best plants" list, are among the select few awarded 15 to NRC licensees nationwide. DCPP's commendations O 16 specifically reference the excellent performance of PG&E 17 personnel and programs, and recognize DCPP as being one of the best operated nuclear plants in the nation. These NRC 18 O commendations of DCPP's overall safety performance are ,
19 20 strong indicators that DCPP's regulatory performance in 21 all areas - including maintenance and surveillance - has O 22 attained a level of excellence among the highest in the 23 United States. Copies of each of the four NRC letters 24 commending DCPP are attached as Exhibit 19.
9 i
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1 b. SALP Ratings j 2 Q23 What have been the NRC SALP ratings of the DCPP 3 maintenance and surveillance programs?
i j 4 A23 (Giffin) The NRC periodically assesses the 5 programmatic performance of DCPP in key functional areas 6 such as maintenance and surveillance through the l
7 Systematic Assessment of Licensee Performance ("SALP")
3
-i 8 program. According to NRC SECY 90-189 (May 25, 1990), the 9 SALP program is an integrated NRC staff effort to ;
10 consolidate available information to support a periodic 3
11 evaluation of a licensee's overall performance. The SALP 12 process is a means of expressing NRC senior management's j 13 observations and judgments on licensee performance. SALP 14 reviews are conducted by a SALP Board which is 15 multidisciplinary in nature. SALP reviews are intended to l
j 16 result in an integrated assessment of licensee 17 performance. SALP Board members consist of a mixture of 18 NRC regional and headquarters personnel. ;
j 19 NRC Manual Chapter 0516 dated September 28, 1990 20 establishes the specific performance criteria applicable 21 to each functional area, and defines the 22 maintenance / surveillance functional area as including:
l 23 "
...[A]ll activities associated with either 24 diagnostic, predictive, preventive or corrective 25 maintenance of plant structures, systems, and 26 components; procurement. control, and storage of 27 components, including qualification controls; 28 installation of plant modifications; and maintenance 29 of the plant physical condition. It includes conduct
- 30 of all surveillance (diagnostic) testing activities as 31 well as all inservice inspection and testing 182 3
D 1 activities. Examples of activities included are instrument calibrations; equipment operability tests;
'g 2 post-maintenance, post-modification, and post-outage 3
4 testing; containment leak rate tests; special tests; 5 inservice inspection and performance tests of pumps 6 and valves; and all other inservice inspection 7 activities."
O 8 Thus, the SALP report provides PG&E - as well as the NRC -
9 with an integrated, multidisciplinary, programmatic 10 evaluation of the effectiveness of a licensee's
- O performance in key functional areas, including maintenance 11
~
12 and surveillance.
13 NRC's SALP ratings have recognized the high quality 14 and effective performance of DCPP's maintenance and 15 surveillance programs. For the most recent SALP review 16 period, July 1, 1991 through December 31, 1992, the NRC lo 17 rated PG&E's performance with six "1" ratings and one "2 18 and improving" rating. The NRC rated l
19 Maintenance / Surveillance at DCPP as Category 1, the f
lO i
20 highest possible rating. Category 1 is defined as a 21 " superior level of performance" based on licensee i
[ 22 management attention and involvement in licensed l0 23 activities (NRC Manual Chapter 0516, Section 08.a). More 24 specifically, the SALP report cited the following bases
- 25 for its " Category 1" rating of DCPP maintenance and
'O 26 surveillance:
l 27 e Virtually trouble-free plant operation evidenced a 1
28 high quality of maintenance work; SO' 29 e The high-level of management involvement in scheduling 30 and planning maintenance and surveillance work 31 maximized safety system availability; O 183
l I
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I e Maintenance and surveillance work was generally of 2 high quality; ,
3
- The training and qualification program for maintenance 4 personnel was strong; ;
'O 5
- Management of outages was marked by an overriding 6 understanding and emphasis of the risk of each job. !
7 Work crews and planners were trained and aware of the O 8 safety significance of the jobs and systems on which 9 they worked; 10
- Work items were well prioritized, with safety-11 significant issues given high priority. Backlog of 12 non-outage safety related work was low; 13
- The involvement and leadership shown by maintenance 14 staff in root cause investigations was a significant 15 strength, as was the integration of maintenance, 16 operations and engineering staff in maintenance and 17 surveillance activities; r
18
- The number of personnel errors was reduced due to the 19 high level of management involvement throughout the -
O 20 organization; and 21
- The maintenance staff overall improved their response i
22 to problems by identifying, anhayzing and correcting i
~O 23 maintenance and surveillance problems promptly. l 24 The SALP report cited four Level IV violations that had i 25 occurred in the maintenance / surveillance area, but i 26 concluded that the concerns associated with each appeared 27 to have been isolated. Some additional minor weaknesses 28 were noted as having occurred early in the SALP assessment
,0 l 184 t
1
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l 1 period. However, the SALP report stated that most had 2 been identified by PG&E immediately upon occurrence, and ,
3 PG&E management involvement was effective in promptly and 4 appropriately correcting the problems. A copy of the SALP
) 5 report is attached as Exhibit 20. ,
6 Previous SALP evaluations also have recognized the r
7 effectiveness of DCPP's maintenance and surveillance
) 8 programs from an overall perspective. The NRC judged 9 DCPP's performance in maintenance / surveillance to be 10 Category 2 (a " good level of performance") in each of the
) 11. three SALP assessment periods spanning August 1987 through i
12 June 1991. It is importan,t to note that every SALP 13 evaluation, including those in which PG&E's ;
) 14 maintenance / surveillance programs have been rated " good" 15 or " superior," contains qualitative assessments of various 16 individual weaknesses and strengths exhibited within the
) 17 particular functional area during the assessment period.
18 This " feedback" from the regulators is subject tc the same 19 formal followup evaluation and corrective action by PG&E
) 20 management as internal problem reports. PG&E maintenance.
21 and surveillance management track each item to closure to 22 assure that corrective actions have been implemented in 23 response to the SALP findings.
24 The SALP ratings and NRC "best plants" list also must 25 be considered in the context of the entire nuclear 26 industry. DCPP's current SALP "1" rating in i
27 maintenance / surveillance places it high among current l
l 28 licensees in that functional area. DCPP is one of only 185 i !
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1 five plants nationwide which was commended by the NRC in
) 2 June 1993 for its overall superior safety performance.
3 This indicates that, in terms of overall regulatory 4 performance, DCPP's maintenance and surveillance program
)
t 5 is at a very high level when " benchmarked" against its '
6 peers in the industry.
D 7 2. NRC Inspection and Enforcement Activities 8 Q24 Please discuss how NRC inspection and enforcement 9 activities evaluate DCPP's maintenance and surveillance !
)'
! 10 programs?
i
- _ 11 A24 (Giffin) Routine and special NRC inspections provide
) 12 continuous information to PG&E on specific aspects of the 13 DCPP maintenance and surveillance programs. For example, 4 14 NRC inspectors routinely review DCPP LERs.and other plant J 15 events. Open items on such events are identified in 16 monthly NRC inspection reports, and PG&E's response to 17 each open item is evaluated in a subsequent NRC inspection
. 18 report. Each LER event is fully evaluated by PG&E through -
19 root cause analysis, corrective actions and/or program t
20 improvements.
)
21 Other NRC inspections and reviews provide PG&E with {
22 in-depth NRC evaluations of important maintenance and a 23 surveillance activities and programs. Examples include:
) 24
- NRC Inspection Report ("IR") 93-08, in which the NRC ,
i 25 reviewed the DCPP 10-year In-service Inspection
("ISI") performed during the Unit 2 fifth refueling I
26
) 186 i
s 0
1 outage, and concluded that PG&E was implementing a 0 2 comprehensive ISI program in accordance with Technical 3 Specification, ASME Code and other NRC and industry 4 requirements; O 5
- IR 91-39, concluding that PG&E was developing an 6 aggressive, well-integrated program for assuring 7 safety-related MOV reliability under Generic Letter O 8 89-10, and at the same time noting certain areas 9 requiring further development; 10
- IR 92-201, concluding that PG&E was implementing a 0 11 strong program for evaluating and minimizing safety 12 risks associated with maintenance activities performed 13 while certain safety systems are unavailable during O 14 refueling outages; 15
- NRC Safety Evaluation (TAC No. M83285), dated 16 September 4, 1992, concluding that PG&E's supplemental
' O 17 reactor vessel surveillance program meets NRC 18 requirements.
19 Other NRC inspections may identify isolated areas in Q 20 PG&E's maintenance and surveillance programs which need to L
l 21 be addressed. PG&E evaluates the safety significance of l
l 22 major items, performs a formal root cause analysis, and
.D implements corrective actions to avoid recurrence.
23 24 Based on a review of thousands of hours of NRC 25 inspection and enforcement activities, and hundreds of 9
26 pages of NRC inspection reports issued over the last two 27 SALP assessment periods, PG&E believes that the regulatory 187
[7
I 1 performance of DCPP's maintenance and surveillance 2 programs has been excellent.
3 III. CONCLUSION
)
4 Q25 Based on the content, performance and history of 5 PGEE's maintenance and surveillance programs for DCPP, are
)
6 the programs sufficiently effective and comprehensive to 7 assure that the public health and safety is protected if 8 PG&E is granted a full 40-year operating license for DCPP 9 as requested by the license amendment at issue in this 10 proceeding?
)
11 A25 (All) Yes.
12 Q26 Does this conclude your testimony?
)
13 A26 (All) Yes.
)
)
)
188
O pgJ.TED COnREEPCNDENCE f August,2,.1993 f hNd !
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UNITED STATES OF AMERICA f NUCLEAR REGULATORY COMMISSION "93 AUS -4 P3 :44 I BEFORE THE ATOMIC SAFETY AND LICENSING BOARD !
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DUCni + . . . -
4U t i f I!.f. h i
'O In the Matter of: Docket Nos. 50-275-OLA
) l
) 50-323-OLA ;
Pacific Gas and Electric Company ) i
) (Construction Period }
(Diablo Canyon Nuclear Power ) Recovery) ;
Plant, Units I and 2) )
0
's f
f TESTIMONY OF PACIFIC GAS AND ELECTRIC COMPANY !
ADDRESSING CONTENTION I: MAINTENANCE AND SURVEILLANCE !
- O ;
1 EXHIBITS l LO !
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D i i
LIST OF EXHIBITS ,
D-
- 1. Professional Qualifications of Bryant W. Giffin
- 2. Professional Qualifications of William G. Crockett
. 3. Professional Qualifications of David A. Vosburg D
- 4. Professional Qualifications of Steven R. Ortore j
- 5. PG&E's License Amendment Request ;
- 6. Program Directive PD MA1 " Maintenance" ?
- 7. Diablo Canyon Maintenance Services Department Organizational i Chart I
- 8. Basic Qualification Program Training Topics 3 9. Advanced or Plant-Specific Training Examples
- 10. List of Maintenance and Surveillance Procedures
- 11. DCPP Work Control Process Overview 3 12. Professional Qualifications of Tedd A. Dillard
- 13. Professional Qualifications of David B. Miklush
- 14. Diablo Canyon Annual Capacity Factor D 15. Diablo Canyon Operating Capacity Factor
- 16. Diablo Canyon Refueling Outage Duration
- 17. Diablo Canyon Corrective Maintenance Backlog 3 18. Diablo Canyon overdue Preventiv Maintenance Items
- 19. NRC Commendation Letters [
- 20. SALP Report l D
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EXHIBIT l'
'O PROFESSIONAL QUALIFICATIONS ;
OF i BRYANT W. GIFFIN O I i
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RESUME i
MANAGER, MAINTENANCE SERVICES O Bryant W. Giffin
- 1. Birthdate - July 22, 1941 0 2. Citizenship - USA ;
- 3. Education
- a. B.S., Electrical Engineering, North Carolina State University, 1966.
O b. Professional Degree, Electrical Engineer, North '
Carolina State University, 1967
- 4. Employment History - Joined PG&E on June 1, 1981. f O "- 1959 - 1967 U.S. Navy enlisted
- b. 1967 - 1980 U.S. Navy Officer r
- c. 1980 - 1981 Senior Engineer, Westinghouse Electric Co. i O
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- d. 1981 - 1983 Senior Engineer, .PG&E l e. 1983 - 1985 Instrument & Controls !
l Maintenance Manager, DCPP ,
t
- f. 1985 - 1987 Supervising Engineer, Nuclear
- lb Operations Support Department, PG&E l
- g. 1987 - 1988 Manager Nuclear Operations Support, PG&E .
- h. 07/01/88 - 09/25/89 Assistant Plant Manager -
Technical Services, DCPP, PG&E ,
- i. 09/25/89 - 07/08/91 Assistant Plant Manager - ,
Maintenance Services, DCPP,
[) PG&E
- j. 07/08/91 - Present Manager, Maintenance Services <
(title change)
- 5. Nuclear Experience ;
p- a. Manager, Nuclear Operations Support Responsible for providing support to the plant in the areas of: project management, naintenance, primary and ,
secondary chemistry, health physics, training, security, and environmental protection. In addition to the above' support functions, I was responsible for D
e nuclear fuel procurement, nuclear analysis, corporate emergency planning, t'ne review of industry operating experience, and contract negotiations and administration for outage services.
- b. Assistant Plant Manager - Technical Services, DCPP Responsible for the Plant Engineering, Computer Engineering, and Licensing Departments. The plants engineering department includes plant system engineering, reactor engineering, surveillance 8 testing, inservice inspection and testing, and design control. The computer engineering department provides engineering support for all plant process computers and maintains the plant information management system and internal communications system. The licensing department performs all plant related licensing
- functions.
- c. Assistant Plant Manager - Maintenance Services, DCPP Responsible for the maintenance, work planning, scheduling, materials, and outage management for both
, DCPP units. The maintenance organization at DCPP is J comprised of three departments: Electrical, Mechanical, and Instrumentation and Controls, and is responsible for maintenance activities for both DCPP Units. The Work Planning and Scheduling Department is responsible for the day to day, forced outage, and planned outage scheduling of all corrective and 3 preventive maintenance activities, design modifications, and testing activities. The Matsrials Department is responsible for the ordering, storage and issuance for all spare parts for DCPP. The Dutage Management organization is responsible for the
, planning and implementation of forced and planned J outage activities for both DCPP units.
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LXNIBIT 2 l O FRorEssionAL ocALIFIcATIons i Or WILLIAM G. CROCKETT O i L
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4 RESUME MANAGER, TECHNICAL AND SUPPORT SERVICES 9 William G. Crockett
- 1. Birthdate - June 30, 1946
- 2. Citizenship - USA
- 3. Education
- a. B.S., Physical Science, California State University, Hayward, 1972.
) b. California State Secondary Teaching Credential, 1973
- c. B.S., Nuclear Engineering / Tech., Oregon State University, 1977.
- 4. Employrent History - Joined PG&E in May 1979.
- a. July 1964 to June 1968 - Electronics Tech. - U.S.
Navy.
- b. August 1973 to June 1975 - High School Teacher -
Math / Science.
- c. July 1977 to May 1979 - Employed by United Nuclear Industries as a Control Room Shift Supervisor.
- d. May 1979 to February 1980 - Employed by PG&E -
O Assigned to DCPP as a Power Production Engineer (nuclear)
- e. February 1980 - Assigned to Operations Department as a Shift Technical Advisor.
O
- f. November 1980 - Promoted to Senior Power Production Engineer of the Operations Department.
- g. January 1985 - Promoted to Maintenance Director of Instrumentation and Controls.
- h. October 1989 to January 1991 - Rotational assignment O as Assistant Plant Manager for Support Services.
- i. January 1991 to December 1991 - Instrumentation and Controls Maintenance Director.
- j. January 1992 to July 1992 - Promoted to Manager of Support Services
- k. July 1992 to June 1993 - Promoted to Manager of Technical Services.
t l
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- 1. June 1993 to present - Manager of Technical and ,
support Services. '
- 5. Nuclear Experience O ,
- a. Control Room Supervisor at the Hanford N-Reactor. The N-Reactor is a dual purpose, 3800 MWT, PWR, owned by the DOE and operated by United Nuclear Industries. '
- b. Power Production Engineer (Nuclear) at DCPP. Engaged O in procedure preparation and startup testing of plant -
systems and equipment.
- c. Shift Technical Advisor at DCPP '
- d. Received NRC Senior Operators License at DCPP, Unit 1, i
.O June 15, 1981 (SOP-3956)
- e. Senior Power Production Engineer (Operations).
Supervise Operations Department Engineering staff. ,
Provide technical and administrative assistance to the Supervisor of Operations. Originated normal, ,
O abnormal, and emergency operating procedures.
- f. Member of the Westinghouse Owner Group Sub-Committee for Emergency Response Guidelines, June 1980 -
June 1984.
'O g. Received Unit 2 Senior Operators License for DCPP, ;
rebruary 13, 1985.
- h. Director of Maintenance for Instrumentation and controls. ,
O i. Manager of Support Services - Responsible for Training, Security, Building and Land Services, Emergency Planning, Fire Protection, and Industrial ,
Safety. -
- j. Manager of Technical Services - Responsible for the O following engineering disciplines: System !
Engineering, Reliability Engineering, Nuclear i Engineering, and Computer Engineering. Provide technical support to the plant staff and the Vice ,
! President and Plant Manager. -
'O 6. Formal Training :
i'
- a. Electronics Training - Class A and Class C schools -
U.S. Navy.
l
- b. Participated and completed all formal training' course
- O leading to Control Room Supervisor Certification at Hanford N-Reactor.
- c. Participated and completed Westinghouse training course for Shift Technical Advisor at DCPP.
O
1,
- d. Completed Thermodynamics, Heat Transfer, and Fluid Flow pre-license review course given by Energy Consultants, Inc. on December 12, 1980
- e. Participated in training courses for preparation of NRC Senior Operators License at DCPP.
- f. Completed Mitigating Core Damage course given by Westinghouse Electric Corporation, July 24, 1981.
3
- 1) Option II, 7 day course, February 1980.
3 2) Option III, 14 day course, August 1990.
- 3) Retraining, 5 day course, June 1981.
- 4) Retraining, 5 day course, April 1982.
9 5) Retraining, 5 day course, May 1983
- h. Annual NRC Operator License and simulator requalification courses at DCPP.
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O EXHIBIT 3 O PROFESSIONAL QUALIFICATIONS OF DAVID A. VOSBURG O
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RESUME WORK PLANNING DIRECTOR 3 David A. Vosburg
- 1. Birthdate - October 22, 1952
- 2. Citizenship - USA g
- 3. Education - B.S., Nuclear Engineering, University of California, Santa Barbara, 1977
- 4. Employment History - Joined PG&E in December 1979.
o a. June 1977 to March 1979 - Employed by the Bechtel
~'
Power Corporation as a Nuclear Engineer assigned to the Arkansas Nuclear One - Unit 2 project during the final construction and initial startup phases of the facility.
m
- b. March 1979 to November 1979 - Employed by the General Electric Company as a Design Engineer involved in the thermal design and environmental qualification of Solid State Nuclear Plant Protection Systems.
- c. December 1979 to April 1980 - Employed by PG&E
- d. April 1980 - Assigned as a Shift Technical Advisor at DCPP.
- e. January 1983 - Promoted to Shift Foreman in the Operations Department.
- ~)
~
- f. August 1986 - Promoted to Engineering Manager.
- g. January 1987 - Transferred to the Operations Department as a Shift Supervisor.
) h. April 1992 - Promoted to Operations Superintendent.
- i. August 1992 - Promoted to Work Planning Director.
- 5. Nuclear Experience
-)
~
- a. Bechtel Power Corporation - Responsible Systems Engineer for the Chemical and Volume Control, Safety Injection, Regenerative Waste and Solid Radwaste Systems. Participated in the development of the Plant Technical Specifications, Fire Hazards Analysis and High Energy Pipe Break Analysis.
9
- b. General Electric Company - Involved in the thermal design and environmental qualification of safety-related electronic equipment used in nuclear power facilities. Additional responsibilities were in D
)
the area of' radiation shielding of electronic equipment. ;
- c. Pacific Gas and Electric, DCPP - Power Production D Engineer (Staff) engaged in revising the Site ,
Emergency Plan and in the preparation of Emergency f Procedures. Worked in the Operations Department throughout the pre-license, startup and power ;
ascension phases of both Unit 1 and 2 as a Shift Technical Advisor,' Shift Foreman, Shift Supervisor and D Operations Superintendent. Work Planning Director responsible for the maintenance work planning and work ,'
l scheduling activities for DCPP during normal operation l
and outage periods.
- d. Senior Reactor Operator License for DCPP maintained D from April 1982 (Unit 1), April 1984 (Unit 2) through l September 1992.
- 6. Formal Training Courses
- a. Westinghouse Shift Technical Advisor Training program J for transient and accidental analysis (27 week program, 1980).
- b. Simulator Training - Westinghouse Nuclear Training
[
Center, Option III (14 day course, 1980).
D c. Westinghouse Operation PWR Training Program, Phase II l (10 week course, 1981).
l l
- d. Westinghouse PWR Simulator Training Program, Phase III (11 week course, 1981).
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O EXHIBIT 4 3 PROFESSIONAL QUALIFICATIONS OF STEVEN R. ORTORE
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O RESUME !
ELECTRICAL MAINTENANCE DIRECTOR ;
O Steven R. Ortore {
t
- 1. Birthdate - February 13, 1951
- 2. Citizenship - USA O
- 3. Education
- a. M.S., Electrical Engineering, Columbia University, N.Y. ,
- b. B.S., Engineering Science, Richmond College, N.Y.
O
- 4. Employment History I
- a. 1974 to 1983 - Burns and Roe - Assigned to Clinch River Breeder Reactor Plant Project, Oradell, New Jersey in various capacities including cognizant Engineer of Radiation Monitoring, HVAC I&C, Security and Fire ;
O
- Detection Systems.
- b. 1983 to 1984 - Burns and Roe - Assigned to Nuclear Engineering Division of Pennsylvania Power and Light at Allentown, Pennsylvania as et gnizant I&C Engineer of in- {
O containment design modifications. ,
I
- c. 1984 to 1987 - Burns and Roe - Assigned to Procurement Specialist Group, and served as Lead I&C Specialist at t DCPP.
O d. 1987 to 1989 - PG&E . Procurement Specialist Supervisor in Materials Services Department at DCPP.
- e. 1989 to 1991 - PGLE Materials Manager, DCPP.
- f. 1991 to 1993 - PG&E Materials Director DCPP (title O change) 9 1993 to Present - Promoted to PG&E Electrical Maintenance Director DCPP
- 5. Nuclear Experience O
All of tha above work experience is nuclear related
- 6. Military 1971 to 1977 U.S. Army Reserves, 24th Military Intelligence at Ft. i O Wadsworth, N.Y.
O ,
a-O EXHIBIT 5 PG&E'S LICENSE AMENDMENT REQUEST O
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Pacific Gas and Electr'c Company M hr S ift G t;; , r.t p.e;r Sr Frams:o C19:1X See V :t pre 3>::t .: g - !
415E'3-4%4 Gent a' t.4 .a;r ,
feier Pu.r Gr.rr :-
July 9,1992 3-PG&E I.etter No. DCL-92-154 ;
U.S. Nuclear Regulatory Commission '
ATTN: Document Control Desk g Washington, D.C. 20555 Re: Docket No. 50-275, OL-DPR-80 .;
Docket No. 50-323, OL-DPR-82 i Diablo Canyon Units 1 and 2 License Amendment Request 92-04 0 40-Year Operating License Application Gentlemen:
Pacific Gas and Electric Company applies for an amendment to Facility Operating i O License Nos. DPR-80 and DPR-82 to change the expiration dates of the full-power ';
licenses for Diablo Canyon Units 1 and 2. TG&E's enclosed request would change l' the expiration date for the Unit 1 Operating License from April 23,2008, to September 22,2021, and the expiration date for the Umt 2 Operating License from December 9,2010, to April 26,2025. These proposed expiration dates would allow-9 for 40 years of operation as permitted by 10 CFR 50.51.
The present operating license terms for Diablo Canyon are based on the NRC policy in effect prior to the 1982 determination by the Commission that the 40-year term of operation may begin upon issuance of tae first operating license, rather than upon .
issuance of the construction permit. Therefore, the present operating license terms .
g for.Diablo Canyon commence with the dates ofissuance of the construcFon permits for Units 1 and 2, April 23,1968, and December 9,1970, respectively. .
Accordingly, the expiration date for the Unit 1 Operating License is April 23 2008, and the expiration date for the Unit 2 Operating License is December 9,2010.
O Since 1982, the Commission has accepted and approved requests to amend exiving operating licenses to change the expiration dates and recover the time between the effective dates of the construction permit and the first operating license. More than 50 such license amendments have been granted by the Comenission. Based on the enclosed request, the proposed 40 year term start dates for Diablo Canyon are -
3 September 22,1981, for Unit I and April 26,1985, for Unit 2, which correspond to ;
the effective dates of the fuel-load / low-power operating licenses for each unit.-
The proposed license term changes do not affect the design, operation, or Technical Specifications of the plant. Based on a review of the Diablo Canyon Final Safety g Analysis Report Update and the associated NRC Safety Evaluation Report and ,
Supplements, PG&E concludes that the proposed changes do not involve significant hazard considerations.
3
Document Control Desk July 9,1992 -
PG&E lett:r No. DCL-92-154 O- ,
PG&E also has determined that the environment will not be adversely affected by the proposed license 'erm changes based on a review of the Diablo Canyon Environmental Report and Suppsments and the NRC Final Environmental Statement and Addendum.
O Therefore, pursuant to 10 CFR 51.30 through 51.35, PG&E believes the proposed changes mquire preparation only of an environmental assessment and finding of no significant impact, ,
and that preparation of an environmental impact statement is not required. 'Ihese actions are l consistent with the Commission's practice for similar amendment requests. t I
O Based on these safety and environmental reviews, PG&E requests that the Diablo Canyon operating license expiration dates be changed from April 23,2008, to September 22,2021, >
for Unit I and from December 9,2010, to April 26,2025, for Unit 2.
Sincerely, O ,
m Gregcry M. Rueger O cc: Edgar Bailey, DHS Ann P. Hodgdon John B. Martin >
Philip J. Morrill Harry Rood O CPUC Diablo Distribution Enclosure O 5408S/85K/EMG/2057 O
O.
O O
NAA Lettet No. DCL 154 g ENCLOSURE UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 3
) Docket No. 50-275 In the Matter of ) Facility Operating License
) No. DPR-80 PACIFIC GAS AND ELECTRIC COMPAh7 )
) Docket No. 50-323
_ Diablo Canyon Power Plant ) Facility Operating License U Units 1 and 2 ) No. D: R-82 I
License Amendment Request No. 92-04 O
Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company hereby applies to amend its Diablo Canyon Power Plant Facility Operating License Nos. DPR-80 and DPR-82. The purpose of tbase license amendments is to change the term of the licenses to permit 40-year operation of each unit from the date of issuance of the operating licenses.
O Information regarding the proposed amendments is provided in Attachments A and B.
These changes have been reviewed and are considered not to involve a significant hazards consideration as defined in 10 CFR 50.92 and to have no significant environmental impacts.
O Further, there is reasonable assuran:e that the health and safety of the public will not be adversely affected by the proposed changes.
Sincerely, O
n u.t ut r Gregory M. Rueger a Subscribed and sworn to before me Howard V. Golub this 7th day of July 1992. Christopher J. Warner Richard F. Locke Attorneys for Pacific Gas and Electric Company O
h/ ryz j
'O Mildred J. WiHfams, Notary Public Christopher J. Warner MI:4.:nt:nznan. resonannt: :.n:nn:nnenen.
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4 9
>O ATTACIIMENT A DIABLO CANYON POWER PLANT UNITS 1 AND 2 40-YEAR OPERATING LICENSE APPLICAT;C;;
O O
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o 540SS 'S5K e
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CONTENTS l r
j 1.0 DESCRIFFION OF A5 FEND 51ENT REQUEST . . ....................I
2.0 BACKGROUND
................................ ,,,,,,,,,,,3 3.0 JUSTI FI C ATI ON . . . . . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . 2 J
D 3.1 Baseload Generation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.2 Electric Rates ................................ ......3 3.3 A ir Emis sions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.4 State and Local Economy . . . . . . . . . . . . ...................3 '
g 4.0 S A FETY EV A LU ATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 .
4.1 Int rod u :t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 I 4.2 Licensing Basis Documents / Programs ........................4 '
4.2.1 FS AR and Technical Specifications . . . . . . . . . . . . . . . . . . . 4 D 4.2.2 Probabilistic Risk Assessment . . . . . . . . . . . . . . . . . . . . . . . 4 i 4.2.3 Surveillan:e and hiaintenance Programs . . . . . . . . . . . . . . . . 5 4.2.3.1 151 and IST Programs ...........................5 4.2.3.2 EQ Program ... .. ........................6 4.2.3.3 hiaintenance Program . .... . ...... .........7 4.3 Plant Operating History ............... ......... .. 10 4.3.1 Operating Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
, 4.3.2 Reliability / Safety-Related Plant hiodifications . . . . . . . . . . . . . I1 D 4.3.3 Regulatory Performance . . . ... . . . . . . . . . . . . . . . . 13 4.4 Assuran:e of Continued Functional Capability of S afety-Related Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 .
t -
4.4.1 Reactor Coolant System Pressure Boundary . . . . . . . . . . . . . 13 4.4.1.1 General ................................... 13 ;
4.4.1.2 Reactor Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.4.1.3 Rea: tor Vessel Internals . . . . . . . . . . . . . . . . . . . . . . . . . 15 t
p 4.4.2 Other hiechanical Components . . . . . . . . . . . . . . . . . . . . . 15 1 4.4.3 Electrical Components . . . . . . . . . . . . . . . . . . . . . . . . . . 16 i 4.4.4 Stru. rural Components . . . . . . . . . . . . . . . . . . . . . . . . . . '6 .
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4.4.4.1 Primary Containment . . . . . . . . .................. 16 4.4.4.2 Other Structures . . ......
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5.0 ENVIRONMENTAL EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 i l
5.1 I ntrod u cti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 l 5.2 Systems and Programs for Environmental ,
O control and Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 !
i 5.2.1 Waste Processing System ........................18 :
5.2.2 ALARA Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 5.2.3 Process and Area Radiation Monitoring System . . . . . . . . . . . 20 5.2.4 Radiological Environmental Monitoring Program . . . . . . . . . . 20 l O 5.2.5 Nonradiological Surveillance Program ................ 21 !
r 5.2.5.1 Environmental Protection Plan . . . . . . . . . . . . . . . . . . . . . 21 i 5.2.5.2 NPD ES Permit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 i 5.3 Environmental Impa:t During Normal Operation ................22 ;
O 5.3.1 0::upational Radiation Exposure . . . . . . . . . . .........22 ,
5.3.2 Offsite Radiation Exposure ....................... 22 ,
5.3.3 Uranium Fuel . . . . . . . . . . . . . . . ................ 24 ;
5.3.4 Spent Fuel Storage ............ ...............25 ,
O 5.3.5 S olid w aste . . . . . . ...... ........... ....... 26 5.3.5.1 Iww Level Rad c .:tive Waste . ....... ............26 5.3.5.2 Spent Fuel . . . . . . . . . . . . . .................... 26 ,
5.3.5.3 Waste Shipping . . . ..... .. ................ 27 l 5.3.5.4 Solid Waste Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . 27 O 8 5.3.6 shermal and Ecological Effects of the Cooling Water System . . 27
5.3.7 Prote
tion of Histori: Properties . . . . . . . . . . . . . . . . . . . . 28 5.4 Exposure from Releases During Postulated Accidents . . . . . ....... . 29 O 5.5 Environmental-Related Plant Modift:ations .............. . .. 33 5.6 Decommissioning .. .... ......... ... .... 34 t
6.0 NO SIGNIFICANT llAZARDS EVALUATION ..... . ......... 35 !
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I TABLES 5-1 Diablo Canyon vs. INPO Industry Goal Average Annual o Occupational Exposure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 52 Comparison of Offsite Appendix I Radiation Exposure Limits l an d A ctu al D ata . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 O 5-3 Annual Volume of Low Level Solid Radioactive Waste Generated at Diablo Canyon Compared with Median PWR . . . . . . . . . . . . . . . . . . . . 26 5-4 Summary of Population Projections for the Diablo Canyon Vicinity ..........30 O FIGURES 5-1 Diablo Canyon Exclusion Area ........ ... .................. 11 l0 5-2 Low Population Zone . . . . . . . . . . . . . . . . ..................... 32 l
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. ATTACHMENT A DIABLO CANYON POWER PLANT UNITS 1 AND 2 40-YEAR OPERATING LICENSE APPLICATION D
1.0 DESCRIPTION
OF AMENDMENT REQUEST This license amendment request (LAR) proposes to revise LI:ense Condition 2.I, " Term of License," in the Diablo Canyon Power Plant (DCPP) Units 1 and 2 Operating Licenses.
The proposed amendment would change the expiration date for the Unit i Operating License DPR 80 from April 23,2008, to September 22,2021, and change the expiration date for the Unit 2 Operating License DPR-82 from December 9,2010, to April 26,2025.
3 Marked-up pages from the licenses are included in Attachment B.
2.0 BACKGROUND
Both the Atomic Energy Act of 1954 and NRC regulations authorize issuance of facility operating licenses 3 for a pericd of up to 40 years. Commencement of the 40-year period from the date ofissuance of the operating license is allowed by 10 CFR 50.51, which states in pan: *Where the operation of a facility is involved the Commission will issue the license for the. term requested by the applicmt or for the estimated useful life of the fa:ility if the Commission determines that the estimated useful life is less than the term requested.* The NRC has established that requests for operation for a 40-yer term from.the 3 date ofissuance of the first operating license may be granted, provided the utility demonstrates reasonable assurance of safety during 40-year fa:ility operation and that the environmental effects of 40-year operation are evaluated in the Environmental Report.
Operating licenses were issued for a term of 40 years starting from the effective date of the construction permit for plants licensed prior to 1982. Beginning in 1982, however, the NRC's policy has been to 3 issue operating licenses for a 40-year term commencing with the effective date of the first operating li:ense. This revised policy in:ludes not only newly licensed plants, but also earlier plants that may apply for amendments to their operating licenses to change the license expiration date to 40 years from the effective date of the operating license.
3 DCPP Units 1 and 2 are currently licensed for 40 years commencing with the effective dates of the constru; tion permits on April 23,1968, and December 9,1970, respectively. Thus, the Unit I license expires at midnight on April 23,2008, and the Unit 2 license expires at midnight on December 9,2010.
This LAR proposes that the 40-year license terms, as permitted by 10 CFR 50.51, begin from the effective dates of the Unit I and Unit 2 first operating licenses. The initial fuel-load / low-power operating D license for Unit I was effective on September 22,1981, and the initial fuel-load / low-power operating license for Unit 2 was effective on April 26,1985. Thus, the proposed license term expiration dates would be September 22,2021, for Unit I and April 26. 2025, for Unit 2.
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!* 3.0 JUSTIFICATION PG&E has assessed the safety aspects of plant design and operation of each DCPP unit for the proposed 40-year operation. Based on this assessment, as provided in Section 4.0 of this atta:hment, PG&E Ig concludes that plant safety will be maintained during the 40-year operation. This conclusion is supported ,
by the following factors (1) the DCPP facility has been designed and analyzed for at least 40 years of ,
operation; (2) the equipment, structures, and materials were purchased or constructed based on operation l of at least 40 years; and (3) inspection and maintenance programs were developed to be applicable for at least 40 years of plant operation.
- O PG&E has reviewed the Diablo Canyon Final Environmental Statement and Addendum and the Environmental Repon and its Supplements to ensure that 40-year operation of each unit, commencing ;
with the effective date of the operating licenses, is consistent with the previously evaluated environmental effects. The results indicate that no additional significant environmental impacts beyond those originally addressed are involved with 40-year operation. A summary of the environmental evaluation is provided in Secti n 5.0 of this attachment.
O PG&E has also performed an assessment of the potential impact on histori properties in accordance with Section 106 of the Histori a1 Preservation Act of 1966, as amended, and in accordance with the provisions of 36 CFR 800, Protection of Historic and Cultural Properties. No signi0: ant impa:t on histori properties was identified that could be asso:iated with the proposed license term changes.
O Benefits expected from the additional period of operation include: (1) continued availability of reliable baseload generation, (2) avoided increase in electric rates to consumers, (3) avoided air emissions, and !
(4) continued benefits to the state and local area economy. Each of these benefits is briefly summarized -
below.
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3.1 BASELOAD GENERATION L
Diablo Canyon provides baseload generation to consumers throughout Northern and Central Califorma.
The plant has prode:ed approximately 100 billion kilowatt hours (kWh) of ele:tricity since beginning ;
O commercial operation in 1985 and is one of the most reliable nuclear power facilities in the country, with a lifetire capacity fa: tor of approximately 77 percent. Accordingly, it is prudent and beneficial to keep i this whable source of power in operation, parti ularly in light of the projected growth of California's ele.triaty demand.
In 1990, PG&E's ele:tricity demand was more than 85 billion kWh. It is expected to grow through
.O 2025. The California Energy Commission in 1992 adopted a new forecast for the PG&E service area L under which electricity demand is projected to in:rease at the rate of about 1.5 percent compounded annually through 2011. Assuming the demand for electricity continues to increase at a 1.5 percent per year rate, the overall increase in demand (1990 to 2025) would be 57 billion kWh or 65 percent. For comparison, in the period 1980-1990, PG&E's actual electri ity demand grew at an annual rate of about '
O 2.6 percent.
Given the projected electricity demand and associated increased capacity requirements through 2025, the construction of some new centralized power plants will probably be required, even assuming the term of
( Diablo Canyon's operating licenses are changed as requested in this LAR. Early retirement of Diablo
! Canyon would significantly exacerbate su:h requirements. Therefore, continued availability of Diablo O Canyon will be of substantial benefit to Northern and Central California.
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3.2 ELECTRIC RATES In addition to providing reliable baseload generation, Diablo Canyon power costs to the customer are l
) anticipated to be competitive with PG&E's overall production costs by 2008. After 2008, Diablo Canyon costs are currently estimated to increase at less than the rate ofinflation or otherwise be competitive with new power plant costs. If the plant is not operated beyond 2008, it is likely that it will be necessary to ,
construct new baseload capacity. Accordingly, by current estimates, continued operation of Diablo Canyon through the proposed license terms would reduce future electric rate increases.
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3.3 AIR E5flSSIONS '
Substantial reduction in air emissions would result from the proposed amendment because the air !
emissions from natural gas or other fossil-fired replacement power source to replace a prematurely closed -
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Diablo Canyon would be greater than from Diablo Canyon. If PG&E were to operate a gas-fired ;
generator instead of Diablo Canyon, an additional annual emission of approximately 14 billion pounds of carbon dioxide,2.5 million pounds of nitrogen oxides, and 70,000 pounds of sulfur-dioxide would ,
occur. Continued operation of Diablo Canyon will avoid these emissions.
) 3.4 STATE AND LOCAL ECONO 51Y The Diablo Canyon plant is the largest private employer and tu source in San Luis Obispo County, California. This prominence is expected to continue through 2025. PG&E fully recognizes its role as a major economic base for the community. Currently, annual propeny and related taxes on Diablo
) Canyon total about 560 million. The majority of these tax revenues goes to San Luis Obispo County and represents about 25 percent of the total county propeny tax collections. The remainder is distributed throughout the rest of the state of California, where it represents about 0.2 percent of the total state propeny tax collections. PG&E currently estimates that Diablo Canyon propeny taxes will remain substantial throughout the operational lifetime of the plant.
) Equally imponant, Diablo Canyon expends approximately 5100 million annually for vendor goods and {
services. The plant has over 1,500 full-time employees as well as 500 contract employees on a ,
year ,d basis with an annual payroll of approximately 5190 million. During refueling ouages, which
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occur approximately every 18 months for each unit. nearly 1,000 additional contract personnel are employed.
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These benefits are anticipated to continue through the propbsed license terms.
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4.0 SAFETY EVALUATION 4.I INTRODUCTION The purpose of this safety evaluation is to determine whether the proposed 40-year operating license terms would adversely affect the health and safety of the public. Most of the information that follows summarizes existing programs and activities that have been previously approved or provided to the NRC.
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Section 4.2 reviews documents and programs that assure continued safety through the operating lifetime of the plant. Section 4.3 summarizes the plant performance and safety record during the initial 7 years of operation. Finally, Section 4.4 reviews the assurances for continued functional capability of safety-related components, systems, and structures through at least 40 years of plant operation.
D These activities and programs provide assurance that continued operation for a 40-year term will be consistent with protection of the health and safety of the public.
4.2 LICENSING BASIS DOCUMENTS / PROGRAMS J
4.2.1 FSAR and Technical Specifications "Ihe Diablo Canyon Final Safety Analysis Report (FSAR) was originally submitted to the NRC (then AEC) in September 1973 to suppon the operating license applications for Units I and 2. The FSAR included facility and system design descriptions, site characteristics, analyses of design basis accidents, and descriptions of plant operations. The NRC review of the FSAR was documented in a Safety Evaluation Repon (SER) issued in October 1974 and SER Supplements I through 34 issued periodically through June 1991. In 1984, a major update of the FSAR (termed the FSAR Update) was submitted to the NRC in accordance with 10 CFR 30.71(e). The FSAR Update is revised annually to reflect changes in plant design, evaluations, and analyses. The latest revision (Revision 7) was submitted to the NRC D in September 1991.
Separate Technical Specifications were issued for Unit 1 'in 1981 and 1984 and for Unit 2 in 1985.
Subsequently, combined Technical Specifications applicable to Units I and 2 were issued in August 1985 l- as Appendix A to the operating licenses. Since that time, PG&E has implemented a comprehensive Technical Specification Improvement Program for Diablo Canyon designed to reduce the likelihood of r reactor trips and forced outages, while maintaining a high level of operational safety. The program l invohes evaluation of plant systems and operations to identify desired improvements to individual i Technical Specifications. The improvements have included eliminating unnecessary on-line testing, I increasing surveillance test intervals for enuipment, revising tolerances and serpoints, eliminating or L relocating unnecessary specifications, supponing plant betterment design changes, and allowing use of 3 improved fuel designs To date, over 60 amendments to the Technical Specifications have been requested 1- and issued.
4.2.2 Probabilistic Risk Assessment The Diablo Canyon probabilistic risk assessment (DCPRA) is a full scope, Level I risk assessment. It has recently been updated and extended to a Level 2 assessment addressing Individual Plant Examination (IPE) requirements, as specified in Generic Letter 88-20 and NUREG-1335.
I In addition to using the DCPRA to meet IPE requirements, the DCPRA is used, as appropriate, for other p goals and objectives, such as to:
- SupponTechnical Specification changes, especially changes to allowed outage times and surveillance test imervals
- Evaluate continued plant operation under nancoriformini conditions by quantifying effects on plant
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- Evaluate certain proposed design changes that are intended to improve plant safety or reliability
- Provide insights to the Training Department for use in developing operator training plans ,
) For these applications, it is necessary to maintain an up-to-date PRA. Thus, a process has been I developed and implemented to update the DCPRA on a periodic basis (every 18 months) to ensure it represents the current plant configuration. The update process will include revisions to:
- DCPP-specific initiating event data
- DCPP component failure rate data
- DCPP maintenan:e duration and frequency data
- DCPRA models and databases based on a review of DCPP design changes and plant procedure 3 changes In summary, the DCPRA wi!! be used to continually enhance and improve operation throughout the proposed 40-year operating license terms.
D 4.2.3 Surveillance and Maintenance Programs In accordance with the Technical Specificatios and the requirements of Title 10 of the Code of Federal Regulations, Diablo Canyon has established programs for maintenance and surveillance of safety-related equipment. These programs include the Inservi:e Inspection OSI) Program, Inservice Testing OST)
) Program, Environmental Qualification (EQ) Program, and Maintenance Program. These programs assure that any significant degradation of plant equipment will be promptly idcntified and corrected throughout the proposed 40-year operating Isense terms.
D 4.2.3.1 ISI and IST Programs The Diablo Canyon ISI and IST Programs were initiated in 1985 for Unit I and 1986 for Unit 2, corresponding to the start of commercial operation of each unit. The ISI and IST Programs comply with the requirements of 10 CFR 50.55a(b)(2) and 50.55a(g). The ISI and IST Programs also comply with the requirements in the Diablo Canyon Technical Specifications. The ISI and IST Programs include j inspection, testing, and maintenance of pressure-retaining components (including their support structures) as required by the American Society of Me:hanical Engineers Boiler and Pressure Vessel (ASME B&PV)
Code for Class 1,2, and 3 systems. Components that are within the scope of the IST Program are Class 1,2, and 3 pumps and valves that are required to perform a specific function in shutting down the reactor j
or mitigating the consequences of an accident.
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! As part of the inservice inspection effort, a Preservice Inspection (PSI) Program for Class 1,2, and 3 l systems was conducted on Unit I which, to the extent practicable, complied with the requirements of
! ASME B&PV Code, Section XI,1974 Edition including the Summer 1975 Addenda. For the PSI piping examinations, the examination technique of Appendix 111 and the acceptance criteria ofIWB-3514, both from the Winter 1975 Addenda of the ASME B&PV Code, Section XI, were used. For Unit 2, a PSI h Program for Class 1,2, and 3 systems was conducted which, to the extent practicable, complied with the requirements of ASME B&PV Code, Se: tion XI,1977 Edition including the Summer 1978 Addenda.
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The first 10-year inspection interval, which began in 1985 for Unit I and 1986 for Unit 2, meets the requirements of the ASME B&PV Cnde Section XI,1977 Edition including the Summer 1978 Addenda.
When examination techniques differ due to code changes between the PSI and ISI examinations, the ISI exarination data are used as the new baseline.
) ne ISI Program in !udes visual, surfa:e, and volumetric examinations. De surface examinations are done with the liquid penetrant or magnetic particle methods. The volumetric examinations are done using the ultrasonic or radiographi examination methods. De objectives of these examinations are to: ,
- Identify unexpected service-induced component degradation, which would be evidenced by surface
) cracks, wear, corrosion, or erosion e Ixcate any evidence of componer t leakage during system pressure or functional tests
- Verify operability of components and integrity of component supports
) Records of inspections completed under the ISI Program are kept in accordance with the requirements of ANSI Standard N45.2.9 and ASME Section XI and are transmitted to the NRC following each inspection. ,
"' m" 50 55a(g) requires revL;on of the ISI and IST Programs as necessary to comply (to the extent
) pr . ._ . man the limitations of design, geometry, and materials for construction of componenu) with .
the edition of the ASME B&PV Code and Addenda in effect and adopted by the NRC 12 months prior to the start of each 10-year inspection interval. These programs ensure that pressure-retaining ,
components will be adequately inspected, tested, and maintained throughout the proposed 40-year operating license terms. .
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4.2.3.2 EQ Program j Environmental qualification (EQ) is a rigorous program of testing, analysis, and maintenance to confirm [
that electri: equipment relied on in the event of an accident will be capable of performing its design safety l'
) function, despite exposure to the harsh environment resulting from the a:cident.
The EQ Program for DCPP complies with the requirements of 10 CFR 50.49,
- Environmental Qualification of Electri: Equipment Important to Safety for Nuclear Power Plants.* As applied to DCPP, 10 CFR 50.49 requires that electric equipment important to safety and located in a harsh environment :
be environmentally qualified, at a minimum, in accordance with IEEE Trial-Use Standard 323-1971 and
) the Category Il positions in NUREG-0588 ("For Comment
- version, dated December 1979). [
In accordance with 10 CFR 50.49 (1), replacement equipment (for equipment that is required to be environmentally qualified) is required to be qualified in accordance with IEEE Standard 3231974 and i the Category I positions in NUREG-0588 (*For Comment
- version, dated December 1979), unless there are sound reasons to the contrary, ne DCPP EQ Program is a continuing program. The master list of equipment required to be qualified ;
is maintained as a controlled engineering drawing and is revised as plant design changes are implemented.
Detailed EQ files do:ument the results of the testing and analyses that substantiate that the equipment will perform as required in accident environments. Surveillance activities are performed to detect adverse !
) trends in aging or performance. Maintenan:e procedures assure that the qualified configuration of i
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3 equipment is restored after maintenan;e. Equipment that is not designed for the entire 40-year operating license terms is refurbished or repla:ed prior to exceeding its qualified life.
Supplements 15 (dated September 1981) and 31 (dated April 1985) to the Diablo Canyon Safety D Evaluation Report (SER) provide the NRC Staffs evaluation of the Diablo Canyon EQ Program. In Supplement 31 to the SER, the Staff concluded that the Diablo Canyon EQ Program is acceptable and that complian:e with 10 CFR 50.49 has been demonstrated. Supplement 31 also noted that the EQ program had been expanded to in !ude Regulatory Guide 1.97 Category 1 and 2 instrumentation. The NRC Staffs findings were premised on the continuing nature of the Diablo Canyon EQ Program (e.g.,
g replacement of equipment prior to expiration ofits qualified life), without regard to the remaining license period. Thus, no Staff findings regarding the adequa:y of the EQ Program are altered by this amendment request.
In summary, the EQ Program ensures that electric equipment important to safety within the scope of 10 CFR 50.49 will be adequately qua!ified and maintained, and thus capable of performing required O safety functions throughout the proposed 40-year operating license terms.
4.2.3.3 Maintenance Program Maintenance is the integrated means of maintaining the plant material condition throughout its 40-year operating life. It consists of those maintenance, surveillan:e, engineering, and operations activities necessary to control normal degradation that o::urs with time and use.
PG&E has developed and implemented comprehensive programs to manage the effect of aging and service wear on Diablo Canyon's systems, stru:tures, and components (SSCs) throughout the 40-year operating 3 life of the plant. The programs provide the methods for inspection, surveillance, and monitoring of the plant to detect aging effects before the loss of system function and to effect maintenance and component replacement practices for mitigating the effects of degradation.
It is recognized that SSCs can deteriorate as the service life of the plant increases. This recognition forms
, the basis for regulatory requirements, industry codes and standards, and the development of PG&E's
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Maintenan:e Program and its component categories. Design features have been incorporated into Diablo Canyon that provide the ability to test, inspect, and perform preventive and corrective maintenan:e on these SSCs. As a reait. Maintenan:e Program pra:tices provide assuran:e that any unexpected degradation in plant equipment will be identified and corrected.
3 The Maintenance Program was developed with the underlying philosophy that it is necessary to have the requisite administrative and technical controls to ensure that maintenance is performed in a timely and safe manner consistent with applicable license and quality control criteria. This philosophy is important a the purpose of the Maintenan e Program is to maintain the continued functionality and safe operational performan:e of all plant equipment throughout the 40-year operating life.
O The Maintenan:e Program is implemented through procedures developed in conjunction with the Diablo Canyon Quality Assurance Program and PG&E's operational philosophy. These procedures incorporat:
relevant information from the Technical Specifications, design basis criteria, and NRC Safety Evaluation Reports. They provide means to monitcr, inspect, maintain, and test plant SSCs in a programmati:
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Maintenance tasks are categorized as preventive maintenance or corrective maintenance. Preventive and corrective maintenance tasks were identified and initiated during the construction phase by PG&E General :
Construction personnel. Maintenance continuity has been maintained throughout the construction phase, system turnover to plant staff. *he period from operational testing through the stan of commercial perati n, and during commercia operation.
O Maintenan:e tasks for equipment (and systems) are largely based on the applicable manufacturer maintenance recommendations, applicable industry standards and experien:e, plant maintenance history, and PG&E experience. Additionally, Westinghouse (the manufacturer of most major plant components) was present onsite during the construction phase and has provided preventive maintenance information 0 and guidelines for the Nuclear Steam Supply System equipment it supplied. The Westinghouse guidelines include recommended chemistry controls and system layup conditions.
In addition to procedural guidance on specific maintenance activities, maintenance procedure's also provide '
for scheduling, implementing, and documenting activities covered by the Maintenance Program. PG&E installed a computer-based Plant Information Management System (PIMS) prior to the stan of commercial O operations to assist plant maintenance and engineering personnel in there 1ctivities. PG&E also committed additional resources to develop state-of-the-art machine shops, maintenance training facilities (which in:lude extensive laboratories for training technicians), spare parts inventories, and management svstems. Staff resources and personnel training are provided to fully implement and use these suppon resources.
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De 151, IST, and EQ Programs, discussed in Sections 4.2.3.1 and 4.2.3.2, are integral parts of the l Mainoan:e Program. Interdisciplinary activities in the areas of surveillan:e, engineering, and operations, along with a description of program enhancements, are presented below.
Surveillance and Predictive Maintenance O
Surveillance test prccedures (STPs) are implemented by the DCPP Maintenance Groups, Plant Engineering, Operations, Chemistry, Radiation Protection, and Plant Suppon Services groups. De STPs include those tests that are required by Technical Specifications, licenses, and other documents regulating the operation and maintenance of the plant. In addition, predictive maintenance programs include O vibration analysis, acousti: analysis, oil analysis, ferrography, thermography, and component monitoring (using both on-line and off-line performance monitoring systems).
Encineerire i Engineering is integrally involved in all aspects of the Maintenance Program. Engineering support is O integral to the maintenance process in assessing the effectiveness of maintenance, in-depth failure analysis, l
configuration control, and equipment qualification. Engineering provides the design and construction :
suppon for plant improvements and the retirement of equipment that is degraded or approaching )
obsolescen:e. I o Engineering also provides evaluation and analysis support for SSCs with specific aging problems that are identified at the plant or through industry experien:e or research. Engineering has generated the programmatic solutions to mitigate these problems and incorporate solutions into maintenance and ;
monitoring practices to permit management and control of the aging process to within specifications.
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Plant operations is also involved in the maintenance of plant equipment. Diablo Canyon procedures and
, pra:ti es emphasize the need to control the manner in which equipment is operated as well as its environment. Equipment life is prolonged thrcugh proper control of equipment environmental factors, rigid chemistry controls, and proper equipment layup when not in service. Operational procedures have been developed to minimize the stress placed on equipment through transients and to prevent error-induced stress initiators. 06er operational pra:tices contribute to prolonging the service life of equipment such as balan ing the run time on redundant equipment, minimizing unnecessary operation,
@ and monitoring and maintaining external environmental factors including temperature, humidity, moisture, and foreign contaminants.
Enhancements to the Maintenance Procram g The Maintenan:e Program requires continual adjustment as plant-specifi: performance is evaluated, industry experien:e is gained, and new methods and technologies develop. The Maintenance Program at DCPP is a living program that has evolved continuously through the construction and operation phases.
Maintenance Program task improvements have been developed as a result of the equipment failure analysis process (root cause analysis and corrective a: tion program), component f ailure reports, evaluation of industry operating experience, and through better understanding of aginq and failure mechanisms D applicable to Diablo Canyon equipment. These program improvements are proceduralized and receive significant engineering support and evaluation.
Examples of program enhancements include:
g Predictive maintenance programs have been expanded and are managed by a dedicated staff. These programs allow greater reliance on condition monitoring to establish preventive maintenance a:tivities.
- A Reliability Cemered Maintenance Pr) ject has been implemented to optimize the maintenan:e program and improve plant / system reliability and availability. This program uses risk-based methods O for identifying the necessity and importance of maintenance on components, allows optimization of maintenance resour:es, and emphasizes the use of predi:tive maintenance methods.
a Conduct of an additional review of tne latest revisions of component vendor manual maintenance re ommendations to update maintenance tasks and validate the component programs.
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- Establishment of a computerized component database on PIMS that provides the basis for the performan:e of preventive maintenan:e tasks. This effort was an enhancement to the PIMS system, which already contained a component database and history section.
- Implementation of a long-term master s:hedule to plan major component overhaul and replacement D strategies.
- Implementation of major preservation program activities, such as those for concrete structures.
Examples of engineered solutions to improve equipment performance through improved maintenance and surveillance in !ude:
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e Implementation of inspection and monitoring for piping deterioration in the component cooling :
water / auxiliary saltwater (CCW/ASW) system.
e implementation of a program to replace salt water piping with lined steel pipe in the intake area.'
e Upgrading of bolting. t
- Performance of piping inspections (e.g., ASW system) with robot cameras.
3 Development and implementation of component diagnostic or inspection programs; for example: l
- Motor-operated valve testing and diagnostics. !
- Emergency diesel generator engine analyzer diagnostic testing.
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- Check valve inspections, j
- Valve internal leakage evaluations using acoustic monitormg.
3 - Reactor coolant pump seal inspections. t Improvement of steam generator maintenance and inspection practices and implementation of design. [
modifications to inspe:t for and mitigate the effects of various aging mechanisms or mechanical i problems related to generator design and construction. !
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? e I Implementation of a piping erosion / corrosion monitoring program in the condensate, feedwater, and r extraction and heater / moisture separator reheater drain systems using EPRI-developed techniques. l N
In summary, the comprehensive Maintenance Program at DCPP ensures that both normal and unexpected i degradation of plant equipment will be promptly identif ed and corrected. i 4.3 PLANT OPERATING HISTORY
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Diablo Canyon has been in commercial vperation for approximately 7 years. During this time, a !
3 significant amount of data and experience has been accumulated that demonstrates the safety and reliability of the plant. Numerous plant modifications have also been made to improve reliability and upgrade safety-related equipment. Since beginning operation, PG&E has invested about 100 million j
do!!ars per year on an aggressive capital equipment upgrade and replacement program. Operating !
performance, plant modifications, and regulatory performance are reviewed in the following sections. j J
4.3.1 Operating Performance The lifetime capacity factor for Diablo Canyon through its first 7 years of commercial operation is l approximately 77 percent. Achievement of reliable operating performance, in the form of a high capacity-
-l fa: tor, has always been a PG&E goal for Diablo Canyon because PG&E is convinced that high capacity ;
3 factors are the result of a well managed, safely operated nuclear plant. Keeping refueling outages as shon !
as possible through sound outage planning, while safely and effectively performing all necessary work, has also contributed to high capacity factors for Diablo Canyon. Overall industry experience supports 3 our conclusion that units with' consistently high capacity factors, relatively short refueling outages, and i Iow forced outage rates are well maintained, follow good operating practices, and thus can be expected {
? to operate at high levels of safety. !
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O in addition to capa:ity fa: tors, there are other performance indicators commonly used to monitor performance: unplanned automatic reactor trips, unplanned automatic safety system actuations, collective radianon exposure, significant events, safety system failures, forced outage rate, and equipment forced 8 outages /1000 critical hours. As indicated in the NRC's quarterly publication of industry performance indicators, Diablo Canyon currently compares favorably with the industry averages for these indicators.
4.3.2 Reliability / Safety-Related Plant Modifications A number of major plant modifications designed to improve reliability or upgrade safety-related equipment have been made during the 7 years since Diablo Canyon began operation. Several others are planned in the near future. Some of the more 0 mificant mc4ifications include:
- Copper Removal - Ris project involved repla einent of all feedwater heaters and retubing of all O moisture separator reheaters. These changes resulted in the removal of essentially all copper from the secondary side of the plant to increase the life of the steam generators.
- Condensate Polisher Addition - To further increase the life of the steam generators, Condensate Polisher System was added to process se:ondary water by ion-exchange. Regeneration of the ion-exchange resin in the condensate polisher system results in additional solid waste and liquid effluents.
g
- Ammonium Hydroxide Storage - To regenerate condensate polisher resin, a 6,000 gallon bulk storage tank for ammonium hydroxide was added.
- SG Blowdown Rate Increase - To improve secondary water chemistry and thus in:rease the expected O life of the steam generators, the blowdown rate has been increased.
- Control Room Upgrade - A detailed control room design review (DCRDR) was performed in a::ordance with the requirements specified in Supplement I to NUREG-0737. Weaknesses in the man-machine interface between control room operators and equipment were identified in the DCRDR.
g Following review and approval by the NRC, control room equipment upgrades have been implememed.
- High Density Spect Fucl Fool Racks - The original fuel ra:ks in ea:h unit's spent fuel pool were repla:ed with high density ra;ks, increasing the capa:ity in ea:h spent fuel pool from 270 to 1324 fuel assemblies.
O
- Improved fuel Design - ne reactor fuel used in ea:h unit is being replaced with' an improved VANTAGE 5 Westinghouse design. Most of the fuel in ea:h reactor will be of the VANTAGE 5 design by early 1993.
- Baron Injection Tank fBIT) Removal-In response to industry experien:e and NRC recommendations, g the BITS in both Diablo Canyon units were removed from service to reduce the potential for boric acid crystallization in ECCS piping and valves whi:h could potentially have degraded safety-related equipment operability.
- Reduced Boric Acid Concentration - The baron con:entration in the Bori: Acid System has been
- redu:ed from 12 to 4 weight percent to reduce the pmential for boric acid crystallization in safety-related components.
540SS/85K
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- Digital Feedwater Control System - A digital feedwater control system was installed to improve feedwater control performance and reliability. The enhanced feedwater control features provided by ,
this system reduces the likelihood of steam generator level-related reactor trips. Implementation of ' the steam generator level median signal select feature prevents interaction between *he feedwater 3 control and reactor protection systems, which allowed deletion of the low feedwater flow reactor trip ; function.
- Computer Replacement - ne original plant pro:ess computer was replaced with one having improved man-ma: hine interface, greater capacity, faster response time, improved print and report capability, 3 improved retrieval of historical data, and complete redundancy to prevent loss of information due to single failure.
- A7WS Ffitigation System Actuation Circuitry (Ah! SAC) - The AMSAC System was installed in both Diablo Canyon units to ensure rea: tor protection during an anticipated transient without scram '
(A~BVS) event that results in the loss of the secondary side heat sink. AMSAC is designed to trip ) the main turbine, initiate auxiliary feedwater flow, and isolate steam generator blowdown and sample lines during an ATWS with a low steam generater level condition.
- Chlorination System Afodifications - Modin;ations to the Chlorine System include (1) the use ofliquid hypo:hlorite to control microbiofouling instead of gaseous chlorine, (2) implementation of continuous chlorination of the auxiliary saltwater system to control macrobiofouling (invertebrate marine life),
3 and (3) possible use of intermittent injection of a chlorine / bromine mixture to prevent macrofouling in the Circulating Water System.
- Fatigue Afonitoring - PG&E is installing an on-line fatigue monitoring system at Diablo Canyon that will continuously analyze plant operational data to tra:k fatigue usage of critical reactor coolant 3 system components.
- Additional Diesc. (Jenerator Addition of a sixth diesel generator will provide ea:h unit witn three dedicated diesel generators. This will enhance reliability of the onsite power distribution system by eliminating dependen;e on a swing diasel generator and the associated procedural complexities.
Installation, testing, and tie in of the sixth diesel is scheduled for completion in the spring of 1993. J
- Plant Process Protection System Upgrade - This project will upgrade the Process Prote: tion System New by repla:ing the existing HAGE.< 7100 equipment with a Westinghouse Eagle 21 system.
steamline break logi: and steam generator low-low level trip time delay options will be in:luded in the upgrade. These changes will improve the reliability and availability of the Process Protection 3 System. De digital micropro:essor-based system with computer-enhanced testing will also minimiz the likelihood of personnel error during surveillance testing. System installation is scheduled for the spring of 1994 for Unit I and the fall of 1994 for Unit 2.
- R7D Bypass Elimination - This project will repla:e the resistance temperature detector (RTD) bypass loop piping with fast response RTDs installed in the hot and cold legs of the Reactor Coolant Syste J Plant downtime and radiation exposures will be reduced and numerous snubbers can be eliminated.
Installation is scheduled for the spring of 1994 for Unit I and the fall of 1994 for Unit 2.
- Radiation Afonitoring System (RhfS) Upgrade - This proje:t will upgrade the present RMS to improve performan:e, reliability, and capability. Most of the work is scheduled to be completed by 19 540SS/S5K 12 3
i G 4.3.3 Regulatory Performance l The latest Systemati: Assessment of Licensee Performance (S ALP) report for Diablo Canyon covered the
- period of January 1990 through June 1991. The S ALP Board found the performance oflicensed activiti-at Diablo Canyon to be very good, in some cases to be superior, and clearly directed toward safe facility operation.
, 4.4 ASSURANCE OF CONTINUED FUNCTIONAL CAPABILITY OF SAFETY-RELATED COMPONENTS Assurance of an acceptable level of safety throughout the proposed 40-year operating license terms is provided by the continued functional capability of safety-related components. These are components associated with systems designed to prevent or mitigate events that could cause a release of radioa:tivity 9 to the environment. The following discussion reviews such components at Diablo Canyon. 4.4.1 Reactor Coolant System Pressure Boundary y 4.4.1.1 General The mechanical components associated with the Rea: tor Coolant System (RCS) pressure boundary include the rentor vessel, pressurizer, portions of the steam generators, piping, valve bodies, and pump casings. The design of these components incluued consideration of potential effects of age-related phenornena such as corrosion, therma! and vibration fatigue, and radiation-indu:ed embrittlement. Consideration of these O effects was also taken into account when the operating limits and surveillance requirements were established in the Technical Specifications. In accordance with the latter requirements, the RCS is included in the Inservice Inspection Program (see Se: tion 4.2.3.1). Components are located su:h that critical areas are sufficiently a essible for the required inspections and/or tests. The potential for corrosion was accounted for by using corrosion resistant materials in the design of the g plant. All mechanical components that are in conta:t with reactor coolant, except the fuel, are either made of or clad with austenitic stainless steel or nickel-based corrosion resistant alloys. The fuel is clad with Zircaloy. The RCS water chemistry is managed to minimize corrosion. Analysis of the coolant is routinely performed to verify that it meets the spe:ifications for chemistry, radioactivity level, and boron concentration. O Components of the RCS pressure boundary are designed to withstand the fatigue effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. During startup and shutdown, the heatup and cocidown rates are limited to less than 100*F/ hour, consistent with system design specifications and the Technical Specifications. Deviations from these conditions are evaluated to determine complian:e g with code requirements. Plant procedures provide for the ongoing collection of data relative to thermal transients. These data are compared with original design requirements to assure the adequacy of the remaining fatigue life of RCS components. PG&E is in the pro:ess of installing an online fatigue monitoring system at DCPP that will D continuously analyze plant operation data to track fatigue usage of the critical RCS components. 540SS/85K j3
1 l 3 i 4.4.1.2 Reactor Vessel i 1 De Diablo Canyon rea: tor vessels were designed and fabricated in accordance with ASME Boiler and I Pressure Vessel Code, Section III,
- Nuclear Power Plant Components.* They were designed for !
D transients considered to envelope design conditions over a 50-year operating period. To ensure the continued integrity of the vessels during operation, an ISI Program (see Section 4.2.3.1) has been in pla:e since plant startup. The ISI Program requires volumetri examination of each a :essible pressure boundary weld at least once in each 10 year interval. The effects of neutron radiation embrittlement of the vessel beltline region are considered in the design 3 and operation of the units. Compliance with all NRC regu:ations governing vessel integrity has been documemed recently in PG&E's response to Generi: Letter 9241 (PG&E Letter No. DCL-92-150, dated June 30,1992). In addition, PG&E has instituted an Embrittlement Management Plan to manage rea: tor vessel embrittlement throughout the entire operating life of Diablo Canyon. O Pressurized Thermal Shock Following Cy:le 1 for ea:h unit, the neutron tluence at the reactor vessel inner wall was reduced by installing increasingly lower neutron leakage cores, thus decreasing the reactor vessel rate of embrittlement and prolonging vessel life. The Diablo Canyon rea: tor psssure vessel beltline materials have been evaluated a: ording to the NRC's Pressurized Thermal Shock (PTS) screening criterion defined O in 10 CFR 50.61. The Referen:e Temperature for Pressurized Thermal Shock, RTm, has been calculated for each weld metal and base metal in the DCPP beltline regions for neutron Ouences corresponding to 40 operating years. '!he RTm for ali materials will not exceed the screening limit of 270*F for base metal and longitudinal welds and 300*F for circumferential welds. Since all materials meet the screening criterion, neither additional Dux reduction nor plant specific PTS analyses are required O to comply with the PTS rule. Details of the PTS evaluation have been submitted to the NRC (PG&E Lwr No. DCL-92-056, dated Mar:h 6,1992). Based on a conservative Cuence projection for 40 operating years, DCPP will also meet the requirements of 10 CFR 50, Appendix G. Charpy Upper Shelf Energies were determined (FSAR Update Tables 5.2-19A,5.2-19B,5.2-21 A, and 5.2-21B) in a::ordance with Regulatory Guide 1.99, Revision 2. All Diablo O Canyon beltline materials will remain above the 50 ft-lb Charpy Upper Shelf Energy fracture toughness requiremem for more than 40 operating years. In addition, reactor vessel pressure-temperature limits will meet 10 Cf R 50, Appendix G requirements for 40-year operation without requiring plant modin:ation or imposing operational restrictions. O Material Surveillance Procram The toughness properties of the rea: tor vessel beltline material will be monitored throughout the proposed 40-year operating license terms with a material surveillan:e program that meets the requirements of 10 CFR 50, Appendix H. O Re surveillan:e test program for DCPP Unit I complies with ASTM E 185-70, the standard in effe:t ) when the vessel was manufa:tured. Although the Unit I surveillance program was designed prior to the j existence of 10 CFR 50, Appendix H, it does contain the significant features required for later I surveillance programs and will effectively monitor vessel embrittlement throughout the requested license period. The program in:!udes a total of eight surveillance capsules. Three of the eight capsules contain p the limiting weld metal and base metal, correlation monitor material, dosimeters, and thermal monitors. 540SS/S5K 14 0 1
3 The remaining five capsules contain the limitmg base metal, but no wel charpy specimens in the capsules are longitudinally oriented. The Unit 2 surveillance program includes six capsules and .conforms All capsules to ASTM contain the limiting weld metal, which is the limiting beltline material. The b O the same heat treatment, and similar level at plate end of life as the limiting plate. of embrittl ure within 3*F) To date, one surveillance capsule from Unit I and two capsules from U g Analysis of these capsules confirms that the measured shifts,in nil um, ductility r for the limiting materials in predicted by Regulatory Guide 1.99, Rev. 2. the vessels are well within two standard , ym shifts deviations In addition to those required surveillan:e programs, a sttpplemental surveill implemented for Unit I beginning with Cycle 6 in 1992. The supplemental pro O surveillance capsules that will provide additional data to better manage the plant operating life. Neutron Dosimetry Program. Additional measures to monitor DCPP y Units O consists of irradiating and evaluating reactor cavity dosimetr i and axial flux gradient wires attached to the metal rede:tive insulation surround Results obtained are used to confirm and complement surveillan:e capsule data. . The overall program to monitor reactor vessel beltline materials is thorough . O degree of embrittlement of beltline materials over the . 4.4.1.3 Reactor VesselInternals 1 g The design of the rea: tor vessel internals meets the intent of Section 111 of th Vessel Code. The material used for fabrication of most of the re treated, unsensitized Type 304 austenitic stainless steel conforming ns.to ASTM specific We;J fabrication was done using procedures and personnel quali6ed in accordance
- ASME Boiler and Pressure Code. Evaluations performed prior to initial plant st O during normal operation, abnormal operational transients, an Periodic inspections performed under the ISI Program ensure that any significa '
vessel timely internals over the proposed 40-year operating license terms will be de manner. D 4.4.2 Other Mechanical Components The passive mechanical components (tanks, pump casings, and valve bodies systems are designed to meet the intent of Regulatory Guides L26 and 1.29. Consid to possible aging effects including corrosion, erosion, and thermal cycling . v ce fatigue The 540SS/85K 15 D
) , life of these passive components is greater than 40 years, as is the service life for the RCS pressure boundary. Nevertheless, such components are included in the plant ISI and Maintenance Programs, so I that any unexpected degradation will be identified and corrected should it occur. Many of the active (moving or rotating) me:hanical components, such as pumps and valves, are expected to wear and are ) therefore periodically tested and maintained under the IST and Maintenance Programs. Degradation of these components will therefore be identified ant. corrected, and component functional capability will be F maintained. In summary, passive mechanical componerIs tre designed such that they are not expected to be replaced over the 40-year operating license terms, d 41e the functional capability of active components will bc ) maintained through inservice testing, mtinter .nce, and/or periedic replacement. 4.4.3 Electrical Components * ) Electrical components that are required to function in a harsh environment (significantly more severe than the environment that would occur during normal plant operation, including anticipated operational , occurrences) during a design basis event are covered by the EQ Program. The EQ Program complies with the requirements of 10 CFR 50.49, as discussed in Section 4.2.3.2. Safety-related electrical equipment at Diablo Canyon that is not covered by the EQ Program is covered 3 by the plant Maintenance Program (Section 4.2.3.3). As required by the Technical Specifications, equipment is tested periodi: ally in a::ordan:e with surveillan:e test procedures. Equipment is repla:ed as required. 4.4.4 Structural Components 4.4.4.1 Primary Containment , The structural integrity of Diablo Canyon'* containment steel liner, the concrete shell, and internal l concrete and steel structures is assured for a 40. year operating period for the following reasons: )
- The containment liner, the con: rete shell, and internal concrete and steel structures were designed and constructed in a: ordance with Amerien Concrete Institute (ACI), American Institute of Steel Constru: tion (AISC), and ASME Section Vlli Codes to withstand a:cident load combinations, including postulated seismic and loss-of-coolant accident events. Rese load combinations have resulted in much stronger primary containment structures than required to support normal operating h loads.
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- Construction was carried out in accordance with strict construction and quality control procedures.
Actual material testing results indicate a high level of quality in construction techniques and materials used to construct the containment structure. For example, the concrete strength tests for the containment structure indi: ate that the in-situ concrete strength is approximately 25 percent higher than the nominal design values used in the design calculations.
- Through the first 7 years of operation at Diablo Canyon, minor maintenance to concrete surfaces and ,
protective coatings has been performed to ensure the continued structural integrity of the primary j containment. The need for this work has been detected through the following actions
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- Inspections and testing required by plant procedures.
Periodic inspections of protective coatings for concrete and steel surfaces required by plant procedures. Periodic performan:e of local leak rate and integrated leak rate tests required by the Technical ) Specifications. The inspection and documentation process is designed to detect unacceptable concrete cracking, concrete spalling, or deterioration of protective coatings. Repairs are made as needed. Root cause evaluations are performed as required to ensure other potential problem areas are addressed.
- A review of the performance of other principal concrete structures in the U.S. nuclear industry has not yielded experience data that would indicate an adverse impact on the structural integrity of the primary containment for the proposed 40-year operating license terms.
- Based on research presented in ACI-SP-ll7 (Long Term Serviceability of Concrete Structures, 3 American Concrete Institute, January 1989) for nuclear power plant concrete structures, service lives of 60 or more years are possible, provided that appropriate preservation activities are performed.
4.4.4.2 Other Structures ) The structural integrity of Diablo Canyon's other critical plant structures, such as the auxiliary and turbine buildings and the intake structure, is assured for the proposed 40-year operating license terms for the following reasons:
- These critical plant structures consist of reinforced concrete and/or structural steel. These structures
) were designed and constructed in a:cordance with ACI and AISC codes to withstand abnormal design load combinations. These load combinations have resulted in much stronger structures than required to support normal operatiry loads.
- Construction was carried out in a:cordance with strict construction and quality control procedures.
Actual material testing results indicate a high level of quality in construction te:hniques and materials 3 used to construct these structures.
- These plant structures and associated protective coatings are periodically inspected and maintained to ensure continued stru:tural integrity. Through the first 7 years of operation, the intake structure has been the only concrete structure to show some degradition. An evaluation by PG&E showed that 3
minor refurbishment would restore the intake to its original condition. The intake structure is currently being refurbuhed to ensure its continued structural integrity. In addition, PG&E is enhancing marine inspe: tion and repair procedures to better preserve the integrity and serviceability
- of the structure. i l
- A review of the performan:e of other principal concrete structures in the U.S. nuclear industry has not yielded experience data that would indicate an adverse impact on the structural integrity of critical plant structures for the proposed 40 year operating license terms.
- Based on research presented in ACI-SP 117 for nuclear power plant concrete structures, service lives
- of 60 or more years are possible, provided that appropriate preservative a:tivities are performed for ,
these structures. 540SS/85K 17 2 1
i ) 5.0 ENVIRONMENTAL EVALUATION
5.1 INTRODUCTION
l PC, 2 has reviewed and assessed information contained in the " Final Environmental Statement related ) to the Nuclear Generating Station, Diablo Canyon Units 1 and 2" (FES) and Addendum, the Environmental Report and the eight Supplemental Environmental Reports, the Final Safety Analysis Report Update, and other studies and data accumulated over the past years of operation to ensure l thorough and complete evaluation of potential environmental issues related to the proposed 40-year operating license terms. The environmental evaluation is divided into six additional sections. In Section 5.2, systems and programs for environmental control and monitoring are evaluated to ensure they meet applicable regulatory criteria and show evidence of continued enhancement and effectiveness. The systems and programs evaluated in:!ude the Waste Processing System, ALARA Program, Process and Area Radiation Monitoring System, Radiological Environmental Monitoring Program, and Nonradiological Surveillan:e - ) Program. Section 5.3 presents an assessment of the environmental effects of plant operation during the proposed 40-year license terms to ensure they remain within the bounds of the FES and applicable regulatory criteria and permits, or where appropriate, upper iimits estab'ished from plant operation to date. Topics ) addressed include occupational and offsite radiation exposures, new fuel requirements and spent fuel storage, generation and transponation of spent fuel and other wastes, thermal and ecological effects of the cooling water system, and the poterti9 impact on histori properties. Section 5.4 considers the offsite exposures from releases during postulated accidents. These exposures were previously evaluated in the original FS AR (presently the FSAR Update) where the results were ) found to be within the guideline values in 10 CFR 100. Using projected population size and distribution in the vicinity of the plant for the roposed 40-year operating license terms, it is shora that 10 CFR 100 criteria will continue to be met. L De potential effects of population in:reases on emergency planning are evaluated in Section 5.5, and ) Section 5.6 summarizes the plant modi 0 cations that are environmentally related. Also, the effe:ts of the proposed 40-year operating license terms on decommissioning are evaluated in Section 5.7. t 5.2 SYSTE51S AND PROGRASIS FOR ENVIRONh1 ENTAL CONTROL AND MONITORING ) 5.2.1 Waste Processing System The Waste Pro:essing System is described in Chapter 11 of the FSAR Update. His system, which has been significantly upgraded since the original design, receives, processes, and safely discharges or ) packages radioactive waste for funher offsite processing or disposal. The basic processes used are decay of radioactive isotopes, particulate filtration, selective ion exchange, dewatering or solidation of wet waste, filtration of gases by HEPA filters, release of low activity wastes, and compa: tion of dry active waste. As described in the FSAR Update, the system consists of components such as liquid and gas storage tanks, pumps, compressors, filters, ion exchange vessels, instruments, piping, and valves.
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3 Efnuent releases from the Waste Processing System have always remained well below the liraits contained in the Te:hnical Spe:in:ations, and within the range projected in the Final Environmertal Statement. System performance is monitored, and design and operational enhancements have ben made, where " feasible, to minine efnuent releases. Some examples of enhancements include optimizing the ion exchange resins ano 61ter pore sizes, modifying the spent resin storage tanks, and processing waste more frequently and to lower activity levels. A summary of the quantities of radioactive liquid and gaseous efCuents and solid waste released from the plant is reported to the NRC in the Semiannual Radioactive Efnuent Release Report. O This system will continue to be used throughout the proposed 40-year operating license terms to ensure wastes are safely and effectively processed. 5.2.2 ALARA Program 3 PG&E's management is strongly committed to maintaining personnel exposures to ionizing radiation at DCPP as low as is reasonably achievable (ALARA). As a result of this commitment, a formal organization has been established to review, evaluate, and re:ommend ALARA actions for all radiological work at DCPP. PG&E has established a goal to maintain the performance indicator of plant collective radiation exposure, in person-rem per year, to be within the Institute of Nuclear Power Operations 3 (ISPO) guidennes. Occup-ional exposures resulting from the proposed 40-year operating license terms will remain within the lists of 10 CFR 20. The DCPP ALARA Prcgram established in response to the requirements of 10 CFR 20.l(c) will contribute, in large part, to minimizing the already low levels of occupational exposure at the plant. O. Diablo Canyon's ALARA Program requires that an estimate of the total dose be provided for all Radiation Work Permits (RWPs). If the estimate is less than 1 person-rem, the ALARA controls are established by the person initiating the RWP. If the estimated total dose is at least I person-rem but less than 10 person rem, the job receives an ALARA review by the Radiation Prote: tion Work Planning 3 Group. All jobs estimated at 10 person-rem or greater require an ALARA review by the Radiation Protection Engineering Group prior to thejob being initiated. Jobs estimated at 25 person-rem or greater, or an indnidual dose of I rem or greater, require additional review and approval by the plant ALARA Committee. s The Plant Staff Review Comminee (PSRC) serves as the plant ALARA Comminee, and is composed of personnel from the various Diablo Canyon departments. A separate group, the Joint ALARA Review Comminee (JARC), comprised of Diablo Canyon and corporate management, meets periodically to assess the effectiveness of exposure control methods in keeping personnel exposures ALARA. Furthermore, the JARC assists in developing ALARA policy and procedures, and monitors the implementation of ALARA measures. D The ALARA Program is assessed after each outage in a report that lists exposures incurred on majorjobs and summarizes the ALARA lessons learned for future reference and application. De report allows the JARC to identify ALARA-related inadequacies in designs or pro:edures used for equipment installation, operation, surveillance, and maintenance. The results of these post-job critiques provide knowledge that can be used to improve designs and reduce exposures on the same or similar jobs in the future. $ Similarly, Diablo Canyon a!so sets an annual plant exposure goal based on the outage report and input 540SS/85K 19 3
( ., . from each plant department. Current exposures are periodically reviewed by management to identify ; adverse trends, thus assuring timely corrective action, when necessary. ; Continued compliance with the a' ARA Program at DCPP ensures that personnel exposures to ionizing
). radiation will be minimized over tne proposed 40-year operating license terms.
5.2.3 Process and Area Radiation Monitoring System ; The Process and Area Radiation Monitoring System is described in Chapter 11 of the FSAR Update. ;
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This system monitors radiation levels associated with process systems and areas at various locations in the containment and in the auxiliary and turbine buildings. It is designed for use during normal operation or postulated accident situations and includes equipment' for detecting, computing, indicating, and ! alarming. Periodic testing and inspection of the system assure its functional readiness. l
) A number of radiation monitors and monitoring systems are also provided on process liquid and gas lines [
I that may serve as discharge routes for radioactive materials. The monitors include the following: e Main steam line radiation monitors
- Steam generator blowdown radiation monitor
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e Plant vent radiation monitors
- Liquid radwaste radiation monitors
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- Fuel handling building radiation monitors
- Condenser off-gas monitors ;
PG&E is implementing a program to upgrade the Radiation Monitoring System to improve performance, j reliability, and capability. These upgrades combined with surveillance test procedure requirements ensure !
) the system will provide effective radiation monitoring over the proposed 40-year operating license terms. t i
5.2.4 Radiological Ensironmental Monitoring Program 1
) The Radiological Environmental Monitoring Program is described in Chapter 11 of the FSAR Update.
The program was established prior to the start of plant operation to determine preoperational background _ i levels. The Radiological Environmental Monitoring Program is designed to validate the adequacy of - i j safeguards inherent in plant design and the effectiveness of dose calculations, based on plant emission data and appropriate meteorological and aquatic dispersion models. Emphasis is placed on control at the 1 source, with follow up and confirmation by environmental surveillance. This is accomplished by.
). continuously measuring radiation levels and airborne radioactivity levels and periodically measuring amounts of radioactivity in samples at various locations surrounding the plant. Results from the.
Radiological Environmental Monitoring Program are reported to the NRC in the Annual Radiological Environmental Operating Report. The several types of sample media used correspond to the possible exposure pathways. These pathways
) are direct radiation, inhalation, and waterborne or airborne ingestion. Direct radiation is measured 540SS/S5K 20 )
O continuously by thermoluminescent dosimeters (TLDs). Airborne radioactivity is collected continuously by passing air through a fiber filter in series with a char:oal absorption media. He filter collects particulate radioa:tivity, and the charcoal collects radiciodine. Waterborne radioa:tivity levels are O monitored by taking drinking water, surfa: vater, and naine samples. Ingested radioactivity is collected by obtaining samples of vegetation, milk, and fish. examining the distribution of radionuclides in the environment and lower trophic levels, comparisons are made with the preoperational data to determine if there are any biological or physical cGmpartments O in nature that are a::umulating radioactivity. Similarly, external radioa:tivity measurements taken after plant operation are compared with the average and range of data obtained in the preoperational program. A se:ond comparison of external radiation measurements is made by selecting TLD stations considered to be out of the radiological influence of the plant and using them as a measure of natural backgmund exposure. Dese referen:e stations are compared with the remaining stations to make estimates of any increases in external exposure levels that can be attributed to operation of DCPP. O To ensure that the program continues to include environmental sample locations most likely to detect plant-related radica:tivity, a land-use census is conducted annually. Changes in milk sampling locations may be required following the census based on relative potential doses or dose commitments and the availability of samples. O Continued environmental monitoring and surveillance under 'his program ensures early detection of any increase in exposures over the proposed 40-year operating license terms. 5.2.5 Nonradiological Sursei!!ance Program O 5.2.5.1 Environmental Protection Plan The Environmental Prote: tion Plan (EPP) was developed to provide for the prote: tion of the environment during constru: tion and operation of DCPP. The EPP is Appendix B to the Facility Operating Li:ense and has three prin:ipal obje:tives: (1) to verify that DCPP is operated in an environmentally ateeptable O manner, as established in the FES and other assessments, (2) to coordinate NRC requirements and maintain consisten:y with other federal, state and local requirements for environmental prote:! ion, and (3) to keep the NRC informed of the environmental effects of fa:ility constru: tion and operation and of q a:tions taken to control those effects. Environmental protection activities required by the EPP are l reported to the NRC in the Annual Environmental Operating Report. Any environmental con: erns O identified in the FES concerning water quality issues are deferred to the Regional Water Quality Control Board (RWQCB), as regulated under the National Pollutant Discharge Elimination System (NPDES) permit. The environmental issues in the EPP are categorized as aquati issues and terrestrialissues. The aquati: issues have been deferred to the RWQCB for resolution. Presently, the RWQCB protects the water O quality in the vicinity of DCPP through the provisions of the NPDES permit. Terrestrialissues include erosion control in the vicinity of the transmission line rights-of-way, the controlled use of herbicides on transmission rights-of-way, and the preservation of areas in the vicinity of DCPP of archeological significan:e and the provision of access to these areas by the Chumash Indian Tribe. The EPP contains specifi: requirements with regard to these issues. O
$40SS!85K ,1 0
I
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5.2.5.2 NPDES Permit ' An NPDES permit, No. CA0003751, to discharge Diablo Canyon ef0uents into the Pacinc Ocean has i been issued through the RWQCB. This permit describes all linuid discharges from the plant to the Pacific j Ocean and Diablo Creek to include cooling water, steam ge,.erator blowdown, makeup water systems, ' l seawater reverse osmosis system blowdown, radioactive waste treatment system, intake screen wash, biology laboratory discharge, and storm water runoff. Specific efnuent limitations and monitoring l requirements are established for various parameters depending on the probable constituents. These l limitations and requirements are based on Titles !!! and IV of the Federal Clean Water Act in accordance I with 40 CFR 122125 and 423, the California Water Code, and California Water Quality Control Plans. 3 Biological investigations, whi:h began in 1966 in the vicinity of Diablo Canyon, are used to support t compliance with NEPA and the Cuan Water Act. These studies are also used to ensure protection and propagation of the balan:ed, indigenous communities of shellfish, fish, and wildlife in the Pacific Ocean in and round the Diablo Canyon vicinity. NPDES permits are issued to Diablo Canyon every 5 years. 3 These permits contain requirements. as appropriate, for contiaued monitoring to ensure protection of the ' environment. The extensive studies to date have not identined any additional con: erns with relation to i the cooling water discharge to the Paci6: 0:ean. Required monitoring studies have been reduced in scope as a: cumulated information supports the absence of any significant effects. , D 5.3 ENVIRONMENTAL IMPACT DURING NORMAL OPERATION 5.3.1 Occupational Radiation Exposure ; Diablo Canyon's o::upational exposure trend and comparison with the INPO 3-year industry average goal for PWRs are shown in Table 5-1. Diablo Canyon's average annual occupational exposures have consistently been below the INPO goal of 288 person-rem. As shown, aggressive implementation of the Al hA Program has resulted in a distinct reduction in exposures for the most recent 3-year perin i compared with the previous two averaging periods. ' Table 5-1 also shows the projected o::upational exposure averages per unit through the year 2000. The D proje: tion is based on resistance temperature detector (RTD) t,ypass elimination in 1994, and continued I operation with an 18-month fuel cycle. Assuming 3.5 person-rem per month for non-outage periods, and , savings of 50 person-rem per outage following RTD bypass elimination, the levelized 3-year average dose is projected"to de:rease to 150 person-rem per year per unit by 1997. Subsequent annual exposure averages can reasonably be expe:ted to remain approximately at this level, and within the INPO 1995 g 3-year industry average goal. These projected exposures are significantly less than the 450 man-rem per , year per unit values estimated by the Staffin the FES Addendum for Diablo Canyon. Given Diablo Canyon's current trend of decreasing refueling outage exposures and continued emphasis on effective refueling and outage management, it is expected that the base occupational exposures established above will serve as a realisti: estimate through the proposed 40-year period of operation. ' ) , 5.3.2 Offsite Radiation Exposure i The proposed 40-y ear operating li:ense terms will have no significant impa:t on the capability of the plant
) to maintain routine releases of radica:tive materials Oiquid and gases) to the environment in compliance with 10 CFR 20.l(c) and 10 CFR 50, Appendix 1. This conclusion is based on the calculation of annua!
l l 540SS/S5K 22
) doses to individuals and population groups over the past 5 years of plant operation as reported to the NRC in Semiannual Effluent Release Reports Table 5-2 provides a comparison of Appendix I radiation exposure limits and a:tual maximum exposures based on plant operating data. 3 J TABLE 5-1 ) Diablo Canyon vs. INPO Industry Goal Average Annual Occupational Exposure ) Refueling Total Dose (nerson-rem ner reactor unit) Year Outages DCPP 3-Yr Average INPO 3-Yr Average Goal 3 1986 1 151 288 1987 1 168 288 1988 2 253 288 1989 1 275 288 1990 1 269 288 ) 1991 1992* 2 1 214 199 288 288 1993* 1 195 288 1994* 2 218 288 1995* 1 202- 185 1996* 1 188 185 3 1997* 2 150 185 1998* 1 150 185 1999* 1 150 185 2000* 2 150 185 )
- Projected, based on:
- 18-month fuel cycle operation
- 3.5 person-rem per non-outage month 1993 based on 80% of 1992 due to dose rate differences between units 3
- 50 person-rem savings per outage due to RTD bypass elimination in 1994
) 5408S/85K 23 )
i i O TABE 5-2 Comparison of Offsite Appendix I Radiation g Exposure Limits and Actual Data H l 10 ) l DCPP 5-Year Percent of Parameter Appendix I Maximum Appendix ! Dose Limits Individual Dose Dose Limit
- 0 (mrem) (mrem)
L. quids s3 0.031 1.04 i Gases s 10 0.212 2.16 lodines and s 15 0.027 0.18
- O Particulates s
With regard to the plant's liquid eftluent, a review of the actual liquid effluent since 1986 indicates that the maximum individual dose averages no more than 0.013 mrem per year. This represents only 0.43 percent of the Appendix I objective doses. From this operating experience, the plant has clearly
- O demonstrated its capability to process waste to a degree that ensures continued compliance with
- 10 CFR 50, Appendix 1, design objectives.
i Gaseous eftluents from Diablo Canyon have also resulted in doses far below Appendix I objectives. Based on recorded noble gas eftluents over the past 5 years, the calculated maximum w) ole body dose i iO was, on average, only 0.07 mrem per year, or approximately 0.7 percent of the Apper. dix I dose objective of 10 mrem per year. Since this average maximum whole body dose occurred at the site bc.mdary, the plar.t's contribution to the total 50-mile population dose is insignificant in comparison to , th.9 from background radiation. j For iodines and particulates, dose calculations based on the actual releases over the past 5 years indicate
- O that the average maximum annual organ dose is only 0.016 mrem per year. This represents just 0.11 percent of the Appentlix 1 dose criterion of 15 mrem per year to any organ. As with the noble gases, i
the r!cses from iodine and particulates also represent an insignificant frxtion of the total person-rem exposure the same 50-mile population receives each year from background radiation. !O Based on the design and performance history of the Waste Processing System, it is expected that offsite radiation exposures will remain within the plant's ALARA criteria through the proposed 40-year operating ' license terms. The projected exposures are also well within the offsite exposures estimated by the Staff , in the Diablo Canyon FES. It is important to note that the ALARA criteria are formally incorporated into the plant's operating Technical Specifications. Furthermore, the plant's contribution to the local l population dose within a 50-mile radius is expected to remain insignificant in comparison to that from
- O background radiation.
1 3
. S.3.3 Uranium Fuel ;O The DCPP reactor fuel is in the form of sintered UO: pellets containing an initial enrichment of U-235 i less than 4.5 percent by weight. The fuel pellets are contained in Zircaloy rods. The fuel parameters ;
540SS/85K 24
- O
)
- meet the screening criteria of 10 CFR St.52(a)(2), except for fuel enrichment, which may be as much as 0.5 weight percent higher in the DCPP fuel rods.
The ensironmental effects of extended fuel burnup and higher initial enrichment are addressed by the ) NRC in a Notice of Environmental Assessment and Finding of No Significant Impact published in tl. Federal Register on February 29,1988 (53 FR 6040). This notice stated that the NRC's environmental , assessment of extended fuel burnup and higher enrichment fuel is complete, and that the environmental 1 impacts summarized in Tables S 3 of 10 CFR 51.51 and S-4 of 10 CFR 51.52 bound the corresponding impacts for burnup levels up to 60 gigawatt-days / metric ton uranium (GWD/MTU) and enrichments up ) to 5 weight percent U-235. The environmental impacts of transportation resulting from the use of extended fuel burnup and higher enrichment fuel were further addressed in the Federal Register on August 11,1988 (53 FR 30355 - NRC Assessment of Environmental Effects of Transportation Resulting From Extended Fuel Enrichment and Irradiation). This notice reiterated the conclusion stated in 53 FR 6040 and further concluded that there ) are no significant adverse radiological or nonradiological impacts associated with the use of extended burnup and/or increased enrichment, and that burnup levels to 60 GWD/MTU and enrichments to 5 weight percent U-235 will not significantly affect the quality of the human environment. Moreover, pursuant to 10 CFR 51.31,
- Determinations Based on Environmental Assessment,' the Commission determined that an environmental impact statement need not be prepared for this action.
)' Each Diablo Canyon reactor comains 193 fuel assemblies. The assemblies consist of fuel rods in a 17x17 . array. About 39 to 46 percent of the fuel assemblies are replaced every refueling. Since issuance of the operating licenses, PG&E has adopted several fuel design changes and improved fuel management schemes. These changes have significantly improved uranium utilization. ) To date, both Diablo Canyon units have been operating with nominal 18-month refueling cycles. It is anticipated that the units will continue to operate on the 18 month refueling frequency through the end , of the proposed 40-year operating license terms. in the Diablo Canyon FES, it was assumed for purposes of estimating the amount of uranium required ) that the plant would operate for 40 years with an 80 percent capacity factor. It was further assumed that the units would be refueled on approximately an annual basis. Since the Diablo Canyon units are refueled approximately every 18 months and improvements in uranium utilization have been made, the total amount of uranium required for the proposed 40-year operating license terms is expected to be less than the amount projected in the FES. ) 5.3.4 Spent Fuel Storage The Diablo Canyon spent fuel pools currently have onsite storage capacity for plant operation through about 2007 while maintaining the capability for a full-core off-load. After 2007, storage space would no . ) longer be available for a full-core off-load. The existing spent fuel storage racks will be filled by 2010. PG&E has a contract with the U.S. Department of Energy for the removal and disposal of Diablo Canyon ' spent fuel. The U. S. Department of Energy currently plans to begin accepting spent fuel for permanent disposal no later than 2010, and is considering providing interim storage in a monitored retrievable storage facility earlier than 2010. ) I 540SSIE5K 25 ) l
- O 5.3.5 Solid Waste 5.3.5.1 Low Level Radioactive Waste O The volume of solid low level radioactive waste generated at DCPP has historically been among the i Iowest in the nuclear power industry. Table 5-3 compares the annual volume of solid waste generated at Diablo Canyon with the PWR industry median. These valaes show that Diablo Canyon has generated i significantly less solid waste compared with most PWRs and has followed the industry trend of reduced volume in re:ent years. During future years of plant operation, reducing solid waste generation will O c ntinue to be emphasized. The maximum solid waste volume for the proposed license terms should be about 75 cubic meters per year average for each unit, which is consistent with the recent rate of solid waste generation.
i TABLE 5-3 , "O Annual Volume of Low Level Solid Radioactise Waste Generated at Diablo Canyon Compared with Median PWR Diablo Canyon Median PWR Year (cubic meterslunit) (cubic meters / unit) 1956 47 198 iO 1987 83 156 1988 105" 128 1989 94 147 { 1990 42 95 jO 1991 95 " 85 Source: INPO, March 1992 Two refueling outages o::urred in these years 2O i i 5.3.5.2 Spent Fuel i The reactor core thermal power for DCPP is 3338 megawatts for Unit I and 3411 megawatts for Unit 2, O satisfying the screening criterion in 10 CFR Sl.52(a)(1), which requires core thermal power to be less than 3800 megawatts. The average level of irradiation of the DCPP discharged fuel assemblies is less than 45,000 megawatt- l days per metri: ton uranium (MWD /MTU). The decay period after fuel is discharged prior to l {o transportation will be greater than 270 deys. The screening criteria of 10 CFR St.52(a)(3) are an average ! level of irradiation not to exceed 33.000 MWD /MTU and no shipments until at least 90 days after discharge from the reactor. 4 540SS/S5K 26
- O
@ l Although the average burnup level of the DCPP fuel assemblies exceeds the screening criterion in 10 CFR 51.52 (a)(2), the environmental effects for fuel burnups to 60,000 MWD /MTU have been evaluated and shown to be acceptable by the NRC. As discussed in Section 5.3.3., the NRC has 8 concluded that there are no significant radiological or nonradiological impacts associated with the use of extended burnup fuel and that fuel burnups to 60,000 MWD /MTU will not significantly affect the quality of the human environment. Furthermore, improved fuel cycle designs should result in less total spent fuel over the proposed 40-year operating license terms than was projected in the FES. O 5.3.5.3 Waste Shipping Radioactive waste shipped from the DCPP site is packaged in solid form in accordance with the requirements of 10 CFR 61 and the screening criterion of 10 CFR 51.52(a)(4). Unitradiated and irradiated fuel from the DCPP site will be shipped in accordance with the requirements of 10 CFR O Sl.52(a)(5). The screening criteria require truct shipment for unirradiated fuel and truck, rail, or barge shipment for irradiated fuel. Depanment of Transportation and NRC regulations provide protection of the public and transport workers from excessive radiation exposure. Protection is achieved by standards and requirements applicable to packaging, limitations on the contents and radiation levels of packages, and procedures to limit the g exposure of persons under normal and postulated accident conditions. Primary reliance for safety in transport of radioactive material is placed on the packaging. The packaging must meet regulatory standards (10 CFR 71 and 49 CFR 173) established according to the type and form of material for containment, shielding, nuclear criticality safety, and heat dissipation. The standards require that the packaging prevent the loss or dispersal of the radioactive contents, retain shielding O efficiency, assure nuclear criticality safety, and provide adequate heat dissipation under normal conditions of transport and under specified accident damage test conditions. 'Ite contents of packages not designed to withstand accidents are limited, thereby limiting the risk from releases that could occur in an accident, and to ensure standards for radiation levels, .emperature, pressure, and containment are met. O 5.3.5.4 Solid Waste Conclusions The amount of nuclear fuel and volume of solid waste resulting from the proposed 40-year operating license terms will continue to be within the limits assumed for the original licensing basis. Because of O improved fuel cycle designs, the total amount of spent fuel produced over the proposed 40-year operating license terms is expected to be less than that originally projected in the FES for DCPP. Based on the above, PG&E concludes that the radiological and environmental impacts from the storage and transportation of irradiated fuel and solid radioactive waste are consistent with the impacts set forth in Table S-4 of 10 CFR 51.52, and the environmental costs will not be significantly affected during the 3 proposed 40-year operating license terms. 5.3.6 Thermal and Ecological Effects of the Cooling Water System The DCPP cooling water system is a once-through system discharging directly into Diablo Cove of the $ Pacific Ocean. The potential ecological effects of the cooling water system are: (1) those resulting from elevated water temperatures in portions of Diablo Cove, (2) entrainment of organisms in the cooling water 540S5/85K 27 D
9 system, (3) impingement of organisms on the intake traveling screens, and (4) scouring effects of the dischrge in the intenidal zone at the point of discharge. 1 These effe:ts have been extensively studied and the study results were considered in issuan:e of the g NPDES Permit and renewals (see Section 5.2.5). The NPDES Permit is conditional upon the discharge complying with provisions of Division 7 of the California Water Code and of the Clean Water Act (as amended or as supplemented by implementing guidelines and regulations) and with any more stringent effluent limitations necessay to implement water quality control plans, to protect beneficial uses, and to prevent nuisan:e. Two of the findings of the California Regional Water Quality Control Bond, Central Coast Region (Bord) as noted in the current NPDES Permit (Bord Waste Dischuge 8 Requirements Order No. 90-09) are:
- Thermal effe:ts on the receiving water, actual temperature i reases, and e:tual temperatures of the discharge are monitored and results correlated and evaluated hs pan of ongoing studies. These studies are being performed as required in Monitoring and Reponing Program No. 90-09. This monitoring
" is developed and reviewed jointly by Board and Department of Fish and Game staff. The results of the studies are useJ to evaluate thermal limits.
- Section 316(b) of the Clean Water Act requires that the location, design, construction and capacity of cooling water intake structures refle:t the Best Technology Available (BTA) for minimizing adverse environmental impa:t.
J An April 25, 1988. study of the cooling water intake structure was submitted to the Bord whi:h concluded the facilities at Diablo Canyon Power Plant reDe:t BTA. Funher, the Monitoring and Reporte.g Program requires PG&E to continue ecological studies as approved by the Executive Of5:er in order to evaluate changes in distribution and abundance of marine plants and animals within the 3 vicinity of the discharge. These operational studies have indicated that the effects of the discharge are consistent with the preoperational studies and mudelling predictions; i.e., that the dischrge would not signWantly affe:t the marine e: ology in the vi:inity of DCPP. The Board and Department of Fish and Same have found the observed th:.nges (mainly in relative abundan:e of species) to be a::eptable. Additional discharge and thermal effects are not anti:ipated based on operational data colle:ted sin:e J 1984. Accordingly, the basis for the Board's order is expected to remain valid when the NPDES Permit is renewed in 1995 and thereafter. 5.3.7 Protection of Historic Properties g' In 1966, Fran:is Riddell conducted a survey for PG&E of approximately 250 a:res to be used as the site for DCPP. In 1968, Greenwood and Associates undenock subsurface investigations for PG&E at six sites within the construction areas for the DCPP facilities and a proposed access road from the plant site to Avila Bea:h. Based on these excavations, Greenwood in '1972 suggested that an area of the DCPP site, SLO-2, was a major village whi:h figured prominently in the social, economic, and political life of the J indigenous occupants of the area. The 1968 excavations at the site resulted in the identincation of one of the oldest culturally strati 6ed sites identiDed to date in San Luis Obispo County. These 1968 studies by Greenwood are discussed in the Diablo Canyon FES where it mentions three archaeologi:al sites in the plant rea were investigated by archaeologists. g in 1978. Greenwood and Associates reponed on a sursey of 90 acres of land thought to be the areal extent of site SLO-2. This repen centained information to be used for the nomination of SLO-2 to the 540SS!S5K ;g O
O National Register of Historic Pla:es. The SLO-2 site, whi:h was determined to be eligible for the National Register of Histori: Pla:es in December of 1982 by application from the Nuclear Regulatory Commission, in:ludes the locations of recorded sites SLO-2, SLO-3, and SLO-8. In a: ordan:e with 36 CFR S00, a determination of no-effect was made by PG&E in hiay 1984 in conjunction with the California Historic Preservation Of6cer. The NRC agreed with the no-effect determination in a transmittal to PG&E on June 25,1984, stating that operation and maintenance of the plant will not affect cultural resources (SLO 2) listed in or eligible for inclusion in the National Register of Histori: Pla:es. 9 Since November 1983, photographs have been taken at regular intervals from 23 stations within the site to monitor any physical changes to the site caused by natural or other processes. Further, a Cultural Resources hianagement Procedure Plan has been drawn up which provides speciS guidelines regarding future management of SLO-2 by PG&E The Archaeological Resources hianagement Plan (ARhfP) and these procedures have been approved by the State Historic Preservation Ofncer as pan of PG&E's compliance with the National Historic Preservation Act (36 CFR 800). In the spring of 1986, unexamined portions (an area of approximately 495 a:res) of the DCPP property were investigated for cultural resources. A total of six prehistoric sites were recorded and/or relocated during this study. This included three previously unrecorded sites, SLO-1161, SLO-1162, and SLO-1163 g and one previously re:orded site, SLO-61. In addition, site forms were completed for Riddell sites 3 (SLO-1159) and 5 (SLO-1160) whi:h had not been recorded with the California Archaeological Survey. PG&E utinues to man:we and prote:t the histt n: propenies at DCPP in consultation with the California State Histori: Preservanon OfG:e and the local Native American communities. As a result of this aggressive management, PG&E con:ludes, as did the NRC in 1984, that operation of DCPP throughout O the 40 year operatir.; li:ense terms will not adversely affe:t any known historic sites. 5.4 EXPOSURE FRO 31 RELEASES DURING POSTULATED ACCIDENTS g The offsite exposure from releases during postulated a::idents has been previously evaluated in the DCPP FS AR Update. The results are a::eptable when compared with the criteria denned in 10 CFR 100. This type of evalunion is a function of four parameters: (1) the types of a:cidents postulated, (2) the radios:tivity release :alculated for ea:h a::ident, (3) the assumed meteorologi:al conditions, ad (4) population distribution versus distan:e from the plant. On the basis of the safety assessment in Section 4.0, it can be concluded that neither the types of accidents nor the calculated radion:tivity releases O will change through the proposed 40-year operating license terms. Furthermore, the site meteorology, as denned in the FSAR Update, is essentially constant. Thus, population is the only time-dependent parameter. It is important to note that there is no expected change in land usage during the license terms that would affect offsite dose cal:ulations. g The population si:e and distribution in the vicinity of the plant have been reviewed several times sin:e J the constru: tion permit was issued: in the FES in 1973 (through the year 2000), in the original FSAR l in 1974, in the 1985 FSAR Update using 1980 census data, again in late 1991, and in a review of the l California Depanment of Finance projections through 2025 performed for the purpose of the proposed li:ense terms. The Department of Finance projections indicated that a compound average growth rate of 215 percent is expe:ted for the 50 mile radius area around Diablo Canyon through the year 2025. 9 Tab!e 5-4 summarizes the projected population size and distributions. 540SS 55K ;9 0
/
TABLE 5-4 J Summary of Population Projections for the Diablo Canyon Vicinity J Area Original FSAR Revised FSAR (miles) (1974) (1985) Current Current 2010 2010 2010 2025 J 0 - 6* 29 26 100 100 6 - 10 18,992 36,126 36,403 46,480 ! O - 10 19,021 36,152 36,503 46,580 10 -50 508,130 438,035 555,108 730,566 3 591,611 777,146 0 -50 527,151 474,187
- Reflects Low Population Zone J The plant exclusion area depicted in Figure 5-1 will remain uninhabited through the proposed license terms. The activities within the exclusion area only pertain to normal plant operation. No part of the !
ex:hision vea will be sold, and no structure will be located within it except those owned by PG&E or , a related company and used in conjunction with normal utility functions. No residences will be permitted on the site. J The Low Pcpulation Zone (LPZ) for DCPP is the area included within a 6-mile radius of the site, as shown in Figun 5-2. This land area was sele:ted because it meets the requirements of 10 CFR 100 with respe:t to proximity to the nearest population center. The major population centers (with populations of 25,000 or more) currently within 50 miles of Diablo 3 Canyon are Lompo: (1990 population 37,649), about 45 miles to the south-south-east; Santa Maria (1990 population 61,284),29 miles to the south-east; and San Luis Obispo (1990 population 41,958),10 miles to the east-north-east. Accordingly,10 miles is currently the Population Center Distance (PCD). However, population projections indicate that the community of Baywood-Los Osos (8 miles to the north) l v ill reach a population of 25,000 in approximately the year 2020. The 2020 projected population for 3 Baywood-Los Osos is 26,844. Accordingly, the PCD would become 8 miles in approximately 2020. i Federal Regulation 10 CFR 100.11(a)(3) provides that the PCD be at least 1-1/3 times the distance from the reactor to the outer boundary of the LPZ. The community of Baywood-Los Osos, at 8 miles north, j will satisfy this rule as the projected PCD in the year 2020, thus requiring no change in the definition of the current LPZ now or during the proposed 40-year operating license terms. The changes projected for the population distribution through 2025 will not impact the boundaries used for existing accident analyses. The current exclusion area boundary, LPZ, and nearest population center , distance will continue to meet the requirements of 10 CFR 100.11(a)(3) for the proposed 40-year license I terms. Accordingly, the proposed 40-year operating license terms will not affect previous conclusions greached in the Staffs SER and FES on the potential environmental effects of offsite releases from ) postulated a::idents. 540SS/S5K 30 )
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I S 5.S ENVIRONMENTAL RELATED PLAhT MODIFICATIONS Several environmental related plant modifications have been made since issuance of the FES and Addendum. These changes funher reduce the environmental impacts associated with DCPP operation and in:lude the following: O e Wastewater Holding and Treatment (%HA T) System - ne WH AT System includes storage tanks and associated equipment for improved storage and treatment of wastewater. He tanks are placed in a concrete retention basin for secondary containment.
- Wastewater Routing - Pumps and piping were added to route wastewater from the turbine building sump to the WHAT System, allowing for treatment prior to discharge.
- Ha:ardous Waste Storage - A building was provided for the storage of containerized hazardous waste prior to offsite shipment. Other areas in the plant are designated for hazardous waste storage. These storage areas are provided with fire protection equipment and secondary conta'mment features to O prevent the release of hazardous waste to the environment.
- OilSpill Prevention - A spill prevention pond was constructed at the 230 kV switch yard to prevent oil from transformer leaks from entering Diablo Creek.
- Seawater Reverse Osmosis System - A Seawater Reverse Osmosis System was provided as an C additional source of makeup water for the plant, effectively replacing the seawater evaporator described in the FES.
- Discharge Structure Modifcations - To reduce the, peak temperature of cooling water being discharged from the plant, modifi:ations were made to the discharge structure. These modifications O allow for better mixing of water between the two units and included cutting openings in the c'entral wall of the discharge structure and modi 6:ation to the weirs.
e intate/ Discharge Temperature Monitoring - An improved temperature monitoring system was installed to measure the cooling water teniperature at the intake and discharge structures. This system g initiates an alarm in the plant control room when a set temperature difference is exceeded.
- Expanded Sewage Treatment - To allow for increases in plant staffmg, an expanded sewage treatment system was constru:ted that provides for secondary treatment of effluent. He system discharges to the cooling water system and is monitored as required by the NPDES permit. An expanded lea:h field was provided as a ba:kup for the secondary treatment system.
O
- Chlorination System Modifcations - Modifications to the Chlorine System include (1) the use ofliquid hypochlorite to control microbiofouling instead of gaseous chlorine, (2) implementation of continuous chlorination of the auxiliary saltwater system to control macrobiofouling (invertebrate marine life),
and (3) possible use of intermittent injection of a chlorine / bromine mixture to prevent macrofouling in the Circulating Water System.
- Dechlorination Process - De:hlorination, by injection of sodium bisulfite, is now used to reduce free chlorine levels in discharged cooling water. .
)
- Additional Steam Equipment - To provide for additional steam needs during plant startup testing, a l p second auxiliary boiler and two underground fuel tanks were added. While the combined air j emissions from both boilers exceed the projections in the FES, both have been permitted by the lo:al l Air Pollution Control District and operate within the District's emission limits. j l
540SS/E5K 33 0
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- Liquid Radwaste Modi # cations - The DCPP Liquid Radivaste System has been modified by providing l l additional processing capability via ion-exchange and filtration. Based on operational problems ,
j i experienced at other plants, the radwaste concentrator was never operated and has been abandoned i in-pla:e. Piping modifications were made to allow for the use of mobile processing equipment if j required. j
- Solid Radwaste Processing - Regulatory changes to allowable solid radioactive waste forms resulted in removal of the installed radwaste solidification equipment. Presently, contractor-provided mobile. i equipment is used to solidify or dewater radioactive waste. j
?
- Solid Radwaste Volume Reducilon - To reduce the volume of solid radioactive waste being produced, two radioactive waste compactors were installed. '!
- Hazardous Materials Storage - A warehouse was provided for storage of hazardous materials needed in plant operation. -
0 :
- Makeup Water Treatment - Contra: tor-supplied equipment is used to treat makeup water used at the 1 plant. I Most of these plant design modifications and changes have had a direct positive impact on the 'l envir nment; f r example, chemical discharges have decreased and spill prevention has improved. :
O Changes have been conducted in accordance with approved procedures, current license conditions, ! Technical Specification requirements, and environmental permits. Additional plant modifications and i changes may be 5plemented during the proposed 40-year operating license terms.~ For example, the ! Radiation Monitoring System is presently being upgraded to improve performance, reliability, and' , capability. Based on past experience, future changes are r.ot expected to have a significant negative j impact on the environment and should provide improved means to monitor potential environmental ! effects. . 5.6 DECOMMISSIONING In accordance with 10 CFR 50.33(k), PG&E submitted a decommissioning report to the NRC that provided reasonable assurance that sufficient funds will be available to decommission the plant. Additionally, PG&E prepared a preliminary decommissionirq cost study for DCPP. This study was last , updated in May 1991 to reflect recent changes in costs and schedules. O. The decommissioning study provides cost, schedule, waste generation / disposition, and radiation exposure , estimates associated with the decommissioning of the DCPP nuclear units following cessation of ; operations. The alternatives evaluated were DECON (prompt removal / dismantling) and SAFSTOR l (mothball with delayed dismantling). The plant retirement dates were taken as 30 years following ; commercial operation. This timeframe was used as input in scheduling analysis. O The decommissioning study did not determine the incremental impact on decommissioning due to the .! proposed 40-year license terms. However, PG&E anticipates that the proposed 40-year operating license
- l. terms will have a negligible incremental environmental and cost impact on decommissioning the DCPP .;
units. A reactor's activation product inventory rapidly builds up and reaches an equilibrium level, so that : ( , after 10 years of operation the inventory is approximately 90 percent of the total expected after 40 years l h of operation. As a result, most of the fa: tors affecting costs for decontamination and decommissioning, !- 1 i ; 540?S/ESK 34 ) C. t 1 p ear- y ,-- W - - - ^ + -
J as well as occupational exposure, arise early in a reactor's operating life. Thus, the proposed 40-year operating license terms will not add significantly to the cost and effort to decommission the plant. Moreover, it was concluded in NUREG-0586, " Final Generic Environmental Impact Statement on
, Decommissioning of Nuclear Facilities," that decommissioning is not expered to significantly impact the environment. The NUREG also categorically excludes power reactors from tne mandatory Environmental Impact Statement requirement for decommissioning. The other key conclusions of this report were:
- Decommissioning at the present time can be performed safely and at reasonable cost.
O e Decommissioning of nuclear facilities is not an imminent health or safety problem. e Decommissioning of a nuclear fa:ility generally has a positive environmental irmact. Therefore, PG&E concludes that any changes in the environmental impact associated with decommissioning after expiration of the proposed 40-year license terms will be negligible.
-,J 6,0 NO SIGNIFICANT HAZARDS EVALUATION PG&E has evaluated the no significant hazard considerations involved with the pwposed amendment, D focusing on the three standards set forth in 10 CFR 50.9:(c) as quoted below:
The Commission may make a final determination, pursuant to the procedures in 6 50.91, that a proposed amendment to an operating license for a facility licensed under i 50.21(b) or i 50.22 or for a testing facility involves no significant hazards q consideration, if operation of the facility in accordance with the proposed amendment
~
would not:
- 1. Involve a signifi: ant increase in the probability or consequences of an accident previously evaluated; or O 2. Create the possibility of a new or different kind of accident from any a:cident previously evaluated; or
- 3. Involve a signifi: ant redu: tion in a margin of safety.
The following evaluation is provided for the no significant hazards consideration standards. 3
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed 40-year operating license terms do not affect the probability or consequences of an J accident previously evaluated since the requested extensions entail no physical change in the plant ecuipment or operating procedures and the FS AR Update safety analyses are based on 40-year plant operation. Surveillance and maintenance pra:tices, as well as other programs such as environmental qualification of equipment, ensure timely identification and correction of any degradation of safety-related plant equipment. The long term integrity of the reactor vessels has been recently reevaluated
, using currently acceptable NRC calculational methods and best available DCPP-specific data. The evaluation results demonstrate, as before, that both reactor vessels are safe for normal operations l
540SS/85K 35 3 I
i n excess of 40 years. Also, the offsite radiation exposures resulting from postulated accidents have O been reanalyzed using population projections for the propos ed 40-year operating license terms. The calculated exposures are not significantly different from those documented in the FSAR Update and are well within 10 CFR 100 guideline values. I herefore, the proposed changes do not involve a signifi: ant increase in the probability or , O consequen:es of an a:cident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The possibility of a new or different kind of a:cident is not created by the proposed 40-year ; O operating license terms since at least 40-years operation was assumed in the design and construction of DCPP Units 1 and 2. The plant Maintenance Program is designed to both maintain and determine the need to replace safety-related components. Thus, any degradation that might possibly create a new or different kind of postulated accident would be detected and corrected before the occurrence of such an event. O Therefore, the proposed changes do not create the possibility of a new or different kind of accident ; from any a cident previously evaluated.
- 3. Does the change involve a signifi: ant reduction in a margin of safety? ;
O The proposed 40-year operating license terms do not involve a signifi: ant reduction in a margin of i safety since degradation of safety-related equipment will be identified and corrected by ongoing ! surveillan:e and maintenance era:tices. Existing programs, routine maintenance, and compliance with Te:hnical Spe:ificaticns assure that an adequate margin of safety is maintained. These activities will remain in effe:t for the duration of the operating licenses. 7 O Therefore, the proposed changes do not involve a significant redu: tion in a margin of safety. i In conclusion, based on the above safety and environmental evaluations, PG&E submits that the activities associated with this license amendment request satisfy the no significant hazards consideration standards of 10 CFR 50.92(c), and, accordingly, a no significant hazards finding is justified. Such a finding is O consistent with NRC Staff approval of numerous appli:ations of this tjpe. PG&E requests that, if ' necessary in a::ordance with the provisions of 10 CFR 50.91(a)(4), the NRC make a final determination that no significant hazards considerations are involved and issue a license amendment in a:cordante with ! this request. O t i lO l O 540SS/85K 36 O
9 ATTACHMENT B O DIABLO CANYON POWER PLANT UNITS 1 AND 2 MARKED UP LICENSE PAGES , Remove Pace Insert Pace
.)
9, Unit 1 License 9, Unit 1 License 7, Unit 2 License 7, Unit 2 License J J D D D D l 9 l 540SS!S5K O
Unit 1 License Nuclear Plant Physical Security Plan," Revision 11 dated May 27, , 1982, as revised July 19, August 12, September 17, 1982; February 4, August 3,1983; January 11, February 6, March 19, April 19, August 29, 1984; "Diablo Canyon Nuclear Plant Guard Training and Qualification ; Plan," Revision 2 dated February 4,1983 as revised August 29, 1984; '
"Diablo Canyon Nuclear Plant Safeguards Contingency Plan," Revision 2 dated February 4,1983 as revised August 3,1983, August 29, 1984.
7' F. Antitrust Pacific Gas and Electric Company shall comply with the antitrust conditions in Appendix C to this license. 3 G. Reporting PG&E shall report any violations of the requirements contained in Sections 2.C(3) through 2.C(10), 2.E and 2.F. of this License within 24 hours. Initial notification shall be made in accordance with the provisions of 10 CFR 50.72 with written follow-up in 3' accordance with the procedures described in 10 CFR 50.73 (b), (c), (d) and (e). H. Financial Prctection PG&E shall have and maintain financial protection of such type and 3' in such amounts as the Co r-ission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims. I. Term cf License " This License is effective as of the date of issuance and shall expire at midnight on Agil 23, 200:. Segfew her 2 '2., 2.o 7 l . FOR THE NUCLEAR REGULATORY COMMISSION
/
D y L - arold R. Denton,' Director Office of Nuclear Reactor Regulation Attachments: 3 1. Appendix A - Technical Specifications
- 2. Appendix B - Environmental Protection Plan
- 3. Appendix C - Antitrust Conditions Date o r Issuance: November 2, 19B4 6
0
1 Unit 2 License
)
) H. Financial Protection PG&E shall have and maintain financial protection of such type and in such amounts as the Comission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover ). public liability claims.
- 1. Tem of License This License is effective as of the date of issuance and shall expire at midnight on Oe e.-?wr ;, 2010. Ap4 / 2.(, Zo2f. ,
FOR THE NUCLEAR REGULATORY COMMISSION
/f k Harold R. Denton, Director
) Office of Nuclear Reactor Regulation i Attachments: !
- 1. Appendix A - Technical Specifications (NUREG-1151)
- 2. Appendix B - Environmental Protection Plan l 1
- 3. Appendix C - Antitrust Conditions '
l Date of Issuance: August 26, 1985 ) ? )
/ )-
EXHIBIT 6 PROGRAM DIRECTIVE PD MA1 - MAINTENANCE l 3 J J' b a D D
) !
1
ce****e************* FOR INFORMATION ONLY ******************** D PACIFIC GAS AND ELECTRIC COMPANY NUMBER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PROGRAM DIRECTIVE PAGE 1 0F 16 TITLE: MAINTENANCE APPROVED: 12/12/92 12/12/92 - SVP&GM, NPGBU DATE EFFECTIVE DATE i 3 SPONSORING ORGANIZATION: DCPP ELECTRICAL MAINTENANCE ; CLASSIFICATION: QUALITY RELATED ! TABLE OF CONTENTS SECTION DESCRIPTION PAGE 3 i 1.0 PROGRAM 0VERVIEW................................................... 2 l 2.0 APPLICABILITY...................................................... 4 ; 3.0 DEFINITIONS........................................................ 4 4.0 SPECIFIC PROGRAM OBJECTIVES........................................ 5 5.0 GENERAL PROGRAM REQUIREMENTS D 5.1 Maintenance Program and Support............................ 6 5.2 Assessment of Maintenance Program Effectiveness........... 7 5.3 Control of Maintenance Activities.......................... 7 5.4 Work Management............................................ 9 5.5 Preventive Maintenance Program............................. 9 5.6 ASME XI Repairs / Replacements.............................. 10 D 6.0 MANAGEMENT PROCESS MODEL 6.1 Mode 1..................................................... 11 1 6.2 Interfaces................................................ 11 - 7.0 RESP 0NSIBILITIES.................................................. 13 i 8.0 IMPLEMENTING DOCUMENTS 8.1 Interdepartmental Administrative Procedures (IDAPS)....... 15 3 8.2 Departmental Level Administrative Procedures (DLAPS)...... 15 , 9.0 REC 0RDS........................................................... 15 10.0 ATTACHMENTS....................................................... 16
11.0 REFERENCES
........................................................ 16 12.0 SPONS0R........................................................... 16 b i l MA100100.SF 1 D
******************** FOR-INFORMATION ONLY ******************** !
PACIFIC GAS AND ELECTRIC COMPANY () NUCLEAR POWER GENERATION BUSINESS' UNIT NUMBER MA1 REVISION O TITLE: MAINTENANCE PAGE 2 0F 16 C) 1.0 PROGRAM OVERVIEW r 1.1 Scope ! This Program Directive addresses NPGBU's program for the planning, , scheduling, and performance of preventive and corrective maintenance on () plant equipment, and the performance of repairs and replacements of
+
components covered by Section XI of the ASME Code. It applies to safety related systems, structures, and components (SSC) and other i specified equipment that affect safe and reliable plant operation. The ' overall objective of the maintenance program is to prevent the i' degradation or failure of plant equipment. The hierarchy of procedures .() associated with this PD is illustrated in Figure 1. ' Effective maintenance is essential to plant safety and to the reliable and efficient operation of DCPP. Effective maintenance can reduce the ' frequency of events and challenges to safety systems as well as enhance f the operability, availability, and reliability of such systems.
- <3 Maintenance also helps to minimize failures of non-safety related
equipment that could result in off-normal plant conditions or affect ! l ' safety system performance. Finally, effective maintenance will help ' ensure that assumptions and margins in the original design basis and other engineering evaluations (e.g., probability risk assessment (PRA), ) individual plant evaluations (IPE), etc.) are maintained. j 'O Good maintenance is-also an important determinant of plant performance. The time required to perform maintenance activities during normal operations and outages, and the ability to implement maintenance prior to loss or significant degradation of equipment function, are critical , to overall plant availability. These considerations provide a ' C) c mpelling basis for implementing effective maintenance management with clear and specific delineation of policy considerations and responsibilities. O i l l lC) .
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER 3 NUCLEAR POWER-GENERATION BUSINESS UNIT REVISION MAI 0 PAGE 3 0F 16
-TITLE: MAINTENANCE 3 Figure 1 i
PD MA1 Hierarchy of Procedures ? i PD MAI
- Maintcnance D
l Y IDAPs o Maintenance Program & Support o Monitoring of Maintenance Program Effectiveness o Preventive Maintenance Program MOV Maintenance o 3 o ASME XI Repairs / Replacements ) l DLAPs o Department Specific Administrative Controls l a MA1001.00.SF 3 J
I
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PACIFIC GAS AND ELECTRIC COMPANY NUMBER 3 NUCLEAR POWER GENERATION BUSINESS UNIT MA1 REVISION 0 i PAGE 4 0F 16 ! TITLE: MAINTENANCE : O 2.0 APPLICABIllTY This PD is applicable throughout the NPGBU and other PG&E organizations that may provide support services for the DCPP maintenance process. Due to the i extensive role of contractors in the performance of maintenance, their i O activities must also conform to the content of this PD. This PD is not applicable to Humboldt Bay Unit 3 (HBPP-3). The maintenance program at HBPP-3 will be in accord with PD HB1, "SAFSTOR - Humboldt Bay Power Plant." 3.0 DEFINITIONS , 3.1 Corrective Maintenance - Actions that restore, by repair, overhaul, or : O replacement. the capability of a failed system, structure, or component , to perform its des *gn function within acceptable criteria. 3.2 Emergency Maintenance - Immediate corrective maintenance required to prevent or mitigate the release of radioactive material, hazards to personnel, or extensive equipment damage. ; O 3.3 Maintenance History Program - A program to document data, provide historical information for future maintenance planning and trending. , 3.4 Preventive Maintenance - Periodic, predictive, or planned maintenance performed prior to failure of a system, structure, or component to O extend its service life by controlling degradation or failure.. > 3.4.1 Predictive Maintenance - A form of preventive maintenance ! performed periodically or continuously to monitor, inspect, i test, diagnose, or trend a system's, structure's, or ' component's performance or condition indicators; results i
- 9. indicate or forecast functional ability or the nature and '
schedule of planned maintenance prior to failure. ! 3.4.2 Periodic Maintenance - A form of preventive maintenance f consisting of servicing, inspection, testing and replacement at i predetermined intervals of calendar time, operating time, or g number of cycles. i 3.4.3 Planned Maintenance - A form of preventive maintenance consisting of refurbishment, overhaul, and replacement that is , scheduled and performed prior to system, structure, or l component failure. j O 3.5 Quality Related - For a definition of Quality Related, please refer to . PD OMS, " Quality Assurance Program." l 3.6 Repair - Restoration of acceptable functional ability in a degraded system, structure, or component that has failed. O MA100100.SF 4 0
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1 PACIFIC GAS AND ELECTRIC COMPANY 3 NUCLEAR POWER GENERATION BUSINESS UNIT NUMBER REVISION MA1 O i 1 PAGE 5 0F 16 TITLE: MAINTENANCE i l O 3.7 Repeat Maintenance (Rework) - Any additional maintenance required, during or following completion of a maintenance activity, because the work package was inadequate, the maintenance perfonned was inadequate, or there was component failure within 3 months of return to service. 3.8 Replacement - Replacing an undegraded, degraded, or failed system, O structure, or component or a part thereof with another that complies with design requirements.. , 3.9 Root Cause - The underlying cause of the problem, failure, or degraded : condition, which, when corrected, eliminates recurrence of the problem. O 3.10 Rush Maintenance - Maintenance repairs on quality related equipment that may need to be started immediately to avoid a plant shutdown due to Technical Specification requirements or potential loss of essential plant equipment. 3.11 Temporary Repair - An action in which the original design configuration O is temporarily modified, but with the concurrence of engineering. 3.12 Troubleshuoting - A means of collecting information to determine the ' cause of a problem and the action required to resolve it. 3.13 High Risk - Ecuipment not removed from service or cleared. Could O result in an unexpected load reduction, a plant transient, or a reportable event. Should not result in a reactor, turbine, or generator trip. 3.14 Very High Risk - Equipment not removed from service or cleared and presents a risk of tripping the plant either directly or as a result of O causing a major plant transient. 4.0 SPECIFIC PROGRAM OBJECTIVES The Maintenance Program is establisheo to meet the following objectives: 9 4.1 Assure that plant $5Cs are capable of performing their intended functions by identifying equipment degradation prior to. impaired ' function or failure, and promptly restoring intended functions. l 4.2 Establish visible management support for the maintenance program by all involved organizations and support functions such as engineering and pr curement. O I L O I MA100100.5F S ! Q , t i
coococcococcocooocco FOR INFORMATION ONLY cocoooco*cocoococooo ' g PACIFIC GAS AND ELECTRIC COMPANY HlHBER HA1 NUCLEAR POWER GENERATION BUSINESS UNIl REVISION O PAGE 6 0F 16 TITLE: MAINTENANCE i 4.3 On a periodic basis, assess the effectiveness of the maintenance program based upon a performance based review of equipment, independent assessments of maintenance performance, and experience and expertise available from industry sources. 4.4 Control the conduct of maintenance such that the safety of personnel is O protected and the plant is maintained in a safe condition at all times. 4.5 Comprehensively plan and coordinate r,aintenance activities. To the extent practical, combine the performance of maintenance with other related maintenance, testing, inspection or modification activities. Programs should seek an optimum relationship between plant availability C and additional reliability provided ty corrective and preventive maintenance. 4.6 Ensure consistent control of the performance of maintenance whether it is performed by plant, non-plant, or contractor personnel. Minimize the need for rework. Promote effective work practices and avoid C repetition of deficient practices through the use of written procedures and instructions and the feedback of experience from the performance of maintenance. The return of equipment to service must ensure that it is fully restored to the correct configuration and post maintenance testing is completed to ensure its operational readiness.
- 4.7 Maintenance activities associated with ASME XI components and supports require special consideration and expertise due to their safety significance. Ensure proper recognition of ASME XI components and supports c.1d control their repair or replacement through adherence to applicable requirements and the use of qualified personnel.
9 4.8 Maintain equipment in a manner that does not compromise its qualification (e.g., environmental, seismic, etc.). 5.0 GENERAL PROGRAM REQUIREMENTS 5.1 Maintenance Program and Support 9 5.1.1 A comprehensive maintenance program shall be implemented to assure that plant equipment is available and capable of performing its intended function reliably and safely. Maintenance is considered to be the aggregate of those actions that prevent the degradation or failure of, and that promptly g restore the intended functions of SSCs. 5.1.2 Adequate facilities, staff, equipment and training shall be provided to support the performance of maintenance activities. MA100100.SF 6 O
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FOR INFORMATION ONLY *****2*a*ce***c-*** PACIFIC GAS AND ELECTRIC COMPANY NUMBER' MA1 ) NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 7 0F 16 TITLE: MAINTENANCE ) 5.1.3 A maintenance history program shall be incorporated in the maintenance program to support determination of root causes of equipment failures and subsequent correction. Provisions should be made for a review to analyze failures or unacceptable performance, and to identify possible trends and corrective actions. Performance after corrective actions should be ) monitored to ensure that deficiencies have been corrected. Provision should also be made for tracking Maintenance Preventable Functional Failures (MPFF) for all SSCs and their corrections. 5.1.4 The maintenance program shall be continuously supported by all ) NPGBU organizations with responsibilities and interfaces clearly defined. Maintenance requirements and documentation shall be incorporated in the modification process. 5.2 Assessment of Maintenance Program Effectiveness ) 5.2.1 The effectiveness of the maintenance program and maintenance activities shall be monitored and evaluated on a periodic basis. Monitoring shall be accomplished through ongoing self-assessment, QA/QC programs, and management reviews of overall program performance. ) 5.2.2 By July 10, 1996 the Maintenance Program will comply with the requirements of 10CFR50.65, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants." This rule, commonly referred to as the Maintenance Rule, requires that'the performance or condition of designated structures, systems and components (SSCs) be monitored against established goals, in a ) manner sufficient to provide reasonable assurance that those SSCs will be capable of performing their intended functions. Alternatively, where monitoring proves unnecessary, it will be permissible to rely upon an appropriate preventive maintenance program. The effectiveness of performance and condition monitoring, and preventive maintenance activities will be ) evaluated at least annually, taking into account, where practical, industry-wide operating experience. 5.2.3 The maintenance program and processes shall be maintained in conformance to all applicable NPGBU commitment requirements. 5.3 Control of Maintenance Activities ) 5.3.1 Maintenance shall be scheduled and planned so as not to compromise the safety of the plant. Planning shall consider the possible safety consequences of concurrent or sequential maintenance, testing or operating activities. h MA100100.SF 7 )
e r a r m e e s :<c- w s w a FOR INFORMATION ONLY err m:scoc-a:meeso o PACIFIC GAS AND ELECTRIC COMPANY HUMBER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 8 0F 16 TITLE: MAINTENANCE 5.3.2 Maintenance activities shall be conducted in a manner that does not defeat an entire safety function. In performing (or planning) preventive maintenance activities, an assessment of the total plant equipment that is out of service should be taken into account to determine the overall effect on g performance of safety functions. 5.3.3 Troubleshooting may be necessary to determine the cause of problems with plant equipment function and the action required to resolve it. Preferably, troubleshooting shoulo be performed on equipment that has been removed from service or cleared. Where circumstances require equipment to remain in service, O. troubleshooting shall be performed in accordance with risk level assessment procedures, and after obtaining appropriate approvals and authorizations. 5.3.4 The maintenance program shall monitor and maintain the status of environmentally qualified equipment. The performance of O maintenance shall be controlled so as not to compromise the qualification of such equipment. 5.3.5 Maintenance activities that could affect the proper functioning of quality related plant equipment shall be performed in acccrdance with procedures, instructions or drawings O appropriate to the circumstances. 5.3.6 Before implementing a temporary repair, an engineering review of the temporary repair shall be performed to determine its co7formance with design basis requirements. The temporary re; air shall also be reviewed for its effects on personnel O safety, equipment safety, and reliability. Temporary repairs shall be tracked after completion and final corrective action implemented as soon as practical. 5.3.7 Contractor and other non-plant personnel shall be properly trained, qualified, and supervised and shall perform O maintenance work under the same controls, procedures, and standards as plant maintenance personnel. 5.3.8 Radiological protection, industrial safety, personnel health and safety, and housekeeping principals shall be integrated into the planning and performance of maintenance activities. , 5.3.9 Equipment maintenance histories and records shall be maintained for structures, systems and components that affect safe and reliable operation. Records shall be retained in accordance with program directive AD10, " Records Management." O MA100100.5F 8 D
eccesaweet.;;;. ea2*** FOR INFORMATION ONLY *******e e t.e e e e e :2 -* PACIFIC GAS AND ELECTRIC COMPANY NUMBER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 9 0F 16 TITLE: MAINTENANCE 3 5.4 Work Management 5.4.1 A work control system (e.g., PIMS Work Order Module) shall be - utilized to provide an integrated method for planning, initiating, tracking, resource allocation, inspection, and completion of maintenance activities. The work control system O should provide management with an accurate status of maintenance planning and outstanding maintenance work. 5.4.2 Maintenance work shall be screened and prioritized, and the , backlog actively managed. ; 3 5.4.3 Maintenance shall be scheduled in coordination with other required work on the same equipment such as testing, inspection . and modifications, to avoid unnecessary removals of equipment from service and to efficiently utilize resources. 5.4.4 The efficiency of maintenance work processes should be 3 periodically evaluated, including monitoring against valid performance indicators. ' 5.4.5 Maintenance work management should include provisions for ' authorizing emergency maintenance and rush maintenance under , appropriate circumstances and with proper levels of 3 authorization. Once the condition leading to the need for emergency or rush maintenance has been addressed and stabilized, normal work control practices should be regained and the initial actions documented. i 5.4.6 Repeat maintenance shall be identified, documented, evaluated 3 and tracked. Corrective actions shall be taken to minimize repeat maintenance including periodic reviews to determine any ' generic implications. ; i 5.5 Preventive Maintenance Program 3 5.5.1 A preventive maintenance (PM) program shall be implemented and shall consider the following types of equipment:
- a. Any installed plant equipment, both Nuclear Steam Supply )
System (NSSS) and balance of plant (BOP), needed for safe l and reliable plant operation. The selection of equipment l g should be guided by reliability based methods, and/or vendor recommendations, and/or industry and/or plant experience, and the requirements of the Maintenance Rule (effective July 10,1996).
- b. Any equipment required to be maintained based on a PG&E g commitment as delineated in the Commitments Management Database (CMD). 1 MA1001QO.SF 9 O' !
l
- .:. u u u n n :. u n e FOR INFORMATION ONLY c e a s e n n a &**Mee*
O PACIFIC GAS AND ELECTRIC COMPANY NUMBER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 10 0F 16 TITLE: MAINTENANCE O-
- c. Any equipment in storage that requires PM to maintain its readiness for plant installation in either of the categories above.
Preventive maintenance includes predictive, periodic, and planned maintenance actions taken to maintain a piece of O equipment within design operating conditions and extend its life. Preventive maintenance should be performed prior to equipment failure or to prevent equipment failure. 5.5.2 A Master Equipment List (Component Data Base) consisting of a database of plant equipment, components and structures, should O be maintained to help in selecting equipment to be included in the PV Program. A PM Equipment List shall be maintained and indicate what maintenance is to be performed and the schedule for performance. 5.5.3 The frequency of PM tasks and applicable grace periods shall be O specified and shall conform to applicable PM commitments. 5.5.4 Predictive maintenance should be integrated into the overall PM program so that planned maintenance can be performed prior to equipment failure. Predictive maintenance should be selectively applied based on the risk significance'of the O equipment and the ability to monitor specific equipment conditions and failure modes. Performance and/or condition. monitoring of equipment may be relied on where such activities would provide assurance of detecting' failures comparable to that provided by recommended PM tasks. O 5.5.5 PM activities shall be performed within established intervals. PM shall be waived or deferred only if justified by plant conditions or appropriate technical reviews. 5.5.6 The PM program should be continuously refined based on established feedback mechanisms, including an annual management O assessment of maintenance program effectiveness. 5.6 ASME XI Repairs / Replacements 5.6.1 Repairs and replacements of components and supports covered by the In-Service Inspection Program Plan for DCPP shall conform o to the ASME Boiler & Pressure Vessel Code Section XI, 1977 Edition including the Summer 1978 Addenda, as well as all applicable QA program requirements. 5.6.2 Repairs and replacements shall be documented and guided by a specific plan and reviewed by responsible personnel including g the Authorized Nuclear Inspector (ANI). MA100100.5F 10 0
.est.n ar.ee2n a****** '
FOR INFORMATION ONLY en a u esen naeae i 3 PACIFIC GAS AND ELECTRIC COMPANY NUMBER- MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 11 0F 16 TITLE: MAINTENANCE ;
}
6.0 MANAGEMENT PROCESS MODEL i 6.1 Moc'el The block diagram model of directive interface with the Maintenance Program is provided as Attachment 10.1. The activities and interfaces 3 contained in the model provide the framework for performance of this process. Implementing procedures shall be written in accordance with this PD. 6.2 Interfaces 3 This section describes each of the principal interfaces and boundaries between this Program Directive and other management processes. 6.2.1 AD4 Housekeeping and Material Condition Housekeeping and Material Condition places specific cont:ols on ] work areas where maintenance is performed to ensure equipment quality and cleanliness requirements are met. 6.2.2 AD5 Inspections ' AD13 Test Control J Inspections and Test Control implement controls on the I performance of post maintenanr.e inspection and testing. 6.2.3 AD7 Work Planning and Management Work Planning and Management addresses controls for preparing, J- scheduling and implementing maintenance work packages. , 6.2.4 AD8 Outage Planning and Management Outage Planning and Management addresses the implementation of maintenance activities during refueling outages and forced D outages. 6.2.5 AD10 Records Management Records Management provides for the retention,. control, processing, and storage of hardcopy and electronic maintenance 3 history records. 6.2.6 CFI Configuration Managemant Configuration Management proddes controls to ensure that equipment is maintained consistent with its design bases and , ) that any impact due to authorized repair / replacements is reflected in the configuration baseline. MA100100.SF 11 J
******+*c ceeecc*** FOR INFORMATION ONLY *x+***exem /
)- PACIFIC GAS AND ELECTRIC COMPANY NUMBER MA1 HUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 12 0F 16 TITLE: KAINTENANCE ) 6.2.7 CF3 Design Control Design control provides the design of the plant and design bases. Maintenance activities must comply with design documents and design bases documents. ) 6.2.8 CF7 Control and Use of Vendor Information Control and Use of Vendor Information directs vendor supplied recommendations for equipment maintenance to the Maintenance Program. ) 6.2.9 MA2 Test, Measurement, and Diagnostic Equipment Control Test, Measurement and Diagnostic Equipment Control provides for the control and calibration of portable and installed M&TE. 6.2.10 MA3 Control of Special Processes Special Processes controls the implementation of repair activities such as grinding, weiding and heat treatment. 6.2.11 OM7 Problem Resolution ) Problem Resolution provides for the evaluation, root cause determination, and disposition of identified equipment problems, including the performance of corrective maintenance and/or repair / replacement. 6.2.12 OP2 Tagging Programs Tagging Programs control the isolation and removal from service of equipment to be maintained and tracks the total amount of equipment out of service. 6.2.13 TS1 %ging Management ) The D CD Aging Management Program addresses the ability to detect 55: deterioration, project the rate of deterioration, and evaluate the uncertainties considered in developing the frequency and selection of maintenance tasks to prevent unacceptable conditions. ) 6.2.14 T54 Operational Reliability Operational Reliability tracks and analyzes equipment maintenance data, provides NPRDS reporting and interfacing, and provides reliability based methods to the maintenance program. ) MA100100.SF 12 ).
t e*s:een 22:c.;c<****** FOR INFORMATION ONLY ********e**:c. eee2c,*
- O PACIFIC GAS AND ELECTRIC COMPANY NUMBER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION 0 i
PAGE 13 0F 16 i TITLE: MAINTENANCE
- O 6.2.15 TQ1 Personnel Training and Qualification 1
Training programs are developed and implemented through Personnel Training and Qualification to ensure properly qualified maintenance personnel. O 7.0 RESPONSIBit.ITIES 7.1 Senior Vice President and General Manaoer, NPGBU - is responsible for establishing the Maintenance Program and for providing management , support and resources for the program's implementation. ; 7.2 Vice President, Diablo Canyon Operations and Plant Manager - is responsible for: 7.2.1 Implementing DCPP assigned activities. O 7.2.2 Providing adequate facilities, staff, and equipment to support the performance of maintenance activities. 7.2.3 Monitoring the effectiveness of the maintenance program and ensuring that appropriate actions are taken to comply with the maintenance rule (10 CFR50.65). O 7.3 ' Manaoer, DCPP Maintenance Services - has overall responsibility for maintenance activities, including the Maintenance Program. In addition, the Manager, Maintenance Services is responsible for: 7.3.1 Managing the backlog of maintenance work and monitoring i scheduled maintenance performance. . 7.3.2 Assuring that PMs are scheduled and coordinated with cms for plant equipment and for preparing work orders for plant maintenance work. ; 7.3.3 Implementing the repeat maintenance program. 7.3.4 Ensuring correct performance of PM tasks assigned to maintenance and approving deferral of tasks beyond their i scheduled dates. ' 7.3.5 Performing PM evaluations of equipment under their control and O maintaining PM equipment lists and data. 7.3.6 Evaluating the effectiveness of the maintenance program and revising the program _as necessary to meet established goals. 7.3.7 Approving the conduct of maintenance troubleshooting activities l O determined to be VERY HIGH RISK. i l MA100100.SF 13 0
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) PACIFIC GAS AND ELECTRIC COMPANY NUMBER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 14 0F 16 TITLE: MAINTENANCE ) : 7.4 Hanager,'DCPP Operations Service - is responsible for authorizing all l maintenance activities that could affect operating or operable ; installed plant equipment and for tracking the total plant equipment ; that is out of service. ) 7.5 Vice President, Nuclear Technical Services - is responsible for overall technical support to DCPP associated with the maintenance process including responsibilities assigned to NECS and N05. 7.6 Manaaer, Nuclear Enaineerino and Construction Services - is responsible , for management of NECS assigned responsibilities for DCPP maintenance ) including technical analyses, support to system and maintenance engineers and incorporating changes to design documents resulting from ' maintenance activities. Upon request of the Work Planning Center, he is responsible for conduct of ASME Section XI repair / replacement work packages. ' ) 7.7 Manacer, Nuclear Operations Support - is responsible for the management of NOS assigned responsibilities that support DCPP maintenance including:
- 7.7.1 Supporting and/or coordinating projects that address improvement to maintenance programs or maintenance concerns of
) plant staff. The Manager maintains technical expertise,- assists DCPP in problem identification and resolution, and assists in the periodic assessment of maintenance program ; effectiveness. In addition, the Manager provides staff engineering support for refueling outages and other special maintenance staffing requirements. ) 7.7.2 Monitoring and reviewing maint.._ ace experiences at other ' nuclear plants for applicability to DCPP. , 7.7.3 Providing safety analysis and Probabilistic Risk Assessment (PRA) support for maintenance activities; 7.7.4 Interfacing with other organizations outside of NPGBU (e.g., EPRI, NUMARC, Westinghouse) and other PG&E Departments (e.g., HVT&S) as required to enhance the performance of maintenance. 7.8 Director., DCPP Ouality Control - is responsible for reviewing quality related maintenance work orders and/or procedures and ) repair / replacement plans and specifying any QC inspection and hold ; points. ) MA100100.SF 14 I 1 ) l 1
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******************** FOR INFORMATION ONLY ******************** i PACIFIC GAS AND ELECTRIC COMPANY NUMBER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 15 0F 16 TITLE: MAINTENANCE O
8.0 IMPLEMENTING DOCUMENTS l 8.1 Interdepartmental Administrative Procedures (IDAPS) Inter-Department Administration Procedures shall be developed to l O address the following aspects of the Maintenance Program: 8.1.1 An IDAP shall be developed that clearly defines the ! organizational structure of the maintenance organization, specifies formal agreements for support of the maintenance program, and outlines the performance standards and
- O expectations of the maintenance organization including !
contractors. s 8.1.2 An IDAP shall be developed to monitor the effectiveness of the j maintenance program. The procedure should incorporate, to the : extent practical, the requirements of the maintenance rule l
- O (10CFR50.65). In any case, the IDAP shall adhere to the !
requirements of 10CFR50.65 by July 10, 1996. Monitoring shall ! i evaluate the ability of the maintenance organization to sustain i adequate SSC performance and maintain systems and equipment in : good working order.
- O 8.1.3 An IDAP shall be developed to clearly define and implement a !
preventive maintenance (PM) program for systems, equipment, and j appropriate spare parts that affect safe and reliable plant ; operation. The PM program should include predictive i maintenance and scheduled activities, management control of ' processes, and documented results and actions. , 8.1.4 An IDAP shall be developed to specify the requirements for maintenance of MOVs and monitoring of MOV performance. l 8.1.5 An IDAP shall be developed to specify detailed controls for the ; implementation of ASME XI repairs / replacements. 4 (Combination of AP C-755 and AP C-756) i 8.2 Departmental Level Administrative Procedures (DLAPS) l Departments responsible for performing activities related to the l !g Maintenance Program that need more specific direction than provided in );
- this PD and related IDAPs shall develop DLAPs to assign l l responsibilities and implement the program requirements specified in
- ~ this PD and associated IDAPs.
9.D RECORDS
- O None A
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) PACIFIC GAS AND ELECTRIC COMPANY NUPEER MA1 NUCLEAR POWER GENERATION BUSINESS UNIT REVISION O PAGE 16 0F 16
' TITLE: MAINTENANCE
') 10.0 ATTACHMENTS 10.1 Block Diagram - Maintenance Program Interface
11.0 REFERENCES
11.1 Code of Federal Regulations, Title 10, Part 50.65 11.2 Code of Federal Regulations, Title 10, Part 50.49. 11.3 INPO 90-008, Maintenance Programs in the Nuclear Power Industry,
) Revision 1, 3/90 11.4 INPO 86-002, Maintenance History Program, 1/86 11.5 INP0 85-032, Preventive Maintenance, Revision 1, 12/88 )' 11.6 INPO 90-015, Performance Objectives and Criteria for Operating and Near-term Operating License Plants, 8/90 11.7 INPO 92-001, " Guidelines for the Conduct of Maintenance at Nuclear Power . Stations, 4/92 ) 11.8 INPO 89-009, Plant Predictive Maintenance, 9/89 11.9 NRC Generic Letter 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance, June 28, 1989 11.10 ASME Boiler and Presstre Vessel Code, Section XI, 1977 Edition, Summer ) 1978 Addenda 11.11 PG&E Licensing Submittal Letter, Chron 004416, ISI Generic Repair / Replacement Plans, October 4, 1982 12.0 SPONSOR Harry Phillips j
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O EXHIBIT 8 AVAILABLE BASIC (OR FUNDAMENTAL) QUALIFICATION TRAINING O
- Instrument & Controls Basic Science (5 weeks) g General I&C Training (5 weeks) ,
Electricity and Electronics (16 weeks) ! Process Instrumentation and' Controls (15 weeks) Systems Overview (2 weeks) , f
- Electrical ;
O Plant orientation (3\ weeks) Initial Electrical Sessions (26 weeks)
- examples - Electrical Theory ,
Print reading Test Equipment l Gr unding O
- Mechanical i
Orientation (1 weeks) Systems (2 weeks) () Fundamentals (8 weeks) Plant Component Training (12 weeks) ,
- examples - Valves i Snubbers
- Pumps !
Air Compressors C) i f O i O b
.O l
I [ O-
O EXHIBIT 9 AVAILABLE ADVANCED OR PLANT-SPECIFIC TRAINING C)-
- Instrument & Controls (total hours available: 25h weeks)
Digital Rod Position Indication Systems , Advanced Digital Feedwater Control System
,3 Containment Hydrogen Monitor
- Electrical (total hours available: 42 weeks)
Meter Calibration ; Circuit Breakers and Switchgear
- M t r Maintenance >
O Air Conditioning and Refrigeration
- Hechanical (total hours available: 13 weeks) r Freeze Seals Auxiliary reedwater Pumps O Traveling Screens Containment Hatches Reactors Coolant Pump Seals
'O t i O O i ! P I) I
/ t
- O ,
+ ;. EXHIBIT 10 O LIST OF 1%INTENANCE AND SURVEILLANCE PROCEDURES O : O I O :. i O i i i i b O 9 i O : 4 O ; 4 0 O i j w-
t
) ?? JUL 93 DIABLO CANYOH POWER PLANT UNITS 1 AND 2 UOLUME SA ELECTRICAL MAINTENANCE PROCEDURES TABLE OF CONTENTS )
NUMBER REV UNIT TITLE a============== === ==== ===.....========..========.=...=..,===.......==...==,=== MP E-2.1 3 1&2 ** CONDENSATE PUMP MOTOR OVERHAUL MP E-2.2 3 1&2 ** CONDENSATE B00 STER PUMP MOTOR OVERHAUL
) MP E-3.1 7 IL2 *** AUXILIARY FEEDWATER ? UMP MOTOR OVERHAUL MP E-3.2 2 1&2 ** AUXILIARY FEED WATER TURBINE PUMP TACH-PAK 3 AND RPM INDICATOR VERIFICATION MP E-5.1 0 1&2 ** HEATER 2 DRAIN PUMP MOTOR OVERHAUL WESTINGHOUSE WORLD SERIES MOTOR MP E-7.3 3 1&2 ** REMOVAL, DISASSEMBLY & REASSEMBLY OF THE OPERATING ) CONTROL ROD DRIVES MF E-7.5A 4 IL2 *** MAINTENANCE OF PRESSURIZER HEATERS, PM MP E-7.5B 0 JL2 ** PRESSURIZER HEATER INSULATOR REPAIR MP E-7.6 7 1&2
- ROUTINE PREVENTIVE MAINTENANCE OF SCR POWER CONTROLLER FOR PPESSURIZER HEATERS MP E-8.1 1 162 ** CENTRIFUGAL CHARGING PUMP MOTOP OVERHAUL MP E-10.1 11 IL2 **RHR PUMP MOTOR OVERHAUL MP E-10.3 1 JL2 ***RHR PUMP MOTOR PARTIAL DISASSEMBLY IN PLACE i
(WESTINGHOUSE MANUFACTURER) MP E-10.4 1 IL2 *** CALIBRATION AND TESTING OF THE RHR MID LOOP CURRENT ALAPM MP E-14.1 5 162 *** COMPONENT COOLING WATER PUMP MOTOR OVERHAUL MP E-17.1 7 IL2 **AUXILIARV SALTWATER PUMP riUf0R OVERHAUL
) MF E-17.2A 3 1&2
- MAINTENANCE OF CIRCULATING WATER PUMP MOTORS MF E-17.28 6 JL2 *** CIRCULATING WATER PUMP MOTOR DISASSEMBLY, INSPECTION, AND REASSEMBLY MP E-23.1 7 IL2 ** MAINTENANCE OF DIESEL GENERATOR MP E-21.4; 4 IL2 ** DIESEL GENERATOP TACH-PAK 3 AND RPM INDICATOR TESTING / CALIBRATION
) MF f-21.5 2 IL2 Mr E-21.6 ** DIESEL GENERATOR AIR CONTROLLER (KW SENSOR) 3 IL2 ** DIESEL GENERATOR ELECTRICAL GOVERNOR MAlNTENANCE AND ADJUSTMENT Mi f-22.1 5 IL2 **HIGH VOLTAGE DC TEST OF MAIN GENERATOR MF E-23.1 11 IL2 ** MAINTENANCE Of REACTOR CONTAINMENT FAN COOLER MOTOR MP E-23.2 0 IL2 *** CONTROL ROD DRIVE MECHANISM EXHAUST FAN MOTOR OVERHAUL MP E-23.3 4 IL2
/ *LUBRICATl0N OF REACTOR CONTAINMENT FAN COOLER MOTORS MP E-23.4 5 IL2 ** ROUTINE PREVENTIVE MAINTENANCE OF INTERNAL HYDROGEN RECOMBINER MP E-35.1 3 IL2 ** TEMPERATURE INSTRUMENTS CALIBRATION CHECK MP E-35.2 7 IL2 **MAlNTENANCE OF TRANSFORMER TEMPERATURE PROBES MP E-35.3 0 IL2
- PRESSURE INSfRUMENTS CALIBRATION CHECK '
) MP E-35.4 1 IL2 ** LEVEL INSTRUMENTS CALIBRATION CHECK MP E-35.5 2 IL2 *lNSTRUMENTATION DEVICE AND SETPOINT LIST MP E-41.lA 14 JL2 ** MAINTENANCE OF WESTINGHOUSE TYPE 08-50 480V REACTOR TRIP AND BYPASS CIRCUll BREAKERS MP E-41.lB 2 1&?
- WESTINGHOUSE DB-50 CIRCUIT BREAKER UVTA REPLACEMENT 1
)
)
i i 22 JUL 93 DIABLO CANYON POWER PLANT l UNITS 1 AND 2 VOLUME 5A ELECTRICAL HAINTENANCE PROCEDURES
) TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ==== ===========================================================
MP E 41.1C 2 IL2 ** MAINTENANCE OF WESTINGHOUSE 08-50 480V MAIN GENERATOR
) FIELD CIRCUIT BREALERS AND ROD CONTROL MG SET OUTPUT BREAKERS MP E-41.2 5 IL2
- MAINTENANCE OF REACTOR TRIP SWITCHGEAR AND MG SET CONTROLLER MP E-41.3 2 IL2 ** CONTROL ROD DRIVE MG SET GENERATOR OVERHAUL MP E-42.1 0 IL2 ** POLAR CRANE PC CARD REMOVAL, STORAGE, AND REINSTALLATION
) MF E-42.2A 0 IL2 ** ELECTRICAL MAINTENANCE OF PLANT CRANES, HOISTS AND MONORAILS MF E-42.25 1 IL2 **fLECTRICAL MAINTENANCE OF POLAR CRANE MP E-42.2C 2 162 ** ELECTRICAL MAINTENANCE OF MANIPULATOR CRANE MP E-42.20 1 IL2 ** ELECTRICAL MAINTENANCE OF SPENT FUEL POOL CRANE MF E-45.1 0 IL? *** ELECTRICAL PENETRATION REPAIR 7)
MP E-50.1 22 1&2 ** MAINTENANCE OF THERMAL OVERLOAD RELAYS AND LINE STARTERS MF E-50.2 5 IL2 ** WESTINGHOUSE TYPE TD-5 RELAY MAINTENANCE Mi E-50.3L 6 IL2 ** WESTINGHOUSE TYPE SA-1 GENERATOR DIFFERENTIAL RELAY MAINTENANCE MP E-5(.4 15 1c? ** GENERAL ELECTPlc TYPE IAC RELAY MAINTENANCE MF E-50.5 1 IL2 ** WESTINGHOUSE TYPE IRV-2 DIRECTIONAL OVERCURRENT RELAY MAINTENANCE 7) MP E-50.6 1 IL2 ** WESTINGHOUSE TYPE " AR" (HIGH SPEED) AUXILI ARY RELAY MAINTENANCE MP t-5U.7 8 IL2 ** GENERAL ELECThlC TYPE IJCV RELAY MAINTENANCE MP E-50.5 6 IL2 ** WESTINGHOUSE TYPE MG-6 RELAY MAINTENANCE MP E-50.9 5 IL2 *** WEST'NGHOUSE TYPE TK RELAY MAINTENANCE MP E-50.10A 0 IL2 **G.E. TYPE lAV54 RELAY MAINTENANCE 7) MF E-50.10b 2 IL2 ** GENERAL ELECTRIC TYPE IAV55 RELAY MAINTENANCE MP E-50.ll 11 IL2 ** WESTINGHOUSE TYPE SC AND SV RELAY MAINTENANCE MP E-50.12 0 lL2 ** GENERAL ELECTRIC TYPE RPM RELAY MAINTENANCE MP E-50.13 3 IL2 **POUTINE PREVENTlvE MAINTENANCE OF STV STATIC OVEREXCITATION RELAi MP E-50.14 1 IL2 ** GENERAL ELECTRIC TYPE CFD RELAY MAINTENANCE 3' MP E-50.15 5 IL2 ** GENERAL ELECTRIC TYPE IJS SYNCHRO-CHECK RELAY MAINTENANCE MP E-50.16 0 IL2 ** WESTINGHOUSE TYPE HRU RELAY MAINTENANCE MP E-50.17 6 IL2 ** WESTINGHOUSE TYPE "CM" PHASE BALANCE RELAY MAINTENANCE MP E-50.18 7 IL2 ** WESTINGHOUSE TYPE KLF LOSS OF FIELD RELAY MAINTENANCE MP E-50.19 8 IL2 ** WESTINGHOUSE TYPE COM-5 RELAY MAINTENANCE MP E-50.2D 0 IL2 ** GENERAL ELECTRIC TYPE HEA RELAY MAINTENANCE MP E-50.21 0 IL2 ** GENERAL ELECTklC TYPE RAV RELAY MAINTENANCE MP E-50.22A 1 IL2 ** WESTINGHOUSE TYPE CV-2 AND CV-7 UNDERVOLTAGE RELAY MAINTENANCE MP E-50.228 0 IL2 ** WESTINGHOUSE TYPE CV-8 AND CV-25 OVERVOLTAGE RELAY
, MAINTENANCE MP E-50.23 0 1L2 **WESTINGHDUSE TYPE SG AUXILIARY RELAY MAINTENANCE I
4D 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SA ELECTRICAL HAINTENANCE PROCEDURES TABLE OF CONTENTS NUMBER REV UNIT TITLE
==== ,,======,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
MF E-50.24 1 IL2 ** WESTINGHOUSE TYPE CRN-1 REVERSE POWER RELAY MAINTf'3NCE MP E-50.25 0 IL2 ** WESTINGHOUSE TYPE C0Q RELAY MAINTENANCE 4D MF E-50.27 0 IL2 ** WESTINGHOUSE TYPE CVQ RELAY MAINTENANCE MP E-50.26 10 162 ** WESTINGHOUSE TYPE C0 RELAY MAINTENANCE MP E-50.29 0 IL2 ** WESTINGHOUSE TYPE AR AND ARD INDUSTRIAL CONTROL RELAY MAINTENANCE MF E-50.3D 10 IL2 ** TESTING TIMING RELAYS MP E-50.31 6 IL2 ** GENERAL ELECTRIC TYPE HFA RELAY MAINTENANCE C) MP E-50.32 0 IL2 ** GENERAL ELECTRIC TYPE HGA RELAY MAINTENANCE MP f -50.33 17 IL2 ** WESTINGHOUSE TYPE SSV-T RELAY MAINTENANCE MP E-50.34 4 ILP **STRUTHERS-DUNN GROUND CURRENT SENSOR RELAY MAINTENANCE M; E-50.36 5 IL2 ** GENERAL ELECTRIC TYPE IFC66 RELAY MAINTENANCE MP E-50.37 0 IL2
- MAINTENANCE OF PIS MODEL ET-1200L CURRENT ALARMS MP E-50.4; 1 IL2 ** ROUTINE PREVENTIVE MAINTENANCE OF GE TYPE "SLJ" C) SYNCHRON!SM CHECK RELAY IS E-50.41; 2 lt? **BASLER ELECTRIC TYPE bel-81 SINGLE SETPOINT DIGITAL FPEQUENCY RELAY MAINTENANCE MF E-5D.415 1 IL? **EASLER ELECTRIC TiFE bel-81 THREE SETPOINT DIGITAL FREQUENCi RELAY MAINTENANCE MP E-50.42 1 IL2 ** WESTINGHOUSE TYPE KLF-1 LOSS OF FIELD RELAY MAINTENANCE
[] MP E-50.47 0 IL2 ** BROWN BOVERI lYPE ITE-500 GROUND OVERCURRENT RELAY MAINTENANCE MP E-50.5D 0 IL2 ** GENERAL ELECTRIC TiPE CEB 0FFSET MHO RELAY MAINTENANCE MP E-50.51 1 ILP ** WESTINGHOUSE TYPE MVH VOLTS / HERTZ RELAY MAINTENANCE MP E-52.1 9 IL2 ** MAINTENANCE Of DRY TYPE LOAD CENTER TRANSFORMER AND 480 VOLT SWITCHGEAR [] MP E-52.2 3 IL2 **MalNTENANCE Of LOW VOLTAGE DRY TYPE TRANSFORMERS MP E-52.3 0 JL2
- MAINTENANCE OF GENERATOR NEUTRAL TRANSFORMER & RESISTOR GR]D MP E -52.4 5 li? ** MAINTENANCE OT INSTRUMENT A-C TRANSFORMER / REGULATOR MP E-52.5 1 IL2 ** POWER TRANSiOEMER RADIATOR REMOVAL & INSTALLATION MP E-53.1 8 IL2 ** MAINTENANCE Of 115 & 480 VOLT MOTORS C) MP E-53.2 14 IL2 *** SPLIT END BEtt, 4000 VOLT MOTOR OVERHAUL MP E-53.7 5 IL2 ** MAINTENANCE OF ITT, GENERAL CONTROLS HYDRAMOTOR ACTUATOR MP E-53.7A 5 IL2 **REPAlR/0VERHAUL ITT CONTROLS HYDRAMOTORS, MODEL NH92, VALVE ACTUATORS FOR LCV-Il0, LCV-lll, LCV-ll3 AND LCV-ll5 MP E-53.8 7 JL2 ** MAINTENANCE Of REACTOR HEAD VENT VALVES MP E-53.BB 0 162 ** TARGET ROCK PARTS REPLACEMENT PROCEDURE
- ) MP E-53.10A 13 IL2 ** PREVENTIVE MAINTENANCE OF LIMITORQUE MOTOR OPERATORS MP E-53.10B 18 IL2 **LIMITORQUE OPERATOR TORQUE SWITCH ADJUSTMENT MP E-53.10C 11 IL2 **LIMITORQUE OPERATOR TORQUE SWITCH REMOVAL AND INSTALLATIOh MP E-53.10D 17 1&P **LIMITORQUE OPERATOR LIMIT SWITCH ADJUSTMENT MP E-53.10E 4 1&2 ** PREVENTIVE MAINTENANCE OF LIMITORQUE 90 DEGREE GEAR DRIVES O l 1
i i
() 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SA ELECTRICAL MAINTENANCE PROCEDURES TABLE OF CONTENTS O HUMBLR REV UNIT TITLE
=============== === ==== ===========================================================
MP E-53.10F 1 1&2 *LIMITORQUE OPERATOR TYPE SMC-04 LIMIT SWITCH GREASE INSPECTION () MP E-53.10G 5 1&2 **LIMITORQUE OPERATOR LlHIT SWITCH REMOVAL, REPAIR AND INSTALLATION MP E-53.101 7 1&2 **LIMITORQUE SMB-000 VALVE OPERATOR MAINTENANCE MP E-53.10K 7 1&2 **LIMITORQUE SMB-0, 1, 2 AND 3 VALVE OPERATOR MAINTENANCE AND DISASSEMBLY MP E-53.10L 0 1&2 **HBC-0 THROUGH HBC-3 POSITION LIMIT STOP SETTING () MP E-53.10M 1 1&2 **LIMITORQUE SMB-00 AND SB-00 VALVE OPERATOR MAINTENANCE MP E-53.10N 1 1&2 **LIMITORQUE SMB-0, 1, 2 AND 3 VALVE OPERATOR MAINTENANCE AND DISASSEMBLY MP E-53.10T 1 1&2 ** WESTINGHOUSE VITALS MOTOR OPERATED VALVE DIAGNOSTIC TEST MP E-53.10V 9 1&2 ** LIBERTY TECHNOLOGY VOTES MOTOR OPERATED VALVE DIAGNOSTIC TESTING (3 - MP E-53.10X 0 1&2 ** TROUBLESHOOTING LIMITORQUE OPERATORS MP E-53.11A 2 1&2 ** PREVENTIVE MAINTENANCE OF ROTORK MOTOR OPERATORS MP E-53.llB 6 1&2 **ROTORK OPERATOR LIMIT / TORQUE SWITCH SETTING AND ADJUSTMENTS MP E-54.2 10 1&2 **HIGH VOLTAGE TESTING 0F ELECTRICAL EQUIPMENT MP E-54.5 5 1&2 ** POTTER & BROMf lELD TYPE CSL AND CSJ VOLTAGE SENSOR RELAY .)( MAINTENANCE MP E-55.5A 11 1&2 *** MAINTENANCE OF BATTERY PACK EMERGENCY LIGHTS INSIDE POWER BLOCK l MP E-57.2A 2 1&2 ** WIRE & CABLE INSTALLATION i MP E-57.2B 16 1&2 ** WIRE AND CABLE TERMINATIONS - MP E-57.2C 1 1&2 ** POWER SEMICONDUCTOR (STUD OR DISK) MOUNTING PROCEDURE 4 () MP E-57.4 7 1&2
- ENVIRONMENTAL QUALIFICATION MAINTENANCE & SURVEY OF CONTAINMENT PENETRAT10NS, CABLE, AND SPLICES MP E-57.5 1 1&2 ** MAINTENANCE Of COMPRESSION 0.Z./GEDNEY TYPE LOCA SEALS MP E-57.6 1 1&2
- ANCHOR BOLT INSTALLATION MP E-57.7B 2 1&2 **lNSTALLATION & REPAIR OF THERMAL-LAG 330 FIRE BARRIERS MP E-57.8A 4 1&2 *** TEMPERATURE MONITORING
() MP E-57.10A 1 1&2 ** MOTOR DRYlHG METHODS MP E-57.10B 4 1&2 *** GENERIC MOTOR CLEAN, INSPECT, LUBE AND TEST PREVENTIVE MAINTENANCE MP E-57.11A 5 1&2
- SAFETY PRECAUTIONS WHEN WORKING ON POTENTIALLY ENERGlZED EQUIPMENT l MP E-57.llB 6 1&2 ** INSTALLING AND REMOVING GROUNDS FNOM DEENERGIZED POWER i' 13 PLANT ELECTRICAL EQUIPMENT 4
MP E-57.12A 0 1&2 **ESNA 600 AMP SEPARABLE INSULATED CONNECTORS DETERM/RETERM MP E-57.13 1 1&2 ** CALIBRATION OF WIRE STRIP 2FRS AND LRIMPERS MP E-57.15 0 1&2 *** MAINTENANCE AND CALIBRATION OF AMMETERS VOLTMETERS, FREQUENCY METERS AND TACH 0 METERS I l) l
C) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 ; VOLUME SA ELECTRICAL KAINTENANCE PROCEDURES TABLE OF CONTENTS O NUMBER REV UNIT TITLE
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MP E-57.16 0 1&2 *** MAINTENANCE AND CAllBRATION OF WATT, VAR, POWER FACTOR METERS AND TRANSDUCERS C) MP E-57.17 0 1&2 ***SYNCHROSCOPE METER FUNCTIONAL TEST MP E-5B.1 2 1&2 ** GENERAL ELECTRIC TYPE SB-1 CONTROL SWITCH MAINTENANCE MP E-58.2 2 1&2 ** WESTINGHOUSE XA SYNCHRONIZER RELAY MAINTENANCE MP E-5B.5 9 1&2 **REPRESSURIZA110N OF CONTAINMENT ELECTRICAL PENETRATIONS MP E-59.lA 3 1&2 **lNSTRUCTIONS FOR ASSEMBLY & TAPE INSULATION OF 4 & 12KV BUS () MP E-60.1A 0 1
- MAIN GENERATOR PROTECTIVE circuli FUNCTIONAL TEST -
CIRCUIT GM1 1 MF E-60.lE O1
- MAIN GENERATOR PROTECTIVE CIRCUlT FUNCTIONAL TEST -
CIRCulT GM2 MP E-60.2A 0 2
- MAIN GENERATOR PROTECTIVE CIRCUIT FUNCTIONAL TEST -
CIRCUIT GM2 () MP E-60.?B 1 2
- MAIN GENERATOR PROTECTIVE CIRCUIT FUNCTIONAL TEST CIRCUlT :
GM2 MP E-60.10 1 1&2
- GENERIC RELAY FUNCTIONAL TEST INSTRUCTIONS ,
MP E-60.11 1 1
- RELAY FUNCTIONAL TEST - STANDBY START-UP TRANSFORMER 11 >
PROTECTION SCHEME MP E-60.12 1 2
- RELAY FUNCTIONAL TEST - STANDBY START-UP TRANSFORMER 21
() PROTECTION SCHEME MP E-60.13 2 1
- RELAY FUNCTIONAL TEST - STANDBY START-UP TRANSFORMER 12 PROTECTION SCHEME MP E-60.14 1 2
- RELAY FUNCTIONAL TEST - STANDBY START-UP TRANSFORMER 22 PROTECTION SCHEME MP E-60.15 0 1
- RELAY FUNCTIONAL TEST - l?KV START-UP BUS 1 BUS
() DIFFERENTIAL RELA) SCHEME MP E-60.16 2 2
- RELAY FUNCTIONAL TEST - 12KV START-UP BUS 2 BUS ,
DIFFERENTIAL RELAY SCHEME - MP E-60.17 3 1 *PELAY FUNCTIONAL TEST - UNIT 1 12KV BUS D DIFFERENTIAL RELAY SCHEME MP E-60.lB 3 2
- RELAY FUNCTIONAL TEST - UNIT 2 12KV BUS 0 DIFFERENTIAL
() RELAY SCHEME MP E-60.19 3 1
- RELAY FUNCTIONAL TEST - UNIT 1 12KV BUS E -DIFFERENTI AL
- RELAY SCHEME MP E-60.20 3 2
- RELAY FUNCTIONAL TEST - UNIT 2 12KV BUS E DIFFERENTIAL RELAY SCHIME ,
MP E-60.21 2 1
- RELAY FUNCTIONAL TEST - UNIT 1 4.16KV BUS "D" DIFFERENTIAL
() RELAY SCHEME MP E-60.22 2 2
- RELAY FUNCTIONAL TEST - UNIT 2 4.16KV BUS "0" DIFFERENTIAL RELAY SCHEME MP E-60.23 2 1
- RELAY FUNCTIONAL TEST - UNIT 1 4.16KV BUS E DIFFERENTIAL RELAY SCHEME
- O O
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C) 22 JUL 93 DIABLO CANYON POWER PLANT l l UNITS 1 AND 2 VOLUME SA ELECTRICAL KAINTENANCE PROCEDURES TABLE OF CONTENTS NUMBER REV UNil TITLE
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MP E-60.24 2 2
- RELAY FUNCTIONAL TEST - UNIT 2 4.16KV BUS "E" DIFFERENTIAL RELAY SCHEME '
C) MP E-60.25 2 1
- RELAY FUNCTIONAL TEST - UNIT 1 4.16KV BUS "F" DIFFERENTIAL RELAY SCHEME MP E-60.26 4 2 *(NON-PERM XPR 3-15-93/SCHUELKE) RELAY FUNCTIONAL TEST -
UNIT 2 4.16 KV BUS F DIFFERENTIAL RELAY SCHEME MP E-63.27 1 1
- RELAY FUNCTIONAL TEST - UNIT 1 4.16KV BUS "G" DIFFERENTIAL RELAY SCHEME
.C) MP E-60.28 2 2
- RELAY FUNCTIONAL TEST - UNIT 2 4.16 KV BUS "G" DIFFERENTIAL RELAY SCHEME i MP E-60.29 2 1
- RELAY FUNCTIONAL TEST - UNIT 1 4.16KV BUS "H" DIFFERENTIAL RELAY SCHEME MP E-60.30 2 2
- RELAY FUNCTIONAL TEST - UNIT 2 4.16KV BUS "H" DIFFERENTIAL RELAY SCHEME
.C) MP L-60.31 2 2
- RELAY FUNCTIONAL TEST - UNIT 2 12KV BUS D START-UP CABLE '
DIFFERENTIAL RELAY SCHEME MP E-60.32 2 2
- RELAY FUNCTIONAL TEST - UNIT 2 12KV BUS E STARl-UP CABLE DIFFERENTIAL RELAY SCHEME MP E-60.33 1 1
- RELAY FUNCTIONAL TEST - 12KV BUS 0 UNDERVOLTAGE TRIP RELAY SCHEME l C) MP E-60.31 1 2
- RELAY FUNCTIONAL TEST - UNIT 2 12KV BUS D UNDERVOLTAGE
- TRIP RELAY SCHEME MP E-60.35 1 1
- RELAY FUNCTIONAL TEST - UNIT 1 12KV BUS E UNDERVOLTAGE l TRIP RELAY SCHEME ,
MP E-60.36 l 0 2
- RELAY FUNCTIONAL TEST - UNIT 2 12KV BUS E UNDERVOLTAGE TRIP RELAY SCHEME
() Mi f-60.37 1 1
- RELAY FUNCTIONAL TEST - UNIT 1 4.16KV BUS D UNDERVOLTAGE TRIP RELAY SCHEME MP E-60.38 1 2
- RELAY FUNCTIONAL TEST - UNIT 2 4.16KV BUS "0" UNDERVOLTAGE TRIP RELAY SCHEME MP E-60.39 0 1
- RELAY FUNCTIONAL TEST - UNIT 1 4.16KV BUS "E" UNDERVOLTAGE TRIP RELAY SCHEME i() MP E-63.40 0 2
- RELAY FUNCTIONAL TEST - UNIT 2 4.16KV BUS "E" i UNDERVOLTAGE TRIP RELAY SCHEME l MP E-60.41 0 1
- FUNCTIONAL TEST - CIRCULATING WATER PUMP 11 AUTOMATIC
! PECLOSE SCHEME MF E-60.42 0 2
- FUNCTIONAL TEST - CIRCULATING WATER PUMP 21 AUTOMATIC RECLOSE SCHEME
,[] MP E-60.43 0 1
- FUNCTIONAL TEST - CIRCULATING WATER PUMP 12 AUTOMATIC RECLOSE SCHEME MP E-60.44 0 2
- FUNCTIONAL TEST - CIRCULATING WATER PUMP 22 AUTOMATIC RECLOSE SCHEME MP E-61.1 3 1&2 **CAllBRATION OF THE MAIN GENERATOR VOLTAGE REGULATOR POWER SYSTEM STABILIZER SIGNAL D
D
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SA ELECTRICAL MAINTENANCE PROCEDURES g TABLE OF CONTENTS NUMBER REV UNIT TITLE
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MP E-61.2 1 1&2 ** TEST FOR STATOR C0ll WATER BLOCKAGE g MP E-61.3A 0 1&2 " MAIN GENERATOR STATOR C0ll LE AK RATE TEST MP E-61.5B 0 1&2 *** MAIN GENERATOR EXCITER ELECTRICALTESTING MP E-61.6B 0 2 *** MAIN GENERATOR LOOP TEST MP E-61.9A 0 1&2 **lS0 LATED PHASE BUS & MOTOR OPERATED DISCONNECT PREVENTATIVE MAINTENANCE MP E-61.10A 0 IL2 ** MAIN GENERATOR CONDITION MONITOR AUT0-SAMPLER INSTRUCTIONS g MP E-61.10B 1 1&2 ** MAIN GENERATOR CONDITION MONITOR CAllBRATION MP E-62.1A 14 IL2 ** MAINTENANCE OF I?KV MAGNE-BLAST CIRCUIT BREAKERS MP E-62.1D 4 IE2 **MAGNE-BLAST CIRCUll BREAKER PREVENTIVE MAINTENANCE (MINI) MP E-62.2A 3 1&? ** MAINTENANCE OF GENERAL ELECTRIC 12KV POWER-VAC circuli BREAKERS MP E-63.lA 15 IL2 ** MAINTENANCE OF 4KV MAGNE-BLAST CIRCUIT BREAKERS O MP E-63.lC 6 l e.2 **0VERHAUL OF 4KV AND 12KV MAGNA-BLAST CIRCUIT BREAKERS Mf E-63.3B 0 1&2 " MAINTENANCE OF POTENTI AL TRANSFORMER CABINETS IN GENERAL ELECTRIC METAL-CLAD 4KV AND 12KV SWITCHGEAR MF E-63.3E O IL2 ** MAINTENANCE OF GENERAL ELECTRIC METAL CLAD 4KV AND 12KV SWITCHGEAR MP E-63.20 0 IL2 *" MICRO-OHM CHECK OF BUS CONNECTIONS FOR 4KV AND 12KV SWITCHGEAR O MP E-64.lA 22 IL? "*AC AND DC MOLDED CASE CIRCUIT BREAKER TEST PROCEDURE MP E-64.lB 3 1&2 *** MOLDED CASE CIRCUIT BREAKER EXERCISE MF E-64.10 2 IL2 *PY PANEL MOLDED CASE circuli BREAKER EXERCISING MP E-64.4 9 IL2 " MAINTENANCE OF FPE TYPE FPS 2 CIRCulT BREAKERS MP E-64.5A 1 IL2 " MAINTENANCE OF ABB, ITE-TYPES K-3000(S) AND K-4000(S) CIRCUIT BREAKERS O MP E-64.6 - 5 IL2 ** MAINTENANCE Of BBC, ITE-TYPES K-600 THRU K-2000 AND K-6005 THRU K-20005 CIRCU1T BREAKERS l MP E-65.lA 13 IL2 ** MAINTENANCE OF WESTINGHOUSE 7.5KVA INVERTERS i MP E-65.lb 2 IL2 *0VERHAUL OF WESTINGHOUSE 7.5 KVA INVERTERS l MP E-65.2A 1 IL2 *'lNSPECTION AND CLEANING AND FUNCTIONAL TESTING OF I UNINTERRUPTIELE POWER SUPPLY (UPS) FOR CHEMICAL LAB AND h MP E-65.3A 0 lL2 AMSAC -
*** MAINTENANCE OF SOLIDSTATE CONTROLS UPS POWER SUPPLY AND l
l BATTERY PACKS FOR TECH SUPPORT CENTER ! MP E-65.4A 0 IL2 *** MAINTENANCE OF SOLIDSTATE CONTROLS UPS FOR THE PPC ! COMPUTER ', MP E-65.5A 0 1&2 " MAINTENANCE Of INVERTER POWER SUPPLY FOR MAIN FEEDWATER O TURBINE SPEED CON 1ROL MP E-65.6A 0 IL2 *" MAINTENANCE OF SOLIDSTATE CONTROLS UPS POWER SUPPLY AND BATTERY PACKS FOR DIGITAL FEEDWATER CONTROL MP E-65.7A 0 1&2 ** MAINTENANCE OF CONTROL ROOM EMERGENCY LIGHTING UNINTERRUPTIBLE POWER SUPPLY AND BATTERY PACKS I Me E-65.sA 1 IL2 *** MAINTENANCE cr C15EREx INVERTER r0R THE P-2000 COMPu1ER D
) 22 JUL 93 DIABLO CANYON POWER PLANT l UNITS 1 AND 2 VOLUME SA ELECTRICAL KAINTENANCE PROCEDURES I
) TABLE OF CONTENTS NUMBER REV UNIT TITLE
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MF E-65.9A 1 IL2 ** MAINTENANCE OF SOLA UNINTERRUPTIBLE POWER SUPPLY FOR THE
- 2) FIRE ALARM PANEL MP E-65.lCA 0 IL2 *** MAINTENANCE OF DOSIMETRY COMPUTER UNINTERRUPTIBLE POWER SUPPLY AND BATTERY PACKS Mi E-65.llA 0 JLP *** MAINTENANCE OF SECONDARY CHEMISTRY SPS/R REGULATING STANDBY POWER SOURCE MP E-65.12A 0 162 *** CLEAN AND INSPECTION OF SECURITY UNINTERRUPTIBLE POWER SUPPLY
- 2) Mf E-67.1 0 IL? ** LOCATING DC GROUNDS WITH THE OC SCOUT BATTERY DISTRIBUTION GROUND LOCATOR Mi L-67.?B C ILP *** CALIBRATION OF NONVITAL 125/250 VDC GROUND DETECTION SYSTEMS MP E-67.?E O lt? **0PERATING INSTRUCTIONS FOR ALBER CELL +PLUS PORTABLE BATTERY TESTER C) MP E-67.2D 1 162 **0PERATING INSTRUCTIONS FOR ALBER BCT-1000 BATTERY CAPACITY TEST SYSTEM Mi E-67.3A IB ILP ** ROUTINE PREVENTIVE MAINTENANCE OF STATION BATTERY CHARGERS MP E-67.3b 0 lL2 ** SIXTY-MONTH PREVENTIVE MAINTENANCE OF STATION BATTERY CHARGERS MP E-67.4B D 16? *** BALANCING SPECIFIC GRAVITY OF LEAD ACID CELLS 8 MP E-67.5B O IL2 **MAINTEtANCE OF BATTERY PACK EMERGENCY LIGHTS OUTSIDE POWER BLOCK M0 E-67.6 1 lL2 *** STATION BATTERY PREVENTIVE MAINTENANCE MP E-67.7 1 1&? ** ROUTINE PREVENTIVE MAINTENANCE OF C&D BATTERY CHARGERS M? E-71.lA 7 IL?
- HEAT T' ACE CABLE TESTING MF E-71.1B 11 IL2 **INSTALLAT]ON AND REPAIR OF THERMJN ECOND TRACE HEATING
- CABLE MF E-71.lC ? IL2 **SSK HEATING CABLE INSTALLATION AND REPAIR MP E-71.2A 2 ILP **CAllBRATION OF BORlC ACID TRACE MONITOR MP E-71.3A 1 IL? **lNSTALLATION AkD REPAIR OF TYPE HG-P ECONO TRACE HEATING CABLE MP E-72.1 2 ILP ***WEErLY CATHODIC PROTECTION SYSTEM MONITORING 8 MP E -72.2 1 IL2 ** MONTHLY CATHODIC PROTECTION SYSTEM MONITORING MP E-99.01 0 2 *MOV FLOW TEST - TD AFW FLOW CONTROL VALVES LCV-106, 107, 108, AND 109 MP E-99.0? 0 ? *M3V FLOW TEST - TDAFW STEAM SUPPL! VALVES FCV-37, 38 AND 95 MF E-99.03 0 2
- DIFFERENTIAL PPESSUFE TESTING OF FCV-355 MP E-99.05 0 2 *M3V FLOW TEST - CCW CONTAINMENT 150 VALVES 2-FCV-356, Gb 2-FCV-749, AND 2-FCV-363 MP E-99.07 0 2 *MOV FLOW TEST - CVCS HEADER AND RECIRC VALVES 8105, 8106, 8107 AND 8108 MP E-99.0B 1 2 *MOV FLOW TEST - RHR SYSTEM VALVES 8703, 8716A/B, AND 8809A/B S
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O 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 5A ELECTRICAL MAlHTENANCE PROCEDURES 1 TABLE OF CONTENTS .O NUMBER REV UNIT TITLE
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MP E-99.10 0 2 *MOV FLOW TEST - 51 INJECTION VALVES 8821A/8, 8835, AND 8802A/B 'O MP E-101A 0 1&2 ** INFRARED THERM 0 GRAPHY INSPECTIONS MP E-102.A 1 1&2
- ACOUSTIC LEAK DETECTION PROGRAM t
O O I 'O F
- O f
O I lO i i l l
'O l
lO I _______._______________________________________________._____.i
) 22 JUL 93 DIABLO CANYON POSER PLANT UNITS 1 AND 2 VOLUME SB MECHANICAL MAINTENANCE PROCEDURES TABLE OF CONTENTS HUMBER REV UNIT TITLE
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MP M-2.1 3 IL2 ** CONDENSATE PUMP & MOTOR REMOVAL MP M-2.2 5 1&2
- CONDENSATE POLISHING VESSEL WELLSCREEN AND RESIN TRAP 3 CLEANING AND MAINTENANCE MP M-2.4 0 1&2 *** CONDENSATE PUMP MAlHTENANCE MP M-3.2 7 1&2 *STE AM GENERATOR AUXILI ARY FEEDWATER PUMP MAINTENANCE (6 AND 9 STAGE)
MP M-3.3 4 IL2 **FEEDWATER PUMP TURBINE CYLINDER COVER AND ROTOR REMOVAL AND REINSTALLAT10N 3 MP M-3.4 3 lE2 ** TURBINE DRIVEN AUXILIARY FEEDWATER PUMP & TURBINE REMOVAL MP M-3.5 3 IL2
- MOTOR DRIVEN AUXILIARY FEEDWATER PUMP & MOTOR REMOVAL MP M-3.7 2 It2 *** TERR) TURBlNE THROTTLE TRIP VALVE (FCV-152) DISASSEMBLY AND REASSEMBLY MF M-3.9 1 IL2 ***FEEDWATER REGULATING VALVE MAINTENANCE MP M-4.1 4 1&2 **HIGH PRESSURE OR LOW PRESSURE TURBINE ROTOR REMOVAL AND 3 REINSTALLATION MP M 4.2 3 1&2 **HIGH PRESSURE OR LOW PRESSURE TURBINE BLADE RING REMOVAL AND REINSTALLATION MP M-4.3 3 IL2 ** TURBINE BEARING COVER REMOVAL AND REINSTALLATION MP M-4.4 3 IL2 **HIGH PRESSURE TURBINE OUTER COVER REMOVAL AND REINSTALLATION 3 MF M-4.5 3 1&2 ** LOW PRESSURE TURBINE OUTER COVER AND CYLINDER COVERS 1 AND 2 REMOVAL REINSTALLATION MP M-4.6 3 162 ** LOW PRESSURE TURBINE CROSSOVER TEE REMOVAL AND REINSTALLATION MF M-4.7 3 IL2 ** MAIN STEAM REHEATER HIGH PRESSURE AND LOW PRESSURE TUBE BUNDLE REMOVAL AND REINSTALLATION 3 MP M-4.9 7 lt2 ** STEAM GENERATOR SAFETY VALVE LIFT POINT SETTING USING STEAM MP M-4.10 13 IL2 *** STEAM GENERATOR SAFETY VALVE RECONDITIONING MP M 4.ll 4 IL2 ** MAIN STEAM SAFETY VALVE SETTING WITH HYDRAULIC ASSIST ML M-4.12 0 1&2 **lNSTALLATION & REMOVAL OF TEMPORARY CROSS-0VER PIPE SUPPORTS 3 MP M-4.13 1 1&2 ** VISUAL INSPECTION OF STEAM GENERATOR TUBESHEET SECONDARY SIDE MP M-4.13A 0 1&2 ** STEAM GENERATOR PRIMARY & SECONDARY MANWAY & HANDHOLE BOLT HOLE REPAIR MP M-4.14 0 JL2 ** AUXILIARY FEEDWATER PUMP TURBINE MAINTENANCE MP M-4.16 1 IL2
- STEAM GENERATOP SAFETY VALVE MANUAL LIFTING'& GAGGING 3 MP M-4.1B 8 1&2 *** VERIFICATION OF LIFT POINT USING FURMANITE'S TREVITEST EQUIPMENT FOR THE MAIN STEAM SAFETY VALVES' MP M-4.19 3 1&2 **HP TURBINE CONTROL 1 VALVE MAINTENANCE MP M-4.20 2 1&2 **HP TURBINE STOP VALVE hAINTENANCE MP M-4.21 1 1&2 **lNTERCEPTOR AND REHEAT STOP VALVE MAINTENANCE MP M-4.25 4 IL2 **AUxlLIARY FEEDWATER PUMP TURBlNE GOVERNDP MAINTENANCE D
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- 2) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SB MECHANICAL MAINTENANCE PROCEDURES
) TABLE OF CONTENTS NUMBER REV UNIT TITLE
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MP M-4.26 0 1&2 *** MAIN STEAM SAFETY VALVE SETPOINT ESTABLISHMENT USING I) MP M-6.1 1 1&2 FURMANITE'S TREVITEST ON A TEST STAND MP M-7.2 ** BUILDING HEATING REB 0ILER TUBE BUNDLE & CASING HANDLING 22 1&2 **lNSTALL REACTOR CLOSURE HEAD MP M-7.3 15 1&2 *** REACTOR VESSEL CLOSURE HEAD REMOVAL MP M-7.4 8 1&2 *** REACTOR HEAD 0-RING REMOVAL AND REPLACEMENT MP M-7.5 13 1&2
** CORE EXIT THERMOC0UPLE N0ZZLE ASSEMBLY (CETNA) OPENIN AND CLOSURE
- 2) MP M-7.6L 3 1&2 ** REACTOR VESSEL UPPER INTERNALS REMOVAL MP M-7.6B 4 1&2 ** REACTOR VESSEL UPPER INTERNALS INSTALLATION MP M-7.7A 1 1&2 *** REACTOR VESSEL LOWER INTERNALS REMOVAL MP M-7.7B 1 1&2 *** REACTOR VESSEL LOWER INTERNALS INSTALLATION MP M-7.10 8 1&2 **lNSTALLATION & TESTING OF THE CAVITY SEALS i MP M 7.11 4 1&2 C) ** CONTROL ROD DRIVE MOTOR COOLING FAN HANDLING MP M-7.13 16 1&2 MP M-7.17 *** FULL LENGTH RCCA DRIVE SHAFT UNCOUPLING AND RECOUPLING 5 1&2 **DISASSEMBL), INSPECTION, ASSEMBLY & INSTALLATION OF REACTOR COOLANT PUMPS MP M-7.19 2 1&2 ** REPLACEMENT OF PRES $URIZER IMMERSION HEATERS MP M-7.20 5 1&2 ** REMOVAL & REASSEMBLY OF PRESSURIZER MANWAY COVER MP M-7.22 14 1&2 *** PRESSURIZER SAFETY VALVE RECONDITIONING
.[) MP M-7.23 3 1&2 ** PRESSURIZER SAFETY VALVE LIFT POINT SETTING USING AIR SET PRESSURE DEVICE MP M-7.25 15 1&2 ** REMOVAL & INSTALLATION OF STEAM GENERATOR SECONDARY MANWAY AND HANDHOLE COVERS MP M-7.27 2 1&2
- MOTOR GENERATOR SET MOTOR & GENERATOR HANDLING MP M-7.?B 3 1&2
- ) ** HANDLING REACTOR HEAD STUDS & NUTS WITH LIFTING BASKET i MP M-7.29 2 1&2 ** REACTOR COOLANT PUMP HATCH COVER HANDLING MP M-7.30 1 1&2 ** REACTOR COOLANT PUMP HANDLING MP M-7.31 5 1&2 ** REACTOR COOLANT PUMP (RCP) MOTOR AND MOTOR STATOR HANDLING MP M-7.32 12 1&2 ***lNCORE INSTRUMENTATION SEAL TABLE PREPARATION MP M-7.33 7 1&2 ** REACTOR COOLANT PUMP FLYWHEEL REMOVAL & REINSTALLATION
- i. MP M-7.36 13 1&2
[) *** PRESSURIZER SAFETY VALVE LIFT POINT SETTING USING STEAM MP M-7.37 4 1&2 ** SERVICE READlNESS AND PERIODIC INSPECTION OF THE REAC10R i VESSEL HEAD AND INTERNALS LIFTING DEVICES MP M-7.38 0 1&2 **THE CLEANING OF INCORE FLUX THIMBLES MP M-7.39 1 1&2
** MECHANICAL CLEANING OF REACTOR CLOSURE HEAD STUD HOLES MP M-7.40 1 1&2 **lNSTALL/ REMOVE REACTOR VESSEL STUD HOLE PLUGS MP M-7.41 2 1&2
[) *** REMOVAL & INSTALLATION OF REACTOR COOLANT PUMP COUPLING MP M-7.42 3 1&2 *** REACTOR COOLANT MOTOR-TO-PUMP ALIGNMENT ! MP M-7.43 6 1&2 *** REMOVAL, INSPECTION & INSTALLATION OF MECHANICAL SEAL-REACTOR COOLANTPUMP (MOTOR IN PLACE) MP M-7.44 1 1&2 ** STEAM GENERATOR PRIMARY MANWAY COVER REMOVAL & INSTALLATION D
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C) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 5B MECHANICAL HAINTENANCE PROCEDURES 4 TABLE OF CONTENTS .O NUMBER REV UNIT TITLE
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MF M-7.45 0 1&2 ** VERIFICATION OF LIFT POINT USING FURMANITE'S TREVITEST EQUIPMENT FOR THE PRESSURIZER RELIEF VALVES () MP M-7.46 4 1&2 ** REACTOR COOLANT PUMP MOTOR INSPECTION & MAINTENANCE MP M 7.47 1 1&2
- CLEANING REACTOR HEAD STUDS MP M-7.48 1 1&2 **CRDM SEAL CLAMP INSTALLATION MP M-7.49 0 1&2 ** CLEANING OF INCORE FLUX THIMBLES, APEX METHOD MP M-7.50 0 1
- REACTOR HEAD VENT TRAIN 1 AND 2 VALVE SEAT LEAKAGE TEST q
- EACH REFUELING CYCLE
() MP M-7.50 0 2
- REACTOR HEAD VENT TRAIN 1 AND 2 VALVE SEAT LEAKAGE MP M-7.53 0 IL2 ** REACTOR COOLANT PUMP MOTOR TEN (10) YEAR INSPECTION PROCEDURE MP M-8.! 4 1&2 ** BORIC ACID TRANSFER PUMP REMOVAL, REPAIR & INSTALLATION MP M-8.2 6 1&2 ** CENTRIFUGAL CHARGE PUMP (CCP) REMOVAL, REPAIR AND REASSEMBLY
.C) MP M-8.3 2 1&2 ** GAS STRIPPER FEED PUMP REMOVAL REPAIR & INSTALLATION
+ MP M-8.4 6 1&2 **RECIPROCATNG CHARGE PUMP MAINTENANCE MP M-8.5 1 1&2 ** LIQUID HOLDUP TANK RECIRCULATION PUMP REMOVAL REPAIR & 4 1NSTALLAT10H , MP M-8.6A 4 1&2 ** FILTER ELEMENT REPLACEMENT FOR REACTOR COOLANT LETDOHN , 1-1, AND 2-1 AND SEALWATER RETURN () MP M-8.6B 2 1&2 ** FILTER ELEMENT REPLACEMENT FOR SEALWATER INJECTION FILTERS MP M-8.6C 2 1&2 ** FILTER ELEMENT REPLACEMENT FOR REACTOR COOLANT LETDOWN FILTERS 1-2 AkD2-2 MP M-8.6D 1 1&2 ** FILTER ELEMENT REPLACEMENT FOR BORIC ACID EVAPORATOR CONCENTRATES AND 10N EXCHANGE FILTER 4 MP M-8.6E 2 1&2 ***FiliER ELEMENT REPLACEMENT FOR THE BORIC ACID EVAPORATOR
.C) CONDENSATE FITTER MP M-8.6F 1 1&2 *** FILTER ELEMENT REPLACEMENT FOR THE BORIC ACID FILTER 4
MP M-8.7 2 1&2 **LET-DOWN & SEAL WATER HEAT EXCHANGER HANDLING MP M-8.8 0 1&2 ** CENTRIFUGAL CHARGING PUMP HANDLING I MP M-9.1 7 1&2 ** SAFETY INJECTION PUMP MAINTENANCE MP M-9.2 5 1&2 ** SAFETY INJECTION PUMP & MOTOR HANDLING () MP M-9.3 1 1&2 ** SAFETY INJEC110N PUMP INTERNAL MP M-10.1 9 1&2 ** RESIDUAL HEAT REMOVAL PUMP MAINTENANCE MP M-10.2 4 1&2 ** RESIDUAL HEAT REMOVAL PUMP MOTOR & IMPELLER HANDLING MP M-10.3 3 1&2 ** RESIDUAL HEAT REMOVAL HEAT EXCHANGER SHELL & TUBES-HANDLING MP M-12.1 1 1&2 ** CONTAINMENT SPRAY PUMP () MP M-13.2 8 1&2 *** FILTER ELEMENT REPLACEMENT FOR REFUELING WATER l PURIFICATION FILTERS MP M-13.4 2 1&2 *** FILTER ELEMENT REPLACEMENT FOR SPENT FUEL PIT AND SPENT FUEL PIT SKIMMER ' MP M-13.5 2 1&2 ** FILTER ELEMENT REPLACEMENT FOR SPENT FUEL Pil RESlN TRAP O l i C l j
9 22 Jul 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SB HECHANICAL HAINTENANCE PROCEDURES TABLE OF CONTENTS O NUMBER REV UNil TITLE
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MP M-13.6 0 IL2 ** SPENT FUEL POOL PUMPS 1-2 & 2-2 (HAYWARD TvlER) DISASSEMBLY, REPAIR & REASSEMBLY S MP M-14.1 2 1&2
- COMPONENT COOLING WATER PUMP DISASSEMBLY REPAIR &
REASSEMBLY MP M-14.2 3 1&2 ** COMPONENT C00LlNG WATER PUMPS MOTOR & PUMP HANDLING MP M-14.3 4 It2 ** COMPONENT COOLING WATER (CCW) HEAT EXCHANGER HEAD, CHANNEL BARREL AND TUBE BUNDLE HANDLING MP M-14.4 0 1E2 ** REACTOR COOLANT PUMP MOTOR OIL HEAT EXCHANGER FLUSHING GD PROCEDURE MP M-16.1 4 IL2 ** MAKEUP WATER TRANSFER PUMP DISASSEMBLY REPAIR & REASSEMBLY MP M-17.1 3 IL2 ** SCREEN WASH PUMP MOTOR & PUMP REMOVAL MP M-17.2 4 1&2 ** AUXILIARY SALT WATER PUMP & MOTOR REMOVAL & REINSTALLATION MP M-17.3 5 1&2 ** CIRCULATING WATER PUMP ROTOR, SHAFT, IMPELLER, STATOR, AND HATCH COVERREMOVAL D MF M-17.4 3 1&2 ** CIRCULATING WATER DISCHARGE VALVE HANDLING MP M-17.5 3 1&2 **lNTAKE BAR RACK REMOVAL AND REINSTALLATION MP M-17.6 3 1&2 ** TRAVELING WATER SCREEN REMOVAL MP M-17.7 3 It? ** CIRCULATING WATER SCREEN GATE & AUXILIARY SALT WATER PUMP SCREEN AND GALE REMOVAL AND REINSTALLATION MP M-17.8 3 1&2 ** AUXILIARY SALT WATER (ASW) PUMP INTAKE BAY GATE REMOVAL D AND REINSTALLATILW MP M-17.9 6 IL2 **AUYlLI ARY SAll WATER PUMP MAINTENANCE MP M-17.ll 3 1&2 ** CIRCULATING WLTER PUMP MAINTENANCE MP M-17.12 1 1&2 *** HEAT TREATMENT OF MISCELLANEOUS SEAWATER PIPING MP M-18.1 5 IL2 ** FIRE PUMP DISASSEMBLY, PrPAIR AND REASSEMBLY MP M-18.2 7 IL2
- TAGGING, CHAPGlNG, INSPECT 30N & HYDROSTATIC TESTING OF D PORTABLE FIRE EXTINGUISHERS MP M-18.6 3 IL2 ** SEMI-ANNUAt PREVENTIVE MAINTENANCE ON PERMANENTLY
]NSTALLED CO2 CYLINDERS FOR THE CIRCULATING WATER PUMP MOTORS (WPSI)
MP M-19.1 2 IL2 ** REACTOR COOLANT DPAIN TANK PUMP DISASSEMBLY AND REASSEMBLY MP M-19.2 1 IL2 ** EQUIPMENT DRAIN RECEIVER PUMP DISASSEMBLY REPAIR & D REASSEMBli MP M-19.3 6 1&2
- FUEL TRANSFER BLANK FLANGE REMOVAL & INSTALLATION MP M-19.4 1 IL2 ** FILTER ELEMENT REPLACEMENT FOR SPENT RESIN TRANSFER FILTERS 0-1 AND 0-2 MP M-20.1 1 IL2 ** PREPARATION FOR MAIN TURBINE LUBE OIL FLUSH MF M-21.1 4 IL2 ** DIESEL FUEL OIL TRANSFER PUMP MAINTENANCE D MP M-21.2 2 IL2
- DIESEL ENGINE GENERATOR AIR COMPRESSOR MAINTENANCE MP M-21.5A 0 IL2 ** DIESEL ENGINE WATER PUMP MAINTENANCE MP M-21.58 0 1&2 ** DIESEL ENGINE WATER PUMP MODEL 22500168-1 MAINTENANCE MP M-21.6 2 IL2 ** DIESEL ENGINE TURBOCHARGER MAINTENANCE MP M-21.7A 0 IL2 *** DIESEL ENGINE FUEL OIL BOOSTER PUMP h N-21.78 0 IL2 ** DIESEL ENGINE 2-3 FUEL Olt BODSTER PUMP O
3
i C) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SB MECHANICAL MAINTENANCE PROCEDURES O TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ====
===========================================================
MP M-21.8 10 1&2 ** DIESEL ENGINE GOVERNOR ACTUATOR MP M-21.9 1 1&2 ** DIESEL ENGINE LUBE Olt PUMP C) MP M-21.9B 0 1&2 **DEG 2-3 LUBE OIL PUMP MP M-21.10 3 1&2 ** DIESEL ENGINE PRE-LUBE OIL PUMP MAINTENANCE MP M-21.ll 1 1&2 ** DIESEL ENGINE FUEL INJECTION PUMP MAINTENANCE MP M-21.12 3 1&2 ** DIESEL ENGINE CYLINDER HEAD MP M-21.13 3 1&2 ** DIESEL ENGINE RADIATOR MP M-21.14 1 1&2
- CLEANING INSPECTION & TESTING OF DIESEL OIL STORAGE TANK
() MP M-21.15 0 1&2 ** DIESEL GENERATORS HEAVY LOAD COMPONENTS REMOVAL MP M-21.16 2 1&2 *** DIESEL ENGINE CAMSHAFT MP M-21.17 2 1&2 *** DIESEL ENGINE TURNING ARRANGEMENT MP M-21.19 3 1&2 ** DIESEL ENGINE GOVERNOR CONTROL LINKAGE MP M-21.20 2 1&2 ** DIESEL ENGINE RADIATOR FAN DRIVE MP M-21.21 2 IL2 ** DIESEL ENGINE RADIATOR COOLING FAN DRIVE RIGHT ANGLE GEAR C) BOX MP M-21.22 1 IL2 ** DIESEL ENGINE OVERSPEED TRIP MECHANISM MP M-21.23 3 1&2 ** DIESEL ENGINE FUEL PUMP CONTROL SHAFT AND CROSSOVER LINKAGE MP M-21.24 6 IE? *** DIESEL ENGINE AEROQUIP HOSE MAINTENANCE MP M-21.25 0 1&2 ** REBUILDING DIESEL STARTING AIR MOTORS C) MP M-21.25B 0 1&2
- EXTERNAL VISUAL MAINTENANCE INSPECTION AND/0R REPLACEMENT OF DEG 2-3 STARTING AIR MOTOR MP M-21.27 3 1&2 *** DIESEL GENERATOR ROTOR REMOVAL AND INSTALLATION MP M-21.28 1 1&2 *** DIESEL GENERATOR CRANKSHAFT DEFLECTION MP M-21.29 0 1&2 **lNJEC OR N0ZZLE/ HOLDER AND HIGH PRESSURE FUEL LINE MAINTENANCE C) Mi M-21.30B IL2 1 **DEG 2-3 POWER ASSEMBLY INSPECTION MP M-21.31 1 1&2 ** MAIN BEARING INSPECTION MP M-21.32 0 1&2 *** DIESEL ENGINE VALVE ADJUSTMENT MP M-22.1 3 1&? ** GENERATOR ROTOR HANDLING MP M-22.2 2 1&2 ** EXCITER HOUSING HANDLING MP M-22.3 3 1&2 ** EXCITER HANDLING C) MP M-23.1 1&2 1 *HEPA FILTER REPLACEMENT MP M-23.2 3 1&2 ** CONTAINMENT FAN COOLER MOTOR HANDLING MP M-23.3 3 1&2 *** AIR PARTICULATE DETECTOR PUMP MAINTENANCE MF M-23.4 10 1&2 ** PREVENTIVE MAINTENANCE OF PLANT VENTILATION FANS, ASSOCIATED DAMPERS AND FILTERS MP M-23.5 1 1&2 **0UTAGE MODIFICATION FOR CONTAINMENT PENETRATION 63 C) MP M-23.7 1 1&2
- BUTTERFLY DAMPERS MAINTENANCE, INSPECTION AND SEAT REPLACEMENT MP M-23.8 0 1&2 *** PREVENTIVE MAINTENANCE OF CONTAINMENT FAN COOLERS AND ASSOCIATED COMPONENTS MP M-23.9 1 1&2 ***0UTAGE MODiflCATION FOR CONTAINMENT PENETRATIONS 58 AND 60 0
C)
22 JUL 93 DIABLO CANVON POWER PLANT UNITS 1 AND 2 i VOLUME SB MECHANICAL HAINTENANCE PROCEDURES ) TABLE OF CONTENTS I NUMBER REV UNIT TITLE
=============== === ==== ===========================================================
+
MP M-23.10 1 1&2 ***0UTAGE MOD. FOR CONTAINMENT PENETRATION 63 (USING ) MP M-42.1 5 IL2 REDUCER / ISOLATION COVER)
*** FUEL TRANSFER SYSTEM INSPECTION & PREVENTIVE MAINTENANCE ,
HP M-45.1 6 1&2 *** CONTAINMENT EQUIPMENT HATCH DOOR OPENING & CLOSING MP M-50.1 1 162 ** RIGGING AND HOISTING EQUIPMENT INSPECTION AND MAINTENANCE MP M-50.2 3 1&2 **P&H MOBILE CRANE HANDLING MP M-50.3 8 1&2 ***0VERHEAD, GANTRY & MOBILE CRANE INSPECTION, TESTING & MAINTENANCE MP M-50.4 4 1&2 ** FUEL HANDLING BUILDING (FHB) CRANE OPERATION AND MOVEABLE WALL RELOCA110N MP M-50.5 1 IL2
- HANDLING NEW FUEL IN SHIPPING CONTAINERS -
MP M-50.7 5 IL2 ** INSTALLING & REMOVlNG THE REACTOR VESSEL HEAD LIFTING DEVICE MP M-50.8 4 IL2 **lNSTALLATION AND REMOVAL OF THE REACTOR VESSEL UPPER AND ) LOWER INTERNALS' LIFTING DEVICE MP M-50.9 9 IL2 ** MISSILE SHIELD HANDLING MP M-50.10 2 It? ** FUEL SHIPPING CASK HANDLING MP M-50.ll 6 IL2 ** SPENT FILTER 1RANSFER CASK 0-1 HANDLING MP M-50.13 0 162 ** PREVENTIVE MAINTENANCE ON THE CONTAINMENT DOME SERVICE ) MF M-50.14 0 IL2 CRANE
** POLAR CRANE SEISMIC GUIDE STRUT ADJUSTMENT MP M-50.16 0 102 *SPECIAL SERVICE HOISTS, JIB CRANES AND MONRAILS INSPECTION MP M-50.17 1 1&2 ** REFUELING CAVITY STAIRCASE HANDLING ,
MP M-50.19 1 IL2 *RIACTOR VESSEL TEMPORARY COVER HANDLING t MP M-50.20 9 IL2 *** LOADING PRE-LOADED LINERS INTO THE NUPAC MODEL 14/210 RADWASTE SHIPPING CASK 3 MP M-50.21 8 IL2 *** LOADING DRUMS IN10 THE NUPAC MODEL 14/210 RADWASTE SHIPPING CASK MP M-50.22 7 1&2 *** PROCESSING WASTE IN THE 10-142 RADWASTE SHIPPING CASK , MP M-50.23 6 IL2 *** LOADING PRL-LOADED LINERS INTO THE NUPAC 10-142 RADWASTE . SHIPPING CASK MP M-50.24 ) MP M-50.25 1 1&2 0 IL2
*** SPENT FILTER TRANSFER CASK 0-4 HANDLING
*** SPENT FILTER TRANSFER CASK 0-3 HANDLING MP M-50.26 1 1&2 *** SPENT FILTER TRANSFER CASK 0-5 HANDLING MP M-50.27 0 1&2 *** SPENT FILTER TRANSFER CASK 0-2 HANDLING FOR CONTAINMENT-SERVICE MP M-51.1 3 IL2
- PACKING VALVES USING BRAIDED TYPE PACKING MP M-51.2 5 lt?
- SITE FABRICAT]DN OF' PRE-FORMED GRAPHITE PACKING RINGS 3 MP M-51.3 5 IL2
- PACKING VALVES IN PLACE USING GRAPHITE TAPE MP M-51.4 4 1&2 ** REPAIR OF ROCKWELL-EDWARD VALVES WITH SEAL-WELDED BONNETS -'
MP M-51.5 12 1&2 ***SETPOINT CAllBRATION OF SAFETY RELIEF VALVES MP M-51.6 4 1&2
- VALVE EXTERNAL INSPECTIONS AND MAINTENANCE MP M-51.7 10 1&2 ***ITT GRINNELL DIAPHRAGM VALVE MAINTENANCE ,
) '
i () 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 , VOLUME SB MECHANICAL MAINTENANCE PROCEDURES l TABLE OF CONTENTS c) ; NUMBER REV UNIT TITLE
=============== === ==== ===========================================================
MP M-51.10 6 1&2 ***lNSTALLATION & ADJUSTMENT OF T-RING ON RCV-11 & 12 AND
.g FCV-660 & 661 MP M-51.11 3 1&2 ***CONTR0MATIC PNEUMATIC SPRING RETURN ACTUATOR MAINTENANCE MP M-51.12 2 1&2 **CONTROMATICS BALL VALVES MP M-51.14 4 1&2
- GENERIC CHECK VALVE INSPECTION MP M-51.15 1 1&2 **VELAN SWING CHECK VALVE DISASSEMBLY INSPECTION &
REASSEMBLY MP M-51.21 4 1&2 **10% ATMOSPHERIC STEAM DUMP VALVE MAINTENANCE E) Mi M-51.22 9 1&2 *** VALVE PACKING USING PRE-FORMED GRAPHITE WITH BRAIDED END RINGS Mi M-51.23 1 1&2 *** VALVE PACKING USING PRE-FORMED GRAPHITE WITH COMPOSITE END RINGS MF M-51.24 1 1&2 **lN PLACE SET PolNT CAllBRATION FOR REllEF VALVES MP M-51.26 2 1&2
- LUBE OIL AND FUEL OIL RECIRCULATION VALVE INSPECTION AND C)
ADJUSTMENT MP M-51.28 2 1&2 ** DIAPHRAGM REPLACEMENT FOR MASONEILAN ACTUATORS SIZES 9,11,13,15,18 & 24 MP M-51.29 0 1&2 ** CRANE (CHAPMAN) TILTING DISC CHECK VALVE ASSEMBLY MP M-51.30 0 1&2 ** FISHER VALVE MANUAL OPERATOR INSTALLATION AND SETUP MP M-51.32 2 1&2 ***ATWOOD & MORRILL SWING DISC CHECK VALVE MAINTENANCE C) MP M-51.33 0 1&2
- CROSBY N0ZZLE PELIEF VALVE MAINTENANCE '
MP M-51.34 1 1&2 **35% AND 40% STEAM DUMP VALVE MAINTENANCE MP M-51.35 0 1&2 **lNTERNAL MEASUREMENT OF WEDGE GATE VALVES MP M-52.1 1 1&2 ** MECHANICAL SEAL REMOVAL & INSTALLATION (BORG-WARNER) MP M-52.2 4 1&2 ** MECHANICAL SEAL PEMOVAL & INSTALLATION (CRANE) MP M-54.1 6 1&2 ** BOLT TOROUlNG C) ' MP M-54.3 4 1&2 *** FREEZE SEALING 0F PIPING MP M-54.4 12 1&2 ** SPIRAL WOUND GASKET REPLACEMENT GUIDE ; MP M-55.1 11 1&2 ** HANDLING /TESilNG/ REPAIR / MAINTENANCE OF GRINNELL HYDRAULIC SNUBBERS MP M-55.2 9 1&2 *** HANDLING / TESTING / MAINTENANCE OF PAUL-HUNROE SNUBBERS MP M-55.3 8 1&2 ** HANDLING / TESTING / REPAIR / MAINTENANCE OF PSA SNUBBERS C) MP M-55.4 6 1&2 ** HANDLING /TES11NG/ REPAIR / MAINTENANCE OF ANCHOR / DARLING SNUBBERS MP M-55.5 5 1&2 ** FABRICATION, INSTALLATION, AND MODIFICATION OF PIPE SUPPORTS MP M-56.1 9 1&2
- SYSTEM PRESSURL TEST MP M-56.2A 0 1 **lNSTALL/ REMOVE CONDENSATE /FEEDWATER SYSTEM DRY LAY-UP
() MP M-56.28 1 2 ***lNSTALL/ REMOVE CONDENSATE /FEEDWATER SYSTEM DRY LAY-UP MP M-56.4 0 1&2
- MANUFACTURE OF BEARINGS BUSHING & SLEEVES-FROM SYNTHETIC MATERIALS MP M-56.6 9 1&2 *** PERSONNEL HATCH PREVENTIVE MAINTENANCE MP M-56.7 4 1&2 ** LUBRICANT SAMPLING, TESTING, AND ANALYSIS
'O i ca 1
) ?? JUL 93 DIABLO CANYON POBER PLANT UNITS 1 AND 2 VULUME SB HECHANICAL MAINTENANCE PROCEDURES j TABLE OF CONTENTS NUMBER REV UNIT TITLE
.............== === ...= . . . . . . . . = = = . . . . = . . . . = . . . . . . . . . . . . . . . . . . . . . . . . = = = = = = . . . . . =
MP M-56.8 0 1&2 **PREVENilVE MAINTENANCE FOR CONTAINMENT EMERGENCY ) MP M-56.9 3 1&2 PERSONNEL HATCH
- CONTROL OF TEMPORARY ON-LINE LEAK SEALING USING VENDORS OR i CONTRACTOR'S MP M-56.10 4 102 *** PIPING FABRIEATION, INSTALLATION, REPAIR, OR SYSTEM ALTERATION MP M-56.11 0 1&2
- DOOR MAINTENANCE MP M-56.13 2 1&2
) MP M-56.15 4 1&2
**0RIFICE FABRICATION & INSTALLATION
*lNTERNAL VALVE CLEANLINESS / LAPPING MP M-56.16 1 1&2
- HEAT EXCHANGER TUBE CLEANING MP M-56.17 0 1&2 **BUNA-N 0-RING FABRICATION MP M-56.18 2 1&2 **EQUOTIP HARDNESS TESTING OF STEELS & STAINLESS STEELS MP M-56.19 0 1&2 ** SHAFT / COUPLING ALIGNMENT FOR ROTATING EQUIPMENT MP M-56.21 1&2
) MP M-58 1 0 1&2
*** SALT WATER HEAT EXCHANGER TUBE AND TUBE SHEET PLUGGING-
- DIESEL OIL TRANSFER FILTER ELEMENT REPLACEMENT MP M-59 0 1&2
- DIESEL LUBRICATING OIL FILTER ELEMENT REPLACEMENT - -
MP M-83 1 1&2 ** WASTE GAS COMPRESSOR DISASSEMBLY REPAIR & REASSEMBLY D )
) l l
l l l 1
i l 8I 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC ILC KAINTENANCE PROCEDURES TABLE OF CONTENTS NUMBLR REV UNil TITLE
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MP l-1.1-1 2 IL2 *SSPS DEMULTIPLEXER SYSTEM VERIFICATION MP l-1.1-2 1 IL2 ** AMP TERMIPOINT CLIP VISUAL INSPECTION AND PULL TESTING I) MP l-1.1-3 0 IL2 *5SPS MULTIPLEXER TEST SWITCH POST MODIFICATION / MAINTENANCE TEST MF 1-1.2-1 9 1&2 *CAllBRATION OF THE WIDE RANGE NUCLEAR INSTRUMENTATION MONITORS NE-51 AND NE-52 MP 1-1.3-4 4 ILP
- CALIBRATION OF THE SPING 3A MICROPROCESSOR CONTROLLED CONTINUOUS AIR MONITOR
[) MP I-1.3-6 0 IL2 **CAllBRATION OF ND66 MULTICHANNEL ANALYZER USED TO CHECK DISCRIMINATOR BIAS DRIFT IN RMS CHANNEL MP l-1.3-B 2 !L? ** SETUP CANBERRA MULTICHANNEL ANALYZER MODEL SE20 TO CHECK DISCRIMINATORBIAS FOR RMS CHANNEL MP I-1.3-10 1 IL2 *CAllBRATION OF THE GROSS FAILED FUEL DETECTOR MP 1-1.4-1A 2 IL2 *(OTSC 6/16/93-PATTERSON)HAGAN SIGNAL SUMMATOR ACCEPTANCE [) TEST PROCEDUPE M 1-1.4-]E 1 162 *HAGAN SIGNAL SUMMATOR HIGH SIGNAL SELECT ACCEPTANCE TEST PROCEDURE Mr 1-1.4-cA 1 IL2 ** DYNAMIC TESTING OF HAGAN LEAD-LAG MODULES MF 1-1.4-2E 3 IL2 *HAGAN LEAD / LAG AND DERIVATIVE AMPLIFIER ACCEPTANCE TEST PROCEDURE [) MF I-1."-3; O IL2 *HAGAN SIGNAL ISOLATOR ACCEPTANCE TEST PROCEDURE Mr 1-1.4-4L 4 ILP *HAGAN ALARM L SIGNAL COMPARATOR ACCEPTANCE TEST PROCEDURE Mi 1-1.4-5; O ]L2 **HAGAN FUNCTION GENERATOR ACCEPTANCE TEST PROCEDURE Mi 1-2.,-Et 0 IL2 *HAGAN PEMDTE/ MANUAL SETPOINT STATION ACCEPTANCE TEST PROCEDURE Mi 1-1.4-7; O IL2 *HAGAN MV/! ACCEPTANCE TEST PROCEDURE [) Mc i-1.4-E- 3 162 **(OTSC 6/16/93--PATTERSON) HAGAN MULTIPLIER / DIVIDER ACCEPTANCE TEST MF 1-1.4-9A 5 162 *HAGAN MANUAL / AUTOMATIC CONTROLLER SYSTEM MF 1-1.4-10: 1 IL2 *HAGAN LOOP POWER SUPPLY ACCEPTANCE TEST PROCEDURE MP l-1.4-llA 0 lLP *HAGAN OPTIMAC ELECTRONIC RECORDER ACCEPTANCE TEST PROCEDURE MP l-1.4-12A 0 lL2 ** REM 3 VAL AND INSTALLATION OF INSTRUMENT PANEL SEISMIC C) BRACING MP l-1.5-1 1 IL2 ** SET MAIN ANNUNCIATOR TIME AND DATE MP l-1.5-2 0 IL?
- ENERGIZING AND DE-ENEPGIZING THE VISUAL ANNUNCIATOR MP l-1.5-3 0 IL2
- ENERGIZING & DEENERGIZING THE OPERATIONS RECORDER MF 1-1.5-4 0 1&2 ** REMOVE & REPLACE PRINTED CIRCUIT CARD (S) IN THE MAIN ANNUNCIATOR DURING OPERATION O MP l-1.5-5 0 IL2 **lSOLATION OF A GROUNDED INPUT IN THE MAIN ANNUNCIATOR MP I-1.5-6 2 lL2 ** FUNCTIONAL TEST OF A MOMENTARY ALARM FOR REFLASHER SEQUENCE - (AN-154C) PRINTED CIRCulT CARD (MAIN ANNUNCIATOR)
MP l-1.5-7 0 IL2
- PROCEDURE FOR WRITING ONTO DRUM OF MAIN ANNUNCIATOR Mi' I-1.5-B 0 IL2
- MAIN ANNUNClATCR VERIFICATION O
O
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC ILC MAINTENANCE PROCEDURES , () TABLE OF CONTENTS NUMBER REV UNIT TITLE
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MP l-1.6-1 0 1&2
- LOGIC TEST OF DIGITAL ROD POSITION INDICATION SYSTEM g) MP l-1.6-2 0 1&2
- PROCEDURE FULL LENGTH ROD CONTROL MECHANISM C0ll RESISTANCE ~ CHECKS MP l-1.6-3 1 1
- ROD CONTROL POWER SUPPLY CALIBRATION AND ALARM VERIFICATION MP l-1.6-3 0 2
- ROD CONTROL POWEP SUPPLY CAllBRATION AND ALARM VER]FICATION t MP l-1.8-6 0 1&2 ** CHANGING Oil FROM THE GEAR BOX OF INCORE DETECTOR DRIVE MECHANISM
() MP l-1.6-7 0 1&2 *** MOVABLE INCORE DETECTOR SYSTEM ROTARY TRANSFER DEVICE MAINTENANCE MF 1-1.10-1 4 1&2 *DRPI-DETECTOR / ENCODER CARD CHECKS MP l-1.10-2 1 1&2 *DRPI - C0ll STACK AND CABLING CHECK MP I-1.13-1 1 1&2 *P-2000 COMPUTER D.E.H. SYSTEM PROGRAMS LOADING PROCEDURE M; l-1.13-2 0 1&2 *P-2000 COMPUTER D.E.H. TIME UPDATE-DISPLAY PROCEDURE MP l-1.13-3 2 1&2 **P-2000 DEH SYSTEM PROGRAM LOADING PROCEDURE USING THE E) ' ML-2000 AUTOMATIC MEMORY LOADER CARD MP I-1.13-4 0 1&2 *P-2000 DEH SYSTEM PROGRAM LOADING PROCEDURE FOR THE TERMINAL EMULATOR PROGRAM Mf I-1.13-5 1 1&2 *P-2000 DEH SYSlEM PROGRAM LOADING PROCEDURE FOR THE GRAPHICS OPTION MP l-1.13-6 0 1&2 () *P2000/DEH SYSTEM PREPARATION FOR OPTIMUM PERFORMANCE PRIOR TO STARTUP MP I-1.13-10 0 1&2 *DEH VALVE MANAGEMENT CURVE VERIFICATION MP l-1.13-11 1 1 *DEH/P2000 TURBINE CONTROL SYS CAPACITOR TEST AND POWER , SUPPLY VERIFICATION > MP l-1.13-11 0 2 *DEH/P2000 TURBlNE SYS CAPACITOR TEST AND POWER SUPPLY () VERIFICATION MP l-1.16-7 1 IL2 ***PPC 16 CHANNEL ANALOG INPUT CARD RTP7334/01 REPLACEMENT
& CALIBRATION MP l-1.16-8 0 1&2
- PLANT PROCESS COMPUTER (PPC) SYSTEM RESTART MP l-1.17-1 0 1&2
- CALIBRATION OF ERFDS-SPDS ANALOG INPUTS i MP l-1.18-1 5 1&2
- GUIDELINES FOR THE SERVICING & TESTING OF THE PYR-A-LARM l FIRE DETECTORS C)
MP l-1.18-2 0 1&2 #
*** FIRE PROTECTION SYSTEM COMPUTER REMOVAE AND RETURN TO SERVICE l MP l-1.21-1 1 1&2 ** SUPPLEMENTARY SEISMIC MONITORING SYSTEM TAPE REPLACEMENT AFTER RECEIVING SEISMIC EVENT {
MP l-1.21-2 0 1&2 ** SUPPLEMENTARY SEISMIC MONITORING SYSTEM TAPE PLAYBACK ONTO CHART PAPER C) MP l-1.22-1 0 1&2
- REPLACEMENT OF THE MAIN TURBINE MOOG VALVE WITH THE UNIT-ON-LINE MP l-1.23-1 2 1&2
- STEAM DUMP SYSTEM PRESTARTUP INSPECTION AND FUNCTIONAL TEST MP l-1.25-1 1 1 *** DIGITAL FEEDWATER CONTROL SYSTEM ENGINEERING CONSOLE i ROUTINE OPERATIONS
'O O
) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 MOLUME SC 1&C KAINTENANCE PROCEDURES ! ) TABLE OF CONTENTS HUMBER REV UNil TITLE massazamam==nn= === ==z= =z= amen =zz=m=munammazassazzamanamaunzzazczz=======mazazamma h MP l-1.25-1 1 2 *** DIGITAL FEEDWATER CONTROL SYSTEM ENGINEERING CONSOLE [ ) ROUTINE OPERATIONS MP l-1.26-1 O 162
- CONDENSATE POLISHER COMPUTER SYSTEM BOOT-STRAPPING '
INSTRUCTIONS MP l-1.27-1 0 1 *lNSTRUMENT PANE LS PI A & PIB COMMON POWER SUPPLIES CAllBRATION , MP l-1.27-1 0 2 *lNSTRUMENT PANELS PIA AND PIB COMMON POWER SUPPLIES ) CAllBRATION MP l-1.28-1 0 1
- MAIN AND BYPASS FEEDWATER REG VALVE PRE-STARTUP FUNCTIONAL ;
TEST MP l-1.28-1 1 2
- MAIN AND BYPASS FEEDWATER REG VALVE PRE-STARTUP INSPECTION MP l-1.29-1 1 1 *BRANDT PI-DPl TRANSMITTER / CONTROLLER ALIGNMENT MP l-1.29-1 1 2 *BRANDT PI-DPT TRANSMITTER / CONTROLLER ALIGNMENT
) MP l-1.30-1 2 1 *10% ADV BACKUP AIR BOTTLE CHANGE-00T AND LEAK TEST MP l-1.30-1 0 2 *10% ADV BACKUP AIR BOTTLE CHANGE-0UT AND LEAK TEST MP l-1.31-1 1 1&2
- FUNCTIONAL TES' VIBRATION AND LOOSE PARTS MONITOR MP I-1.31-2 1 1&2
- CHANNEL CAllBRATION AND FUNCTIONAL TEST VIBRATION AND LOOSE PARTS MONITOR .
MP l-1.32-1 2 1 *RADAT10N MONITORING SYSTEM KAYE DATA LOGGER PROGRAMMING ) AND CAllBRATION MP l-1.32-1 1 2
- RADIATION MONITORING SYSTEM KAYE DATA LOGGER PROGRAMMING.
AND CAllBRATION MP l-1.33-1 1 1 *40% CONDENSER STEAM DUMP VALVE STROKE FUNCTIONAL TEST MP l-1.33-1 1 2 *40% CONDENSER STEAM DUMP VALVE STROKE FUNCTIONAL TEST MP l-1.34-1 0 1 $35% AlnOSPHERIC STEAM DUMP VALVE STROKE FUNCTIONAL TEST , ) MP l-1.34-1 0 2 *35X ATMOSPHERIC STEAM DUMP VALVE STROKE FUNCTIONAL TEST ; MP I-1.35-1 0 IL2
- ACCESSING EMERGENCY ASSESSMENT AND RESPONSE SYSTEM (EARS)
INPUTS MP I-1.36-1 1 1 *** MAIN TURBINE DEH DC BUS, AUTO-STOP OIL AND E-H HEADER FUNCTIONAL TEST MP l-1.36-1 1 2 *** MAIN TURBlNE CONTROL INTEGRATED FUNCTIONAL TEST l ). _MP l-1.37-1 1 1
- CONTAINMENT H2 MONITORS CELL 82 AND CELL 83 TEMPORARY POWER JUMPERS MP l-1.37-1 1 2
- CONTAINMENT H2 MONITORS CELL 82 AND CELL 83 TEMPORARY POWER SUPPLIES MP l-1,37-2 1 1
- CONTAINMENT PERSONNEL AND EMERGENCY HATCH ANNUNCIATOR JUMPER . ,
1 MP l-1.37-2 0 2
- CONTAINMENT PERSONNEL AND EMERGENCY HATCH ANNUNCIATOR _
JUMPER E MF 1-1.37-3 1 1
- CONTAINMENT EVACUATION ALARM JUMPER INSTALLATION AND -
REMOVAL ! MP l-1.37-3 0 2
- CONTAINMENT EVACUATION ALARM JUMPER INSTALLATION AND REMOVAL
) o
O 22 JUL 93 DIABLO CANYON POWER PLANT ; UNITS 1 AND 2 VOLUME SC I&C MAINTENANCE PROCEDURES TABLE F CONTENTS O NUMBER REV UNIT TITLE
=============== =y= ==== = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = - - - - - - = = = = = = = = = = = = = = = =
MP l-1.37-4 0 1 *** MAIN TURBINE AUTO-STOP OIL /EHC INTERFACE VALVE PCV-23 JUMPER O MP I-1.37-4 0 2 *** MAIN TURBINE AUT0-STOP OIL /EHC INTERFACE VALVE PCV-23 JUMPER MP I-1.39-1 0 1 *VALIDYNE TO PPC VERIFICATION AND VALIDATION MP I-1.39-1 0 2 *VAllDYNE TO PPC VERIFICATION AND VALIDATION MP I-1.40-1 0 1&2 *RADLOG DOWNLOAD INSTRUCTIONS FOR VICTOREEN DIGITAL RADIATION MONITOR SYSTEMS O MP I-1.40-2 0 1&2
- FLOW CONTROLLER AllGNMENT VICTOREEN SERIES #B44-223-10 MP I-2.1-0 3 1&2 ** MASTER LIST FOR COAX AND TRIAX CABLE WORK MP I-2.1-1 1 1&2 ** COAX 1AL AND TRIAX1AL CABLE TERMINATION PROCEDURE MP 1-2.1-2 1 1&2 **C0 AXIAL & TRIAXIAL CABLE & CONNECTOR ELECTRICAL TESTS ~
MP I-2.1-3 2 1&2 ** COAXIAL & TRIAXIAL CARLE CONNECTOR BOOTING MP I-2.1-4 0 1 ***NIS TRIAXIAL CONNECTOR INSTALLATION, CABLE TESTING, AND O RAYCHEM INSTALLATION MP l-2.1-4 1 2 ***NIS TRIAX1AL CONNECTOR INSTALLATION CABLE TESTING, AND RAYCHEM INSTALLATION MP I-2.2-1 5 1&2 ** ENVIRONMENTALLY QUALIFIED CONDUIT SEALS MP l-2.3-1 2 1&2
- FIELD WIRING SPLICE ENVIRONMENTAL QUALIFICATION MP I-2.5-1 5 1&2
- CALIBRATION OF PLANT PRESSURE GAUGES O MP I-2.7-1 12 1&2 ** TRANSMITTER ENVIRONMENTAL- QUALIFICATION MP I-2.8-1 2 1&2 ***DPU INSTRUMENT VALVE MANIPULATION DURING CALIBRATION AND SENSING LINE FILLING MP l-2.9-1 1 1&2 ** TESTING BARTON CAPILLARY SYSTEMS >
MP l-2.9-2 1 1&2 **CAPILIARY SYSTEM FILLING & PROOF TESTING MP l-2.9-3 0 1&2 ** CAPILLARY FILL VAPOR EVACUATION O Mp i-2.11-1 1 1&2 ** DIFFERENTIAL PRESSURE INDICATING SWITCH ENVIRONMENTAL QUALIFICATl0N MP l-2.12-1 5 1&2 **MODEL EA180 POSITION SWITCH MAINTENANCE & ENVIRONMENTAL : QUALIFICATION MP I-2.12-2 2 1&2 **NAMCO CONNECTOR 0-RING REPLACEMENT FOR ENVIRONMENTAL QUALIFICATION O MP l-2.12-3 0 1&2 **MODEL EA740\EA750 POSITION SWITCH MAINTENANCE & ENVIRONMENTAL QUAllFICATION , MP I-2.12-4 0 1&2 **MODEL EA170 POSITION SWITCH MAINTENANCE & ENVIRONMENTAL QUAllflCATION MP l-2.13-1 0 1&2 ** GAMMA-METRICS JUNCTION BOX ENVIRONMENTAL QUALIFICATION i MP l-2.13-2 0 1&2
- WESTINGHOUSE INCORE THERMOCOUPLE REFERENCE JUNCTION BOX i O ENVIRONMENTAL QUALIFICATION l MP l-2.14-1 3 1&2
- TEMPERATURE SENSOR CALIBRATION CHECK, THERM 0 COUPLES &
i THERMOMETERS MP l-2.14-2 3 1&2
- REPLACEMENT OF REACTOR COOLANT RTD'S
(: MP l-2.17-1 1 1&2 ***Gul0ELINES FOR PRINTED CIRCUIT BOARD REPAIR AND MODIFICATION
O 22 JUL 93 DIABLO CANYON PO!!ER PLANT l UNITS 1 AND 2 ) VOLUME SC I&C MAINTENANCE PROCEDURES
]
TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ==== =========================================================== ,
MP I-2.21-1 6 1&2 *** REACTOR HEAD INSTRUMENTATION REMOVAL AND REINSTALLATION- ! MP I-2.22-1 0 1&2 ** REPAIR & INSTALLATION OF CORE EXIT THERMOCOUPLE CONNECTORS O MP l-2.23-1 0 1&2
- TESTING MSA PRESSURE DEMAND REGULATORS !
MP l-2.23-2 0 1&2 ** CALIBRATION OF MSA POWERED AIR PURIFYING RESPIRATOR (PAPR) FLOW TESTER MP l-2.24-1 3 1&2
** INSTALLATION OF METALLIC COMPRESSION FITTINGS ON i INSTRUMENT TUBING
. MP I-2.24-3 0 1&2 **SYNFLEX HOSE CONNECTOR MAKE UP ' O MP l-2.24-4 2 1&2
- PRESSURE TESTING OF INSTRUMENT TUBING SYSTEMS MP I-2.25-1 2 1&2 ** CONTROL VALVE TRAVEL & BENCH SET ADJUSTMENTS MP l-2.25-2 1 1&2 **lTT-GRINNELL AIR OPERATED VALVE SET UP MP l-2.28 5 1 ***CALIB, ACTIVATION AND REMOVAL FROM SERVICE OF RX VESSEL REFUELING LEVEL INDICATION SYS (RVRLIS) ,
MP l-2.28 2 2 ***CALIB, ACTIVATION & REMOVAL FROM SERVICE OF RX VESSEL O REFUELING LEVEL INDICATION SYS (RVRLIS) MP I-2.29-1 1 1&2 ** CAPACITOR CAPACITANCE AND LEAKAGE TESTING MP l-2.30-1 1 1&2
- FUNCTIONAL TEST OF THE FEEDWATER HEATERS & MSR DUMP LEVEL CONTROL SYSTEMS MP l-2.30-2 2 1
- FUNCTIONAL TEST OF THE FEEDWATER HEATERS AND MSR ALARM, ;
' TRIP, AND CONTROL SYSTEMS ' O MP l-2.30-2 0 2
- FUNCTIONAL TEST OF THE FEEDWATER HEATERS AND MSR ALARM, TRIP, AND CONTROL SYSTEMS "
MP l-2.31-1 3 1
- FUNCTIONAL CHECK OF MAIN FEEDWATER PUMP 1-1 CONTROLS MP I-2.31-2 3 1
- FUNCTIONAL CHECK OF MAIN FEEDWATER PUMP 1-2 CONTROLS '
MP l-2.31-3 4 2
- FUNCTIONAL CHECK OF MAIN FEEDWATER PUMP 2-1 CONTROLS MP l-2.31-4 5 2
- FUNCTIONAL CHECK OF MAIN FEEDWATER PUMP 2-2 CONTROLS i O MP l-2.32-1 0 1 *** SOURCE / INTERMEDIATE RANGE DETECTOR NE-31/NE-35 REPLACEMENT MP l-2.32-1 1 2 *** SOURCE / INTERMEDIATE RANGE DETECTOR NE-31/NE-35 REPLACEMENT MP l-2.32-2 2 1 *** SOURCE / INTERMEDIATE RANGE DETECTOR NE-32/NE-36 REPLACEMENT :
O MP l-2.32-2 0 2 *** SOURCE / INTERMEDIATE RANGE DETECTOR NE-32/NE-36 REPLACEMENT MP l-2.33-1 1 1&2
- CALIBRATION OF BREATHING AIR MONITOR LOW PRESSURE ALARM i SWITCH i MP l-3.1-1 4 1&2
- ROUTINE MAINTENANCE OF HAGAN OPTIMAC RECORDERS ,
! MP l-3.1-2 0 1&2
- ROUTINE MAINTENANCE ON BAILEY TYPE WM RECORDERS -
O MP I-3.1-3 2 1&2
- ROUTINE MAINTENANCE ON FISCHER & PORTER SERIES 51 RECORDERS !
MP l-3.1-4 1 1&2
- ROUTINE MAINTENANCE ON ESTERLINE ANGUS MODEL A RECORDERS
- MP l-3.1-5 1 1&2
- ROUTINE MAINTENANCE ON WESTRONICS SERIES E RECORDERS (PEN TYPE)
MP I-3.1-6 2 1&2 *ROUTIhE MAINTENANCE ON WESTRONICS SERIES E RECORDERS . (MULTIP01.NT TYPE) P : L i
O 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC ILC KAINTENANCE PROCEDURES g TABLE OF CONTENTS NUMBER REV UNIT TITLE
============ == === ==== ===========================,===============================
MP l-3.1-7 1 1&2 ** ROUTINE MAINTENANCE ON L & N SPEEDOMAX M & W PEN TYPE g RECORDERS MP I-3.1-8 2 1&2 ** ROUTINE MAINTENANCE ON L & N SPEEDOMAX H & W MULTIPOINT RECORDERS MP l-3.1-9 1 It? ** ROUTINE MAINTENANCE ON HONEYWELL ELECTRONIK & TURB0 GRAPH MULTIPOINT RECORDERS MP l-3.1-10 1 1&2
- ROUTINE MAINTENANCE ON TAYLOR MODEL 90J RECORDERS
) MP 1-3.1-11 1 1&2 ** ROUTINE MAINTENANCE ON ESTERLINE ANGUS MODEL E1124E MULTIPOINT RECORDEP MP l-3.1-12 0 1&2
- ROUTINE MAINTENANCE ON BECKMAN MODEL ROS RECORDER MP l-3.1-13 0 1&?
- ROUTINE MAINTENANCE ON CONCORD MODEL 5103 DIGITAL PRINTER MP I-3.1-14 0 1&2
- ROUTINE MAINTENANCE ON THE LEEDS & NORTHRUP SPEEDOMAX 2500 MULTIPOINT RECORDERS
- ) MP 1-3.1-15 0 1&2 ** ROUTINE MAINTENANCE ON L&N 430 SERIES RECORDERS MP l-3.1-16 2 1&2 ** ROUTINE MA]NTENANCE ON THE TRACOR WESTRONICS DDR10 DIGITAL DATA RECORDERS Mi 1-3.1-17 1 It? ** ROUTINE MAINTENANCE ON THE TRACOR WESTRONICS DDR10/MB DIGITAL DATA RECORDERS WITH A MX60 MULTIPLEXER MP l-3.1-18 2 1&2
- ROUTINE MAINTENANCE ON FOXBORO E27R SERIES ELECTRONIC
- ) INDICATING RECORDERS MP l-3.1-19 2 1&2
- ROUTINE MAINTENANCE ON TRACOR WESTRONIC E3 & P SERIES MINIATURE STRIP CHART RECORDER MP l-3.1-21 1 1&2
- ROUT]NE MAINTENANCE ON L&N 100 SERIES RECORDERS MP 1-3.1-22 0 1&2
- ROUTINE MAINTENANCE ON MOLYTEK MODEL 2702-5 RECORDER MP l-3.1-23 1 lt? ** ROUTINE MAINTENANCE ON L&N SPEEDOMAX 165/250 SERIES
- ) MJLTIP0lNT RECORDERS MT 1-?.4-1 0 1&2 ' CLEAN FILTERS ON P250 & P2000 COMPUTERS MD l t.4-lo 1 1&2
- CALIBRATION CHECK OF DILLON LOAD CELL USED FOR RCCA DRAG TEST FlxTURE MP l-4.4-?; 4 1&? ** CALIBRATION CHECK Of LOAD CELL & READOUT FOR FUEL HANDLING BRIDGE CRANE
- ) MP l-6.2 0 1&2 **EGEG DEW POINT SENSOR PHOTORESISTIVE DEVICE REPLACEMENT MP l-B.B 1 1&2 *EGEG DEWPOINT CHECK MP I-EA01 2 1&2 ***CAllBRATION OF GENERAL RESISTANCE RTD-100 - V0tTAGE RATIO METHOD MD l-EA04 2 1&2 ***CAllBRATION OF GENERAL RADIO 1433 DECADE RESISTORS -
VOLTAGE RATIO METHOD
- ) MP l-EA05 0 1&2 *CAllBRATION OF LEEDS & NORTHRUP MODEL 8686 MILLIVOLT POTENT 10ME1ER Mi= I-EADB 1 1&2 **CAllBRATION OF GEN RAD TiPES 1531-AB & 1538-A STRDBOTACS MP I-EA09 3 1&2 *** CALIBRATION OF 100 PPM LOAD RESISTORS MP I-EA10 1 1&2 ***CAllBRATION Of SENSITIVE RESEARCH MODEL ESD ELECTROSTATIC VOLTMETERS O
O
/
(3 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC I&C KAINTENANCE PROCEDURES g) TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ==== =========================================================== ,
MP I-EAll 0 1&2 *** CALIBRATION OF GENERAL RADIO 1433 DECADE RESISTORS - CURRENT COMPARISON METHOD () MF 1-EA12 2 1&2 ***CAllBRATION Of FLUKE MODEL 845AB HIGH IMPEDANCE VOLTMETER NULL DETECTOR MP l-EA13 0 1&2
- CALIBRATION OF TELEDYNE ANALYTICAL MODEL 235TG THERMAL CONDUCTIVITY ANALYZER MP l-EA14 1 1&2 *** CALIBRATION OF GUILDLINE MODEL 9975 DC COMPARATOR RESISTANCE BRIDGE
() MP l-EA15 2 1&2 *** CALIBRATION OF PYROTRONICS SCU-9 FIRE DETECTOR ! SENSITIVITY TEST SET MP l-EA17 2 1&2 ***CAllBRATION OF GAS TECH MODEL 1220-110210 GAS DETECTION SYSTEM MP l-EAIB 3 1&2 *CAllBRATION OF KINEMETRICS MODEL FC-1 FIELD CALIBRATOR MP l-EA19 0 1&2 *CAtlBRATION OF ANALYSIS & MEASUREMENT SERVICES ERT-1 RTD 13 LCSR TEST INSTRUMENT MP l-EA21 1 1&2 *CAllBRATION OF FLUKE MODEL 5220A TRANSCONDUCTANCE AMPLIFIER MP l-EA22 1 1&2
- CALIBRATION OF COMPUTER PRODUCTS MODEL RTP7394/01 16 CHANNEL A/D CARD CALIBRATOR MP l-EA23 0 1&2 *** CALIBRATION OF GENERAL RESISTANCE RTD-100 - CURRENT .
COMPARISON METHOD (3 MP I-EA24 0 1&2
- CALIBRATION OF DCPP WIND SPEED CAllBRATOR MP I-EA25 0 1&2
- CALIBRATION OF THE RVLIS RTD TESTER MP l-ED02 2 1&2
- CALIBRATION OF FLUKE 8810A DIGITAL MULTIMETER MP l-ED03 2 1&2
- CALIBRATION OF FLUKE MODEL 8060A DIGITAL MULTIMETER MP l-ED04 0 1&2
- CALIBRATION OF FLUKE HYDRA DATA LOGGER MP l-EDOS 1 1&2 ***CAllBRATION OF FLUKE MODEL 8600A DIGITAL MULTIMETER 33 MP l-EDD6 1 1&2 ***CAllBRATION OF RONAN MODEL X86 PORTABLE CAllBRATOR MP l-ED07 0 1&2 *** VERIFICATION OF SPECTRACOM 8164 NBS RECEIVER AND 8171A WWVB SYNCHRONIZED CLOCK ,
MP l-ED09 1 1&2 *** CALIBRATION OF TRANSMATION MODEL 1040 DIGITAL CAllBRATOR MP l-ED10 1 1&2 *** CALIBRATION OF RONAN MODEL X85 PORTABLE CALIBRATOR MP l-ED13 0 1&2
- CALIBRATION OF STOPWATCHES (3 MP l-E017 0 1&2 ***CAllBRATION OF FLUKE MODEL 5440B DIRECT VOLTS CAllBRATOR i MP l-ED21 1 1&2 ***CAllBRATION OF ANALYSIS AND MEASUREMENT SERVICES ELC-1 '
LCSR ANALYZER SOFTWARE MP I-ED22 3 1&2
- CALIBRATION OF MULTI-AMP MODEL SST-2 DIGITAL TIMER MP l-ED23 4 1&2
- CALIBRATION OF FLUKE MODEL 8842A DIGITAL MULTIMETER ;
MP I-ED?4 0 1&2 ** CALIBRATION OF GEN RAD MODEL 1546 DIGITAL STROBOTAC .(3 MP l-ED28 1 1&2 *** CALIBRATION OF EBERLINE MODELS MP-1 AND MP-2 MINI PULSERS MP I-ED29 0 1&2 *CAllBRATION OF DORIC MODEL 130A C-METER-MP I-ED30 2 1&2 ***CAllBRATION OF NUCLEAR CONSULTING SERVICES F-1000-HD
- MODEL SAE HALIDE DETECTOR MP l-ED32 2 1&2 *CAllBRATION OF EG&G ORTEC 776 COUNTER / TIMER MP l-ED33 0 1&2 *CAllBRATION OF METRABYTE DASH-16 A/D CARD i
() i l. l
3 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC I&C KAINTENANCE PROCEDURES TABLE OF CONTENTS NUMBLR REV UNIT TITLE
=============== === ====
===========================================================
MP I-ED37 1 1&2
- CALIBRATION OF EQUIPMENT TIMER MP l-ED40 1 1&2 I) MP l-ED41 0 1&2
- CALIBRATION OF FLUKE MODEL 45 DIGITAL MULTIMETER
- CALIBRATION OF HONEYWELL OMNILIGHT MODEL 8M37 RECORDER MP l-ED42 1 1&2
*CAllBRATION OF LUDLUM MODEL 2000 PORTABLE SCALER MP I-MA01 0 1&2
- CALIBRATION OF MECHANICAL HOLTEST MICROMETERS MP l-MA02 1 1&2
- CALIBRATION OF TORQUING DEVICES MP l-MA03 0 1&2 *CAllBRATION OF PRATT & WHITNEY MODEL C SUPERMICROMETER MP I-MA04 0 1&2 *** CALIBRATION OF LEVELS
!) MP l-MA05 1 1&2
- CALIBRATION OF MAGNETIC COATING THICKNESS GAUGES MP l-MA06 0 1&2 *CAllBRATION OF G0/NO G0 GAGES MP l-MA07 0 1&2 *** CALIBRATION OF FORCE MEASURING INSTRUMENTS USING MOREHOUSE PROVING RINGS MP l-MA08 1 1&2
- CALIBRATION OF INSIDE MICROMETERS MP l-MA09 4 1&2
[] MP l-MA10 0 1&2
- CALIBRATION OF HYDRAULIC TORQUE WRENCHES
*** CALIBRATION OF OUTSIDE MICROMETERS MP 1-MA11 1 1&2
- CALIBRATION OF DIAL INDICATORS MP l-mal 2 0 1&2
- CALIBRATION OF THICKNESS GAGES MP l-MA13 1 1&2
- CALIBRATION OF WIRE TERMINAL CRIMPING TOOLS MP l-MA14 0 1&2
- CALIBRATION OF HANGING SCALES MP l-MAIS 3 1&2
- CALIBRATION OF MICROMETER HEADS 4D MP I-MA16 1 1&2
- CALIBRATION OF FORNEY DUAL BEAM PLATFORM SCALE MP l-MA17 0 1&2
- CALIBRATION OF AMP TERMI-POINT PULL TEST TOOL MP l-MAlB 3 1&2
- CALIBRATION OF OHAUS MODEL 2610G TRIPLE BEAM BALANCE MP l-MA19 0 ;c2
- CALIBRATION OF MICROMETER DEPTH GAGES MP l-MA20 0 1&2
- CAL]BRATION OF DIAL & VERNIER CALIPERS MP l-MA21 2 1&2
- CALIBRATION OF ADJUSTABLE & NON-ADJUSTABLE PARALLELS S MP l-MA?2 0 1&2 *CAllBRATION OF BELT TENSION CHECKERS MP l-MA23 1 1&2 *CAllBRATlON OF STARRETT MODEL 716 DIAL INDICATOR CALIBRATOR MP l-MA24 1 1&2 *CAllBRATION OF DIRECT READING HEIGHT GAGES MP l-MA25 1 1&2 *** CALIBRATION OF END MEASURING RODS MP l-MA26 0 1&2 *CAllBRATION OF TURB]NE TAPER GAGES MP l-MA28 0 1&2
- CALIBRATION OF GILSON MODEL HS-300P PLATFORM SCALE ED MP 1-MA29 0 1&2 *CAllBRATION OF HOWE-RICHARDSON MODEL 54002 PLATFORM SCALE MP l-MA30 0 1&2 *CAllBRATION OF EDLUND MODEL E80 ELECTRONIC SCALE MP l-MD02 1 1&2
- CALIBRATION OF AMTEK ACCUFORCE DIGITAL FORCE GAUGES MP l-MD03 3 1&2
- CALIBRATION OF AKO TORQUE SPECIALTIES SERIES TSD TORQUE CALIBRATION SYSTEM MP l-PA01 0 1&2
- CALIBRATION OF HEISE MODEL C CM AND CMM PRESSURE GAUGES di MP 1-PA03 0 1&2 *CAllBRATION OF ANALOG AND DIGITAL PRESSURE GAUGES FP l-PA04 0 1&2 *CAllBRATION OF WALLACE & TIERNAN MODEL FA-187 MERCURY MANOMETER MP l-PA05 0 1&2 ** CALIBRATION OF ANALOG VACUUM GAUGES MP l-PA06 0 1&2 *** CALIBRATION Of LOW ACCURACY COMPOUND PRESSURE GAUGES MP l-pal 2 0 1&2 *CAtlBRA110N Of DWYER MODEL 1425 H00t. GAUGE 9
0
i 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC ILC MAINTENANCE PROCEDURES
) TABLE OF CONTENTS l NUMBER REV UNIT TITLE =============== === ==== ===========================================================
MP l-PA13 0 IL2
- MAINTENANCE AND OPERATION OF DWYER MODEL 1425 HOOK GAUGE
- ) MP l-PA14 0 1&2 *CAllBRATION OF DWYER MAGNEHELIC DIFFERENTIAL PRESSURE GAUGES MP l-PA16 1 IL2 ***CAllBRATION OF HIGH ACCURACY COMPOUND PRESSURE GAUGES MP l-PAlB 0 IL2 *CAllBRATION OF VAllDYNE DP15TL PRESSURE TRANSDUCERS MP l-PD02 1 IL2
- CALIBRATION OF DIGITAL GAUGE PRESSURE INDICATORS MP I-PD03 1 IL2
- CALIBRATION OF DIGITAL DIFFERENTIAL PRESSURE GAUGES &
) VACUUM GAUGES MP l PD04 0 IL2 ** CALIBRATION OF HEISE MODELS 730A-07 & 730B-07 DIGITAL PORTABLE PNEUMATIC CALIBRATORS MP l-FDOS 1 IL2 *CAllBRATION OF HONEYWELL STD SERIES SMART PRESSURE TRANSMITTERS MP l-PC01 1 102 ***CAllBRATION OF EBERLINE E-140 PORTABLE SURVEY INSTRUMENT ) MP l-RCO2 1 IL2 ***CAllBRATION OF EBERLINE E-140N COUNT RATE METER MP I-RCO3 1 IL2 *CAllBRATION OF EBERLINE RM-14 COUNT RATE METER MP l-RC04 1 IL2 *CAllBRATION OF EBERLINE RM-15 COUNT RATE METER MP 1-RC05 1 IL2 *CAllBRATION OF EBERLINE RM-20-1 COUNT RATE METER WITH AC-3 FRDBE MP l-RC07 0 IL2 *** AUTO-CAllBRATION OF THE BICRON ALMS AUTOMATED LAUNDRY ) MONITORING SYSTEM MP l-RCOB 0 IL2 *CAllBRATION OF EBERLINE RM-3C-4 RADIATION MONITOR MP l-RC09 0 IL2 *** CALIBRATION OF EBERLINE PRM-5-3 COUNT RATE METER WITH SPA-3 PROBE MP l-RC10 0 IL2 *** CALIBRATION OF EBERLINE MODEL PRM-6 PULSE RATE METER WITH PA-3 PROBE ) MP 1-Fell 0 1&2 *** CALIBRATION OF EBE9LINE MODEL MS-2 MINI SCALER MP l-Ril? O IL2 *** CALIBRATION OF EBERLINE MODEL BC-4 BETA COUNTER '
MP 1-RCl3 0 IL2
- CALIBRATION OF EBERLINE MDDEL SAC-4 SCINTILLATION ALPHA COUNTER MP l-PCl4 2 IL2
- ELECTRONIC CAllBRATION OF EBERLINE AMS-3 BETA AIR MDNITOR MP l-RCIS 5 IL2 ***CAllBRATION OF EBERLINE PCM-1B PERSONNEL CONTAMINATION M3NITOR
) MP l-RC16 2 IL2 *** CALIBRATION OF EBERLINE TCM-1A TOOL CONTAMINATION MONil0R MP I-RC17 0 IL2
- CALIBRATION OF IRT MODEL PRM-110 PORTAL RADIATION MONITOR MP l-RClB 2 IL2 ***150 TOPIC CAllBRATION OF EBERLINE MODEL IM-11 IODINE MONITOR MP l-PC19 0 IL2 *** CALIBRATION OF EBERLINE RM-14S COUNT RATE METER FOR gg FRISKER PROBES MP l-RC21 0 IL2
- CALIBRATION OF EBERLINE RM-20-1 COUNT RATE METER WITH FR!SKER PROBES MP 1-RC23 0 IL2 ***CAllBRATION OF EBERLINE RM-14SA COUNT RATE METER FOR FRISKER PROBES MP l-RD01 2 IL2 *** CALIBRATION OF EBERL]NE MODEL RD-2 ION CHAMBER SURVEY g INSTRUMENT O
i 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC 1&C KAINTENANCE PROCEDURES l g, TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ==== =====================,==,==================================
MP I-RD02 2 IL2 *** CALIBRATION OF EBERLINE R0-2A ION CHAMBER SURVEY 8' INSTRUMENT MP l-FD03 4 IL2 *** CALIBRATION OF EBERLINE 61128 TELETECTOR G-M SURVEY INSTRUMENT MP I-PDD4 2 IL2 *CAtlBRATION OF EBERLINE MODEL PNR-4 PORTABLE NEUTRON REM COUNTER MP I-PDOS 1 IL2 *** CALIBRATION OF EBERLINE EC4-X PORTABLE AREA MONITOR WITH
- 7) DAl-6 G-M DETECTOR MP 1-RD06 1 IL2 *** CALIBRATION OF EBERLINE E-520 GEIGER COUNTER WITH HP-270 PROBE MP l-RDD7 0 ]L2
- CALIBRATION OF EBERLINE ESP-2 WITH HP-220A PROBE AND 60 FOOT CABLE Mi 1-PDOB 2 IL2 *** CALIBRATION OF EBERLINE ASP-1 PORTABLE SURVEY INSTRUMENT b) WITH HP-270 PROBE MP 1-RD09 1 IL2 *** CALIBRATION OF EBERLINE E-530N GEIGER COUNTER WITH HP-220A PROBE MP I-RD10 1 IL2
- CALIBRATION OF EBERLINE MODEL R0-7 ION CHAMBER SURVEY INSTRUMENT MP I-RD12 1 IL2 *** CALIBRATION OF EBERLINE ASP-1 WITH HP-290 PROBE AND 60
) MP 1-RD14 0 ]L2 FOOT CABLE
*** CALIBRATION OF EBERLINE RMS-Il SYSTEM WITH THE HP-310 PROBE MP l-RD15 2 IL2 *** CALIBRATION OF NMC MODEL GA-2TO GAMMA ALARM SYSTEM MP l-RD16 0 IL2 *CALIBRAT10N OF EBERLINE ESP-2 WITH VARIOUS DETECTORS MP l-RD17 0 IL2 *** CALIBRATION OF XETEX MODEL 330A TELESCAN MP l-RD20 IL2
) 1 *** CALIBRATION OF JOHNSON 2000W EXTENDER G-M SURVEY INSTRUMENT MP l-PD25 1 IL2 *** CALIBRATION 0F DDSITEC PR-2 PORTABLE REMOTE MONITOR MP l-RD26 2 ]L2
- CALIBRATION OF DOSITEC MODEL 502A DIGITAL ALARMING DOSIMETER MP l-PD27 1 IL2
- CALIBRATION OF DOSITEC MODEL AR-20 REMOTE MONITOR WITH
) MP l-RD30 3 IL2 UNDERWATER PROBE
*CALIBRATIOk 0F XETEX MODEL 415 B-1 DIGITAL ALARMING DOS 1 METER MP 1-RD31 1 IL2 *** CALIBRATION OF XETEX MODEL 412AT/503AR TELEDOSE SYSTEM MF 1-FFD1 0 IL2 *** CALIBRATION OF RADECO H-609V VARIABLE FLOW AIR SAMPLER MP l-RF02 0 IL2 *** CALIBRATION OF RADECO H-809C PORTABLE BATTERY POWERED AIR SAMPLER
) MP l-PF03 0 162 ** CALIBRATION OF RADECO H-809B2 PORTABLE BATTERY POWERED AIR SAMPLER MP l-RF04 0 IL2 *** CALIBRATION OF RADECO HD-28 CONSTANT FLOW AIR SAMPLER MP I-RF05 0 IL2 *** CALIBRATION OF RADEC0 HD-28B CONSTANT FLOW AIR SAMPLER i MP l-RF06 0 IL? ***CAllBRATION OF BENDIX BDX-60 CONSTANT FLOW PUMP MP l-Rf07 0 IL2 *** CALIBRATION OF Hl-Q MRV-14-C CONSTANT FLOW AIR SAMPLER O
4D 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC ILC KAINTENANCE PROCEDURES TABLE OF CONTENTS
)
NUMBER REV UNIT TITLE
=============== === ==== =======================,=======================, ==========
MP I-RF08 0 IL2 ***CALIBRAT]ON 0F EBERLINE AMS-3 BETA AIR MONITOR ROTAMETER MP I-RF09 0 IL2 ***CAllBRATION OF EkMET ISA-44RAL-0D CARBON MONOXIDE AND 2) OXYGEN MONITOR MP l-RF11 0 IL2 *** CALIBRATION OF RE-ll&l2 GRAB SAMPLE CART ROTAMETERS MP I-RF12 0 IL2 *** CALIBRATION OF EBERLINE RAP-1 REGULATED AIR PUMP MP l-RF13 0 IL2 ***CAllBRATION OF EBERLINE HDDEL IM-ll IODINE MONITOR ROTAMETER MP l-RF15 0 1&2 *** CALIBRATION OF RADECO HD-29A CONSTANT FLOW AIR SAMPLER
- 2) MP l-RXOl 102 1 ***lNITIAL ON-SITE CALIBRATION OF PORTABLE RADIATION DETECTION INSTRUMENTS MP l-RXO2 1 IL2 *** CALIBRATION OF RADIDACTIVE WASTE SCALE MP l-TA02 0 162
- CALIBRATION OF LIQUID-IN-GLASS THERMOMETERS MP I-TA03 0 102
- CALIBRATION OF OMEGA SERIES CJ THERM 0 COUPLE COLD-JUNCTION COMPENSATOR
- 2) MP 1-TA05 0 IL2
- CALIBRATION OF GULTON MODEL Z-55 THERMOCOUPLE RECORDER MF 1-TA06 1 IL2
- CAL]BPATION OF DIAL THERMOMETERS MP l-TA07 1 IL2 ** CALIBRATION OF TRANSMATION MODEL 1013 THRICE REFERENCE CELL ML I-TA09 0 102 ** CALIBRATION OF THERM 0 COUPLES MP l-TA10 0 IL2 **CAllBRATION OF FLUKE MODEL Y2001 THERM 0 COUPLE MULTIPOINT
- 2) SELECTOR MP l-TDC2 0 IL2 **CAllBRATION OF FLUKE MODEL 2175A DIGITAL THERMOMETER Mf' 1-TD03 0 IL2 ***CAllBRATION OF OMEGA OMNI-CAL TEMPERATURE CALIBRATOR MP l-TD04 1 IL2 **CAlllBRATION OF FLUKE MODEL 2168A DIGITAL THERMOMETER MF 1-TD05 0 IL2 *CALIBRATIOk 0F TEMPERATURE INDICATORS WITH DEDICATED IMMERSION OR SURFACE PROBE (S)
- 2) MP l-lD10 0 lL2
- CALIBRATION OF FLUKE MODEL 2190A DIGITAL THERMOMETER MP I-TDll 1 IL2
- CALIBRATION OF WAHL MODEL 392 SERIES DIGITAL THERMOMETERS MF 1-TD17 0 IL2 *** CALIBRATION OF WAHL MODEL 2500M DIGITAL THERMOMETER MP l-TDlB C IL2 **CAllBRATION Of FLUKE MODEL 2180A (TYPE 2) DIGITAL THERMOMETER MP 1-TD19 2 IL2
- CALIBRATION OF GulLDLINE MODEL 9734 CONSTANT TEMPERATURE C) BATH MP 1-TD20 1 IL2 ***CAllBRATION OF FLUKE MODEL 2300A-002 THERM 0 COUPLE SCANNER MP 1-TD21 1 IL2 *CAllBRATION OF TRANSMATION 1060, 1061, 1062, AND 1064P THERM 0 COUPLE CALIBRATORS MP 1-3-L110 2 1 *** STEAM GENERATOR l-1 AUX FW SUPPLY LEVEL CONTROL CHANNEL LCV-110 CAllBPAT}0N tb MP l-3-Lllo 1 2 *** STEAM GENERATOR 2-1 AUX FW SUPPLY LEVEL CONTROL CHANNEL LCV-110 CALIBRATION MP l-3-Lill 2 1 *** STEAM GENERATOR l-2 AUX FW SUPPLY LEVEL CONTROL CHANNEL LCV-lll CALIBRATION MP l-3-Llll 1 2 *** STEAM GENERATOR 2-2 AUX FW SUPPLY LEVEL CONTROL CHANNEL LCV-Ill CAllBFATION O
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) TABLE OF CONIENTS NUMBER REV UNIT TITLE
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MP l-3-L113 1 1
*** STEAM GENERATOR 1-4 AUX FW SUPPLY LEVEL CONTROL CHANNEL ;
C) LCV-ll3 CALIBRATION MP l-3-L113 1 2 *** STEAM GENERATOR 2-4 AUX FW SUPPLY LEVEL CONTROL CHANNEL LCV-113 CAllBRATION MP I-3-L115 1 1 *** STEAM GENERATOR 1- AMX FW SUPPLY LEVEL CONTROL CHANNEL LCV-115 CALIBRATION ' MP l-3-L115 1 2 ***S1EAM GENERATOR 2-3 AUX FW SUPPLY LEVEL CONTROL CHANNEL C) LCV-115 CAllBRATION MP l-3-P433 0 1
- AUXILIARY FEEDWATER PUMP 1-2 DISCHARGE PRESSURE CHANNEL PT-433 CAllBRATION MP l-3-P434 0 1
- AUXILIARY FEEDWATER PUMP 1-3 DISCHARGE PRESSURE CHANNEL PT-434 CALIBRATION MP l-4-F48 2 1
- STEAM GENERATOR l-1 BLOWDOWN FLOW CHANNEL FT-4B CALIBRATION MP l-4-F4B 0 2
, C)
- STEAM GENERATOR 2 1 BLOWDOWN FLOW CHANNEL FT-48 CALIBRATION MP l-4-F49 1 1
- STEAM GENERATOR l-2 BLOWDOWN FLOW CHANNEL FT-49 CALIBRATION MP l-4-F49 0 2
- STEAM GENERATOR 2-2 BLOWDOWN FLOW CHANNEL FT-49 CAllBRATION MP I-4-F51 1 1
- STEAM GENERATOR 1-4 BLOWDOWN FLOW CHANNEL FT-51 CALIBRATION MP l-4-F51 0 2 '
- STEAM GENERATOR 2-4 BLOWDOWN FLOW CHANNEL FT-51 CALIBRATION MP l-4-F60 2 1
- STEAM GENERATOR 1-3 BLOWDOWN FLOW CHANNEL FT-60 CALIBRATION MP I-4-F60 1 2
() MP l-7-L56 1 2
- STEAM GENERATOR 2-3 BLOWDOWN FLOW CHANNEL FT-60' CALIBRATION
*RVRLIS WIDE RANGE INDICATION CHANNEL LT-56 CALIBRATION MP l-7-L462 0 1 *** PRESSURIZER LEVEL CHANNEL LT-462 (COLD) CALIBRATION MP l-7-L462 0 2
- PRESSURIZER LFVEL CHANNEL LT-462 (COLD) CAllBRATION MP l-7-Y.1 0 1 ***CONFIGURAT:>N CONTROL OF RCP VIBRATION /VELOCIT'. PROBES MP l-7-Y.1 0 2 ***CONFIGUR\TIO.9 CONTROL OF RCP VIBRATION / VELOCITY PROBES MP l-7-Y412H 1 1
- ROD SPEED CALIBRATION C) MP l-7-Y412H 0 2
- ROD SPEED CALIBRATION MP I-B-F110 C 1
- BORIC ACID FLOW CHANNEL FT-110 CALIBRATION MP l-B-F110 2 2
- BORIC ACID FLOW CHANNEL FT-110 CALIBRATION MP l-B-F115 0 1 *RCP 1-4 SEAL WATER INLET FLOW CHANNEL FT-115 CALIBRATION MP l-B-F115 0 2 *RCP 2-4 SEAL WATER INLET FLOW CHANNEL FT-115 CAllBRATION MP l-E-F116 0 1 *RCP l-3 SEAL WATER INLET FLOW CHANNEL FT-116 CALIBRATION C) MP I-8-Fil6 0 2 *RCP 2-3 SEAL WATER INLET FLOW CHANNEL FT-116 CALIBRATION MP l-B-F134 2 1 *RCS LETDOWN HEAT EXCHANGER OUTLET FLOW CHANNEL FT-134 CAllBRATION MP I-8-F134 1 2 *RCS LETDOWN HEAT EXCHANGER OUTLET FLOW CHANNEL FT-134 CAllBRATION '
MP l-8-F143 0 1 *RCP 1-2 SEAL WATER INLET FLOW CHANNEL FT-143 CALIBRATION C) MP l-8-F143 1 2 *RCP 2-2 SEAL WATER INLET FLOW CHANNEL FT-143 CAllBRATION MP l-8-F144 0 1 *RCP l-1 SEAL WATER INLET FLOW CHANNEL FT-144 CALIBRATION MP l-B-F144 0 2 *RCP 2-1 SEAL WATER INLET FLOW CHANNEL FT-144 CALIBRATION ! HP 1-B-L102 4 1
- BORIC ACID STORAGE TANK 1-1 LEVEL CHANNEL LT-102 CALIBRATION O
I
i ) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME SC I&C MAINTENANCE PROCEDURES i TABLE OF CONTENTS ) NUMBER REV UNIT TITLE
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MP l-8-L102 1 2
- BORIC ACID STORAGE TANK 2-2 LEVEL CHANNEL LT-102 CALIBRATION
) MP l-B-L106 4 1
- BORIC ACID STORAGE TANK 1-2 LEVEL CHANNEL LT-106 CAllBRATION MP l-6-L106 1 2
- BORIC ACID STORAGE TANK 2-1 LEVEL CHANNEL LT-106 CAllBRATION MP l-8-P142 0 1
- CHARGING PUMPS DISCHARGE HEADER PRESSURE CHANNEL PT-142 ,
CALIBRATION ) MP l-8-P142 0 2
- CHARGING PUMPS DISCHARGE HEADER PRESSURE CHANNEL PT-142 CALIBRATION MP l-10-F640 0 1 *RHR HEAT EXCHANGER OUTLET (RCS HDT LEGS) FLOW CHANNEL FT-640 CALIBRATION MP l-10-F640 0 2 *RHR HEAT EXCHANGER OUTLET (RCS HOT LEGS) FLOW CHANNEL FT-640 CAllBRATION
) MP l-10-T639 MP I-10-T639 0 0 1 2
*RHR HX 1-1 OUTLET TEMPERATURE CHANNEL TE-639 CALIBRATION
*RHR HX 2-1 OUTLET TEMPERATURE CHANNEL TE-639 CAllBRATION MP l-10-1649 0 1 *RHR HX 1-2 OUTLET TEMPERATURE CHANNEL TE-649 CALIBRATION MP l-10-T649 0 2 *RHR HX 2-2 OUTLET TEMPERATURE CHANNEL TE-649 CAllBRATION '
MP l-12-F930 0 1
- SPRAY ADDITIVE TANK 1-1 TO EDUCTORS CHANNEL FT-930 CAllBRATION 3 MP l-12-F930 0 2
- SPRAY ADDITIVE TANK 2-1 TO EDUCTORS CHANNEL FT-930 !
CALIBRATION < MP I-14-F65 1 1
- COMPONENT COOLING WATER SUPPLY HEADER B FLOW CHANNEL FT-65 CALIBRATION MP l-14-F65 1 2
- COMPONENT COOLING WATER SUPPLY HEADER B FLOW CHANNEL FT-65 CALIBRATION 3 MP l-14-F68 1 1
- COMPONENT COOLING WATER SUPPLY. HEADER A FLGW CHANNEL FT-6B l CAllBRATION MP l-14-F68 1 2
- COMPONENT COOLING WATER SUPPLY HEADER A FLOW CHANNEL FT-68
! CAllBRATION l MP I-14-F69 1 1
- COMPONENT COOLING WATER SUPPLY HEADER C FLOW CHANNEL FT-69 l CALIBRATION
} MP I-14-F69 0 2
- COMPONENT COOLING WATER SUPPLY HEADER C FLOW CHANNEL FT-69
- CALIBRATION f MP l-14-F70 0 1
- CONTAINMENT FAN COOLER l-1 FLOW CHANNEL FT-70 CALIBRATION MP l-14-F70 0 2
- CONTAINMENT FAN COOLER 2-1 FLOW CHANNEL FT-70 CAllBRATION MP l-14-F71 0 1
- CONTAINMENT FAN COOLER 1-2 FLOW CHANNEL FT-71 CALIBRATION MP l-14-F71 0 2
- CONTAINMENT FAN COOLER 2-2 FLOW CHANNEL FT-71 CAllBRATION 7 MP l-14-F72 0 1
- CONTAINMENT FAN COOLER 1-3 FLOW CHANNEL FT-72 CALIBRATION l MP l-14-F72 0 2
- CONTAINMENT FAN ^00LER 2-3 FLOW CHANNEL FT-72 CAllBRATION MP l-14-F73 0 1
- CONTAINMENT FAr COOLER 1-4 FLOW CHANNEL FT-73 CAllBRATION .
} MP I-14-F73 0 2
- CONTAINMENT FAN COOLER 2-4 FLOW CHANNEL FT-73 CALIBRATION MP l-14-F74 0 1
- CONTAINMENT FAN COOLER 1-5 FLOW CHANNEL FT-74 CAllBRATION MP i-14-F74 0 2 *CONTAINMEN1 FAN COOLER 2-5 FLOW CHANNEL FT-74 CALIBRATION
)
i
i ) 22 JUL 93 DIABLO CANYON POWER PLANT ! UNITS 1 AND 2 ' VOLUME SC ILC MAINTENANCE PROCEDURES TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ==== =======================--- =============,==================
MP l-14-T6 1 1 *CCW HEAT EXCHANGER 1-1 OUTLET TEMPERATURE CHANNEL TE-6 ) MP l-14-T6 1 2 CALIBRATION
*CCW HEAT EXCHANGER 2-1 OUTLET TEMPERATURE CHANNEL TE-6 CALIBRATION MP l-14 '7 1 1 *CCW HEAT EXCHANGER l-2 OUTLET TEMPERATURE CHANNEL TE-7 CALIBRATION MP I-14-T7 1 2 *CCW HEAT EXCHANGER 2-2 OUTLET TEMPERATURE CHANNEL TE-7 CALIBRATION 3 MP l-17-T10 0 1
- CIRCULATING WATER PUMP l-1 DISCHARGE TEMPERATURE CHANNEL TE-10 CAllBRATION MP l-17-T10 0 2 *CIPCULATING WATER PUMP 2-1 DISCHARGE TEMPERATURE CHANNEL TE-10 CALIBRATION MP I-17-Til 0 1
- CIRCULATING WATER PUMP l-2 DISCHARGE TEMPERATURE CHANNEL i TE-ll CALIBRA110N D MP l-17-T11 0 2
- CIRCULATING WATER PUMP 2-2 DISCHARGE TEMPERATURE CHANNEL TE-Il CALIBRATION MP I-18-L38 3 1&2
- FIREWATER STORAGE TANK LEVEL CHANNEL LT-38 CAllBRATION MP l-19-L99 0 1
- REACTOR COOLANT DRAIN TANK 1-1 LEVEL CHANNEL LT-99 CALIBRATION MP I-19-L99 0 2
- REACTOR COOLANT DRAIN TANK 2-1 LEVEL CHANNEL LT-99 D CALIBRATION MP l-21-P.1 1 1
- DIESEL ENGINE CRANKCASE LOW VACUUM PRESSURE SWITCH CALIBRATION MP l-21-P.1 1 2
- DIESEL GENERATOR CRANKCASE LOW VACUUM PRESSURE SWITCH CALIBRA110N MP l-23-T.1 0 1
- ELECTRICAL & ESF EQUIPMENT ROOM AREA TEMPERATURE 3 MONITORING CHANNELS CAllBRATION MP I-23-T.1 0 2 *(OTSC 6/22/93--SHAFFER) ELECTRICAL & ESF EQUIPMENT ROOM AREA TEMPERATURE M3NITORING CHANNELS CAllBRATION MP l-23-Y.1 1 1
- CONTAINMENT PERSONNEL HATCH LEAK DETECTION MONITOR CAllBRATION MP l-23-Y.1 2 2
- CONTAINMENT PERSONNEL HATCH LEAK DETECTION MONITOR D CALIBRATION MP l-40-M569 0 1&2 *** BACKUP METEOROLOGY WIND DIRECTION, WIND SPEED, AND AIR TEMPERATURE / DELTA-T MP l-40-Y53B 0 1&2 *** PRIMARY METEOROLOGY SYSTEM PRECIPITATION CHANNEL YT-538 CAtlBRATION D
D l
'e l
C) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES ' TABLE OF CONTENTS NUMBER REV UNIT TITLE
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STP I-1A 41 1
- ROUTINE SHIFT CHECKS REQUIRED BY LICENSES STP 1-1A 40 2
- ROUTINE SHIFT CHECKS REQUIRED BY LICENSES C) STP l-18 35 1 *** ROUTINE DAILY CHECKS REQUIRED BY LICENSES STP I-1B 22 2 *** ROUTINE DAILY CHECKS REQUIRED BY LICENSES STP l-1C 33 1
- ROUTINE WEEKLY CHECKS STP l-lC 20 2
- ROUTINE WEEKLY CHECKS STP l-10 35 1
- ROUTINE MONTHLi CHECKS REQUIRED BY LICENSES t STP I-lD 19 2
- ROUTINE MONTHLY CHECKS REQUIRED BY LICENSES C) STP I-lE 3 1&2
- ANALOG CHANNEL OPERATIONAL TESTS REQUIRED PRIOR TO EACH REACTOR STARTUP ,
STP l-2A 4 1&2 *CAllBRATION OF FLUX DEVIATION & MISCELLANEOUS CONTROL & INDICATION CHANEL STP l-2Al 3 1&2
- REMOVAL OF A FLUX DEVIATION & MISCELLANEOUS CONTROL INDICATION CHANNEL FROM SERVICE C) STP I-2A2 5 1&2
- CALIBRATION PROCEDURE FOR FLUX DEVIATION & MISC CONTROL &
INDICATION CHANNEL STP 1-2A3 4 1&2
- REINSTATEMENT PROCEDURE FOR FLUX DEVIATION & MISCELLANEOUS CONTROL & INDICATION CHANNEL STP I-28 19 1&2
- NUCLEAR POWER RANGE CHANNEL ANALOG CHANNEL ANALOG CHANNEL OPERATIONAL TEST C) STP I-2C 5 1&2 **CAllBRATION OF POWER RANGE CHANNEL STP I-2Cl 14 1&2
- REMOVAL OF POWER RANGE CHANNEL FROM SERVICE STP I-2C2 21 1&2 *** CALIBRATION PROCEDURE FOR POWER RANGE CHANNEL STP l-2C3 14 1&2
- REINSTATEMENT OF POWER RANGE CHANNEL TO SERVICE STP l-2D 2B 1&2 ***(0TS" 05/14/93 - BREWER) NUCLEAR POWER RANGE INCORE/EXCORE CAllBRATION C) STP l-2E 1 1&2
- CALIBRATION OF COMPARATOR & RATE CHANNEL STP I-2El 2 1&2
- REMOVAL AND RE]NSTATEMENT OF COMPARATOR AND RATE CHANNEL STP l-2E2 4 1&2 SCAllBRATION PROCEDURE FOR COMPARATOR AND RATE CHANNEL STP I-2U 4 1&2 ** RESETTING HIGH FLUX RANGE BISTABLES FOR POWER RANGE CHANNELS STP l-2V 2 1&2
- POWER RANGE CHANNEL HIGH . LTAGE VERIFICATION
() STP l-3A 10 1&2
- NUCLEAR INTERMEDIATE RANGE CHANNEL ANALOG CHANNEL OPERATIONAL TEST STP l-3B 2 1&2 *CAllBRATION OF INTERMEDIATE RANGE CHANNELS STD l-381 2 1&2
- REMOVAL OF INTERMEDIATE RANGE CHANNEL FROM SERVICE STP l-382 12 1&2
- CALIBRATION PROCEDURE FOR INTERMEDIATE RANGE CHANNEL STP l-3V 1 1&2 *lNTERMEDIATE RANGE CHANNEL HIGH VOLTAGE VERIFICATION C) STP I-4A 15 1&2
- ANALOG CHANNEL OPERATIONAL TEST NUCLEAR SOURCE RANGE STP l-48 4 1&2
- CALIBRATION OF SOURCE RANGE CHANNELS STP I-481 8 1&2
- REMOVAL OF A SOURCE RANGE CHANNEL FROM SERVICE STP I-4B2 9 1&2 ***CAllBRATION PROCEDURE FOR SOURCE RANGE CHANNEL STP l-4B4 6 1&2 ** DETERMINATION OF SOURCE RANGE DETECTOR CHARACTERISTIC CURVES FOR THE WESTINGHOUSE LOW NOISE PREAMPLIFIERS O
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) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS ) NUMBER REV UNIT TITLE
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STP I-4C 2 1&2
- CALIBRATION OF AUDIO COUNT RATE - SCALER TIMER CHANNEL STP 1-4Cl 2 1&2 '
)
- REMOVAL OF AN AUDIO COUNT RATE - SCALER TIMER CHANNEL FROM SERVICE STP 1-4C2 4 1&2 *CAllBRATION PROCEDURES FOR AUDIO COUNT RATE / SCALER TIMER CHANNEL STP I-4C3 3 1&2
- REINSTATEMENT PROCEDURE FOR AUDIO COUNT RATE / SCALER TIMER CHANNEL STP l-5A 38 1&2 *(OTSC 5/25/93-O'NEIL) ANALOG CHANNEL OPERATIONAL TEST OF
) OTDT, OPDT, T (AVG) AND DELTA-T STP l-581 9 1&2
- REMOVAL FROM SERVICE OF OTDT, OPDT, T(AiG) AND DELTA-T l CHANNELS STP 1-582 22 1
- CHANNEL CALIBRATION OF OTDT, OPDT, T(AVG), AND DELTA-T STP l-5B2 17 2
- CHANNEL CAllBRATION OF OTDT, OPDT, T(AVG), AND DELTA-T STP l-5B3 15 1&2
- CALIBRATION OF T(HOT) AND T(COLD) MV/I MODULES
) STP l-585 5 1&? **RCS NARROW RANGE RTD TIME RESPONSE TESTING STP l-5B6 6 1&2
- RETURN TO SERVICE OF OTDT, OPDT, T(AVG) AND DELTA-T CHANNELS STP l-587 3 1&2 ** CHANNEL CALIBRATION RTD BYPASS LOOP FLOW STP l-6A 6 1&2
- FUNCTIONAL TEST OF PRESSURIZER PRESSURE CHANNELS STP l-6B 7 1&2 *CAllBRATION OF PRESSURIZER PRESSURE CHANNELS PROTECTION, 3 SAFEGUARDS & REMOTE SHUTDOWN MONITORING FUNCTIONS (INCLUDING SENSOR RESPONSE TEST)
STP 1-6B1 7 1&2 ** REMOVAL OF PRESSURIZER PRESSURE CHANNEL FROM SERVICE STP l-6B2 9 1&2 *CAllBRATION OF PRESSURIZEP PRESSURE PROTECTION & SAFEGUARDS FUNCTIONS STP l-6B3 7 IL2
- CALIBRATION OF PRESSURIZER PRESSURE TRANSMITTERS
) STP l-6B4 5 1&2
- RESPONSE TIME TESTING OF PRESSURIZER PRESSURE TRANSMITTERS STP I-6B5 6 1&2 ** RETURN TO SERVICE PRESSURIZER PRESSURE CHANNEL STP l-7A 6 1&2
- ANALOG CHANNEL OPERATIONAL TEST, PRESSURIZER LEVEL CHANNELS STP l-7B 5 1&2 ' CHANNEL CALIBRATION - PRESSURIZER LEVEL CHANNELS PROTECTION - REMDTE SHUTDOWN & POST-ACCIDENT MONITORING ,
FUNCTIONS ) STP l-781 9 1&2 *** REMOVAL FROM SERVICE PRESSURIZER LEVEL CHANNEL STP l-782 5 1&2
- CALIBRATION ANALOG ELECTRONICS, PRESSURIZER LEVEL PROTECTION CHANNELS STP l-783 7 1&2 *** CALIBRATION - PRESSURIZER LEVEL TRANSMITTER STP 1-785 6 1&2
- RETURN TO SERVICE PRESSURIZER LEVEL CHANNELS STP l-8A 17 1&2
- ANALOG OPERATIONAL TEST REACTOR COOLANT FLOW CHANNELS
) l STP I-8B 7 1&2 *CAllBRATION OF REACTOR COOLANT FLOW CHANNELS PROTECTION ' FUNCTIONS l STP l-861 7 1&2
- REMOVAL OF REACTOR COOLANT FLOW CHANNEL FROM SERVICE l STP l-8B2 11 1&2
- CALIBRATION OF REACTOR COOLANT FLOW PROTECTION & CHANNELS STP 1-8B5 5 1&2
- REINSTATEMENT OF REACTOR COOLANT FLOW CHANNELS TO SERVICE
) )
) E22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS )
NUMBER REV UNIT TITLE
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STP l-8C 5 1 ** CALIBRATION AND RESPONSE TIME TESTING OF REACTOR COOLANT FLOW TRANSMITTERS ') STP l-8C 4 2 ***CAllBRATION AND RESPONSE TIME TESTING OF REACTOR COOLANT FLOW TRANSMITTERS SIP l-9A 10 1&2
- TRIP ACTUATING DEVICE OPERATIONAL TEST 12KV UNDERVOLTAGE, UNDERFREQUENCY STP l-98 4 1&2 ** CALIBRATION - 12KV UNDERVOLTAGE AND UNDERFREQUENCY CHANNELS PROTECTION AND SAFEGUARDS FUNCTIONS
) STP l-10 3 1&2
- FUNCTIONAL TEST OF RCP BREAKER CHANNELS STP l-llA 17 1&2
- FUNCTIONAL TEST OF STE AM GENERATOR WATER LEVEL PROTECTION CHANNELS STP I-llB 7 1&2
- CALIBRATION OF STM GENERATOR LEVEL CHANNELS PROTECTION AND SAFEGUARDS & POST-ACCIDENT MONITORING FUNCTIONS (INCLUDING SENSOR RESPONSE TEST)
) STP l-1181 10 1&2 *** REMOVAL OF STEAM GENERATOR LEVEL CHANNEL FROM SERVICE SlP l-1182 13 1&2
- CALIBRATION OF STEAM GENERATOR LEVEL PROTECTION AND SAFEGUARD CHANNELS STP l-1183 8 1&2 *CAllBRATION OF STEAM GENERATOR LEVEL TRANSMITTERS SlP I-1184 3 1&2
- RESPONSE TIME TESTING OF STEAM GENERATOR LEVEL SIP l-1185 8 1&2 *** REINSTATEMENT OF STEAM GENERATOR LEVEL CHANNELS TO
) SERVICE STP l-12A 9 1&2
- ANALOG CHANNEL OPERATIONAL TEST OF STEAM FLOW AND PRESSURE CHANNELS STP l-12B 5 1&2
- CHANNEL CAllBRATION STEAM GENERATOR FEEDFLOW, STEAM FLOW &
PRESSURE CHANNELS STP l-1281 14 1
- REMOVAL FROM SERVICE STEAM GENERATOR FEEDFLOW, STEAMFLOW
) AND PRESSURE CHANNELS STP I-12B1 2 2
- REMOVAL FROM SERVICE STEAM GENERATOR FEEDFLOW, STEAMFLOW AND PRESSURE CHANNELS STP l-1222 7 1
- CALIBRATION ANALOG ELECTRONICS STEAM GENERATOR PRESSURE STP 1-1282 1 2
- CALIBRATION ANALOG ELECTRONICS STEAM GENERATOR PRESSURE STP 1-1283 6 1 *CAllBRATION ANALOG ELECTRONICS STEAM GENERATOR PRESSURE
) (FLOW COMPENSATING)
STP l-12B3 0 2 *CAllBRATION ANALOG ELECTRONICS STEAM GENERATOR PRESSURE (FLOW COMPENSATING) STP l-12B4 10 1
- CALIBRATION - ANALOG ELECTRONICS STEAM GENERATOR FEEDFLOW STP l-12B4 2 2 *CAllBRATION - ANALOG ELECTRONICS STEAM GENERATOR FEEDFLOW S1P l-1265 15 1
- CALIBRATION ANALOG ELECTRONICS STEAM GENERATOR STEAM FLOW
) STP l-12B5 6 2
- CALIBRATION ANALOG ELECTRONICS STEAM GENERATOR STEAM FLOW STP l-1286 15 1
- CALIBRATION - COMPARATORS STEAM GENERATOR FCEDFLOW, STEAMFLOW AND PRESSURE CHANNELS STP I-12B6 5 2 *CAllBRATION - COMPARATORS STEAM GENERATOR FEEDFLOW, STEAMFLOW AND PRESSURE CHANNELS STP l-12B7 20 1&2 ***CAllBRATION TRANSMITTERS STEAM GENERATOR STEAM FLOW
) )
- 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS O
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STP 1-12BB 4 1&2
- RESPONSE TIME TEST STEAM GENERATOR FLOW TRANSMITTERS STP l-1289 11 1 **CAllBRATION TRANSMITTER STEAM GENERATOR FEEDFLOW
- STP l-1289 1 2 **CAllBRATION TRANSMITTER STEAM GENERATOR FEEDFLOW STP I-12B10 6 IL2 **CAtlBRATION OF STEAM PRESSURE TRANSMITTERS STP l-12B11 2 IL2 ** RESPONSE TIME TESTING STEAM GENERATOR PRESSURE TRANSMITTERS STP l-12B12 12 1
- RETURN TO SERVICE STEAM GENERATOR FEEDFLOW, STEAM FLOW AND PRESSURE A CHANNEL 9 STP l-12B12 2 2
- RETURN TO SERVICE STEAM GENERATOR FEEDFLOW, STE AMFLOW &
PRESSURE CHANNELS STP 1-13A 3 IL2
- CALIBRATION OF AUTOSTOP OIL PRESSURE CHANNELS STP l-13B 5 IL2 *CAllBRATION OF TURBlNE STOP VALVE POSITION CHANNELS STP l-13C 4 ILP
- FUNCTIONAL TEST OF PEACTOP TRIP FROM TURBINE TRIP STP l-13D1 4 162 ** RESPONSE TIME TEST FOR SSPS TRAIN A MASTER RELAY J ACTUAT10N TO luPBlNE TRIP SIP l-1302 4 IL2 ** RESPONSE TIME TEST FOR SSPS TRAIN B MASTER RELAY ACTUATION TO TURBINE TRIP SIF 1-15: 8 IL2 ' ANALOG CHANhEL OPERATIONAL TEST OF CONTAINMENT PRESSURE CHANNELS SlP l-15B 4 IL?
- CALIBRATION OF CONTAINMENT PRESSURE CHANNELS (INCLUDING D SENSOR RESPONSE TEST)
STP l-1581 6 IL2
- REMOVAL OF CONTAINMENT PRESSURE CHANNEL FROM SERVICE STF 1-15B2 3 IL2
- CALIBRATION PROCEDURES (PROTECTION AND SAFEGUARDS) FOR CONTAINMENT PRESSURE CHANNELS STP l-15B3 3 IL2 ** CALIBRATION OF CONTAINMENT PRESSURE TRANSMITTERS STP l-15B4 4 IL2
- RESPONSE TIME TESTIN3 0F CONTAINMENT PRESSURE TRANSMITTERS D SiF 1-15Bt S IL2
- REINSTATEMENT OF CONTAINMENT PRESSURE CHANNELS STP l-16Al 11 IL2
- REMOVAL FROM SERVICE OF THE SSPS FOR ACTUATION LOGIC TESTING DURING MODES 1, 2, 3 OR 4
.it 1-16 2; E IL?
- ACTUATION LOGli TEST OF PROTECTION SYSTEM LOGIC, INCLUDING MASTER RELAYS AND REACTOR TRIP BREAKERS (MODE 1, 2, 3, OR 4)
J STP l-16:2E B IL2
- ACTUATION LOGIC TEST OF PROTECTION SYSTEM. LOGIC, INCLUDING MASTER RELAYS AND REACTOR TRIP BREAKERS'(MODE 1, 2, 3 OR 4)
STP I-16A3 9 1&2
- REINSTATEMENT OF THE SSPS AFTER ACTUATION LOGIC TESTING AND/OR MAINT DURING MODES 1,2,3 OR 4 STP l-16B 13 1 *** TESTING OF SAFETY INJECTION RESET TIMER AND SLAVE RELAY D K602 STP l-16B 0 2 *** TESTING OF SAFETY INJECTICN RESET TIMER AND SLAVE RELAY K602 STP l-16C 13 1&2
- TRIP ACTUATING DEVICE OPERATIONAL TEST OF MANUAL INITIATION AND BLOCKING REACTOR PROTECTION AND ENGINEERED SAFEGUARDS
.O O
l^ ) 22 JUL 93 DIABLO CANYON POSER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS NUMBER REV UNIT TITLE
===============
=== ==== ===========================================================
STP 1-1601 10 1&2
- REMOVAL FROM SERVICE OF THE SSPS FOR ACTUATION LOGIC TESTING & OR MAINTENANCE DURING MODE 5 OR 6 3 SlP l-1602 11 1&2
- ACTUATION LOGIC TEST OF PROTECTION SYSTEM LOGIC INCLUDING MASTER RELAYS & REACTOR TRIP BREAKERS (MODES 5 OR 6)
STP l-16D3 8 1&2
- REINSTATEMENT OF THE SSPS AFTER ACTUATION LOGIC TESTING AND/OR MAINTENANCE DURING MODES 5 OR 6 STP l-16D4 3 1&2 *RECONFIGURING AN SSPS TRAIN IN MODES 5 OR 6 STP l-18F1 7 1&2
- FUNCTIONAL TEST OF STEAM GENERATOR BLOWDOWN TANK VENT D
sip l-16F2 (RM-27) AND DISCHARGE MONITORS (RM-23) 9 1&2
- STEAM GENERATOP BLOWDOWN TANK LIQUID AND VENT GAS MONITORS
- CALIBRATION (RM-23 AND RM-27)
STP l-ISM 1 6 1&2
- CONTROL ROOM A]R INTAKE MONITOR FUNCTIONAL TEST (RM-25 &
26) STP l-lEM2A 8 1&2 **CAllBRATION OF C0kTROL ROOM VENTILATION SYSTEM INTAKE D RADIATION MONITORS RE-25 OR RE-26 S1P l-18M2B 2 1&2
- REMOVAL OF CONTROL ROOM VENTILATION INTAKE RADIATION MONITORS FROM SERVICE (RE-25 OR RE-26)
SIP I-16M2C 1 1&2
- REINSTATEMENT OF CONTROL ROOM VENTILATION SYSTEM INTAKE RADIATION MONITORS TO SERVICE (RE-25 OR RE-26) blP I-18N1 7 1&2
- FUNCTIONAL TEST OF CONTAlhMENT HIGH RANGE RADIATION
,D MONITORS RM-30, 31 STP l-18P1 4 1&2
- FUNCTIONAL TEST OF OlLY WATER SEPARATOR DISCHARGE LIQUID MONITOR RM-3 STP I-18P2 3 1&2 *0lLY WATER SEPARATOR EFFLUENT MONITOR RM-3 REMOVAL FROM SERVICE STP l-18P3 0 1&2 *0lLY WATER SEPARATOR EFFLUENT DISCHARGE MONITOR RM-3:
D ELECTRONIC AL}GNMENl STP l-18P4 1 1&2 *0lLY WATER SEPARATOR EFFLUENT DISCHARGE MONITOR RM-3: RADIATION SOURCEPRESENTATION (ISOTOPIC) CALIBRATl0N S1P I-18P5 0 1&2 *0ILY WATEP SEPARATOR EFFLUENT MONITOR RM-3: RETURN TO SERVICE STP l-18xl 3 1&2
- FUNCTIONAL TEST OF RMS 11 GAS DECAY TANKS RM-41, 42, 43
[) SlP l-18X2A 2 1&2
- ELECTRONIC AllGNMENT OF RMS 11 DETECTORS - GAS DFrAt TANKS (RE-41, 42, 43)
STP 1-18X2B 2 1&2 ' GAS DECAV TANKS RM-41, RM-42, AND RM-43 REMOVAL FROM SERVICE STP l-1Ex2C 2 1&2
- REINSTATEMENT 10 SERVICE: GAS DECAY TANKS RM-41, RM-42 AND RM-43 D STP l-18X2D 3 1&2 ** SOURCE CAllBRATION OF THE DETECTORS FOR THE GAS DECAY TANK MONITORS (RE-41,42, & 43)
SlP I-18Y1 3 1&2 ** FUNCTIONAL TEST OF THE TSC (RM-66,67,82) AND TSC LAB (RM-68,69,E83) PARTICULATE, IODINE & NOBLE GAS RADIATION MONITORS d D
O 22 JUL 93 DIABLO CANYON POWER PLANT UN1TS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS O HUMBER REV UNIT TITLE
=======,,======
=== . . = = = = = = = = = = = . . = = = = = = . = . . . . . . . . . . = = = . . . . . . . . . = = . . . = = = = = = =
STP l-18Y2A 2 1&2 ** ELECTRONIC ALIGNMENT OF THE TSC RM-66, 67, 82 & TSC LABORATORY RM-68, 69, 83 PARTICULATE 10 DINE & NOBLE GAS C) RADIATION HONITORS STP I-18Y2B 2 1&2
- REMOVAL OF TSC (RM-66, 67 & 82) & TSC LABORATORY (RM-68, 69 & 83) PARTICULATE, 10 DINE AND NOBEL GAS RADIATION MONITORS FROM SERVICE (NONQ)
STP l-18Y2C 0 1&2
- REINSTATEMENT OF THE TSC RM-66 RM-67 RM-82 & TSC LABORATORY RM-68 RM-69 & RM-83 PARTICULATE - IODINE &
() NOBLE GAS RADIATION MONITORS TO SERVICE STP 1-18Y2D 8 1&2 ** SOURCE PRESENTATION / CALIBRATION FOR RM-66, 67, 68, 69, 82 AND 83 STP I-18Z1 2 1&2
- FUNCTIONAL TEST OF RMS TSC AREA RADIATION MONITORS RM-60, 61, 62, 63, 64, 65 STP l-18Z2A 1 1&2 ** ELECTRONIC ALIGNMENT OF TSC AREA RADIATION MONITORS
[] (RE-60, 61, 62, 63 64, 65) STP I-18Z2B 2 1&2
- REMOVAL OF TSC AREA RADIATION MONITORS FROM SERVICE (RE-60, 61, 63, 64 AND 65)
STP I-1822C 0 1&2
- REINSTATEMENT TO SERVICE OF TSC AREA RADIATION MONITORS RE-60, 61, 62, 63, 64, 65 STP l-18Z2D 4 1&2 ** CALIBRATION OF TSC AREA RADIATION MONITORS RE 60-65 ep SlP I-18AA1 7 1&2 *FUNCT. TEST OF PLT. VENT & IODINE GRAB SAMPLER A'_ ARA MNTRS (RM 34, 35) & POST ACCIDENT SAMPLE ROOM AREA MONITOR RM-48 STP l-18AA2A 2 1&2
- ELECTRONIC ALIGNMENT OF PLANT VENT & IODINE GRAB SAMPLER ALARA MONITORS (RE-34,35) & THE POST ACCIDENT SAMPLE ROOM AREA MONITOR (RM-48)
STP I-18AA2B 1 1&2
- REMOVAL OF PLANT VENT & IODINE GRAB SAMPLER ALARA MONITORS e (RM-34,35) & THE POST ACCIDENT SAMPLE ROOM AREA MONITOR (RM-48)
STP 1-18AA2C 0 1&2
- REINSTATEMENT OF PLANT VENT & IODINE GRAB SAMPLER ALARA MONITORS (RM-34 35) & THE POST ACCIDENT SAMPLE ROOM AREA MONITOR (RM-48) TO SERVICE STP l-18AA2D 3 1&2 ** RADIATION SOURCE CALIBRATION OF RM-34 & 35, PLANT VENT gp ALARA & RM-48 SENTRY SYSTEM AREA MONITOR;.
STP 1-18BB1 8 1&2
- FUNCTIONAL TEST OF THE HIGH RADIATION PLANT VENT GROSS GAMMA MONITOR (RM-29)
STP l-18BB2A 2 1&2 ** ELECTRONIC AL]GNMENT OF THE HIGH RADIATION PLANT VENT GROSS GAMMA MONITOR RE-29 STP l-18BB2B 2 1&2
- REMOVAL OF THE HIGH RADIATION PLANT VENT GROSS GAMMA gp MONITOR FROM SERVICE (RM-29')
STP 1-18BB2C 1 1&2 " REINSTATEMENTS OF THE HIGH RADIATION PLANT VENT GROSS GAMMA MONITOR TO SERVICE (RM-29) STP l-18BB2D 5 1&2 ** CALIBRATION OF RM-29 PLANT VENT HIGH RANGE MONITOR STP l-29 14 1&2
- EMERGENCY SIGNALS AND COMMUNICATIONS SYSTEMS FUNCTIONAL i
TEST O O
) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS )
NUMBER REV UNIT TITLE
=============== === ==== ===========================================================
STP l-29B 5 1&2 *NRC FTS2000 PHONE CHECK STP l-33A 14 1&2
- REACTOR TRIP AND ESF RESPONSE TIME TEST
) STP I-33B7 1 2 *** REACTOR TRIP AND ESF LOGIC TRAIN A RESPONSE TIME FOR REFUELING OUTAGEFIVE STP l-33BB 1 1 *** REACTOR TRIP AND ESF LOGIC RESPONSE TIME FOR TEST SEQUENCE EIGHT TRAIN 8 STP l-33C 5 1&2 ** TIME RESPONSE TESTING OF REACTOR TRIP BREAKERS STP I-34 9 1&2 *** FUNCTIONAL TEST: FIRE DETECTION SYSTEM ) STP l-34A 8 1 *** FIRE DETECTION SYSTEM DETECTOR FUNCTIONAL PANEL A STP l-34A 1 2 *** FIRE DETECTION SYSTEM DETECTOR FUNCTIONAL PANEL A STP l-34B 7 1 *** FIRE DETECTION SYSTEM DETECTOR FUNCTIONAL PANEL B STP 1-34B 0 2 *** FIRE DETECTION SYSTEM DETECTOR FUNCTIONAL PANEL B STP l-34C 1 1
- FIRE DETECTION SYSTEM SUPERVISORY FUNCTIONAL STP l-34C 0 2
- FIRE DETECTION SYSTEM SUPERVISORY FUNCTIONAL
) STP I-34D 4 1 *** FIRE DETECTION SYSTEM DETECTOR FUNCTIONAL PANEL D STP 1-340 1 2 *** FIRE DETECTION SYSTEM DETECTOR FUNCTIONAL PANEL D STP l-34E 1 1&2 *** PLANT ELEVATOR SMOKE DETECTOR FUNCTIONAL TEST t STP l-34G 4 1
- FIRE DETECTION SYSTEM PANEL VOLTAGE TEST STP l-34G 1 2
- FIRE DETECTION SYSTEM PANEL VOLTAGE TEST STP l-34H 3 1 *** FIRE DETECTION SYSTEM RTD CAllBRATION CHECK
) STP l-34H 0 2 *** FIRE DETECTION SYSTEM RTD CALIBRATION CHECK i STP l-34] 1 1&2 *** PORTABLE DETECTION SYSTEM INSTALLATION TESTING AND OPERATION PROCEDUPE STP I-34K 1 1&2 *** PORTABLE DETECTION SYSTEM MONTHLY TEST STP l-35A 5 1&2 ** CHANCEL FUNCTIONAL TEST CONTROL ROOM CHLORINE MONITORS STP I-35B 2 1&2 ** CONTROL ROOM CHLORINE MONITORS MAINTENANCE ? STP I-37A 1 1&2 *KINEMETRICS SMA-3 SEISMIC MONITORING SYSTEM FUNCTIONAL TEST STP I-37B1 0 1&2 *CAtlBRATION OF KINEMETRICS SMA-3 SEISMIC HONITORING SYSTEM STP I-3782 0 1&2 *CAllBRATION OF ENGDAHL PSR-1200 PEAK SHOCK RECORDERS STP l-37C 7 1&2 ** RECORDING TRIAX1AL ACCELEROMETER CHANNEL CHECK .
STP l-37D 2 1&2
- PEAK ACCELERATION RECORDER MODEL PAR 400 INSPECTION & !
SCRIBE PLATE CHANGEOUT J STP l-39 5 1&2
- CALIBRATION OF RAW WATER STORAGE RESERVOIR LEVEL CHANNELS 36 & 37, LOOP 16-13A & B STP l-40 12 1&2
- CHANNEL CALIBRATION MAIN CIRCULATING WATER DELTA T STP l-42 7 1&2
- ROD POSITION DEVIATION MONITOR FUNCTIONAL TEST STP I-43B 3 162
- CALIBRATION OF ROD BANK INSERTION LIMIT COMPUTER I S1P l-4381 2 1&2
- REMOVAL OF RAD LIMIT COMPUTER CHANNEL FROM SERVICE
'b STP l-43B2 6 1&2
- CALIBRATION PROCEDURES (CONTROL) FOR ROD LIMIT COMPUTER j CHANNELS STP l-43B3 3 1&2
- REINSTATEMENT PROCEDURE FOR ROD LIMIT COMPUTER CHANNELS STP l-46A 12 1&2
- CHANNEL CALIBRATION (ZERO & SPAN) CONTAINMENT HYDROGEN MONITOR CH. 82 (83) l
)
O 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS O NUMBER REV UNIT TITLE
=============== === ==== ===========================================================
STP l-46B 13 1&2 **CAllBRATION TEST (EQUIPMENT VERIFICATION) CONTAINMENT HYDROGEN MONITOR CH. 82 (83) () *** WATER STORAGE TANK LEVEL CHANNEL 40 - LOOP 16-14A STP l-47 7 1&2 STP l-48 11 1&2 *CALIB 0F REFUELING WATER STORAGE TANK LEVEL CHANNELS 920, 921, & 922 STP I-52 4 1 ** CALIBRATION OF CONTAINMENT TEMPERATURE CHANNELS STP 1-52 4 2 ** CALIBRATION OF CONTAINMENT TEMPERATURE CHANNELS
- CHANNEL CAllBRATION FIRST STAGE PRESSURE CHANNEL 505 STP l-53 18 1
-) ( STP I-53 1 2
- CHANNEL CAllBRATION FIRST STAGE PRESSURE CHANNEL 505 STP l-54 14 1
- CHANNEL CAllBRATION FIRST STAGE PRESSURE CHANNEL 506 STF 1-54 2 2
- CHANNEL CAllBRATION FIRST STAGE PRESSURE CHANNEL 506 STP I-55 13 1&2 ** CALIBRATION Of STEAM GENERATOR WIDE-RANGE CHANNELS 501, 502, 503, 504 1 STP l-5) 5 1&2
- CALIBRATION OF EMERGENCY BORATE FLOW CHANNEL 113 l
() SIP I-58 3 162 **CAllBRATION OF PRIMARY WATER FLOW CHANNEL 111, LOOP 8-97A. AND 8-97B STP I-62 4 1E2 ** CALIBRATION OF REACTOR CAVITY SUMP LEVEL CHANNELS 62 STP I-65 2 IL2
- FUNCTIONAL TEST OF CONTAINMENT FAN COOLERS COLLECTION MEASUREMENT SYSTEM STP I-66 5 1&2 **CAllBRATION OF VOLUME CONTROL TANK LEVEL CHANNEL 114 C) **CAllBRATION OF VOLUME CONTROL TANK LEVEL CHANNEL 112, STP l-67 2 IL2 LOOP 8-84A -
STP l-6BC 3 lE2
- PRESSURIZER POWER OPERATED RELIEF VALVE PCV-455C ACTUATION TIMING STP l-69C 3 lL2
- PRESSURIZER POWER OPERATED RELIEF VALVE PCV-456 ACTUATION TIMING c) ** CALIBRATION AND OPERATIONAL TEST OF THE SEISMIC TRIP SiF 1-72B B IL2 CHANNELS i STP l-79A 3 IL2 ** FUNCTIONAL TEST OF WASTE GAS SYSTEM OXYGEN ANALYZERS 75 (76)
STP l-79B 5 1&2 ** CALIBRATION Of WASTE GAS SYSTEM OXYGEN ANALYZER CHANNEL 75 (76) LOOP 24-17 C) 4 IL2 *SUBC00 LED MARGIN MONITOR FUNCTIONAL CHECK STP l-80A STP I-808 6 IL2 *CAllBRATION OF THE SUBC00 LED MARGIN MONITOR STP l-81B 10 IL2 *CAtlBRATION OF WIDE RANGE TEMPERATURE CHANNELS STP l-83A 5 1&2 *CAllBRATION OF PRESSURIZER SAFETY VALVE RELIEF LINE RTD CHANNELS 465, 467 AND 469 STP l-838 1 IL2 *0PERATIONAL TEST OF THE ACOUSTIC MONITORS FOR PRESSURIZER () SAFETY VALVE POSITION INDICATION CHANNELS 116, 117, AND 138 STP I-83C 2 1&2 ** CALIBRATION OF PRESSURIZER PORV RELIEF LINE TEMPERATURE CHANNEL 463 STP l-84A 6 1&2
- FUNCTIONAL TEST OF CONTROL ROOM PRESSURIZATION SYSTEM CHLORINE M3NITORS j)
O
D ?2 JUL 9s DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE Of CONTENTS 3 NUMBER REV UNIT TITLE
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STP l-84B 3 1&2 ** PRESSURIZATION SYSTEM CHLORINE MONITORS CALIBRATION TEST STP l-87A 3 1&2
- FUNCTIONAL CHECK OF THE REACTOR VESSEL LEVEL
) INSTRUMENTATION SYSTEM STP l-878 0 1&2
- CHANNEL CALIBRATION REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM STP l-87B1 4 1&2
- REMOVAL FROM SERVICE REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM STP l-8782 4 1&2 ** REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM CAllBRATION
- ) OF LEVEL TRANSMITTERS SlP l-8783 2 1&2 **CAllBRATION Of ANALOG INPUT OUTPUT CHANNELS REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM STP l-87B4 1 1&2
- RETURN TO SERVICE REACTOR VESSEL LEVEL INSTRUMENTATION SYSTEM SlP l-8SA 0 1&2 **lNSTRUMENT & CONTROLS PERIODIC VALVE CHECKLIST
- 7) STP l-89 4 1&2 ** CALIBRATION OF CONTAINMENT WIDE RANGE LEVEL CHANNELS 942A
& 943A STP l-90 4 1&2 **CAllBRATION OF CONTAINMENT WIDE RANGE PRESSURE TRANSMITTER PT-938 AND PT-939 STP l-91B 11 1&2 ** CHANNEL CALIBRATION OF THE THERM 0 COUPLE MONITORING SYSTEM STP l-92 3 1&2 ** AMSAC FUNCTIONAL TEST STP I-92B 1 1&2 ***AMSAC EXTERNAL INPUT CALIBRATION CHECK STP I-100A 10 1
- CONTAINMENT AIR PARTICULATE / GAS RADIATION MONITOP RM-11/RM-12 FUNCTIONAL TEST ST P l-100 A 1 2
- CONTAINMENT AIR PARTICULATE / GAS RADIATION MONITOR RM-11/RM-12 FUNCTIONAL TEST STP l-1008 4 1
- CALIBRATION OF CONTAINMENT AIR RAD 10 GAS MONITOR RM-12
- STP l-100B 0 2
- CALIBRATION OF CONTAINMENT AIR RADIOGAS MONITOR RM-12 STP l-100B1 6 1
- CONTAINMENT AIR RAD 10 GAS MONITOR RM-12 REMOVAL FROM SERVICE STP l-100B1 1 2 *** CONTAINMENT AIR RAD 10 GAS MCNITOR RM-12 REMOVAL FROM SERVICE STP l-100B2 1 1
- ELECTRONIC ALIGNMENT OF CONTAINMENT AIR RAD 10 GAS MONITOR RM-12
- STP l-100B2 0 2
- ELECTRONIC ALIGNMENT OF CONTAINMENT AIR RAD 10 GAS MONITOR PM-12 STP l-100B3 3 1
- RADIATION SOURCE PRESENTATION (ISOTOPIC) CALIBRATION OF CONTAINMENT RADIOGAS MONITOR RM-12 STP l-100B3 0 2
- RADIATION SOURCE PRESENTATION (ISOTOPIC) CAllBRATION OF CONTAINMENT RAD 10 GAS MONITOR RM-12
- STP l-100B4 6 1
- REINSTATEMENT TO SERVICE: CONTAINMENT AIR RAD 10 GAS MONITOR 1 RM-12 STP l-100B4 1 2 *** REINSTATEMENT TO SERVICE: CONTAINMENT AIR RAD 10 GAS MONITOR RM-12 STP l-102A 5 1&2
- FUNCTIONAL TEST OF LIQUID RADWASTE DISCHARGE MONITOR RM-18 STP l-1026 2 1&2
- CALIBRATION OF THE LIQUID RADWASTE DISCHARGE MONITOR RM-18 9
C) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 UOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS
)
NUMBER REV UNIT TITLE
====.....==.... === ==== ........=.....=.......====...............=..........= .===.
STP l-102B1 2 1&2
- LIQUID RADWASTE DISCHARGE MONITOR RM-18: REMOVAL FROM SERVICE
[) STP l-10282 2 1&2
- ELECTRONIC ALIGNMENT OF THE LIQUID RADWASTE DISCHARGE MONITOR RM-18 STP l-10283 5 1&2 *RADI ATION SOURCE PRESENTATION (ISOTOPIC) CALIBRATION OF LIQUID RADWASTE DISCHARGE MONITOR RM-18 STP I-102B4 1 1&2
- REINSTATEMENT TO SERVICE: LlQUID RADWASTE DISCHARGE MONITOR RM-18
[) STP I-102B5 1 1&2
- SEMIANNUAL DISCRIMINATOR CHECK USING MULTICHANNEL ANALYZER
- LIQUID RADWASTE DISCHARGE MONITOR RM-18 STP l-10286 2 1&2 *TO RESET HIGH ALARM SETPOINT FOR LIQUID RADWASTE DISCHARGE RADIATION MONITOR RM-18 STP I-103A 3 1&2
- FUNCTIONAL TEST OF GAS DECAY TANK GAS DISCHARGE MONITOR RM-22
[) STP l-1038 1 1&2 *CAllBRATION OF THE GAS DECAY TANK GAS DISCHARGE MONITOR RM-22 STP l-103B1 0 1&2
- GAS DECAY TANK GAS DISCHARGE MONITOR RM REMOVAL FROM SERVICE STP l-103B2 0 1&2
- GAS DECAY TANK GAS DISCHARGE MONITOR RM ELECTRONIC ALIGNMENT '
[) STP I-10333 3 1&2
- RADIATION SOURCE PRESENTATION (ISOTOPIC) CAL OF GAS DECAY TANK GAS DISCHARGE MONITOR RM-22 STP I-103B4 0 1&2
- GAS DECAY TANK GAS DISCHARGE MONITOR RM REINSTATEMENT TO SERVICE STP l-104A 3 1&2
- FUNCTIONAL TEST OF COMPONENT COOLING WATER PUMP HEADERS LIQUID PROCESS MONITORS RM-17A & 17B
,[] STP l-104B 1 1&2
- COMPONENT COOLING WATER PUMP DISCHARGE LIQUID PROCESS MONITORS RM-17A & 178 - CAllBRATION STP I-104B1 1 1&2
- COMPONENT COOLING WATER PUMP DISCHARGE HEADERS LIQUID PROCESS MONITORS RM-17A & RM-17B-REMOVE FROM SERVICE SlP l-104B2 2 1&2
- COMPONENT COOLING WATER PUMP DISCHARGE HEADER LIQUID RAD MONITORS RM-17A & RM-17B: ELECTRONIC ALIGNMENT
[) STP l-104B3 2 1&2
- COMPONENT COOLING WATER PUMP DISCHARGE HEADi LIQUID PROCESS MONITORS RM-17A & 178 - RADIATION SE RCE PRESENTATION - ISOTOPlc - CALIBRATION STP l-104B4 1 1&2
- REINSTATEMENT TO SERVICE-COMPONENT COOLING WA".tR PUMP DISCHARGE HEADERSLIQUID PROCESS MONITORS RM-17A & 17B STP l-104B5 3 1&2
- COMPONENT COOLING WATER PUMP DISCHARGING HEADERS MONITORS
) RM-17A/17B: SEMIANNUAL DISCRIMINATOR CHECK USING A MULTICHANNEL ANALYZER
! STP l-107A 6 1&2
- FUNCTIONAL TEST MISCELLANEDUS AREA RADIATION MONITORS l
RM-1,2,4,6,7 & 10 ' STP I-107B 1 1&2 *CAllBRATION Of MISCELLANE0US AREA RADIATION MONITORS RM-1,2,4,6,7,10 9 l l h
' l D
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 UOLUME 6 SURVEILLANCE TEST PROCEDURES D TABLE OF CONTENTS . NUMBER REV UNIT TITLE
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STP I-107B1 2 1&2
- REMOVAL FROM SERVICE - MISC AREA RAD MONITORS RM-1, 2, 4 ,
D 6, 7 & 10 - STP l-107B2 1 1&2
- ELECTRONIC ALIGNMENT OF MISCELLANEOUS AREA RADIATION MONITORS RM-1, 2, 4, 6, 7 & 10 STP I-107B3 3 1&2
- RADIATION SOURCE PRESENTATION - ISOTOPIC - CALIBRATION OF MISCELLANEOUSAREA RADI ATION MONITORS RM-1,2, 4, 6, 7, 8 & ,
10 3 STP l-107B4 1 1&2
- REINSTATEMENT TO SERVICE - MISCELLANEOUS AREA RADIATION MONITORS RM-1, 2, 4, 6, 7, 8 & 10 STP I-108A 2 1&2
- MISCELLANEOUS AIR PARTICULATE MONITORS RM-13 AND RM-21 FUNCTIONAL TEST STP I-108B 1 1&2
- CALIBRATION OF MISCELLANEOUS AIR PARTICULATE MONITORS RM-13 & RM-21 3 STP l-108B1 2 1&2 ** MISCELLANEOUS AIR PARTICULATE MONITORS RM-13 & RM REMOVAL FROM SERVICE STP l-108B2 0 1&2
- ELECTRONIC ALIGNMENT OF MISCELLANEOUS AIR PARTICULATE MONITORS RM-13 AND RM-21 STP 1-108B3 4 1&2
- MISCELLANEOUS AIR PARTICULATE MONITORS RM-13 & RM-21:
RADIATION SOURCE PRESENTATION (ISOTOPIC) CALIBRATION 3 STP I-108B4 2 1&2 ** REINSTATEMENT TO SERVICE: MISCELLANEOUS AIR PARTICULATE MONITORS RM-13 & RM-21 51P l-108BS 2 1&2
- MISCELLANEOUS AIR PARTICULATE MONITORS RM-13 & RM-21:
SEMIANNUAL DISCRIMINATOR CHECK USING MULTICHANNEL ANALYZER STF 1-111A 4 1&2
- FUNCTIONAL TEST OF STEAM GENERATOR BLOWDOWN SAMPLE EFFLLENT LIQUID MONITOR RM-19 3 STP 1-1118 1 1&2 *CAllBRATION OF STEAM GENERATOR BLOWDOWN SAMPLE LIQUID MONITOR RM-19 STP l-11181 2 1&2
- STEAM GENERATOP BLOWDOWN SAMPLE LIQUID MONITOR RM REMOVAL FROM SERVICE S1P 1-11182 0 1&2
- STEAM GENERATOR BLOWDOWN SAMPLE LIQUID MONITOR RM ELECTRONIC ALIGNMENT 3 STP l-11183 3 1&2
- RADIATION SOURCE PRESENTATION (ISOTOPIC) CALIBRATION OF STEAM GENERATOR BLOWDOWN SAMPLE LIQUID MONITOR RM-19 STP l-111B4 2 1&2
- REINSTATEMENT TO SERVICE - STEAM GENERATOR BLOWDOWN SAMPLE LIQUlD MONIT0; '" 19 STP 1-111B5 2 1&2
- SEMIANNUAL DIS's A10R CHECK USING A MULTICHANNEL ANALYZER - STEA" ;LNERATOR BLOWDOWN SAMPLE LIQUID MONITOR RM-19 3 STP l-ll8A 10 1&2 *** FUNCTIONAL TEST Of CONTROL ROOM PRESSURIZATION SYSTEM RAD]ATION MON] TORS RM-51, 52, 53 AND 54 STP l-118B 2 1&2
- CALIBRATION OF CONTROL ROOM PRESSURIZATION SYSTEM RADIATION MONITORS RM-51 & 52 & 53 & 54 STP l-11881 2 1&?
- REMOVAL FROM SERVICE OF THE CONTROL ROOM PRESSURIZATION SiSTEM RADIATll'N MON] TORS RM-51 & 52 & 53 & 54 7
i
i
}
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS ,,) NUMBER REV UNIT TITLE
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STP 1-118B2 4 1&2 **IS0 TOPIC CAllBRATION OF CONTROL ROOM PRESSURIZATION SYSTEM RADIATION MONITORS RM-51, 52, 53 & 54 c) STP l-11883 2 1&2 *RE]NSTATING OF CONTROL ROOM PRESSURIZATION SYSTEM SUCTION AIR RADIATIONMONITORS RM-51 & 52 & 53 & 54 STP l-119A 8 1&2 *** FUNCTIONAL TEST: FUEL HANDLING BUILDING AREA RADIATION MONITORS RIS-58/RIS-59 STP I-119B 2 1&2
- CALIBRATION OF FUEL HANDLING BUILDING AREA RADIATION MONITORS RIS-58 & RIS-59
() STP l-119B1 2 1L2
- FUEL HANDLING BUILDING AREA RADIATION MONITOR RIS-58 &
RIS-59 REMOVAL FROM SERVICE STP l-11962 2 IL2
- ELECTRONIC ALIGNMENT - FUEL HANDLING BulLDING AREA RADIATION MONITORS RIS-58 & RIS-59 STP 1-119B3 4 1&2 ** RADIATION SOURCE PRESENTATION (150.) CAL. FUEL HDLG.
BLDG. AREA RAD MONITORS RIS-5B & RIS-59 C) STP l-119B4 2 1&2
- REINSTATEMENT TO SERVICE: FUEL HANDLING BUILDING AREA RADIATION MONITOR RIS-58 AND RIS-59 STP l-120B 4 1 *CAllBRATION OF CONTAINMENT AIR PARTICULATE MONITOR RM-11 STP l-120B 0 2 *CAllBRATION OF CONTAINMENT AIR PARTICULATE MONITOR RM-ll STP l-120B1 B 1
- CONTAINMENT AIR PARTICULATE MONITOR RM-ll REMOVAL FROM SERVICE C) STP l-120B1 1 2 *** CONTAINMENT AIR PARTICULATE MONITOR RM-ll REMOVAL FROM SERVICE STP I-120B2 2 1
- ELECTRONIC ALIGNMENT OF CONTAINMENT AIR PARTICULATE MONITOR RM-ll STP 1-120B2 0 2
- ELECTRONIC AllGNMENT OF CONTAINMENT AIR PARTICULATE MONITOR RM-ll C) SlP l-120B3 5 1
- RADIATION SOUREE FRESENTATION (ISOTOPIC) CAllBRATION OF CONTAINMENT AIR PARTICULATE MONITOR RM-11 STP l-120B3 0 2
- RADIATION SOURCE PRESENTATION (ISOTOPIC) CAllBRATION OF CONTAINMENT AIR PARTICULATE STP l-120B4 6 1
- REINSTATEMENT TO SERVICE: CONTAINMENT AIR PARTICULATE MONITOR RM-11 C) STP l-120B4 1 2 *** REINSTATEMENT TO SERVICE: CONTAINMENT. AIR PARTICULATE MONITOR RM-11 STP l-120B5 4 1 *SEMINANNUAL CHECK FOR DISCRIMINATOR DRIFT USING A MULTICHANNEL ANALYZER: CONTAINMENT AIR PARTICULATE DETECTOR RM-11 STP l-120B5 0 2
- SEMIANNUAL CHECK FDP DISCRIMINATOR DRIFT USING A C) MULTICHANNEL ANALYZER: CONTAINMENT AIR DETECTOR RM-11 ,
STP l-3-F50 1 1 *AUxILI ARY FEEDWATER TO STEAM GENERATOR l-1 FLOW CHANNEL F1-50 CALIBRATION STP l-3-F50 0 2
- AUXILIARY FEEDWATER TO STEAM GENERATOR 2-1 FLOW CHANNEL FT-50 CALIBRATION l
- 5) l l
l C) ; i
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ==== ===========================================================
l STP l-3-F77 1 1 *AUXILI ARY FEEDWATER TO STEAM GENERATOR l-2 FLOW CHANNEL g- FT-77 CAllBRATION STP l-3-F77 0 2 ' AUXILIARY FEEDWATER TO STEAM GENERATOR 2-2 FLOW CHANNEL FT-77 CAllBRATION STP l-3-F78 1 1
- AUXILIARY FEEDWATER TO STEAM GENERATOR l-3 FLOW CHANNEL FT-78 CALIBRATION
! STP 1-3-F7B 0 2
- AUXILIARY FEEDWATER TO STEAM GENERATOR 2-3 FLOW CHANNEL *
) FT-78 CALIBRATION STP l-3-F79 1 1
- AUXILIARY FEEDWATER TO STEAM GENERATOR 1-4 FLOW CHANNEL l FT-79 CAllBRATION STP l-3-F79 0 2 *AUXlLIARY FEEDWATER TO STEAM GENERATOR 2-4 FLOW CHANNEL FT-79 CAllBRATION STP I-4-F53.A 1 1
- STEAM GENERATOR BLOWDOWN EFFLUENT LINE FLOW CHANNEL FT-53 FUNCTIONAL TEST k
l SlP l 4-F53.A 2 2
- STEAM GENERATOP BLOWDOWN EFFLUENT LINE FLOW CHANNEL FT-53 FUNCTIONAL TEST l STP l-4-F53.B 1 1
- STEAM GENERATOR BLOWDOWN EFFLUENT LINE FLOW CHANNEL FT-53 CALIBRATION STP l-4-F53.B 2 2
- STEAM GENERATOR BLOWDOWN EFFLUENT LINE FLOW CHANNEL FT-53 CAllBRATION
? STP l-4-PCVl9 3 1 ***10% STEAM DUMP VALVE PCV-19 CALIBRATION STP l-4-PCVl9 2 2 ***10% STEAM DUMP VALVE PCV-19 CALIBRATION STP l-4-PCV20 21 ***10# STEAM DUMP VALVE PCV-20 CAllBRATION STP l-4-PCV20 2 2 ***10% STEAM DUMP VALVE PCV-20 CAllBRATION STP l-4-PCV21 2 1 ***10f STEAM DUMP VALVE PCV-21 CALIBRATION STP l-4-PCV21 1 2 ***10% STEAM DUMP VALVE PCV-21 CALIBRATION ) STP 1-4-PCV22 2 1 ***10% STEAM DUMP VALVE PCV-22 CALIBRATION STP l-4-PCV22 1 2 ***10% STEAM DUMP VALVE PCV-22 CALIBRATION l STP 1-7-M455C.A 2 1 *PORV PCV-455C LTOP CHANNELS PT-403 AND TE-433B FUNCTIONAL TEST STP l-7-M455C.A 2 2 *PORV PCV-455C LTOP CHANNELS PT-403 AND TE-433B FUNCTIONAL TEST STP l-7-M455C.B 0 1 *PORV PCV-455C LTOP CHANNELS PT-403 AND TE-433B CALIBRATION STP l-7-M455C.B 2 2 *PORV PCV-455C LTOP CHANNELS PT-403 AND TE-433B CALIBRATION
- STP l-7-M456.A 1 1 *PORV PCV-456 LTOP CHANNELS PT-405 AND TE-423B FUNCTIONAL l TEST l STP l-7-M456.A 1 2 *PORV PCV-456 LTOP CHANNELS PT-405 AND TE-423B FUNCTIONAL l TEST
? STP l-7-M456.B 0 1 *PORV PCV-456 L10P CHANNELS PT-405 AND TE-423B CALIBRATION STP l-7-M456.B 2 2 *PORV PCV-456 LTOP CHANNELS PT-405 AND TE-423B CAllBRATION STP l-7-PCV455C 0 1
- PRESSURIZER PORV PCV-455C POSITION INDICATION CHANNEL CALIBRATION STP l-7-PCV455C 1 2
- PRESSURIZER PORV PCV-455C POSITION INDICATION CHANNEL CAllBRATION D
C) . 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCC TEST PROCEDURES C) TABLE OF CONTENTS NUMBER REV UNIT TITLE
===== ========= ===, ==== ===========================================================
STP l-7-PCV456 0 1
- PRESSURIZER POPV PCV-456 POSITION INDICATION CHANNEL C) CALIBRATION STP l-7-PCV456 0 2
- PRESSURIZER POPV PCV-456 POSITION INDICATION CHANNEL CAllBRATION STP l-7-PCV474 0 1
- PRESSURIZER POPV PCV-474 POSITION INDICATION CHANNEL CAllBRATION STP l-7-PCV474 0 2 *PRESSUR12ER PORV PCV-474 POSITION INDICATION CHANNEL
() CAllBRATION SlP l-B-F128 3 1
- CHARGING PUMPS DISCHARGE HEADER FLOW CHANNEL FT-128 CALIBRATION STP l-8-F128 2 2
- CHARGING PUMPS DISCHARGE HEADER FLOW CHANNEL FT-128
' CAllBRATION SlP l-9-F917 1 1
- CHARGING PUMPS DISCHARGE (RCS INJECTION) FLOW CHANNEL '
(? FT-917 CAllBRATION l STP l-9-F917 1 2
- CHARGING PUMPS DISCHARGE (RCS INJECTION) FLOW CHANNEL FT-917 CAllBRATION STP l-9-F91B 0 1
- SAFETY INJECTION PUMP 1-1 DISCHARGE FLOW CHANNEL FT-918 CALIBRATION S1P l-9-F91B 0 2
- SAFETY INJECTION PUMP 2-1 DISCHARGE FLOW CHANNEL FT-918 CALIBRATION
()
- SAFETY INJECTION PUMP 1-2 DISCHARGE FLOW CHANNEL FT-922 STP l-9-F922 0 1 i CALIBRATION '
STP l-9-F922 0 2
- SAFETY INJECTION PUMP 2-2 DISCHARGE FLOW CHANNEL FT-922 CAllBRATION STP l-9-F970A 1 1
- RESIDUAL HEAT EXCHANGER 1-1 OUTLET NARROW RANGE FLOW
() CHANNEL FT-970A CALIBRATION STP l-9-F970A 1 2
- RESIDUAL HE AT EXCHANGER 2-1 OUTLET NARROW RANGE FLOW CHANNEL FT-970A CAllBRATION STP l-9-F970B 1 1
- RESIDUAL HEAT EXCHANGER l-1 OUTLET WIDE RANGE FLOW CHANNEL l
l FT-970B CAtlBPATION i STP 1-9-F970B 1 2
- RESIDUAL HEAT EXCHANGER 2-1 OUTLET WIDE RANGE FLOW CHANNEL FT-970B CAllBRATION
'() STP 1-9-F971A 1 1
- RESIDUAL HEAT EXCHANGER 1-2 OUTLET NARROW RANGE FLOW ;
CHANNEL FT-971A CAllBRATION ' STP l-9-F971A 1 2
- RESIDUAL HEAT EXCHANGER 2-2 OUTLET NARROW RANGE FLOW'
+ CHANNEL FT-971A CALIBRATION STP l-9-F971B 1 1
- RESIDUAL HEAT EXCHANGER 1-2 OUTLET WIDE RAhGE FLOW CHANNEL FT-971B CAllBPATION
~()
STP l-9-F971B 1 2
- RESIDUAL HEAT EXCHANGER 2-2 OUTLET WIDE RANGE FLOW CHANNEL FT-971B CAllBRATION STP I-9-L920 0 1
- REFUELING WATER STORAGE TANK 1-1 LEVEL CHANNEL LT-920 .l i
CAllBRATION 0 1 j STP l-9-L921
- REFUELING WATER STORAGE TANK 1-1 LEVEL CHANNEL LT-921 CALIBRATION l
() 1 l C) -
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES I ) TABLE OF CONTENTS ; NUMBER REV UNIT TITLE
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STP l-9-L940 0 1
- CONTAINMENT RECIRCULATION SUMP NARROW RANGE LEVEL CHANNEL LT-940 CAllBRATION STP l-9-L940 0 2
- CONTAINMENT RECIRCULATION SUMP NARROW RANGE LEVEL CHANNEL LT-940 CAllBRATION STP l-9-L941 0 1 *CONTAlHMENT RECIRCULATION SUMP NARROW RANGE LEVEL CHANNEL LT-941 CALIBRATION STP I-9-L941 1 2
- CONTAINMENT RECIRCULATION SUMP HARROW RANGE LEVEL CHANNEL
) STP l-9-L950.A 0 1 LT-941 CALIBRATION
- ACCUMULATOR NAPROW RANGE LEVEL CHANNELS LC-950A/B THRU LC-957A/B FUNCTIONAL STP l-9-L950.A 0 2
- ACCUMULATOR NARROW RANGE LEVEL CHANNELS LC-950A/B THRU LC-957A/B FUNCTIONAL '
STP l-9-L950.B 0 1 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-950 CAllBRATION STP l-9-L950.B ) STP 1-9-L951.B 1 0 2 1
*** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-958 CALIBRATION
*** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-951 CALIBRATION STP l-9-L951.B 1 2 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-951 CAllBRATION STP l-9-L952.B 0 1 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-952 CALIBRATION ,
STP l-9-L952.B 1 2 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-952 CAllBRATION STP l-9-L953.B 0 1 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-953 CAllBRATION STP I-9-L953.B 1 2 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-953 CALIBRATION 3 STP l-9-L954.B 0 1 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-954 CAllBRATION STP I-9-L954.B 1 2 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-954 CALIBRATION STP I-9-L955.B 0 1 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-955 CALIBRATION , STP l-9-L955.B 1 2 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-955 CALIBRATION STP l-9-L956.B 0 1 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-956 CALIBRATION STP l-9-L956.B 1 2 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-956 CAllBRATION 3 STP l-9-L957.B 0 1 *** ACCUMULATOR NARROW RANGE LEVEL CHANEL LT-957 CALIBRATION STP l-9-L957.8 1 2 *** ACCUMULATOR NARROW RANGE LEVEL CHANNEL LT-957 CALIBRATION STP l-9-L955 0 1 *** ACCUMULATOR l-1 WIDE RANGE LEVEL CHANNEL LT-958 CAllBRATION STP l-9-L958 1 2 *** ACCUMULATOR 2-1 WIDE RANGE LEVEL CHANNEL LT-958 CALIBRATION l 3 STP l-9-L959 0 1 *** ACCUMULATOR l-2 WIDE RANGE LEVEL CHANNEL LT-959 CALIBRATION STP l-9-L959 1 2 *** ACCUMULATOR 2-2 WIDE RANGE LEVEL CHANNEL LT-959 CALIBRATION STP l-9-L960 0 1 *** ACCUMULATOR l-3 WIDE RANGE LEVEL CHANNEL LT-960 . CALIBRATION , 3 STP l-9-L960 1 2 *** ACCUMULATOR 2-3 WIDE RANGE LEVEL CHANNEL LT-960 CALIBRATION ; STP I-9-L961 0 1 *** ACCUMULATOR l-4 WIDE RANGE LEVEL CHAENEL LT-961 l CALIBRATION STP l-9-L961 1 2 *** ACCUMULATOR 2-4 WIDE RANGE LEVEL CHANNEL LT-961 ; CAllBRATION i
I 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TCST PROCEDURES TABLE OF CONTENTS g, NUMBER REV UNil TITLE
=============== === ==== ===========================================================
STP l-9-P960.A 0 1
- ACCUMULATOR PRESSURE CHANNEL PC-960A/B THRU PC-967A/B FUNCTIONAL 8I STP l-9-P960.A 0 2
- ACCUMULATOR PRESSURE CHANNEL PC-960A/B THRU PC-967A/B FUNCTIONAL STP l-9-P960.B 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-960 CAllBRATION STP l-9-P960.6 0 2
- ACCUMULATOR PRESSURE CHANNEL PT-960 CAllBRATION STP I-9-P961.8 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-961 CALIBRATION STP I-9-P961.B 0 2
- ACCUMULATOR PPESSURE CHANNEL PT-961 CALIBRATION
[) STP 1-9-P962.B 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-962 CAllBRATION STP I-9-P962.6 0 2
- ACCUMULATOR PRESSURE CHANNEL PT-962 CAllBRATION STP l-9-P963.8 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-963 CALIBRATION STP l-9-P963.B 0 2 *ACCUMULATOP PRESSURE CHANNEL PT-963 CALIBRATION STP I-9-P964.B 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-964 CALIBRATION STP I-9-P964.B 0 2
- ACCUMULATOR PRESSURE CHANNEL PT-964 CALIBRATION C) STP l-9-P965.B 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-965 CALIBRATION STP l-9-P965.B 0 2
- ACCUMULATOR PRESSURE CHANNEL PT-965 CAllBRATION STP l-9-P966.B 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-966 CAllBRATION STP l-9-P966.B 0 2 *ACCUMJLATOR PRESSURE CHANNEL PT-966 CAllBRATION STP l-9-P967.B 0 1
- ACCUMULATOR PRESSURE CHANNEL PT-967 CALIBRATION STP I-9-P967.B 0 2
- ACCUMULATOR PRESSURE CHANNEL PT-967 CAllBRATION.
C) STP I-19-F40 2 1
- CONTAINMENT STRUCTURE SUMP l-1 DISCHARGE FLOW CHANNEL FT-40 CALIBRATION STP I-19-F40 1 2
- CONTAINMENT STRUCTURE SUMP 2-1 DISCHARGE FLOW CHANNEL FT-40 CAllBPA110h STP l-19-F41 3 1
- CONTAINMENT STRUCTURE SUMP 1-2 DISCHARGE FLOW CHANNEL FT-41 CALIBPATION
[) STP l-19-f41 1 2 *CONTA,1NMENT STPUCTURE SUMP 2-2 DISCHARGE FLOW CHANNEL FT-41 CAllBRATION STP l-19-F42 0 1
- REACTOR CAVITY SUMP l-1 DISCHARGE FLOW CHANNEL FT-42 CALIBRATION STP l-19-F42 0 2
- REACTOR CAVITY SUMP 2-1 DISCHARGE FLOW CHANNEL FT-42 CALIBRATION K) STP l-19-F43 0 1
- REACTOR COOLANT DRAIN TANK 1-1 DISCHARGE FLOW CHANNEL FT-43 CAllBRATION STP l-19-F43 0 2
- REACTOR COOLANT DRAIN TANK 2-1 DISCHARGE FLOW CHANNEL FT-43 CAllBRAT10N STP l-19-F243.A 1 IL2
- LIQUID RADWASTE EFFLUENT L]NE FLOW CHANNEL FIT-243 FUNCTIONAL TEST 4 C) STP l-19-f243.8 1 IL2 *** LIQUID RADWASTE EFFLUENT LINE FLOW CHANNEL FIT-243 I CAllBRATION STP I-19-L60 0 1
- CONTAINMENT STRUCTURE SUMP l-1 LEVEL CHANNEL LT-60 l CAllBRATION STP l-19-L60 0 2 *CONTAINHENT STRUCTURE SUMP 2-1 LEVEL CHANNEL LT-60 CAllBRATION 9
e
/
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS 7 NUMBER REV UNIT TITLE
... m========== ==. ==== ..==...====.===.=......=..======..........=.......=
STP l-19-L61 0 1
- CONTAINMENT STPUCTURE SUMP l-2 LEVEL CHANNEL LT-61 l I
CAllBRATION ) STP l-19-L61 0 2
- CONTAINMENT STRUCTURE SUMP 2-2 LEVEL CHANNEL LT-61 CAllBRATION STP l-27-F251.A 1 1&2 *0lLY WATER SEPARATOR SYSTEM FLOW CHANNEL FT-251 FUNCTIONAL TEST STP l-27-F251.B 0 1&2 *0lLY WATER SEPARATOR SYSTEM FLOW CHANNEL FT-251 CALIBRATION l STP l-39-F12.B 2 1
- PLANT VENT IS0 KINETIC SAMPLE SYSTEM SKID CAllBRATION
) STP 1-39-F12.B 3 2
- PLANT VENT IS0 KINETIC SAMPLE SYSTEM SKID CAllBRATION STP I-39-R14R.A 1 1
- PLANT VENT NOBLE GAS RADIATION MONITOR RM-14R FUNCTIONAL '
TEST STP l-39-R14R.A 0 2
- PLANT VENT NOBLE GAS RADIATION MONITOR RM-14R FUNCTIONAL TEST STP I-39-R14R.B 2 1
- PLANT VENT DISCHARGE REDUNDANT NOBLE GAS RM-14R RADIATION
) MONITOR CAllBRATION STP l-39-R14R.B 2 2
- PLANT VENT DISCHARGE REDUNDANT NOBLE GAS RM-14R RADIATION MONITOR CALIBRATION STP l-39-R14.A 1 1
- PLANT VENT NOBLE GAS RADI ATION MONITOR RM-14 FUNCTIONAL TEST ST P l-39-R14. A 0 2
- PLANT VENT NOBLE GAS RADIATION MONITOR RM-14 FUNCTIONAL
) TEST STP I-39-R14.B 2 1
- PLANT VENT DISCHARGE NOBLE GAS RM-14 RADIATION MONITOR CAllBRATION STP l-39-R14.B 1 2
- PLANT VENT DISCHARGE NOBLE GAS RM-14 RADIATION MONITOR CALIBRATION STP l-39-R14.C 1 2
- HEAT 1 RACE CONTROLLER CAllBRATION PLANT VENT RM-14/14R
) SlP l-39-R15R.A 2 1 *CDNDENSER AIR f]ECTOR DISCHARGE RM-15R FUNCTIONAL TEST STP 1-39-R15R.A 0 2
- CONDENSER AIR EJECTOR 9ISCHARGE RM-15R FUNCTIONAL TEST STP I-39-R15.A 2 1
- CONDENSER AIR EJECTOR DISCHARGE RM-15 FUNCTIONAL TEST ST P l-39-R15. A 0 2
- CONDENSER AIR EJECTOR DISCHARGE RM-15 FUNCTIONAL TEST STP l-39-R15.B 6 1
- CONDENSER AIR EJECTOR DISCHARGE RM15/15R RADIATION MONITOR CAllBRATION l
) STP l-39-R15.B 1 2 *(OTSC 6-25-93/G0FF) CONDENSER AIR EJECTOR DISCHARGE l I RM15/15R RADIATION ' STP l-39-R24R.A 1 1
- PLANT VENT IODINE RADIATION MONITOR RM-24R FUNCTIONAL TEST STP I-39-R24R.A 0 2
- PLANT VENT 10 DINE RADIATION MONITOR RM-24R FUNCTIONAL TEST STP l-39-R24R.B 2 1
- PLANT VENT DISCHARGE 10 DINE REDUNDANT RM-24R RADIATION MONITOR CALIBRATION
) STP l-39-R24R.B 2 2
- PLANT VENT DISCHARGE 10 DINE REDUNDANT RM-24R RADIATION MONITOR CAllBRATION STP l-39-R24.A 1 1
- PLANT VENT IODINE RADIATION MONITOR RM-24 FUNCTIONAL TEST STP 1-39-R24.A 0 2
- PLANT VENT 10 DINE RADIATION MONITOR RM-24 FUNCTIONAL TEST STP l-39-R24.8 2 1
- PLANT VENT DISCHARGE 10 DINE RM-24 RADIATION MONITOR CAtlBRATION
) 1
O 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES () TABLE OF CONTENTS HUMBER REV UNIT TITLE
=============== === ==== ===================================----====================
STP I-39-R24.B 1 2
- PLANT VENT DISCHARGE 10 DINE RM-24 RADIATION MONITOR
() CAllBRATION STP l-39-R28R.A 1 1
- PLANT VENT PARTICULATE RADIATION MONITOR RM-28R FUNCTIONAL TEST STP l-39-R28R.A 0 2
- PLANT VENT PARTICULATE RADIATION MONITOR RM-28R FUNCTIONAL TEST STP I-39-R2BR.B 2 1
- PLANT VENT DISCHARGE PARTICULATE RM-28R REDUNDANT RADIATION MONITOR CALIBRATION
'() STP l-39-R28R.B 2 2
- PLANT VENT DISCHARGE PARTICULATE RM-28R REDUNDANT RADIATION MONITOR CAllBRATION STP I-39-R28R.C 2 1 *** PLANT VENT DISCHARGE REDUNDANT NORMAL RANGE IS0 KINETIC FLOW CONTROL CALIBRATION STP l-39-R28R.C 1 2
- PLANT VENT DISCHAPGE REDUNDANT NORMAL RANGE 150 KINETIC
() FLOW CONTROL CAllBRATION SlP l-39-R28.A 2 1
- PLANT VENT PARTICULATE RADI ATION MONITOR RM-28 FUNCTIONAL TEST STP l-39-R28.A 0 2
- PLANT VENT PARTICULATE RADIATION MONITOR RM-28 FUNCTIONAL TEST STP l-39-R28.B 2 1
- PLANT VENT DISCHARGE PARTICULATE RM-28 RADIATION MONITOR
() CALIBRATION STP l-39-R26.B 1 2
- PLANT VENT DISCHARGE PARTICULATE RM-28 RADI ATION MONITOR
) CAllBRATION STP l-39-R28.C 2 1 *** PLANT VENT DISCHARGE NORMAL RANGE SKID IS0 KINETIC FLOW CONTROL CALIBRATION ! STP l-39-R28.C 1 2
- PLANT VENT DISCHARGE NORMAL RANGE SKID IS0 KINETIC FLOW CONTROL CAllBRATION
!() STP l-39-R30.B 0 1
- CONTAINMENT HIGH RANGE AREA MONITOR RM-30 CAllBRATION STP l-39-R30.B 0 2
- CONTAINMENT HIGH RANGE AREA MONITOR RM-30 ChLIBRATION ST P 1-39-R31.B 0 1
- CONTAINMENT HIGH RANGE AREA MONITOR RM-31 CAllBRATION l STP l-39-R31.B 0 2
- CONTAINMENT HIGH RANGE AREA MONITOR RM-31 CAllBRATION
! STP l-39-R34.A 1 1
- PLANT VENT AREA RADI ATION MONITOR RM-34 FUNCTIONAL TEST 33 STP I-39-R34.A 1 2
- PLANT VENT AREA RADI ATION MONITOR RM-34 Fl'NCTIONAL TEST i STP I-39-R34.B 2 1
- PLANT VENT AREA RADI ATION MONITOR RM-34 CALIBRAT10N l STP l-39 R34.B 2 2
- PLANT VENT AREA RADIATION MONITOR RM-34 (ALIBRATION l STP l-39-R40.A 0 1
- PLANT VENT HIGH RANGE GRAB SAMPLER RX-40 TUNCTIONAL TEST l STP l-39-R40.A 1 2
- PLANT VENT HIGH RANGE GRAB SAMPLER RX-40 FUNCTIONAL TEST I STP l-39-R44A.A 1 1
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44A FUNCTIONAL TEST I(3 STP l-39-R44A.A 0 2
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44A FUNCTIONAL TEST STP I-39-R44A.B 2 1
- CONTAINMENT VENTILATION EXHAUST RM-44A RADIATION MONITOR CALIBRATION l
STP l-39-R44A.B 0 2
- CONTAINMENT VENTILATION EXHAUST RM-44A RADIATION MONITOR CALIBRATION t)
I l i O
i 1 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES () TABLE OF CONTENTS NUMBER REV UNIl TITLE
=============== === ==== ===========================================================
STP l-39-R44A.C 4 1
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44A TIME i
() RESPONSE TEST STP l-39-R44A.C 4 2
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44A TIME RESPONSE TEST STP l-39-R448.A 1 1
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44B FUNCTIONAL TEST STP l-39-R44B.A 0 2
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44B FUNCTIONAL TEST
() STE l-39-P44B.B 2 1
- CONTAINMENT VENTILATION EXHAUST RM-44B RADIATION MONITOR CALIBRATION ETP l-39-R448.B 0 2
- CONTAINMENT VENTILATION EXHAUST RM-44B RADIATION MONITOR .
CALIBRATION STP l-39 R44B.C 4 1
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44B TIME RESPONSE TEST
() S1P l-39-R448.C 3 2
- CONTAINMENT VENTILATION EXHAUST RAD MONITOR RM-44B TIME RESPONSE TEST STP l-39-R71.A 0 1 *** FUNCTIONAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-71 STP l-39-R71.A 0 2 *** FUNCTIONAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-71
() STP l-39-R71.B 0 1 *** MAIN STEAM LINE RADIATION MONITOR RM-71 CALIBRATION STP l-39-R71.8 1 2 *** MAIN STEAM LINE RADIATION MONITOR RM-71 CALIBRATION STP l-39-R72.A 0 1 *** FUNCTIONAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-72 STP l-39-R72.A 0 2 ***FUNC'10NAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-72 () STP l-39-R72.B 0 1 *** MAIN STEAM LINE RADIATION MONITOR RM-72 CALIBRATION STP 1-39-R72.B 1 2 *** MAIN STEAM LINE RADIATION MONITOR RM-72 CALIBRATION STP l-39-R73.A 0 1 *** FUNCTIONAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-73 STP l-39-R73.A 0 2 *** FUNCTIONAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-73 c) STP l-39-R73.B 0 1 *** MAIN STEAM LINE RADIATION MONITOR RM-73 CALIBRATION STP l-39-R73.B 1 2 *** MAIN STE AM LINE RADI ATION MONITOR RM-73 CALIBRATION STP l-39-R74.A 0 1 *** FUNCTIONAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-74 STP l-39-R74.A 0 2 *** FUNCTIONAL TEST OF MAIN STEAM LINE RADIATION MONITOR RM-74 () ST P l-39-R74.B 0 1 *** MAIN STEAM LINE RADIATION MONITOR RM-74 CAllBRATION STP I-39-R74.B 1 2 *** MAIN STE AM LINE RADI ATION MONITOR RM-74 CAllBRATION STP I-39-R87.B 2 1 *** PLANT VENT DISCHARGE NOBLE GAS HIGH RANGE RM-87 RADI ATION MONITOR CAllBRATION STP l-39-RB7.8 2 2
- PLANT VENT DISCHARGE NOBLE GAS HIGH RANGE RM-87 RADIATION MONITOR CALIBPATION
)
.O
22 JUL 93 DIABLO CANYOH POWER PLANT UNITS 1 AND 2 VOLUME 6 SlfRVEILLANCE TEST PROCEDURES TABLE OF CONTENTS
)
NUMBER RLV UNIT TITLE
=============== ===. ==== ===========================================================
3 1&2 "* PRIMARY METEOROLOGY WIND DIRECTION, WIND SPEED, AND A(R STP l-40-M559.B TEMPERATURE / DELTA-T CAllBRATION
) SlP l-52-M.1 2 1&2
- EMERGENCY RESPONSE FACILITIES DATA SYSTEM FUNCTIONAL CHEC AND VERIFICATION STP V-2 7 1&2
- EXERCISING AND POSITION VERIFICATION OF POWER OPERATED VALVES STP V-2A1 0 1&2
- AUXILIARY SALTWATER CROSS CONNECT HEADER VALVES STP V-2A2 0 1&2
- AUXILIARY SALTWATER CROSSTIE VALVE FCV-601
) STP V-2A3 0 1&2 *AUXILI ARY SALTWATER DEMUSSELING VALVES STP V-2E 5 1&2
- EXERCISING AND POSITION VERIFICATION OF VALVES 9001A AND 9001B STP V-2B2 3 1&2 *EXERCISlNG AND POSITION VERIFICATION OF VALVES 9003A AND 9003B STP V-2C 8 1
- CHARGING INJECTION LINE VALVES 8803A AND 8803B
-J 2 2
- CHARGING INJEC110N LINE VALVES 8803A AND 8803B STP V-2C STP V-2D 5 1&2 *RHR PUMP RECIRCULATION VALVES STP V-202 2 1&2
- EXERCISING AND POSITION VERIFICATION OF VALVES 8700A AND 87008 STP V-2D3 1 1&2
- EXERCISING AND POSITION VERIFICATION OF VALVES 8701 AND 8702
) STP V-2E1 0 1&2
- CONTAINMENT VENTILATION VALVE FCV-661 (OC)
STP V-2E2 0 1&2
- CONTAINMENT vet <TILAT10N VALVE RCV-12 (OC)
STP V-?E3 0 1&2
- CONTAINMENT VENTILATION VALVES FCV-663 AND FCV-664 (OC)
STP V-2f 4 1&2
- COMPONENT COOLING WATER VALVES STP V-2G B 1&2
- SPRAY ADDITIVE TANK OUTLET VALVES STP V-2H 9 1&2
- MISCELLANEOUS AUXILIARY BUILDING VALVES
) SlP V-211 4 1&2
- CHARGING AND LETDOWN VALVES STP V-212 3 1&2
- CHARGING AND LETDOWN VALVES STP V-2J1 1 1&2 *0UTSIDE CONTAINMENT SAMPLE ISOLATION VALVES STP V-2J2 1 1&2 *0UTSIDE CONTAINMENT COMPONENT COOLING WATER ISOLATION VALVES STP V-2J3 4 1&2 *0UTSIDE CONTAINMENT RESIDUAL HEAT REMOVAL ISOLATION VALVES
) STP V-?J4 2 1&2 *0UTSIDE CONT AINMENT STEAM GENERATOR ISOLATION VALVES STP V-2J5 1 1&2 *0UTSIDE CONTAINMENT RADWASTE ISOLATION VALVES STP V-2J6 5 1&2 *0UTSIDE CONTAINMENT SAFETY INJECTION AND CHARGING INJECTION ISOLATION VALVES STP V-2J7 1 1&2 *0UTSIDE CONTAlhMENT MISCELLANEOUS ISOLATION VALVES STP V-2J8 0 1&2
- EXERCISING OUTSIDE CONT AINMENT LETDOWN ISOLATION VALVE
) CVCS-8152 STP V-2K 5 1&2 *(OTSC 07/15/93 - NELSON) CVCS BORATION VALVES FCV-110A,
' FCV-111A, FCV-1108B, FCV-111B, AND CVCS-8104 STP V-2L 5 1&2
- EXERCISING PRESSURIZER SPRAY VALVES STP V-2M1 1 1&2
- PRESSURIZER RELIEF TANK TO REACTOR COOLANT DRAIN TANK VALVE 8031
) )
22'JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUML 6 SURVEILLANCE TEST PROCEDURES , TABLE OF CONTENTS !O NUMBER REV UNIT TITLE 1 4
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, STP V-2M2 0 1&2
- SIS HOT LEGS 1 AND 2 TEST ISOLATION '
' STP V-2M3 0 1&2 *RHR RETURN TO HOT LEG VALVE 8703 lO STP V-2M4 0 1&2 *RCS TO EXCESS LETDOWN VALVES 8166 AND 8167 STP V-2N1 0 1&2
- EXERCISING STEAM GENERATOR BLOWDOWN VALVES FCV-760,
- FCV-761, FCV-762, FCV-763
- STP V-2N2 0 1&2
- EXCESS LETDOWN VALVES HCV-123 AND CVCS-8143
*lNSIDE CONTAINMENT LEAK TEST ISOLATION VALVES i STP V-201 1 1&2
- STP V-202 2 1&2 *lNSIDE CONTAINMENT COMPONENT COOLING WATER ISOLATION VALVES i O STP V-203 2 1&2 *lNSIDE CONTAINMENT SAMPLE ISOLATION VALVES STP V-204 1 1&2 *lNSIDE CONTAINMENT CHILLED WATER AND PRT ISOLATION VALVES
*INSIDE CONTAINMENT RADWASTE ISOLATION VALVES :
STP V-205 1 1&2 ' STP V-206 2 1&2
- EXERCISING AND POSITION VERIFICATION OF VALVES 8982A AND 8982B STP V-2P 3 1&2
- EXERCISING ACCUMULATOR CHECK VALVE LEAK TEST VALVES AND
- O REACTOR VESSEL FLANGE LEAK OFF VALVE
- STP V-2P1 1 1&2
- EXERCISING ACCUMULATOR ISOLATION VALVE SIP V-20 7 1&2
- CONTAINMENT VENTILATION VALVES (IC)
STP V-25 7 1&2 *CVCS VALVES : STP V-211 3 1&2
- PRESSURIZER POWER OPERATED RELIEF VALVES STP V-2T2 11 1 *** EXERCISING & POSITION INDICATOR VERIFICATION OF REACTOR O VESSEL HEA3 VENT VALVES RCS-1-8078A, B, C, AND D STP V-2T2 4 2 *** EXERCISING AND POSITION INDICATOR VERIFICATION OF REACTOR VESSEL HEAD VENT VALVES RCS-2-8078A, B, C, AND D STP V-2U1 A 0 1&2
- EXERCISING S/G NO.1 MSIV AND BYPASS VALVE STP V-2U1B 0 1&2
- EXERCISING S/G NO. I 10% STEAM DUMP VALVE PCV-19 STP V-2U1C 1 1&2 *EXERCISlNG S/G NO. 1 FEEDWATER ISOLATION AND CONTROL VALVES O STP V-201D 1 1&2
- EXERCISING S/G NO. 1 AFW SUPPLY VALVES LCV-106 AND LCV-110 t STP V-202A 0 1&2
- EXERCISING S/G NO. 2 MSIV AND BYPASS VALVE STP V-2U2B 0 1&2
- EXERCISING S/G NO. 2 10X STEAM DUMP VALVE PCV-20 >
STP V-202C 0 1&2
- EXERCISING S/G NO. 2 FEEDWATER ISOLATION AND CONTROL VALVES STP V-2U2D 1 1&2
- EXERCISING S/G NO. 2 AFW SUPPLY VALVES LCV-107 AND LCV-111 STP V-202E O 1&2
- EXERCISING S/G NO. 2 AFW TURBINE STEAM SUPPLY VALVE FCV-37 '
O STP V-203A 0 1&2
- EXERCISING S/G NO. 3 MSIV AND BYPASS VALVE STP V-203B 0 1&2
- EXERCISING S/G NO. 3 10% STEAM DUMP VALVE PCV-21 STP V-2U3C 0 1&2
- EXERCISING S/G NO. 3 FEEDWATER ISOLATION AND CONTROL VALVES '
STP V-203D 1 1&2
- EXERCISING S/G NO. 3 AFW SUPPLY VALVES LCV-108 AND LCV-115 STP V-?U3E O 1&2
- EXERCISING S/G NO. 3 AFW TURBINE STEAM SUPPLY VALVE FCV-38 STP V-204A 0 1&2
- EXERCISING S/G NO. 4 MSIV AND BYPASS VALVE O STP V-204B 0 1&2
- EXERCISING S/G NO. 4 10% STEAM DUMP VALVE PCV-22 0 1&2 STP V-204C
- EXERCISING S/G NO. 4 FEEDWATER ISOLATION AND CONTROL VALVES
- STP V-204D 1 1&2
- EXERCISING S/G NO. 4 AFW SUPPLY VALVES LCV-109 AND LCV-113 STP V-2VSA 0 1&2
- EXERCISING AFW VALVES FCV-436 AND FCV-437 1 STP V-205B 0 1&2
- EXERCISING AFW STEAM SUPPLY VALVES FCV-95 AND FCV-152
'O O
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22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS 3 NUMBER REV. UNIT TITLE
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STP V-2V1 2 1&2
- EXERCISING AND POSITION VERIFICATION OF VALVES 8804A AND
) STP V-2V2 2 1&2 8804B
- EXERCISING AND POSITION VERIFICATION OF VALVES 8974A AND 8974B STP V-2W 2 1&2
- BORIC ACID STORAGE TANK RECIRCULATION VALVES STP V-2X 12 1&2
- EXERCISING CONTAINMENT HYDROGEN PURGE ISOLATION VALVES STP V-2Y 11 1&2
- REGENERATIVE HEAT EXCHANGER LETDOWN INLET AND LETDOWN ORIFICE ISOLATION VALVES 3 STP V-2Z 4 1&2 *PWST TO CHARGING PUMP SUCTION VALVES STP V-3 9 1&2
- EXERCISING SAFETY RELATED VALVES - GENERAL PROCEDURE STP V-3El 9 1&2
- EXERCISING VALVE FCV-110A BORIC ACID TO BLENDER STP V-3E2 3 1&2
- EXERCISING VALVE FCV-110B BLENDER TO CHARGING HEADER STP V-3E5 5 1
- EXERCISING VALVES 8104 AND 8445, EMERGENCY BORATE VALVES STP V-3E5 1 2
- EXERCISING VALVES 8104 AND 8445, EMERGENCY BORATE VALVES 3 STP V-3E6 2 1&2
- EXERCISING CVCS-FCV-128 CENTRIFUGAL CHARGING PUMP FLOW CONTROL VALVE STP V-3E7 3 1
- VERIFICATION OF BORIC ACID FLOWPATH FROM BORIC ACID TANKS TO RCS STP V-3E7 1 2
- VERIFICATION OF BORIC ACID FLOWPATH FROM BORIC ACID TANKS TO RCS 3 STP V-3E9 3 1
- EXERCISING CHARGING INJECTION LINE VALVES, SI-1-8801A AND SI-1-8801B STP V-3E9 2 2
- EXERCISING CHARGING INJECTION LINE VALVES, SI-2-8801A AND SI-2-8801B STP V-3E11 0 1&2
- EXERCISING CHAkGlNG INJECTION ISOLATION VALVES, SI-8803A AND SI-8803B 3 SlP V-3fl 8 1&2
- EXERCISING VALVE FCV-495 ASW PUMP 2 CROSSTIE VALVE STP V-3F2 8 IL2
- EXERCISING VALVE FCV-496 ASW PUMP 1 CROSSTIE VALVE STP V-3F3 9 1&2
- EXERCISING VALVE FCV-601, UNITS 1 AND 2 ASW CROSSTIE STP V-3F4 8 1&2
- EXERCISING VALVE FCV-602 CCW HX NO. 1 SALTWATER INLET STP V-3F5 7 1&2
- EXERCISING VALVE FCV-603 CCW HX NO 2 SAL 1 WATER INLET STP V-3G1 5 1&2
- EXERCISING FCV-410, GAS DECAY TANK TO PLANT VENT 3 STP V-3H3 4 1&2
- EXERCISING VALVE FCV-355, CCW HEADER C SUPPLY ISOLATION STP V-3H4 4 1&2
- EXERCISING VALVE FCV-356 CCW SUPPLY TO RCP'S AND REACTOR VESSEL SUPPORT COOLERS STP V-3HS 4 1&2
- EXERCISING FCV-357, RCP THERMAL BARRIER CCW RETURN 150 OUTSIDE CONTAINMENT S1P V-3H6 4 1&2 *EXERCISlNG FCV-363, RCP OIL COOLER CCW RETURN 150 OUTSIDE 3 CONTAINMENT STF V-3H7 4 1&2 *EXERCISlNG VALVES FCV-364 & FCV-365 RHR HEAT EXCHANGER CCW RETURN VALVES STP V-3HB 7 1&2
- EXERCISING FCV-430 & FCV 431, CCW HEAT EXCHANGER OUTLET ISOL VALVES
) D
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22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS () NUMBER REV UNIT TITLE
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STP V-3H9 6 1&2
- EXERCISING VALVE FCV-750, RCP THERMAL BARRIER CCW RETURN ISOLATION INSIDE CONTAINMENT
() SlP V-3H10 4 1&2
- EXERCISING VALVE FCV-749, RCP OIL COOLER CCW RETURN ,
ISOLATION INSIDE CONTAINMENT STP V-3H11 12 1&2
- EXERCISING VALVES LCV-69 AND LCV-70 MAKE-UP WATER TO CCW SURGE TANK STP V-3H12 2 1&2
- EXERCISING RCV-16, CCW SURGE TANK VENT VALVE c)
STP V-311 9 1&2
- EXERCISING VALVES 8998A AND 8998B, CONTAINMENT SPRAY EDUCTOR CHECK VALVES '
STP V-312 4 1&2
- SPRAY ADDITIVE TANK OUTLET VALVES CS-8994A AND CS-8994B STP V-313 4 1&2
- FULL STROKE EXERCISE OF CONTAINMENT SPRAY VALVES CS-9001A AND CS-9001B STP V-314 2 1&2
- EXERCISING RHR OUTLET TO SPRAY HEADER VALVES 9003A AND i
9003B c) STF V-331 7 1&2 *EXERCISlNG THE BLOCK VALVES TO THE PRESSURIZER PORV'S i VALVES RCS-8000A 80008, 8000C STP V-3J2 3 1&2
- EXERCISING PRESSURIZER POWER OPERATED RELIEF VALVES PCV-455C 456 & 474 STP V-3K2 4 1&2
- EXERCISING RCP SEAL RETURN ISOLATION VALVES CVCS-8100 AND '
CVCS-8112 c) ' SlP V-3L3 3 1&2
- EXERCISING VALVES 8141A, B, C & D RCP NO 1 SEAL LEAKOFF SlP V-3r4 4 1&2
- EXERCISING VALVE CVCS-8142 REACTOR COOLANT PUMPS N0 1 SEAL BYPASS STP V-3s5 12 1&2
- EXERCISING VALVES 8146 NORMAL CHARGING & 8147 ALTERNATE r CHAPGING AND CHECK VALVES 8378A B.C AND 8379 A, B I STP V-3n6 4 1&2 *EXERCISlNG VALVES 8166 AND 8167 REACTOR COOLANT SYSTEM c)
EXCESS LETDOWN , STi V-3K7A 11 1&2 *EXERCISlNG LETDOWN ORIFICE STOP VALVE CVCS-8149A, B, AND C 3 STP V-3K7B 3 1&2
- EXERCISING LETDOWN ISOLATION VALVE 8152 STP V-3f7C 1 1&2
- STROKE TIMING OF PRESSURIZER LEVEL CONTROL VALVES CVCS-LCV-459, AND 460 l STP V-3KB 5 1&2
- EXERCISING VALVES 8377, 8145 & 8148 PRESSURIZER AUXILIARY C)
SPRAY STP V-3K9 4 1&2
- EXERCISING VALVES 8105 & 8106, CENTRIFUGAL CHARGING PUMP RECIRCULATION STP V-3K10 6 1&2
- EXERCISING OF REFUELING WATER SUPPLY TO CHARGING PUMP SUCTION CHECK VALVE 8924 STP V-3K11 2 1&2
- EXERCISING RWST TO CHARGING PUMP VALVES SI-8805A AND C) SI-8805B STP V-3K12 0 1&2
- EXERCISING VCT OUTLET ISOLATION VALVES LCV-112B & LCV-112C STP V-3K13 5 1&2
- EXERCISING CVCS CHARGING ISOLATION VALVES 8107 AND 8108 STP V-3L1 11 1&2
- EXERCISING VALVES 8802A AND 88028, SAFETY INJECTION PUMP DISCHARGE ISOLATION TO RCS HOT LEGS O
O
) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS 3 NUMBER REV UNIT TITLE
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STP V-3L2 5 1&2
- EXERCISING VALVES 8821A & 8821B SAFETY INJECTION PUMP DISCHARGE TO RCS COLD LEGS 3 STP V-3L3 7 1&2
- EXERCISING VALVES 8807A & 88078 SAFETY INJECTION CHARGING PUMP SUCTION CROSSTIE STP V-3L4 6 1&2
- EXERCISING VALVES 8808A, B, C, AND D ACCUMULATOR DISCHARGE ISOLATION VALVES STP V-3L5 9 1&2
- EXERCISING VALVES 8809A AND 8809B, RHR TO COLD RCS LEGS STP V-3L6 7 1&2
- EXERCISING VALVE 8835, SAFETY INJECTION COLD LEGS ISOLATION 3 STP V-3L10 6 1&2 *EXERCIS]NG VALVES 8923A AND 89238, SAFETY INJECTION PUMP SUCTION VALVES
$TF V-3L13 8 1&2
- EXERCISING VALVE 8976, RWST TO SAFETY INJECTION PUMP STP V-3Ll4 7 1&2
- EXERCISING VALVE 8980, RWST TO RHR PUMP SUCTION STP V-3L15 5 1&2
- EXERCISING SAFETY INJECTION PUMP RECIRCULATION TO RWST VALVES 8974A AND 8974B 3 STP V-3L16 1 1&2
- EXERCISING RHR DUTLET TO SI AND CHARGING PP VALVES 8804A AND 8804B STP V-3L17 2 1&2
- EXERCISING CONTAINMENT RECIRC SUCTION VALVES 8982A AND 8982B STP V-3M1 5 1&2 *EXCERCISING VALVES PHR-FCV-641A AND RHR-FCV-641B, RHR PUMP RECIRCULATION VALVES 3 STP V-3M2 10 162
- EXERCISING VALVES RHR-HCV-637 AND RHR-HCV-638 RHR HEAT EXCHANGER OUTIET VALVES STP V-3M4 7 1&2
- EXERCISING RHR PUMP SUCTION VALVES 8700A AND 8700B STP V-3M5 6 1&2
- EXERCISING VALVES RHR-8701 AND RHR-8702, REACTOR COOLANT LOOP 4 TO RHR PUMP SUCTION STP V-3M6 8 1&2
- EXERCISING VALVE PHR-8703, RHR TO RCS HOT LEGS 1 AND 2 3 STP V-3M7 7 1&2
- EXERCISING VALVES RHR-6716A AND RHR-8716B, RHR TRAIN CROSSTIE VALVES STP V-301 3 1&2
- EXERCISING VALVES DEG-121. DEG-135 AND DEG-149 DIESEL TURBO AIR COMPRESSOR DISCHARGE CHECK VALVES STP V-302 4 1&2
- EXERCISING VALVES DEG-214, 225, 236, 247, 258 AND 269, DIESEL STARTING AIR COMPRESSOR DISCHARGE CHECK 3 STP V-303 5 1&2 *EXERCISlNG VALVES LCV-85 THRU 90, DIESEL-fuel OIL DAY TANK LEVEL CONTROL STP V-3P1 11 1&2 *** EXERCISING MAIN FEEDWATER REGULATING VALVES AND BYPASS VALVES STP V-3P2 5 1&2
- EXERCISING MAlh FEEDWATER ISOLATION VALVES FCV-438, 439, 440 AND 441 3 STP V-3P3 10 1
- EXERCISING MAlh FEEDWATER CHECK VALVES STP V-3P3 2 2
- EXERCISE MAIN FEEDWATER CHECK VALVES STP V-3P4 7 1&2 ***EXERCISlNG AFW VALVES FOR ALTERNATE AUXILI ARY FEEDWATER SUPPLIES STP V-3P5 7 1&2
- EXERCISING AND TIMING OF VALVES LCV-106, 107, 108, AND 109 AUXILIARY FEEDWATER PUMP DISCHARGE D
i 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES O TABLE OF CONTENTS NUMBER REV UNil TITLE
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STP V-3P6 8 IL2
- EXERCISING VALVES LCV-110, 111, 115 AND 113 AUXILIARY O FEEDWATER PUMP DISCHARGE STP V-301 10 IL2 = EXERCISING FIRE SUPPRESSION SYSTEM SECTIONALIZING.
ISOLATION, AND SUPPLY VALVESS , STP V-3P1 7 IL2 *EXERCIS!NG 10% ATMOSPHERIC DUMP VALVES PCV-19, 20, 21, 22 STP V-392 6 IL2
- EXERCISING MAIN STEAM ISOLATION VALVES STP V-3R3 2 IL2
- EXERCISING STEAM GENERATOR BLOWDOWN INSIDE CONTAINMENT o STP V-3R4 1 IL2 ISOLATION VALVES FCV-760, FCV-761 FCV-762, FCV-763
- EXERCISING MAIN STEAM ISOLATION BYPASS VALVES FCV-22, FCV-23, FCV-24 AND FCV-25 STP V-3R5 7 IL2
- EXERCISING STEAM SUPPLY TO AUXILIARY FEEDWATER PUMP TURBINE STOP VALVE,FCV-95 STP V-3R6 2 IL2
- EXERCISING STEAM SUPPLY TO AUXILIARY FEEDWATER PUMP 1 3 TURBINE ISOLATION VALVES, FCV-37 AND FCV-38 STP V-351 6 IL2
- EXERCISING PHASE A CONTAINMENT ISOLATION VALVES STP V-352 5 1&2
- EXERCISING PHASE A CONTAIN'iENT ISOLATION VALVES (STEAM '
GENERATOR BLOWDOWN) STP V-353 8 IL2 *EXERCISlNG PHASE A CONTAINMENT ISOLATION VALVES (RCDT - SUMP ISOLATION) g STP V-3S4 9 IL2
- EXERCISING PHASE A CONTAINMENT ISOLATION VALVES (GROUP K613)
STP V-355 3 IL2 *EXERC] SING PHASE A CONTAINMENT ISOLATION VALVES (ECCS TEST) l STP V-356 0 IL2
- EXERCISING PHASE A CONTAINMENT ISOLATION VALVE FCV-361 STP V-357 4 1 *EXERC151NG PHASE A CONTAINMENT ISOLATION VALVE RCS-8029 l
STP V-357 1 2 *EXERC:3ING PHASE A CONTAINMENT ISOLAT10N VALVE RCS-2-8029 9 STP V-3Se 1 IL2 *EXERC] SING PHASE A CONTAINMENT ISOLATION VALVES STP V-3Tl 6 IL2 *EXERCISlNG CONTAINMENT ATMDSPHERE SAMPLING VALVES l STP V-313 11 IL2 *EXERCISlNG OF CONTAINMENT H2 SAMPLE VALVES AND CONTAINMENT H2 EXTERNAL RECOMBINER VALVES STP V-3T4 7 IL2 *EXERCISlNG OF CONTAINMENT ATMOSPHERE SAMPLE (POST LOCA) l VALVES !g STP V-3T5 5 102
- EXERCISING REACTOR CAVITY SUMP SAMPLE SUPPLY CONTAINMENT ISOLATION VALVES FCV-696 AND FCV-697 l STP V-3T6 3 1&2
- EXERCISING CONTAINMENT VENTILATION ISOLATION VALVES l FCV-660, FCV-661, FCV-662, FCV-663, FCV-664. RCV-11 &
l RCV-12 l STP V-3UI 2 IL2
- EXERCISE AND LEAK CHECK OF HYDRAZINE MIXING CHECK VALVE MU-1555 l3 STP V-302 5 1
- EXERCISING VALVES 1-965, 0-968,1-971 AND 1-1565 MAKE-UP WATER CHECKS STP V-302 7 2
- EXERCISING VALVES 2-965, 0-970, 2-971 AND 2-1565, MAKE-UP WATER CHECKS i STP V-4A 13 1
- FUNCTIONAL TEST OF RHR CHECK VALVES 1 g STP V-4A 4 2 *fuhCT10NAL TEST Of RHR CHECK VALVES O l
O 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 5URVEILLANCE TEST PROCEDURES TABLE OF CONTENTS c) NUMBER REV UNIT TITLE
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c STP V-4B 9 1
- FUNCTIONAL TEST OF THE ECCS CHECK VALVES AT COLD SHUTDOWN
~ C) STP V-4B 3 2
- FUNCTIONAL TEST OF THE ECCS CHECK VALVES AT COLD SHUTDOWN STP V-sal 1 1&2
- EMERGENCY CORE COOLING SYSTEM CHECK VALVE LEAK TEST, POST-REFUELING / POST-MAINTENANCE VALVES 8819 A-D AND 8956 A-D STP V-5A2 1 1&2
- EMERGENCY CORE COOLING SYSTEM CHECK VALVE LEAK TEST, POST-REFUELING / POST-MAINTENANCE VALVES 8948 A-D AND 8818 A-D c) ,
STP V-5C 7 1&2
- EMERGENCY COF.i COOLING SYSTEM HOT LEG CHECK VALVE LEAK TEST STP V-C 9 1
- VERIFICATION uf PROPER ALIGNMENT OF PENETRATION TLST CONNECTION VAtVES, SPARE INSTRUMENT LINES AND EQUIPMENT iATCH FOR CONTAINMENT INTEGRITY STP V-6 7 2 VERiflCATION OF PROPER ALIGNMENT OF PENETRATION TEST -
CONNECTION VALVES, SPARE INSTRUMENT LINES AND EQUIPMENT C) HATCH FOR CONTAlHMENT INTEGRITY STP V-75 15 1&2
- TEST OF ENGINEERED SAFEGUARDS, VALVE INTERLOCKS AND RHR i PUMP TRIP FROM RWST LEVEL CHANNELS STP V-7C 8 1 *** LEAK TEST OF RHR SUCTION VALVES 8701,AND 8702 STP V-7C 5 2 *** LEAK TEST OF RHR SUCTION VALVES 8701 AND 8702 STP V-70 4 1&2 *** LEAK TEST OF RHR PUMP DISCHARGE TO HOT LEG ISOLATION c) VALVE, RHR-8703 ;
STP V-7E 5 1&2 *** LEAK TEST OF S1 PUMP DISCHARGE TO HOT LEG ISOLATION VALVES 8802A AND 8802B STP V-8 6 1&2 ** MAIN STEAM ISQLATION VALVE, MSIV BYPASS, AND STM GEN D' E dDOWN VALVE TIME RESPONSE DETERMINATION , STP V-9 9 1&2
- P: n { LING CVI SYSTEM OPERABILITY DEMONSTARTION
() STP V-Il 7 1&2 *:r' AINMENT ISOLATION PHASE B VALVES FCV-355, FCV-356, i;v-357, FCV-363, FCV-749, AND FCV-750 STP V-13A 0 1 ***CCW FLOW BALANCING STP V-13A 0 2 *CCW FLOW BALANCING STP V-14 6 1&2 ***ECCS THROTTLE VALVE POSITION VERIFICATION S1P V-15 8 1 ***ECCS FLOW BALANCE TEST C) Slo v-15 3 2 ***ECCS FLOW BALANCE TtST STP V-16 4 1&2
- CONTAINMENT PURGE VALVE ACTUATION ON A CONTAINMENT VENTILATION ISOLATION TEST SIGNAL-STP V-18 6 1&?
- CHECK VALVE INSPECTION STP V-19 4 1&2
- FUNCTIONAL TEST OF RHR ALARM - VALVES 8701/8702 NOT FULLY CLOSED IN CONJUNCTION WITH HIGH RCS PRESSURE C) STP V-21 1 152 ** LEAK TEST OF SPRrt ADDITIVE TANK OUTLET LINE CHECK VALVES 1 8998A AND 89988 STP V-23 0 1&2
- HOT STROKE TEST OF PCV-19, 20, 21 AND 22 STP V-24 0 1&2
- LEAK TEST OF SAFETY INJECTION PUMP RECIRCULATION VALVES 8974A AND 8974B STP V-600 11 1 *** GENERAL CONTAINMENT ISOLATION VAtVE LEAK TESTS.
() C)
i l 22 JUL 93 DIABLO CANYON POWER PLANT l UNITS 1 AND 2 l l VOLUME 6 SURVEILLANCE TEST PROCEDURES . I TABLE Of CONTENTS ) NUMBER REV UNIT TITLE
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STP V-600 6 2 *** GENERAL CONTAINMENT ISOLATION VALVE LEAK TESTS : STP V-619 8 1 *** PENETRATION 19 CONTAINMENT ISOLATION VALVE LEAK TESTING ) STP V-619 3 2 *** PENETRATION 19 CONTAINMENT ISOLATION VALVE LEAK TESilNG ' STP V-620 1 1 *** PENETRATION 20 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-620 1 2 *** PENETRATION 20 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-621 2 1 *** PENETRATION 21 CONTAINMEN1 JSOLATION VALVE LEAK TESTING STP V-621 2 2 *** PENETRATION 21 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-623 7 1 *** PENETRATION 22 AND 23 CONTAINMENT ISOLATION VALVE LEAK ) TEST STP V-623 1 2 *** PENETRATION 22 AND 23 CONTAINMENT ISOLATION VALVE LEAK TEST STP V-630 12 1 *** PENETRATION 30 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-630 8 2 *** PENETRATION 30 CONTAINMENT ISOLATION VALVE LEAK TESTING ' STP /-631 4 1 *** PENETRATION 31 CONTAINMENT ISOLATION VALVE LEAK TESTING ) 4 2 *** PENETRATION 31 CONTAINMENT ISOLATION VALVE LEAK TESTING r STP V-631 STP V-635 13 1 *** PENETRATION 35 CONTAINMENT ISOLATION VALVE LEAK TESTING , STP V-635 10 2 *" PENETRATION 35 CONTAINMENT ISOLATION VALVE LEAK TESTING : STP V-641 9 1 *** PENETRATION 41, 42, 43 AND 44 CONTAINMENT ISOLATION VALVE LEAK TESTING ; STP V-641 5 2 *** PENETRATIONS 41, 42, 43, AND 44 CONTAINMENT ISOLATION ' ) VALVE LEAK TESTING STP V-645 13 1 *** PENETRATION 45 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-645 8 2 *** PENETRATION 45 CONTAINMENT lSOLATION VALVE LEAK TESTING STP V-646 10 1 "* PENETRATION 46 AND 47 CONTAINMENT ISOLATION VALVE LEAK TESTING 5TP V-646 5 *" PENETRATION 46 AND 47 CONTAINMENT ISOLATION VALVE LEAK ) TESTING STP V-649 10 1 "* PENETRATION 49 CONTAINMENT IRLATION VALVE LEAK TESTING r STP V-649 4 2 *** PENETRATION 49 CONTAINMEdT ISOLATION VALVE LEAK TESTING STP V-650 9 1 *** PENETRATION 50 CONTAINMENT ISOLATION VALVE LEAK TESTING , STP V-650 4 2 *** PENETRATION 50 CONTAINMENT ISOLATION VALVE LEAK TESTING ST P V-651 A 3 1 *** PENETRATION 51A CONTAINMENT ISOLATION VALVE LEAK TESTING , ) STP V-651A 3 2 *" PENETRATION 51A CONTAINMENT ISOLATION VALVE LEAK TESTING-STP V-651B 8 1 *** PENETRATION 51B CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-651B 6 2 *** PENETRATION SIB CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-651C 3 1 *** PENETRATION 51C CONTAINMENT ISOLATION VALVE LEAK TESTING ' STP V-651C 2 2 '"PEhETRATION SIC CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-651D 2 1 *** PENETRATION 51D CONTAINMENT ISOLATION VALVE LEAK-TESTING- ). STP V-651D 2 2 "*PENEfRAT10N 51D CONT AINMEhI ISOLATION VALVE LEAK TESTING STP V-652A 2 1 "* PENETRATION 52A CONTAINMENT ISOLATION VALVE LEAK TESTING. STP V-652A 4 2 *** PENETRATION 52A CONTAINMENT ISOLATION VALVE LEAK TESTING 2 1 *** PENETRATION 52B CONTAINMENT ISOLATION VALVE LEAK TESTING ; S'P V-652B STP V-652B 2 2 *" PENETRATION 52B CONTAINMENT ISOLATION VALVE LEAK TESTING ' STP V-65?D 2 1 *" PENETRATION 52D CONTAINMENT ISOLATION VALVE LEAK TESTING ) )
$ 22 JUL 93 DIABLO CANY0H POWER PLANT , UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS 3 HUMBER REV. UNIT TITLE
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STP V-652D 2 2 ***PENETRAT10N 52D CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-654 8 1 *** PENETRATION 54 CONTAINMENT ISOLATION VALVE LEAK TESTING 9 STP V-654 4 2 *** PENETRATION 54 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-656 10 1 ***PENETRAITON 56 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-656 2 2 ***PENETRAT10N 56 CONTAINMENT ISOLATION VALVE LEAK TESTING ! STP V-657 7 1 *** PENETRATION 57 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-657 6 2 *** PENETRATION 57 CONTAINMENT ISOLATION VALVE LEAK TESTING 2 1 *** PENETRATION 59A CONTAINMENT ISOLATION VALVE LEAK TESTING g STP V-659A 2 2 *** PENETRATION 59A CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-659A STP V-659B 2 1 *** PENETRATION 59B CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-659B 2 2 *** PENETRATION 598 CONTAINMENT ISOLATION VALVE LEAK TESTING , STP V-659E 2 1 ***PENTRATION 59C CONTAINMENT ISOLATION VALVE LEAK TESTING ' STP V-659C 2 2 *** PENETRATION 59C CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-661 9 1 *** PENETRATION 61 CONTAINMENT ISOLATION VALVE LEAK TESTING g 0 2 *** PENETRATION 61 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-661 STP V-662 6 1 *** PENETRATION 62 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-662 0 2 ***PENETRAT10N 62 CONTAINMENT ISOLATION VALVE LEAK TESTING ' STP V-663 8 1 *** PENETRATION 63 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-663 1 2 *** PENETRATION 63 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-668 9 *** PENETRATIONS 68 AND 69 CONTAINMENT ISOLATION VALVE LEAK g 1 TESTING STP V-668 4 2 *** PENETRATIONS 68 AND 69 CONTAINMENT ISOLATION VALVE LEAK TESTING i STP V-670 8 1 *** PENETRATION 70 CONTAINMENT ISOLATION VALVE LEAK TESTING SlP V-670 4 2 *** PENETRATION 70 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-671 11 1 *** PENETRATION 71 CONTAINMENT ISOLATION VALVE LEAK TESTING g STP V-671 6 2 *** PENETRATION 71 CONTAINMENT ISOLATION VALVE LEAK TESTING j STP V-676A 2 1 ***PENTRATION 76A CONTAINMENT ISOLATION VALVE LEAK TESTING SlP V-676A 2 2 *** PENETRATION 76A CONTAINMENT ISOLATION VALVE LEAK TESTING , STP V-676B 3 1 ***PENTRATION 76B CONTAINMENT ISOLATION VALVE LEAK TESTING I STP V-676B 3 2 *** PENETRATION 76B CONTAINMENT' ISOLATION VALVE LEAK TESTING
]
- j. STP V-678 7 1 *** PENETRATIONS 52 AND 78 CONTAINMENT ISOLATION VALVE LEAK TESTING J
STP V-678 6 2 *** PENETRATIONS 52 AND 78 CONTAINMENT ISOLATION VALVE LEAK TEST 1NG STP V-679 9 1 *** PENETRATION 79 CONTAINMENT ISOLATION VALVE LEAK TESTING SlP V-679 6 2 *** PENETRATION 79 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-681 2 1 *** PENETRATION 81 CONTAINMENT ISOLATION VALVE LEAK TESTING ? STP V-681 3 2 *** PENETRATION 81 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-682A 2 1 *** PENETRATION 82A CONTAINMENT ISOLATION VALVE LEAK TESTING , STP V-682A 3 2 *** PENETRATION 82A CONTAINMENT ISOLATION VALVE LEAK TEST ! STP V-682B 2 1 *** PENETRATIONS 82B AND B2C CONTAINMENT ISOLATION VALVE LEAK TESTING .
i O 22 JUL 93 DIABLO CANYOH POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES S TABLE OF CONTENTS NUMBLR REV UNIT TITLE
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STP V-6B2B 3 2 *** PENETRATIONS 82E AND 82C CONTAINMENT ISOLATION VALVE e LEAK TESTING STP V-68?D 2 1 *** PENETRATIONS 82D AND 83 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-682D 2 2 *** PENETRATIONS 82D AND 83A CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-683 12 1 *** PENETRATION 83 CONTAINMENT ISOLATION VALVE LEAK TESTING STP V-683 5 2 *** PENETRATION 83B CONTAINMENT ISOLATION VALVE LEAK TESTING () 5 ILP SlP P-1A *(OTSC 07/21/93 - SPENCER) PERFORMANCE TEST OF SAFETY INJECTION PUMPS
;iP F-Jb 16 ILP *** ROUTINE SURVEILLANCE TEST OF SAFETY INJECTION PUMPS S1P P-24 2 162 ** PERFORMANCE OF CENTRIFUGAL CHARGING PUMPS SIP P-2b 15 IL2 *** ROUTINE SURVEILLANCE TEST OF CENTRIFUGAL CHARGING PUMPS STP P-3A 9 IL2 *** PERFORMANCE TEST OF RESIDUAL HEAT REMOVAL PUMPS
() IL2 *** ROUTINE SURVEILLANCE TEST OF RHR PUMPS SlP P-33 22 STP P-4B 16 IL2 *** ROUT!NE SURVEILLANCE TEST OF CONTAINMENT SPRAY PUMPS STP P-SA 8 IL2 *** PERFORMANCE TEST OF MOTOR-DPlVEN AUXILLARY FEED PUMPS STP P-5B ?U IL2 ***ROUTINF SJRVEILLANCE TEST OF MOTOR-DRIVEN AUXILIARY FEEPWi ~ ' r PUMPS STP P-6A 4 IL2 **(OTf: 37/21/93 -SPENCER) PERFORMANCE TEST OF STEAM-DRIVEN () AUXlLIAR) FEED PUMP TP P-6B 27 1&2 * ** (OT SC 07/21/93 - NELSON) ROUTINE SURVEILLANCE TEST OF TURBINE-DRIVEN AUXlLIARY FEEDWATER PUMP STP P-6C E 1 *0VERSPEED TRIP OF STE AM ORIVEN AUX 1LI ARY FEED PUMP STP P-6C 4 2 *0VERSPEED TRIP OF STEAM DRIVEN AUXILIARY FEED PUMP STP P-70 7 IL2 ***PERf 0RMANCE TEST OF AUX 1LI ARY SALTWATER PUMP
- ) STP P-7B 24 IL2 *** ROUTINE SURVEILLANCE TEST OF AUXlLIARY SALTWATER PUMPS STP P-EA 0 ILP
- PERFORMANCE TEST OF COMPONENT COOLING WATER PUMPS STP P-8B 24 IL2 *** ROUTINE SURVElLLANCE TEST OF COMPONENT COOLING WATER PUMP STP P-llB 4 IL2 ** SPENT FUEL POOL PUMP l-1(2-1) AND HEAT EXCHANGER PERFORMANCE VERIFICATION
() STP P-llc 5 1&2 *** SPENT FUEL POOL PUMP l-2 (2-2) PERFORMANCE VERIFICATION STP P-1281 4 IL2 *** ROUTINE SURVElLLANCE TEST OF DIESEL FUEL OIL TRANSFER PUMP 0-1 STP P-12B2 4 IL2 *** ROUTINE SURVEILLANCE TEST OF DIESEL FUEL DIL TRANSFER PUMP 0-2 STP P-1253 4 IL2
- ROUTINE SURVEltLANCE OF PORTABLE FUEL OIL TRANSFER PUMP
) STP P-13A 10 1L2 ***(OTSC 07/21/93 - SPENCER) FIRE PUMPS PERFORMANCE TEST SlP P-13B 17 IL2 *** FIRE PUMPS 0-1 AND 0-2 ROUTINE SURVEILLANCE STP P-13C 3 1&2 ** FIRE PUMPS 0-3 AND 0-4 PERFORMANCE TEST STP P-14B 38 1 *** ROUTINE SURVEILLANCE TEST OF BORIC ACID TRANSFER PUMPS STP P-14B 15 2 *** ROUTINE SURVEILLANCE TEST OF BORIC ACID TRANSFER PUMPS STP P-15A 0 1&2
- PERFORMANCE TEST OF MAKE-UP WATER TRANSFER PUMPS g STP P-15B 12 IL2 *** ROUTINE SURVEILLANCE TEST OF MAKEUP WATER TRANSFER PUMPS 4
h 22 JUL 93 DIABLO CANYOH POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS NUMBER REV UNIT TITLE ;
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STP P-17B 20 IL2 *** SURVEILLANCE TEST OF RECIPROCATING CHARGING PUMP STP P-23A 2 1&2 *** ACCELERATION TIMING OF SAFETY RELATED PUMPS ACTUATED BY I
) SSPS TRAIN A STP P-23B 3 1&2 ** ACCELERATION TIMING OF SAFETY RELATED PUMPS ACTUATED BY SSPS TRAIN B STP P-23C 7 1&2 *** ACCELERATION TIMING OF TURBINE DRIVEN AUXILIARY FEEDWATER PUMP STP P-24 6 1&2 ***(OTSC 7/21/93-NELSON) TESTING 0F THE PORTABLE LONG TERM ) COOLING PUMPS STP P-1A 10 1&2
- EXERCISING FULL-LENGTH CONTROL RODS STP R-1B 13 1&2 ** ROD DROP MEASUREMENT STP R-lC 4 1&2
- DIGITAL ROD POSITION INDICATOR FUNCTIONAL TEST STP R-281 4 1&2 *PPC OPERATOR HEAT BALANCE
_ STP R-2B2 2 1&2
- MANUAL OPERATOR HEAT BALANCE ,
3 STP R-2C 8 1&2
- DETERMINATION OF REACTOR POWER LEVEL BELOW 5%
STP R-3A 8 1&2 *USE OF FLUX MAPPING EQUIPMENT STP R-3B 7 1&2
- FLUX MAP DATA TRANSMISSION STP R-3C 11 1&2
- FLUX MAP DATA REDUCTION STP R-3D 12 1&2
- ROUTINE MONTHLY FLUX MAP STP R-3E 4 1&2 ** REPLACEMENT Of AN INCORE DETECTOR
') STP R-4 6 1&2
- EVALUATION OF CERE REACTIVITY STP R-b 5 1&2 *BURNUP TRACKING '
STP R-6 5 1&2 ** LOW POWER RELOAD PHYSICS TEST STP R-7A 8 1&2
- DETERMINATION OF MODERATOR TEMPERATURE COEFFICIENT AT HZP, BOL S1P R-7B 11 1&?
- DETERMINATION OF MODERATOR TEMPERATURE COEFFICIENT AT POWER
) STP R-7C 0 1&2 ** ROD WITHDRAWAL LIMITS CALCULATION ,
STP R-BA 6 1&2
- REACTOR C00LAN1 SYSTEM OPERATIONAL PRESSURE LEAK TEST STP R-8C 1 1&2
- CONTAINMENT WALKDOWN FOR EVIDENCE OF BORIC ACID LEAKAGE STP R-9 1 1&2
- DETERMINATION OF ROD POSITION USING THE MOVEABLE INCORE DETECTOR SYSTEM (MIDS)
STP R-10 6 1&2
- REACTOR COOLANT SYSTEM LEAKAGE EVALUATION
) STP R-10C 9 1 ** REACTOR COOLANT SYSTEM WATER INVENTORY BALANCE STP R-10C 1 2 ** REACTOR COOLANT SYSTEM WATER INVENTORY BALANCE STP R-100 0 1&2 *RCP #3 SEAL LEAKAGE EVALUATION STP R-ll 5 1&2
- QUADRANT POWER TILT RATIO DETERMINATION BY MOVABLE INCORE DETECTORS STP R-l? 4 1&2
- BORON CONCENTRATION DURING REFUELING
) STP R-13 8 1&2 ** NUCLEAR POWER RANGE INCORE/EXCORE POWER RANGE MULTIPLE-POINT CALIBRATION STP R-13A 1 1&2 **BEGINNING OF' CYCLE INCORE-EXCORE CAllBRATION DATA STP R-13B 0 1&2
- NUCLEAR POWER RANGE INCORE/EXCORE POWER RANGE MULTIPLE-POINT CALIBRATION STP R-14 2 1&2 *FER10DIC ADJUSTMENT OF FEEDWATER FLOW N0ZZLE COEFFICIENTS
) '
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0 DIABLO CANYON POWER PLANT 22 JUL 93 UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS () NUMBER REV UNIT TITLE
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STP P-15 0 1&2 ** SPENT FUEL RACK ABSORBER SURVEILLANCE () STP R-17 9 1&2
- ESTIMATED CRITICAL POSITION (ECP)
STP R-19 11 1&2
- SHUTDOWN MARGIN DETERMINATION STP R-20 13 1&2
- BORIC ACID INVENTORY STP R-21 1 1&2
- REACTOR INTERNALS VIBRATION MONITORING STP R-22 2 IL2
- THIMBLE TUBE INSPECTION PROGRAM STP R-23 8 1&2 ** DETERMINATION & UPDATE OF TARGET AX1AL FLUX DIFFERENCE
() STP R-24 4 1&2 ** REMOVAL AND LOADING OF IRRADIATED REACTOR VESSEL MATERIAL SURVEILLANCE CAPSULES STP R-25 12 1&2
- CALCULATION OF QUADRANT POWER TILT RATIO STP R-26 11 1&2 *RCS PRIMARY COOLANT FLOW MEASUREMENTS STP R-27 8 1&2 *lNCORE THERMOCOUPLES AND RCS RTD CROSS CAllBRATION STP R-28 6 1&2
- DETERMINATION OF HIGH FLUX AT SHUTDOWN ALARM SETPolNT STP R-30 7 1&2 ** RELOAD CYCLE INITI AL CRITICALITY
() 3 1&2 ** ROD WORTH MEASUREMENTS USING ROD SWAP METHOD STP R-31 STP R-32 2 1&2 ** POD WORTH MEASUREMENTS USING BORAT10N OR DILUTION STP R-40 10 1&2
- RELOAD POWER ASCENSION TESTING STP R-41 3 1&2 *RE ACTOR COOL ANT SYSTEM TEMPERATURE INSTRUMENTATION DATA STP R-42 3 IL2
- STEAM AND FEEDWATER FLOW CAllBRATION DATA STP R-43 2 1&2
- REACTOR COOLANT SYSTEM REFERENCE TEMPERATURE DATA
() 7 162 *** AIR TEST OF CONTAINMENT SPRAY N022LES STP M-1 STP M-2 7 1 *** CONTAINMENT HYDROGEN PURGE VENTILATION SYSiLM - DOP AND HAllDE PENETRATION TESTS STP M-2 0 2 *** CONTAINMENT HYDROGEN PURGE VENTILATION SYSTEM - 00P AND HALIL PENETRATION TESTS STP M-3A 8 IL2 ** AUXILIARY BUILDING VENTILATION SYSTEM - DOP & HALIDE c) PENETRATION TEST STP M-3B 7 1&2 *** AUXILIARY VENTILATION SYSTEM - DAMPERS M2A AND M2B LEAK RATE TEST STP M-3C 8 IL2 *** AUXILIARY BUILDING VENTILATION SYSTEM CHARC0AL PREHEATER TEST STP M-4 14 IL2 *POUTINE SURVEILLANCE TEST OF THE AUXILIARY BUILDING i() SAFEGUARDS AIR FILTRATION SYSTEM STP M-5 12 IL2
- ROUTINE SURVEILLANCE TEST OF THE FUEL HANDLING BUILDING VENTILATION SYSTEM STP M-6A 15 IL2
- ROUTINE SURVEILLANCE TEST OF CONTROL ROOM VENTILATION SYSTEM STP M-6B 3 IL2
- CONTROL ROOM VENilLATION SYSTEM CHARCOAL PREHEATER TEST C) 13 162 *** CONTAINMENT INTEGRATED LEAKAGE RATE TEST (ILRT), TYPE A STP M-7 S~lP M-7E O 2 *** CONTAINMENT PENETRATION VALVE LINEUP FOR THE INTEGRATED' LEAKAGE RATE TEST (ILRT)
STP M-8Al 1 1 ***0VERALL LEAK RATE TESTING OF THE PERSONNEL AIR LOCK STP M-BAl 0 2 ***0VERALL LEAK RATE TESTING OF THE PERSONNEL AIR LOCK SIP M-8A2 0 1 ***0VERALL LEAK RATE TESTING OF THE EMERGENCY AIP LOCK () O
) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES r TABLE OF CONTENTS 3 HUMBER REV UNIT TITLE
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STP M-8A2 0 2 ***0VERALL LEAK RATE TESTING 0F THE EMERGENCY AIR LOCK J STP M-8B 8 1&2 *LEAKRATE TESTING OF ELECTRICAL PENETRATIONS , STP M-8C1 1 1&2
- LEAK RATE TESTING OF THE EQUIPMENT HATCH SEAL l '
STP M-8C2 0 1
- LEAK RATE TESTING OF THE FUEL TRANSFER TUBE SEAL STP M-8C2 0 2
- LEAK RATE TESTING OF THE FUEL TRANSFER TUBE SEAL STP M-8C3 0 1
- LEAK RATE TESTING PENETRATION 58 MINI EQUIPMENT HATCH SEAL STP M-8C3 0 2
- LEAK RATE TESTING PENETRATION 58 MINI EQUIPMENT HATCH SEAL .
STP M-8C4 0 1
- LEAK RATE TESTING OF PENETRATION 60 MINI EQUIPMENT HATCH 3 SEAL STP M-8C4 0 2
- LEAK RATE TESTING OF PENETRATION 60 MINI EQUIPMENT HATCH SEAL STP M-8D 6 1
- COMBINED TYPE B AND C LEAKAGE EVALUATION '
STP M-8D 1 2
- COMBINED TYPE 8 AND C LEAKAGE EVALUATION STP M-bel 0 1
- PERSONNEL AIR LOCK 000R INTERLOCK VERIFICATION 3 0 2
- PERSONNEL AIRLOCK DOOR INTERLOCK VERIFICATION STP M-8El STP M-SE2 0 1
- EMERGENCY AIR LOCK DOOR INTERLOCK VERiflCATION STP M-8E2 0 2
- EMERGENCY AIRLOCK DOOR INTERLOCK VERIFICATION i
STP M-8F1 1 1 ***ALRM LEAK RATE TESTING OF PERSONNEL AIR LOCK SEALS STP M-8F1 1 2 ***ALRM LEAK RATE TESTING OF PERSONNEL AIR LOCK SEALS , STP M-8F2 0 1 ***PLTM LEAK RATE TESTING OF PERSONNEL AIR LOCK SEALS 3 0 2 ***PLTM LEAK RATE TESTING OF PERSONNEL AIR LOCR SEALS STP M-8F2 STP M-8G 5 1 ** LEAK RATE TESTING OF EMERGENCY AIR LOCK SEALS STP M-8G 1 2 *** LEAK RATE TESTING OF EMERGENCY AIR LOCK SEALS STP M-8H 0 2
- LEAK RATE TESTING SPARE INSTRUMENT TEST LINES l
STP M-9A 25 1&2 *** DIESEL ENGINE GENERATOR ROUTINE SURVEILLANCE TEST STP M-9B 11 IL2 *0VERSPEED TRIP TEST OF DIESEL GENERATORS 3 12 1&2
- DIESEL GENERATOR LOAD REJECTION TEST STP M-90 i STP M-9E 5 IL2
- DIESEL GENERATOR TRIP CIRCulTRY BYPASS VERIFICATION STP M-93 16 IL2
- DIESEL GENERATOR 24-HOUR LOAD TEST '
STP M-9h 6 IL2
- DIESEL GENERATOR INTERDEPENDENCE TEST S1P M-91 6 IL2
- DIESEL GENERATOR TESTING FREQUENCY DETERMINATION +
STP M-9L 13 1&2
- DIESEL GENERATOR SHUTDOWN LOCK 0UT RELAY TEST 3 STP M-9M 5 1
- VERIFICATION OF AUTO-CONNECTED LOADS LES$ THAN 2750 KW STP M-9M 3 2
- VERIFICATION OF AUTO-CONNECTED LOADS LESS THAN 2750 KW
- DIESEL GENERATOR OPERABILITY VERIFICATION ';
STP M-9x 2 IL2 STP M-9Y 1 It2 *** DETERMINATION OF DIESEL LOADING TIME ON A REAL DEMAND SITUATION STP M-10A 8 IL2 " DIESEL FUEL 01L STORAGE TANK INVENTORY 3 9 IL2
- DIESEL FUEL Olt ANALYSES STP M-10B STP M-11A 8 IL2 ** MEASUREMENT OF STATION BATTERY PILOT CELL VOLTAGE AND '
SPECIFIC GRAVITY STP M-llB 11 1&2
- MEASUREMENT OF STATION BATTERY VOLTAGE AND SPECIFIC GRAVITY ,
STP M-11C 8 1&? ** STATION BATTERY TERMINAL RESISTANCE MEASUREMENT AND i INSPECTION 3 1
C1 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS () NUMBER REV UNIT TITLE
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STP M-llCl2 0 1 ** BATTERY TERMINAL RESISTANCE MEASUREMENT AND INSPECTION FOR STATION BATTERY 12 c) STP M-11C21 0 2 *** BATTERY TERMINAL RESISTANCE MEASUREMENT AND INSPECTION FOR STATION BATTERY 21 STP M-12A11 0 1 *** STATION BATTERY 11 PERFORMANCE TEST STP M-12A12 0 1 *** STATION BATTERY 12 PERFORMANCE TEST STP M-12A13 0 1 *** STATION BATTERY 13 PERFORMANCE TEST STP M-12A21 1 2 *** STATION BATTERY 21 PERFORMANCE TEST () STP M-12A22 0 2 *** STATION BATTERY 22 PERFORMANCE TEST STP M-12A23 0 2 *** STATION BATTERY 23 PERFORMANCE TEST STP M-12B 6 1&2 *** BATTERY CHARGER PERFORMANCE TEST STP M-12Cll 0 1 ** STATION BATTERY 11 SERVICE TEST STP M-12Cl2 0 1 ** STATION BATTERY 12 SERVICE TEST STP M-12Cl3 1 1 *** STATION BATTERY 13 SERVICE TEST c) 0 2 ** STATION BATTERY 21 SERVICE TEST STP M-12C21 STP M-12C22 1 2 *** STATION BATTERY 22 SERVICE TEST STP M-12C23 2 2 *** STATION BATTERY 23 SERVICE TEST STP M-13A 2 1&2 *0FFSITE POWER TO 4KV VITAL BUSSES STP M-13B1 4 1&2 *ENGD SFGDS AUTO TIMERS SETTING VERF LOADS STARTED SSPS i RELAY K608 TRAIN A () STP M-1362 4 1&2 *ENGD SFGDS AUTO TIMERS SETTING VERF LOADS STARTED SSPS RELAY K608 TRAIN B STP M-13B3 4 1&2 *ENGD SFGDS AUTO TIMERS SETTING VERF LOADS STARTED SSPS RELAY K609 TRAIN A STP M-1384 4 1&2 *ENGD SFGNS AUTO TIMERS SETTING VERF LOADS STARTED SSPS RELAY K-609 TRAIN B C) SlP M-13C 5 1&2
- ENGINEERED SAFEGUARDS AUTOMATIC TIMERS SETTING VERIFICATION FOR CONTAINMENT SPRAY PUMPS STP M-13F B 1&2 ***4KV BUS F NON-51 AUTO-TRANSFER TEST STP M-13G 7 1&2 ***4KV BUS G NON-SI AUTO-TRANSFER TEST STP M-13H 5 1&2 ***4KV BUS H NON-SI AUTO-TRANSFER TEST STP M-15 19 1&2 *lNTEGRATED TEST OF ENGINEERED SAFEGUARDS AND DIESEL c) GENERATORS STP M-16 6 1&2
- SAFEGUARDS ACTIVE COMPONENT OPERATION BY SLAVE RELAY OPERATION STP M-16A 19 1&2 *0PERATION OF TRAINS A AND B SLAVE RELAYS K603 (SAFETY INJECTION) AND K605 (PHASE A ISOLATION)
STP M-16C 10 1&2 *0PERATION OF TRAIN A SLAVE RELAY K608 (SAFETY INJECTION) () STP M-16D 11 1&2 *0PERATION OF TRAIN B SLAVE RELAY K608 (SAFETY INJECTION) STP M-16E 5 1&2 *0PERATION OF TRAIN A SLAVE RELAYS K609 (SAFETY-lNJECTION)- K633A-(MOTOR DRIVEN AFW PUMP START) STP M-16F 4 1&2 ***(OTSC 07/10/93-ROLLER) OPERATION OF TRAIN'B SLAVE RELAYS 1 K609 (SAFETY INJECTION) K633 (MOTOR DRIVEN AFW PUM ,0 13- H
l l l 3 22 JUL 93 DIABLO CANYOH POWER PLANT l UNITS 1 AND 2 l VOLUME 6 SURVEILLANCE TEST PROCEDURES l
, TABLE OF CONTENTS J j
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STP M-16G 8 1&2 **0PERATION OF TRAIN A & B SLAVE RELAYS K610 AND K617 (BLOCK NON-PPOTECTION GRADE PUMP STARTS) 3 SIP M-161 9 1&2 *0PERATION OF TRAINS A AND B SLAVE RELAY K606 (PHASE A CONTAINMENT ISOLAT10N) STP M-16] 7 1&2 *0PERATION OF TRAIN B SLAVE RELAY K612 (CONTAINMENT PHASE A ISOLATION) STP M-16K 10 1&2 *0PERATION OF SSPS TRAINS A&B SLAVE RELAY K613 (CONTAINMENT PHASE A ISOLATION)
,3 STP M-lf 8 1&2 *0PERATION OF TRAINS A AND B SLAVE RELAY K614 (PHASE A CONTAINMENT ISOLATION)
SlF M-m 7 1&? *0PERATION OF TRAIN A & B SLAVE RELAYS K622 AND K625 STP M-16N 11 1&2 *0PERATION OF TRAINS A AliD B SLAVE RELAYS K632 AND K634 STP M-160 9 1&2 *0PERATION OF TRAINS A AND B SLAVE PELAY K631 (P-12 STEAM
, DUMP INTERLOCF) s SIP M-16P1 9 }&2
- CONTINUITY TESTING OF TRAIN A/B SLAVE RELAYS K627, K628 AND K635 STP M-16P2 8 1&2
- CONTINUITY TESTING OF SSPS TRAINS A & B SLAVE RELAYS K601,K620, K636 & K621 STP M-16P3 9 1&2
- CONTINUITY TESTING OF SSPS TRAINS A & B SLAVE RELAYS K616 AND K623 (MA]N STEAM ISOLATION) 3 STP M-16P4 7 1 r.2
- CONTINUITY TESTING OF SSPS TRAINS A&B SLAVE RELAYS K611 K607 & K619 STP M-1601 3 1&2
- SLAVE RELAY FUNCTIONAL TESTING OF RCP TRIP CIRCulTS STP M-1602 3 1&2
- SLAVE RELAY FUNCTIONAL TEST OF UNIT TRIP 86G RELAYS STP M-16Q3 2 1&2
- FUNCTIONAL TESTING OF BUSSES H, F AND G AUT0 TRANSFER TO STARTUP POWER 3 STP M-16Q4 6 1&2
- FUNCTIONAL TESTING OF MAIN TURBINE TRIP, MAIN FEEDWATER PUMP TRIP AND MAIN FEE 0 WATER ISOLATION VALVES STP M-16Q5 5 1&2
- FUNCTIONAL TESTING OF RCP SEAL RETURN CONTAINMENT ISOLATION VALVES 8100AND 8112 STP M-1606 3 1&2
- FUNCTIONAL TESTING OF CCW HEADER C ISOL. VALVES FCV-355, FCV-356, FCV-363, FCV-357, FCV-749, AND FCV-750 J STP M-16BA 1 1&2 *0PERATION OF TRAIN A SLAVE RELAY K604 (SAFETY INJECTION)
SlP M-16BB 1 1&2 *0PERATION OF TRAIN B SLAVE RELAY K604 (SAFETY INJECTION) STP M-16HA 1 1&2 ***0PERATION OF SSPS TRAIN A SLAVE RELAYS K644 AND K645 (CONTAINMENT SPRAY PUMP 2) STP M-16HB 1 1&2 ***0PERATION OF SSPS TRAIN B SLAVE RELAYS K644 AND K645 (CONTAINMENT SPRAY PUMP 1) O STP M-17Al 2 1
- FUNCTIONAL TEST OF THE COMMUNICATION SYSTEM STP M-17Al 0 2
- FUNCTIONAL TES1 0F THE COMMUNICATION SYSTEM STP M-17A2 3 1
- FUNCTIONAL TEST Of THE COMMUNICATION SYSTEM IN CONTAINMENT STP M-17A? 0 2
- FUNCTIONAL TEST OF THE COMMUNICATION SYSTEM ]N CONTAINMENT STP M-17B1 3 1 *** FUNCTIONAL TEST OF EMERGENCY DC LIGHTING SYSTEM STP M-17B1 3 2 *** FUNCTIONAL TEST OF EMERGENCY DC LIGHTING SYSTEM 9
3
I D 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 UOLUME 6 SURVEILLANCC TEST PROCEDURES TABLE OF CONTENTS D-NUMBER REV UNIT TITLE
=============== === ==== === =================================
STP M-1782 2 1 *** FUNCTIONAL TEST OF EMERGENCY DC LIGHTING SYSTEM IN CONTAINMENT 3 2 2 *** FUNCTIONAL TEST OF EMERGENCY DC LIGHTING SYSTEM IN STP M-17B2 CONTAINMENT STP M-1783 3 1 *** FUNCTIONAL TEST OF UNINTERRUPTIBLE POWER SUPPLY FOR CONTROL ROOM LIGHTING SYSTEM STP M-1783 3 2 *** FUNCTIONAL TEST OF UNINTERRUPTIBLE POWER SUPPLY FOR CONTROL ROOM LIGHTING SYSTEM [) 2 1 *** BATTERY SERVICE TEST FOR UNINTERRUPTIBLE POWER SUPPLY STP M-17B4 BATTERY PACK FOR CONTROL ROOM LIGHTING SYSTEM i STP M-1784 2 2 *** BATTERY SERVICE TEST FOR UNINTERRUPTIBLE POWER SUPPLY BATTERY PACK FOR CONTROL ROOM LIGHTING SYSTEM STP M-17C1 5 1&2
- FUNCTIONAL TEST OF THE EMERGENCY BATTERY OPERATED LIGHTING.
SYSTEM 3 STP M-17C2A 1 1 *** DISCHARGE TEST OF THE EMERGENCY BATTERY OPERATED LIGHTING SYSTEM STP M-17C2A 1 2 *** DISCHARGE TEST OF THE EMERGENCY BATTERY OPERATING LIGHTING SYSTEM STP M-17C2B 1 1 *** DISCHARGE TEST OF THE EMERGENCY BATTERY OPERATED LIGHTING SYSTEM 3 STP M-17C2B 1 2 *** DISCHARGE TEST OF THE EMERGENCY BATTERY OPERATED LIGHTING SYSTEM STP M-17C3 1 1 ** CHECK OF THE EMERGENCY AC LIGHTING SYSTEM FOR SAFE SHUTDOWN ROUTE ILLUMINATION STP M-17C3 1 2 **CHECV 0F THE EMERGENCY AC LIGHTING SYSTEM FOR SAFE SHUTDOWN ROUTE ILLUMINATION 3 STP M-1B 2 1&2
- FUNCTIONAL TESTING OF TURBINE TRIP FROM REACTOR TRIP (P-4)
CIRCUIT STP M-19A 10 1&2 *HALON STORAGE T ANK WEIGHT & PRESSURE VERIFICATION STP M-19B 12 1&2 **HALON FIRE SUPPRESSION SYSTEM FUNCTIONAL TEST STP M-21A 13 1&2
- MAIN TURBINE / GENERATOR FUNCTIONAL TESTS STP M-21B 10 1&2
- MAIN TURBINE SEMI-ANNUAL OVERSPEED TESTS 3 STP M-21C 17 1&2
- MAIN TURBINE VALVE TESTING SlP M-22A 2 1&2
- FUNCTIONAL TESTING OF TURBINE TRIP FROM AMSAC OUTPUT RELAY ,
ACTUATION ST/ M-??B 2 1&2
- FUNCTIONAL TESTING OF MOTOR-DRIVEN AUXILI ARY FEEDWATER PUMPS START AND BLOWDOWN ISOLATION'FROM AMSAC OUTPUT RELAY ACTUATION ;
3 STP M-22C 2 1&2
- FUNCTIONAL TESTING OF TURBINE DRIVEN AUX FEEDWATER PUMP STARTS FROM AMSAC OUTPUT RELAY ACTUATION STP M-25A 9 1&2
- LEAK RATE. TEST FOR UNDERGROUND FUEL Olt STORAGE TANKS STP M-2SB 0 1&2
- PERFORMANCE TEST FOR TLS-250 MONITOR STP M-26 11 1&2 ***ASW SYSTEM FLOW MONITORING
) l ) 1 i
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES g, TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== ==- ==== ===========================================================
STP M-264 1 IL2 ***FCV-601, ASW UNIT 1 AND 2 CROSS-TIE DIVIDING VALVE, FLOW g TEST STP M-27 13 IL2 *** FUEL HANDLING SYSTEM INTERLOCK VERIFICATION AND FUNCTIONAL STP M-39Al 2 IL2 *** ROUTINE SURVEILLANCE TEST OF DIESEL GENERATOR 1-1 (2-1) ROOM CARBON D10X1DE FIRE SYSTEM OPERATION STP M-39A2 3 1&2 *** ROUTINE SURVEILLANCE TEST OF DIESEL GENERATOR l-2 (2-2) ROOM CARBON DIOXIDE FIRE SYSTEM OPERATION f STP M-39A3 1 IL2 *** ROUTINE SURVEILLANCE TEST OF DIESEL GENERATOR l-3 ROOM CARBON 010x10EFIRE SYSTEM OPERATION STP M-39B 10 102 *** ROUTINE SURVEILLANCE TEST OF CABLE SPREADING ROOM CARBON D10X1DE FIRES) STEM OPERATION STP M-39C 6 102 *** ROUTINE SURVEILLANCE TEST OF TURBINE LUBE Oll ROOM CARBON 010X10E FIRE SYSTEM OPERATION
- 7) STP M-39D 4 1&2 *** ROUTINE SURVEILLANCE TEST OF CARBON DIOXIDE HOSE REELS STP M-39E 1 IL2 *** ROUTINE SURVEILLANCE TEST OF TURBINE GENERATOR BEARING
#10 CARBON 010X1DE FIRE SYSTEM OPERATION STP M-41 10 1 *** FUEL HANDLING BUILDING VENTILATION SYSTEM - DOP AND HALIDE PENETRATION TESTS STP M-41 0 2 *** FUEL HANDLING BUILDING VENTILATION SYSTEM - DOP AND ) HALIDE PENETRATION TESTS STP M-41A 0 IL2
- MEASUREMENT OF FUEL HANDLING BUILDING RELATIVE PRESSURE STP M-42 5 IL2 ** LOAD TEST - MANIPULATOR CRANE & AUX 1LIARY HOIST STP M-43 8 1 ** FUEL HANDLING BUILDING OVERHEAD CRANE INTERLOCK VERIFICATION STP M-43 0 2 ** FUEL HANDLING BUILDING OVERHEAD CRANE INTERLOCK VERIFICATION STP M-45A 4 1 *** CONTAINMENT INSPECTION PRIOR TO ESTABLISHING CONTAINMENT INTEGRITY STP M-45A 3 2 *** CONTAINMENT INSPECTION PRIOR TO ESTABLISHING CONTAINMENT INTEGRITY STP M-45B 2 1
- CONTAINMENT INSPECTION WHEN CONTAINMENT INTEGRITY 15
) ESTABLISHED .
STP M-45B 2 2
- CONTAINMENT INSPECTION WHEN CONTAINMENT INTEGRITY IS ESTABLISHED STP M-45C 1 1 *0UTAGE MANAGEMENT CONTAINMENT INSPECTION STP M-45C 1 2 *0UTAGE MANAGEMENT CONTAINMENT INSPECTION STP M-46 1 IL2 **EXPLOS10 METER SUPVEY OF GENERATOR, EXCITER & HYDROGEN g SYSTEM STP M-4s 8 1 *** GENERATOR Alf TEST STP M-48 0 2 *** GENERATOR AIR TEST STP M-51 15 1
- ROUTINE SURVEILLANCE TEST OF CONTAINMENT FAN COOLER UNITS STP M-51 4 2
- ROUTINE SURVEILLANCE TEST OF CONTAINMENT FAN COOLER UNITS
) I D
i I O 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLC OF CONTENTS () NUMBER REV UNIT TITLE
=============== === ==== =======================================- ---- =n===========
SlP M-53 7 1&2 ** CONTROL ROOM VENTILATION SYSTEM - DOP /.ND HALIDE PENETRATION TESTS () STP M-54 9 1&2
- MEASUREMENT OF REACTOR COOLANT PUMP SEAL INJECTION FLOW .
STP M-55 5 1&2
- RECORDING OF CYCLIC FATIGUE OR TRANSIENTS STP M-56 4 1&2
- VERIFICATION OF FLOW PATH FROM CONTAINMENT SPRAY ADDITIVE TANK 1-1 (2-1)
STP M-60 6 1&2
- PERIODIC INSPECTION OF AB0VEGROUND Olt TANKS
() STP M-63A 13 1&2
- FIRE WATER SYSTEM TEST STP M-64 4 1&2
- DELUGE SYSTEM FUNCTIONAL TEST STP M-65 8 1&2
- SPRINKLER / DELUGE SYSTEM VISUAL VERIFICATION STP M-66A 0 1
- DELUGE SYSTEM N0Z2LE PROOF TEST STARTUP TRANSFORMERS STP M-66A 0 2
- DELUGE SYSTEM N0ZZLE PROOF TEST STARTUP TRANSFORMERS ,
STP M-668 0 1
- DELUGE SYSTEM N0ZZLE PROOF TEST MAIN AND AUXILIARY TRANSFORMERS
() STP M-66B 2 2
- DELUGE SYSTEM N0ZZLL PROOF TEST MAIN AND AUXILIARY TRANSFORMERS STP M-67A 13 1&2
- WEEKLY FIRE VALVE INSPECTION STP M-67B1 1 1
- CONTAINMENT FIRE HOSE REEL STATION INSPECTION STP M-6782 1 2
- CONTAINMENT FIRE HOSE REEL STATION INSPECTION STP M-67B3 3 1
- CONTAINMENT FIREWATER FLOW PATH VALVE INSPECTION
() 4 2
- CONTAINMENT FIREWATER FLOW PATH VALVE INSPECTION i STP M-6784 STP M-67C 4 1&2
- MONTHLY HOSE REEL STATION AND HYDRANT INSPECTION '
STP M-69A 12 1&2
- MONTHLY FIRE EXTINGUISHER INSPECTION '
SlP M-69B 4 ILP
- MONTHLY CO2 HOSE REEL & DELUGE VALVE INSPECTION '
STP M-69C 4 1&2 *MONTHl' FIRE EXTINGUISHER INSPECTION OUTSIDE THE PROTECTED AREA
'() **lNSPECTION OF FIRE BARRIER PENETRATIONS STP M-70 15 1&2 I SIP M-71 5 1&2
- FIREWATER SYSTEM FLOW TEST STP M-72 2 1&2 *** EXERCISING VLVS FCV-200, 201 & 203 THRU 207, FWP TURBINES, MN TURBINE BEARINGS & H2 SEAL Olt DELUGE STP M-73 2 1&2 *** AUTO ST ART OF CCW PUMPS ON LOW PRESSURE STP M-74 2 IL2 ** AUTO START OF THE ASW PUMPS ON LOW PRESSURE
'() 14 1&2 ***4KV VITAL BUS UNDERVOLT AGE RELAY CAllBRATION SIP M-75 STP M-76 6 1&?
- PERIODIC TURBINE VALVE DISASSEMBLY '
STP M-77 10 IL2
- SAFETY AND RELIEF VALVE TESTING STP M-7EA 11 1&?
- SNUBBER VISUAL INSPECTION SIP M-766 10 1&?
- SNUBBER FUNCTIONAL TESTING STP M-78C 0 1&2
- TRANSIENT EVENT EVALUATION ,
'() SIP M-79 10 IL2 *lND00R FIRE HOSE lhSPEC110N, GASKET REPLACEMENT AND l RERACKING , STP M-80A 7 1&2 *0UTDOOR FIRE HOSE ANNUAL OPERABILITY AND HYDROSTATlc TEST STP M-80B 8 1&2 *lNDDOR FIRE HOSE OPERABILITY AND HYDROSTATIC TEST STP M-80C 1 1&2 *LONG TERM COOLING WATER PUMP HOSE ANNUAL OPERABILITY AND HYDROSTATIC TEST () l I O l
D DIABLO CANYOH POWER PLANT 22 JUL 93 UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS
)
NUMBER REV UN11 TITLE
=============== ===. ==== ===========================================================
STP M-81A 3 IL2 ***DIESFL ENGINE GENERATOR INSPECTION (18 MONTH INTERVALS)
- ) STP M-818 1 IL2
- DIESEL ENGINE GENERATOR INSPECTION (36 MONTH INTERVALS)
STP M-81C 2 1&2
- DIESEL ENGINE GENERATOR INSPECTION (54 MONTH INTERVALS)
STP M-81D 0 1&2
- DIESEL ENGINE GENERATOR INSPECTION (72 MONTH INTERVALS)
STP M-81E 1 IL2
- DIESEL ENGINE GENERATOR INSPECTION (90 MONTH INTERVALS)
STP M-81F 0 102
- DIESEL ENGINE GENERATOR INSPECTION (108 MONTH INTERVALS)
STP M-81G 1 IL2
- DIESEL ENGINE GENERATOR INSPECTION (ELECTRICAL 18-MONTH
- ) INTERVALS)
STP M-81H 0 IL2
- DIESEL ENGINE GENERATOR INSPECTION (ELECTRICAL 36 MONTH INTERVALS)
STP M-811 0 lt?
- DIESEL ENGINE GENEPATOR INSPECTION (ELECTRICAL 72 MONTH INTERVALS)
SlP M-83a 13 IL2
- PENETRATION OVERCURRENT PROTECTION STP M-83B 9 1L2
- PENETRATION CONDUCTOR OVERCURRENT PROTECTION MAINTENANCE c)
- MOTOR OPERATED VALVE THERMAL OVERLOAD RELAY CALIBRATION STP M-84B 5 1L2 STP M-84F 1 'L2
- BUS F MOTOP OPERATED VALVE THERMAL OVERL0AD BYPASS RELAY TRIP ACTUATION DEVICE OPERATIONAL TEST STP M-845 2 iLP
- SUS G MOTOR OPERATED VALVE THERMAL OVERLOAD BYPASS RELAY TRIP ACTUATION DEVICE OPERATIONAL TEST STP M-84H IL2
- BUS H MOTOR OPERATED VALVE THERMAL OVERLOAD BYPASS RELAY
[) 3 TRIP ACTUATION DEVICE OPERATIONAL TEST STF M-86 7 ILP
- LEAK REDUCTION OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIDACTIVE MATERIALS FOLLOWING AN ACCIDENT (NUREG-0578-TMI-2)
ST P M-86A 13 152 *NUREG-0578: SAFEYY INJECTION SYSTEM LEAK REDUCTION g, STP M-86C 6 , ***NUREG-0578: RHR PUMP 1 SUBSYSTEM LEAK REDUCTION SlP M-86C 1 2 ***NJREG-0578: RHR PUMP 1 SUBSYSTEM LEAK REDUCTION STP M-86D 8 3 ***NUPEG-0578: RHR PUMP 2 SUBSYSTEM LEAK REDUCTION STP M-86D 1 2 ***NUREG-0578: RHR PUMP 2 SUBSYSTEM LEAK REDUCTION STP M-86E 11 1 *NUPEG-0578 NSS SAMPLING SYSTEM LEAK REDUCTION STP M-86E 3 2 *NUREG-0578 NSS SAMPLING SYSTEM LEAK REDUCTION gp STP M-86r 1 2 ** GASEOUS RADWASTE SYSTEM LEAK REDUCTION - INITIAL PLAN STP M-86G 7 IL2 *NUREG-0578: CHARGING SYSTEM (SUCTION) LEAK REDUCTION AND LEAK CHECK OFCVCS-8c40, VCT OUTLET CHECK TO CHARGING PUMPS SUCTION STP M-87 1 IL2 *0PERATIONAL LEAK INVENTORY OF SYSTEMS OUTSIDE CONTAINMENT THAT ARE PARTOF THE ECCS POST-LOCA RECIRCULATION FLOW PATH g SlP M-87A 0 IL2 *SPECIAL ECCS POST-LOCA RECIRCULATION PATH COMPONENT LEAKAGE MEASUREMENT STP M-88A 6 IL2
- ELECTRICAL HYDROGEN RECOMBINERS FUNCTIONAL TEST (EHRS 1, 2)
STP M-8BB 3 1&2 **CAllBRATION OF TEMPERATURE INDICATORS FOR ELECTRIC HYDROGEN RECOMBINERS (EHR$ 1,2) SlP M-88C 10 IL2 ** CHANNEL CAllBRATION OF INTERNAL HYDROGEN RECOMBINER POWER METER, HEATER INSPECTION AND HEATUP TEST q, i I e 1
3 22 JUL 93 DIABLO CANYOH POMER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES
- ) TABLE OF CONTENTS NUMBER REV UNIT TITLE
=============== === ==== ===========================================================
STP M-89 16 1 *ECCS SYSTEM VENTING
- ) STP M-89 4 2 *ECCS SYSTEM VENTING STP M-90 3 1&2 ** SURVEILLANCE OF DIABLO CANYON BREAKWATERS STP M-91 7 1&2
- DIESEL FUEL OIL TRANSFER SYSTEM, PIPING AND COMPONENT INSPECTION STP M-91A 1 1&2
- DIESEL FUEL OIL STORAGE TANKS INSPECTION AND CLEAN 1HG STP M-92 7 1 *** LOST LOCA SAMPLING CENTER VENTILATION SYSTEM - DOP AND
- ) HALIDE PENETRATION TESTS STP M-92 0 2 *** POST LOCA SAMPLING CENTER VENTILATION SYSTEMS - DOP AND HALIDE PENETRATION TESTS ETF M-93A 6 1 ***18 MONTH SURVEILLANCE - CONTAINMENT FAN COOLER SYSTEM STP M-93A 2 2 ***18 MONTH SURVEILLANCE - CONTAINMENT FAN COOLER SYSTEM STP M-93B 0 IL2 *10DlhE REMOVAL UNIT FILTER TEST
- ) STP M-94 1 IL2 *LAUhDRY FACILITY VENTILATION - DOP PENETRATION TEST STP M-95 4 IL2
- SOLID RADWASTE STORAGE FACILITY VENTILATION TEST STP M-96 2 1&2 **TSC VENTILATION SYSTEMS - DOP AND HALIDE PENETRATION TESTS STP M-97 2 1&2
- RESISTANCE TEMPERATURE DETECTOR BYPASS LOOP FLOW MEASUREMENT STP M-96 3 1
- COMPARISON OF FINAL FEEDWATER FLOW N0ZZLES TO ASME FLOW N0ZZLES c)
STP M-96 4 2
- COMPARISON OF FINAL FEEDWATER FLOW N0ZZLES TO ASME FLOW N0ZZLES STP M-99 5 1&2 *** TEST OF BACKUP NITROGEN ACCUMULATOR SYSTEM CHARGING HEADER FLOW CONTROL VALVE HCV-142 STP M-100 5 1&2 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO 9351A AND 9351B, RCS SAMPLE ISOLATION VALVES FOR LOOPS 1 AND 2
(') *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO RCS SAMPLE STP M-101 3 1 INSIDE CONTAINMENT ISOLATION VALVE 9356A SlP M-101 3 2 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO RCS SAMPLE JNSIDE CONTAINMEN1 ISOLATION VALVE 9356A STP M-102 7 1&2 *** TEST OF BACKUP NITROGEN ACCUMULATOR SYSTEM TO SPRAY VALVE AND CHARGING VALVES 8145, 8146, AND 8147 () STP M-103 3 1 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO RCS SAMPLE OUTSIDE CONTAINMENT ISOLATION VALVE 9356B STP M-103 2 2 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO RCS SAMPLE OUTSIDE CONTAINMENT ISOLATION VALVE 9356B STP M-104 7 IL2 *** TEST OF BACKUP NITROGEN ACCUMULATOR SYSTEM TO SPRAY VALVE BYPASS VALVE CVCS-8148 c) 6 1&2 *** TEST OF BACKUP NITROGEN ACCUMULATOR SYSTEM TO STP M-105 PRESSURIZER POWER OPERATED RELIEF VALVE PCV-455C STP M-106 6 1&2 *** TEST OF BACKUP NITROGEN ACCUMULATOR SYSTEM TO PRESSURIZER POWER OPERATED RELIEF VALVE PCV-456 l STP M-107 2 1&2 ** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO FCV-364 RHR HEAT l EXCHANGER CCW RETUPh VALVE g) 1 C)
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 I VOLUME 6 SURVEILLANCE TEST PROCEDURES J TABLE OF CONTENTS l NUMBEk REV. UNIT TIlLE
=============== === ==== = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = = =
STP M-10B 2 IL2 ** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO FCV-365 RHR HEAT
) EXCHANGER CCW RETURN VALVE STP M-109 4 IL2 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO FCV-602, CCW HEAT EXCHANGER NO. 1 SALTWATER INLET VALVE STP M-110 4 IL2 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO FCV-603, CCW HEAT EXCHANGER NO. 2 SALTWATER INLET VALVE STP M-111A 1 1 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO MAIN STEAM a
ISOLATION STP M-Illa 2 2 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO MAIN STEAM ISOLATION VALVE FCV-41 STP M-!!1B 1 1 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO MAIN STEAM ISOLATION VALVE FCV-42 STP M-l!1B 2 2 *** TEST OF BACKUP A]P ACCUMULATJR SYSTEM TO MAIN STEAM
- ) ISOLATION VALVE FCV-42 STP M-111C 1 1 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO MAIN STEAM ISOLATION VALVE FCV-43 STP M-111C 2 2 *** TEST OF BACKUP AIR ACCUMUALTOR SYSTEM TO MAIN STEAM ISOLATION VALVE FCV-43 STP M-111D 1 1 *** TEST OF BACKUP AIR ACCUMULATOR SYSTEM TO MAIN STEAM g' ISOLATION VALVE FCV-44 STP M-111D 2 2 *** TEST OF BACKUP AIR ACCUMUALTOR SYSTEM TO MAIN STEAM ISOLATION VALVE FCV-44 STP M-114 5 1 ***VERIFICAT]ON OF SEISMIC QUALIFICATION OF CONTAINMENT SPFAV PlPING STP M-114 0 2 ***VER]FICATION 0F SEISMIC QUALIFICATION OF CONTAINMENT SPRAY PlPING 3 SlP M-115 1 IL2
- TEST FOR LOW FLOW ALARM ON CONTAINMENT PURGE FAN E-3 STP M-116 1 IL2 *BLOWUOWN AND TEST RUN OF BREATHING AIR SYSTEM STP M-23S 1 1&2 ** LEAK TEST OF SAFELY INJECTION PIPING DOWNSTREAM OF 8802A&B STP G-6A 7 1&2 ** CALIBRATION & USE OF Aiu '1 REACTIVITY COMPUTER STP G-8B 3 IL2
- SETTING UF REACTIVITY COMF ,f ER
'g STP G-8C 5 IL2 **0PERAl10NAL CHECKOUT OF THE REACTIVITY COMPUTER STP G-8D 7 1&2 ***CAllBRATION AND USE OF THE DIGITAL REACTIVITY COMPUTER STP G-BE O 1&2 ***CAllBRATION AND USE OF THE MULTI-CHANNEL DIGITAL ! REACTIVITY COMPUTER WITH MIDS DETECTORS STP G-9 7 IL2
- GENERAL HEPA FILTEP BANK PENETRATION TEST STP G-10 6 IL2
- GENERAL CHARCOAL FILTER BANK PENETRATION TEST
^g STP G-ll 4 IL2 *** PROCEDURE FOR OBTAINING CHARCOAL FILTER MEDIA FOR LABORATORY TESTING (METHYL IODINE) STP G-12 3 IL2 ' OPERATION OF THE PORTABLE LEAK TEST MONITOR STP G-13 3 IL2 **0PERATION & USE OF PRESSURE TIME RESPONSE TESTER STP G-14 10 IL2 *0PERABILITY DETERMINATION OF POST ACCIDENT SAMPLING PROGRAM STP G-15B 2 IL2 *** DETERMINATION OF VALVE STROKE TIMES WITH EQUIPMENT TIMERS 9 D
/
J 22 JUL 93 DIABLO CANVON POWER PLANT UNITS 1 AND 2 UOLUME 6 SURVEILLANCE TEST PROCEDURES TABLE OF CONTENTS [) NUMBER REV UNil TITLE
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STP G-16 2 1&2 ' ALTERNATE METHODS OF MONITORING FOR HIGH-RANGE POST-ACCIDENT RADIATION MONITORS 3 STP G-17A 4 1&2
- EMERGENCY EQUIPMENT FOR ENVIRONMENTAL MONITORING INVENTORY SURVEILLANCE STP G-178 4 1&2
- EVACUATION KIT INVENTORY SURVEILLANCE STP G-17C 4 1&2
- HOSPITAL KIT INVENTORY SURVEILLANCE STP G-17D 3 1&2 *K1 INVENTORY 3 STP G-17E 6 1&2
- EMERGENCY PROCEDURES PHONE NUMBER VERIFICATION STP G-17F 2 1&2
- CONTROL ROOM EMERGENCY EQUIPMENT AND SUPPLIES STP G-17G 4 1&2 ** TECHNICAL SUPPORT CENTER EMERGENCY EQUIPMENT AND SUPPLIES STP G-17H 2 1&2 *0PERATIONAL SUPPORT CENTER EQUIPMENT INVENTORY SURVEILLANCE STP G-171 8 1&2
- EMERGENCY OPERATIONS FACILITY (EOF)
STP G-17] 3 1&2 *DECONTAMINAT10N FACILITY EQUIPMENT SURVEILLANCE 3 STP G-17K 6 1&2 *** JOINT MEDIA CENTER PHONE OPERABILITY CHECK STP G-17L 4 1&2 **0NSITE ASSEMBLY AREAS EMERGENCY INSTRUCTIONS STP G-17N 1 1&2
- SITE EMERGENCY SIGNAL-0UTLYING AREAS SURVEILLANCE STP G-17P 1 1&2 *TSC RADIATION MONITOR CHECK STP G-17Q 0 1&2 *ERDS CHECK STP G-18 0 1&2 ** SITE EMERGENCY SIGNAL AND FIRE SIGNAL AUDIBILITY AND FUNCTIONAL TEST -lNSIDE THE PROTECTED AREA 3 STP G-19 0 1&?
- RAD 10 ACTIVE SOURCE LEAK TESTING STP S-1A 5 1&2 ** MEASUREMENT OF STATION SECURITY SYSTEM BATTERY PILOT CELL VOLTAGE AND SPECIFIC GRAVITY STP S-le 6 1&2
- MEASUREMENT OF STATION SECURITY SYSTEM BATTERY VOLTAGE AND SPEClrlC GRAVITY 3 STP S-lC 2 1&2
- SECURITY BATTERY TERMINAL RESISTANCE MEASUREMENT &
INSPECTION STP C-1 2 1&2
- SPRAY ADDITIVE SYSTEM CHEMICAL INVENTORY STP C-2 4 1&2
- ACCUMULATOR CHEMICAL CONCENTRATION STP C-3 0 1&2
- SPENT FUEL POOL BORON CONCENTRATION STP X-1 1 1&2
- VISUAL EXAMINATION OF THE REACTOR VESSEL INTERIOR STP X-2 0 1&2
- VISUAL INSPECTION OF REACTOR UPPER INTERNALS AND FUEL j 3 ASSEMBLY GUIDE PINS STP X-100 0 1&2 *151 HYDROSTATIC PRESSURE TEST PROGRAM STP X-101 0 1
- REACTOR COOLANT SYSTEM PRESSURE BOUNDARY HYDROSTATIC PRESSURE TEST (ASME CLASS 1)
STP X-107A 0 1
- REACTOR COOLANT PUMP l-1, SEAL WATER INJECTION / BYPASS HYDROSTATIC TEST (ASME CLASS 1) 3 STP X-107B 0 1
- REACTOR COOLANT PUMP 1-2, SEAL WATER INJECTION / BYPASS ,
HYDROSTAllC PRESSURE TEST (ASME CLASS I) -l STP X-107C 0 1
- REACTOR COOLANT PUMP l-3, SEAL WATER INJECTION / BYPASS j HYDROSTATIC PRESSURE TEST (ASME CLASS !)
STP X-107D 0 1
- REACTOR COOLANT PUMP l-4, SEAL WATER INJECTION / BYPASS HYDROSTATIC PRESSURE TEST (ASME CLASS I) 7 D
l j
O 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES
- C) TABLE OF CONTENTS ,
1 NUMBER REV UNIT TITLE
=============== === ==== =========================================================== ;
STP X-200 0 1&2 *151 INSERVICE / FUNCTIONAL PRESSURE TEST PROGRAM C) - STP X-209 0 1 *151 HYDROSTATIC PRESSURE TEST OF RCS SAMPLE LINES (LOOPS 1 AND 4. HOT 1EGS) " STP X-210 0 1 *151 HYDROSTATIC PRESSURE TESI 0F PRESSURIZER SAMPLE LINES S-1673/8 AND S-1674-3/8 STP X-211 0 1 *151 HYDROSTATIC PRESSURE TEST OF 0F EXCESS LETDOWN HEAT EXCHANGER 1-1 TUBE SIDE AND INLET /0UTLET PIPE () STP X-213 0 1
- REACTOR COOLANT PUMP SEAL WATER BYPASS PIPING FROM R0-34, 35, 36 & 40 TO VALVE CVCS-1-8142 STP X-214 0 1 *151 HYDROSTATIC PRESSURE TEST OF RCP SEAL WATER INLET PIPING STP X-218 0 1 *151 HYDROSTATIC PRESSURE TEST OF CVCS RECIPROCATING CHARGING PUMP l-3
' C) STP X-226A 0 1 *151 HYDROSTATIC PRESSURE TEST OF ACCUMULATOR INJECTION LOOP #1 STP X-226B 0 1 *151 HYDROSTATIC PRESSURE TEST OF ACCUMULATOR INJECTION LOOP #2 STP X-226C 0 1 *15] HYDROSTATIC PRESSURE TEST OF ACCUMULATOR INJECTION LOOP #3
() STP X-226D 0 1 *151 HYDROSTATIC PRESSURE TEST OF ACUMMULATOR INJECTION LOOP #4 I STP X-227 3 IL2
- SIS TEST SYSTEM PIPING PRESSURIZATION FOR ISI INSPECTION STP X-229 0 1 *151 HYDROSTATIC PRESSURE TEST OF' ACCUMULATOR SAMPLE LINES STP X-234 0 1 ** CHARGING INJECT]ON SYSTEM PIPING PRESSURIZATION STP X-234 0 2 ** CHARGING INJECTION SYSTEM PIPING PRESSURIZATION i C)
STP X-258 2 1 *151 PRESSURE TEST OF CONTAINMENT SPRAY PUMP DISCHARGE PIPING STP X-258 1 2 *ISI PRESSURE TEST OF CONTAINMENT SPRAY PUMP DISCHARGE PIPING STP X-261 0 1 *151 PRESSURE TEST OF SPRAY ADDITIVE TANK AND PIPING STP X-262A 0 1 *151 HYDROSTATIC PRESSURE TEST OF CONTAINMENT. FAN COOLER
- C) 1-1 CCW SUPPLY AND RETURN PIPING ~
STP X-2628 0 1 *151 HYDROSTATIC PRESSURE TEST OF CONTAINHENT FAN COOLER l-2 CCW SUPPLY AND RETURN PIPING STP X-262C 0 1 *151 HYDROSTATIC PRESSURE TEST OF CONTAINMENT FAN COOLER r 1-5 CCW SUPPLY AND PETURN PIPING STP X-263A 0 1 *151 HYDROSTATIC PRESSURE TEST OF CONTAINMENT FAN COOLER c) 1-3 CCW SUPPLY AND RETURN PIPING
- STP X-2635 0 1 *151 HYDROSTATIC PRESSURE TEST OF CONTAINMENT FAN COOLER ,
1-4 CCW SUPPLY AND RETURN PIPING STP X-264A 0 1 *151 HYDROSTATIC PRESSURE TEST OF RCP CCW OIL COOLER SUPPLY LINE CONTAINMENT PENETRATION 19 STP X-264B 0 1 *151 HYDROSTATIC PRESSURE TEST OF RCP CCW OIL COOLER RETURN LINE CONTAlhMENT PENETRATION 20 IC) O
i 1) 22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES
- D TABLE OF CONTENTS NUMBLR REV UNIT TITLE
=============== === ==== ===========================================================
STP X-265 0 1 *151 HYDROSTATIC PRESSURE TEST OF RCP CCW THERMAL BARRIER J RETURN LINE CONTAINMENT PENETRATION 21 STP X-266 0 1 *151 HYDROSTATIC PRESSURE TEST OF EXCESS LETDOWN HEAT EXCHANGER l-1 SHELL SIDE STP X-268 1 1 *ISI PRESSURE TEST OF N2 SUPPLY TO STEAM GENERATORS STP X-268 1 2 *151 PRESSURE TEST OF N2 SUPPLY TO STEAM GENERATORS SIP X-271 1 1 *151 PRESSURE TEST OF PRESSURIZER RELIEF TANK NITROGEN
) SUPPLY L}NE STP )-272 0 1 *151 HYDROSTATIC PRffSURE TEST OF PRIMARY WATER SUPPLY P] PING TO PRESSURllER RELIEF TANK STP X-275 1 1 *151 PRESSUFE TEST OF ACCUMULATOR TANKS NITROGEN SUPPLY HEADER SlP /-279 0 1 *ISI PRESSUPE TEST OF AUX STEAM CONTAINMENT PENETRATION ) STP X-279 0 2 *151 PRESSURE TEST OF AUX STEAM CONTAINMENT PENETRATION SlF x-280 0 1 *ISI PPESSURE TEST OF FIREWATER SUPPLY PIPING FROM FP-1-FCV-633 10 FP-1-180 STP X-284 0 1 *151 PRESSURE TEST OF NITROGEN TO REACTOR COOLANT DRAIN TANK 1-1 STP X-310 0 1 *10 YEAR 151 HYDROSTATIC TEST OF CCW SUPPLY AND RETURN ) HEADERS A, B AND C AND ASSOCIATED COMPONENTS SlP x-310 0 2 *10 iEAR ISI HYDR 0 STATIC TEST OF CCW SUPPLY AND RETURN HEADERS A, B AND C AND ASSOCIATED COMPONENTS ilP /-3!O? O 2 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW CONTAINMENT SUPPti HEADER SlP x-310L 0 1
- 10 Y E Ar 15] HYDROSTATIC PRESSURE TEST OF CCW RCP l-2 UPPER
- D AND LOWER BEARING LUEE Oll COOLER SUPPLY AND RETURN PIPING S'0 x-31GF 0 2 *10 TEAR 151 HYDROSTATIC PRESSURE TEST OF CCW RCP 2-2 UPPER AND LOWER BEAPING LUBE OIL COOLER SUPPLY AND RETURN PIPING S1P x-3105 0 1 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW RCP l-3 UPPER AND LOWER BEARING LUBE DIL COOLER SUPPLY AND RETURN PIPING STP X-3105 0 2 *10 YEAR 151 HYDROSTATIC PRESSURE TEST OF CCW RCP 2-3 UPPER
- ) AND LOWER BEARING LUBE OIL COOLER SUPPLY AND RETURN PIPING STP X-3101 0 1 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW RCP l-4 UPPER AND LOWER BEARING LUBE OlL COOLER SUPPLY AND RETURN PIPING SlP X-3101 0 2 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW RCP 2-4 UPPER AND LOWER BEARING LUBE OIL COOLER SUPPLY AND RETURN PIPING STP x-310U 0 1 *10 YEAR 151 HYDROSTATIC PRESSURE TEST OF CCW RCP l-1 UPPER
) AND LOWER BEARING LUBE OIL COOLER SUPPLY AND RETURN PlPlNG STP x-310U 0 2 *10 YEAR 151 HYDROSTATIC PRESSURE TEST Of CCW RCP 2-1 UPPER AND LOWER BEARING LUBE OIL COOLER SUPPLY AND RETURN PIPING STP X-310V 0 2 *10 YEAR 151 HYDROSTATIC PRESSURE TEST OF CCW REACTOR VESSEL COOLING PLATES STP X-311A 1 1 *10 YEAR 151 HYDROSTATIC PRESSURE TEST OF CCW RCP 1-2 9 1HERMAL BARRIER AND RETURN HEADER D
22 JUL 93 DIABLO CANYON POWER PLANT UNITS 1 AND 2 VOLUME 6 SURVEILLANCE TEST PROCEDURES
- 7) TABLE OF CONTENTS HUMBER REV UNIT TITLE
=============== ===- ==== ===========================================================
STP X-31]A 0 2 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW RCP 2-2
) STP X-311B THERMAL BARRIER AND RETURN HEADER 1 1 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW RCP 1-3 THERMAL BARRIER PIPING STP X-311B 0 2 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW RCP 2-3 THERMAL BARRIER PIPING STP x-311C 1 1 *10 YEAR 151 HYDROSTATIC PRESSURE TEST OF CCW RCP 1-4
- 7) THERMAL BARRIER PIPING 51P x- 311C 0 2 *10 YEAR ISI HYDROSTATIC PRESSURE TEST OF CCW RCP 2-4 THERMAL BARRIER PIPING STP x-311D 1 1 *10 YEAR ISI HfDR0 STATIC PRESSURE TEST OF CCW RCP 1-1 THERMAL BARRIER PIPING STP X-3110 0 2 *10 YEAR 15] HYDROSTATIC PRESSURE TEST OF CCW RCP 2-1
- 7) THERMAL BARRIER PIPING STP X-400 1 IL2 *151 TEST PROGRAM FOR CLASS 1 AND CLASS 2 COMPONENTS STP X-500 1 lt2 *151 TEST PROGRAM FOR CLASS 1 PIPING SYSTEMS AND CLASS 2 PIPE SUPPORTS STP X-6DO 1 IL2 *151 TEST PROGRAM FOR CLASS 2 PIPING SYSTEMS EXCEPT SUPPORTS SlP X-700 1 IL2 *ISI TEST PROGRAM FOR CLASS 3 COMPONENTS g SlP X-B00 0 IL2 *lSI TEST PROGRAM FOR STEAM GENERATOR TUBES STP x-900 0 IL2 *151 TEST PROGRAM FOR CLASS 1 AND CLASS 2 COMPONENTS REQUIRING SUCCESSIVE INSPECTION STP X-2105 0 1 *151 PRESSURE TEST OF SERVICE AIR PENETRATION STP X-2105 0 2 *ISI PRESSURE TEST OF SERVICE AIR PRESSURE STP X-2106 0 1 *151 PRESSURE TEST OF INSTRUMENT AIR SUPPLY HEADER 4,
STP PEP 00L 0 IL2
- DEVELOPMENT AND IMPLEMENTATION GUIDELINES FOR PERFORMANCE ENGINEERING PROCEDURES STP PEP 00B 0 IL2
- PERFORMANCE MONITORING BASELINE / PERIODIC PROGRAM STP PEP 00C 0 1&2 *VlBRATION IMPLEMENTATION AND MONITORING GUIDELINES STP PEP 03-01 0 1
- MAIN FEEDWATER PUMP l-2 DISCHARGE CHECK VALVE TEST STP PEP 03-02 1 IL2
- DETERMINATION AND VAL 10AT10N OF THE N0ZZLE FOULING FACTOR STP PEP 04-01 0 IL2 *10% ATMOSPHERIC DUMP VALVES OPERATIONAL CHECK OF BALANC]NG CHAMBERS (PCV-19, 20, 21 & 22)
SlP PEP 04-R 2 1&2
- MAIN FEEDWATER TURBINE OVERSPEED TRIP TEST STP PEP 05-01 0 1&2 ***FEEDWATER HEATER PERFORMANCE EVALUATION STP PEP l-80.01 0 1&2
- CALIBRATION OF THE ABL-4000 RESPIRATORY AIR LINE MONITORS AT BLAST / PAINT FACILITY STP PEP M-12A15 1 1 ** BATTERY 15 PERFORMANCE TEST iD STP PEP M-12A16 1 1 ** BATTERY 16 PERFORMANCE TEST STP PEP M-12A17 0 1 *** BATTERY 17 PERFORMANCE TEST STP PEP M-l?A25 0 2 ** PLANT ENGINEERING PROCEDURE BATTERY 25 PERFORMANCE TEST STP PEP M-12A26 0 2 ** PLANT ENGINEERING PROCEDURE BATTERY 26 PERFORMANCE TEST STP PEP M-12A27 0 2 *** BATTERY 27 PERFORMANCE lEST
, SIP PEP R-1 3 }&2
- RETRIEVAL DF UNDERWATER DEBRIS O
D 22 JUL 93 DIABLO CANYON POWER PLANT ) UNITS 1 AND 2 ! VOLUME 6 SURVEILLANCE TEST PROCEDURES , 1 TABLE OF CONTENTS 3 ! l HUMBER REV UNIT TITLE l
=============== === ==== =========================================================== l STP PEP 20-01 1 16.2 ***EH CONTROL SYSTEM HP ACCUMULATOR PRECHARGE PRESSURE CHECKS 3
i 3 3 3 D D j 1 I O
O 1 l l l DIABLO CANYON LOOP TEST REPORT (NOT INCLUDED) l O For Both Units, including Unit 1 and Unit 2 unique tests, Common Area Tests and Unit 1 & 2 Combined, the total number of loop O tests is approximately 3,200. Due to the size of this exhibit (approximately 150 pages) a copy of the entire list is not included with this testimony. However, a copy of the first page of the list is attached for illustration purposes. The complete list will be furnished at the hearing for inclusion in the record. O O O ; O O 1 0 ; i ' j O
l ) DCPP Loon Test Report l 1 Unit Test Number Rev Approval Title 2-69 0 08/14/91 COMPLT BACKWASH SUMP 0-1 LEVEL CHANNEL ) 0 LS-Il4 A/B CALIBRATION O 2 124 0 09/07/90 COhTLT PRECOAT FILTER DEMIN 0-1 BACKWASH AIR FCV-1154 CALIBRATION O 2 125 0 07/12/90 COMPLT PRECOATFILTER DEMIN 01 DRAIN FCV-1155 CALIBRATION 0 2-127 0 07/26/92 COMPLT BACKWASH WATER FLUSH AND FILL TO 3 PRECOAT FILTER DEhEN 0-1 FCV-1157 CALIBRATION O 2-132 0 01/27/93 COhPLT PRECOAT FILTER DEhDNERALIZER 0-1 INLET FCV-1162 CALIBRATION O 2-153 A 2 06/04/93 COMPLT PRECOAT FILTER DEhuN 0-2 OUTLET FLOW CHANNEL FT-1251 ) CALIBRATION O 2-16? O 05/06/92 COMPLT PRECOAT FILTER DENUNERALIZER 0-2 DRAIN FCV-1255 CALIBRATION 0 2 300 0 02/05/91 COMPLT BACKWASH SLURRY TANK 0-1 LEVEL CHANNEL LI-300 CALIBRATION 0 6-4 A 0 01/24/92 COMPLT AUXILIARY BOILER O-1 WATER LEVEL ) CONTROL CHANNEL LS-346 CALIBRATION O 6-4 B 4 01/24/92 COMPLT AUXILIARY BOILER 0-1 WATER LEVEL CONTROL CHANNEL LS-344 ' CALIBRATION 0 6-7A 1 05/13/93 COMPLT AUX BLDG AUX STEAM SUPPLY HEADER ) PRESSURE CHANNEL PC-113 CALIBRATION 0 610D 0 06/14/90 COMPLT REBOILER 0-1 DRIP POT LEVEL SWITCH CHANNEL CALIBRATION 0 6-12B 0 07/16/90 COMPLT WASTE CONC 0-1 DRIP POT LEVEL SWITCH CHANNEL CALIBRATION O 6-13B 0 06h4/90 COMPLT REBOILER 0-1 LEVEL SWITCH CHANNEL CALIBRATION 0 6-14B 0 07/16/90 COMPLT BUILDING HEATING DRAIN RECUVER 0-1 LEVEL SWITCH CHANNEL CALIBRATION 0 6-23A 0 03/30/87 COMPLT AUXILIARY BOILER FUEL OIL PP ) 0-1 INTK 0 6-604 0 07/16/90 COMPLT PACKAGE BOILER 0-2 HIGH , AND LOW WATER LEVEL ALARM CHANNEL CALIBRATION O 6-606 0 07/16/90 COMPLT PACKAGE BOILER 0-2 LOW WATER F.O. ' CLTT OFF CHANNEL CALIBRATION )- 0 6-608 0 07/23/90 COMPLT PACKAGE BOILER 0-2 LO LO WATER LEVEL CHANNEL CALIBRATION 0 6-970 0 05/08/91 COMPLT PACKAGE BOILER 0-2 FUEL OIL STORAGE TANK LEVEL CHANNEL LT-970 CALIBRATION 0 8-7A 0 04/27/88 COMPLT LIQUID HOLD-UP TANK 0-1 LEVEL ) 0 8-7B 0 12/13/88 COMPLT LIQUID HOLD-UP TANK 0-1 LEVEL 0 8 7C 0 11/18'88 COMPLT LlQUID HOLD UP TANK 0-1 LEVEL Page 1 of XX
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EXHIBIT 12 0~ PROFESSIONAL QUALIFICATIONS OF .O - TEco A. oILLAno
- O O
O. O O-iD
-l LO i
! 1 l e iO . ux _ _= _ _ _. . .. -____-__ - - - -
LJ - RESUME O- Tedd A. Dillard
- B.S., Mechanical Engineering, LeTourneau College, Longview Texas, 1969.
- Westinghouse, 1969 through 1973.
- Florida Power and Light, 1973-through present.
- Mechanical Maintenance Department Head, St. Lucie Nuclear.
Plant, 1972 through 1983. O e Superintendent of Maintenance, St. Lucie Nuclear Plant, 1983 through 1988.
- INPO Maintenance Assistance Visit, Peer Evaluator, Ft.
Calhoun Plant, 1987. O
- INPO Maintenance Assistance Visit, Corporate Evaluator, Turkey Point Nuclear Plant, 1987.
- NUMARC Advisory Committee on Industry Response to NRC Maintenance Policy, 1988 through 1989.
O-
- NUMARC Advisory Committee on Industry Guideline for Maintenance Rule, 1991 through 1992.
- Supervisor, Component Programs, Nuclear Division, Florida Power.& Light Company, 1992 through present.
O O. O-O 1 1 u
)
m . _ - - - - _ - _ - _ - - - - - - - - - _ - - - - - - _ - - - - - - - - - - _ _ - - - - - - - _ - _ _ J
O: EXHIBIT 13 0: PROFESSIONAL QUALIFICATIONS OF DAVID B. MIKLUSH O l' (Q - . [- -t o 1 O. i O 1 'O- H i l O' '
.0 .
O' o, 2 :-- _ - _ - __ - - _ - _ _ _ - . - _ - _ - - - - _ - - - - - _ - _ . _ _ _
) l L RESUME i MANAGER, OPERATION SERVICES David B. Miklush
- 1. Birthdate - January 29, 1950
- 2. Citizenship - USA
- 3. Education kng s 2 l b. Registered Professional Engineer, Mechanical,
! California #18199. [ l 4. Employment History - Joined PG&E in July 1982.
- a. September 1972 to April 1976 - General Atomic Company.
Participated in the Technical Graduate Program at General Atomic with three 6-month assignments in manufacturing design engineering, and site startup at Fort St. Vrain from August 1974 to April 1976 in j construction and operations.
- b. April 1976 to February 1978 - General Electric Company.
Responsible Design Engineer for the BWR refueling, fuel handling, and auxiliary service bridger,. Assignment consisted of verification of vendor hardware designs and initial design of the fuel grapple for BWR 6. 3
- c. February 1978 to June 1980 - PG&E at Diablo Canyon Power Plant. Power Production Engineer.
- d. June 1980 to August 1982 - Senior Power Production Engineer (Nuclear). Supervised 5 engineers in 3 preparation of surveillance test procedures and conduct of plant equipment testing.
- e. August 1982 to August 1983 - Assistant Maintenance Supervisor.
- f. August 1983 to June 1988 - Maintenance Manager of l Mechanical and Electrical Maintenance. Supervised 1
department consisting of 15 engineers, 17 supervisors, and 130 journeymen.
) 9 June 1988 to September 1989 - Assistant Plant Manager / Maintenance Services. Responsible for plant maintenance at Diablo Canyon consisting of mechanical, i electrical, instrumentation and controls, maintenance workplanning center, and materials services.
l i
t.
- h. September 1989 to. July-1991 - Assistant Plant Manager / Operation Services. Responsible for~ plant p operations at Diablo Canyon consisting of-Operations, Radiation Protection, and Chemistry Departments.
- i. July 1991 to Present - Manager, Operation Services (title change).
g 5. Nuclear Experience
- a. Fort St. Vrain - Participated in initial core-loading; shift operations engineer during low-power physics tests to 2 percent power.
p b. General Electric - Design of nuclear fuel handling.and. servicing equipment.
- c. Diablo Canyon - Power Production Engineer (Nuclear) engaged in procedure preparation and startup testing of various plant systems and equipment.
)
- d. Westinghouse Phase 2, three-month classroom training on SNUPP's plant. Westinghouse Phase-3, two-month simulator training on SNUPP's plant. Certified SRO-with Westinghouse Training Center, August 1981.
)
- e. Senior Reactor Operator's License,-March 1982 to October 1988 (retired).
- f. INPO Senior Nuclear Plant Manager's Course-(5 weeks),.
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l l )~ ' EXHIBIT-19 i 5 l . I F ) NRC COMMENDATION LETTERS I f a r l l. I
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/ o-UNITED STATES l
g' [" gr -Ek NUCLEAR REGULATORY COMMISSION l W ASHINGTON. D. C. 20555 ' n j
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JUN 2 2 993 O Docket tios. 50-275 & 50-323 Mr. Richard A. Clarke Chairman of the Board and Chief Executive Officer O Pacific Gas and Electric Company 77 Beale Street > San Francisco, California 94106
Dear Mr. Clarke:
O On June 15-16, 1993, fiRC senior managers met to evaluate the nuclear safety performance of operating reactors, fuel facilities, and other materials licensees. The fiRC conducts this meeting semiannually to determine if the safety performance of the various licensees exhibits sufficient weaknesses to warrant increased tiRC attention. In addition, at this meeting, senior managers identify specific plants that have demonstrated a level of safety O performance that deserves formal NRC recognition. At the June 1993 Senior Management Meeting, the Diablo Canyon nuclear power plant (Units 1 and 2) was identified as having achieved a high level of safety performance and as a result met criteria for recognition of its performance. I am pleased to note that Diablo Canyon has again been identified as a good performer. O In identifying such plants, fiRC senior managers perform an evaluation of performance in many areas including operational safety, self-assessment, problem resolution, and plant management organization and oversight. The f4RC recognizes that to ac.ieve the level of performance demonstrated by the Diablo Canyon nuclear power plant, there must be management involvement in O all phases of plant activities, the staff must be dedicated and knowledgeable and fully supportive of plant activities, and a commitment to safety must i exist throughout the organization. We commend you and your staff for achieving this high level of safety performance. Your achievement is a positive example to the industry. O The greatest challenge that you now face is to maintain this level of performance and not to rest on past achievements. Continued management involvement and support, and dedicated efforts from your staff to identify and O O .O
I y i ) Mr. Richard A. Clarke > i promptly correct problems are necessary for you to continue to meet this 3 difficult challenge. Sincerely,
/
J mes M xecutive Director for Operations cc: See next page J D D D 9 , ) 3-
[ no:4o,% UNITED STATES NUCLEAR REGULATORY cot 4 MISSION I g -
,i WASWNGTON, D C.20555 February 5, 1993 k..... f I Docket lios. 50-275 & 50-323 Mr. Richard A. Clarke Chairman of the Board and a' Chief Executive Officer Pacific Gas and Electric Company 77 Beale Street San Francisco, California 94106
Dear Hr. Clarke:
~
On January 26-28, 1993, liRC senior managers met to evaluate the nuclear safety-performance of operating reactors, fuel facilities, and other materials licensees. The fiRC conducts this meeting semiannually to determine if the safety performance of the various licensees exhibits sufficient weaknesses to warrant increased f!RC attention. In addition, at this meeting, senior a~ canagers identify specific plants that have demonstrated a level of safety performance that deserves formal fiRC recognition. At the January 1993 Senior Management lfeeting, the Diablo Canyon nuclear power plant (Units I and 2) was identified as having achieved a high level of safety performance and as a result met criteria for recognition of its performance. I am pleased to note that Diablo Canyon has again been identified as a good performer. O In identifying such plants, fiRC senior managers perform an evaluation of performance in many areas including operational safety, self-assessment, problem resolution, and plant management organization and oversight. The tiRC recognizes that to achieve the level of performance demonstrated by O ' the Diablo Canyon nuclear power plant, there must be management involvement in all phases of plant a .tivities, the staff must be dedicated and knowledgeable
~
and fully supportive of plant activities, and a commitment to safety must exist throughout the organization. We commend you and your staff for achieving this high level of safety performance. Your achievement is a positive example to the industry. O The greatest challenge that you now face is to maintain this level of performance and not te rest on past achievements. Continued management involvement and support, and dedicated efforts from your staff to identify and O 9 i
j O Mr. I ard A. Clarke February 5,1993 O. promptly correct problems are necessary for you to continue to meet this difficult challenge. Sincerely, - O , , xecutive Director for Operations O cc: See next page O 'O O _ O l } 3
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/ )p j- UNITED STATES ,
i 8 j NUCt. EAR REGULATORY COMMISSION l WAaNWc7DN. D C.actes ! 4 9*****,C June 2;).1992 i O Docket No. 50-275 f Decket No. 50-323 O. , Nr. Richard A. Clarke Chairman of the Board and l Chief Executive Officer Pacific Gas and Electric Co:pany . 1 77 Beale Street , O San Francisco, California 94106
Dear Mr. Clarta:
On June 15 and 15, 1992, NRC senior managers met to evaluate the nuclear i safety performance of operating reacters, feel facilities, and other caterials O licensees. The NDC conducts this meeting semiannually to detemine if the safety perfernance of the various licensees exhibits sufficient weaknesses to warrant increased NRC attention. In addition, at this meeting, senior managers icentify specific plants that have deconstrated a level of safety ' performance that deserves formal NRC reccanition At the June 1992 Senior i Hanagerent Meeting, the Diablo Canyon nuc$ ear pow.er plant was identified as ' having achieved a high level of safety performance and met criteria fori O recognitten of its perfor ance. I am pleased to note that Diablo Canyen has again been identified as a good performer, and I consider this a noteworthy accceplish:snt. In identifying such plants, senior managers perform an evaluation of perfemance in cany areas including operational safety, self-assessment, O pr:blem resolution, ar.d plant management crganization and oversignt. , NRC senier managn.ent recognizes that managecent involvement in all phases of plant eceratten, the dedicated and knowledgeable staff that supports- ' plant ' activities, and the cor.raltment to safety throughout the organization are necessary to achieve the level of performance demonstrated by the Diablo O canyon nuclear po-er plant. We comend you and your staff for acMeving a high level of safety performance. Your achievement is the result of dedicated efferts frca your staff and is a positive example to the industry. , O O ' O
Richard A Clarke , i Y-n The greatest chalienge that you now face is to maintain this level of It perforuance and not to rest on past achievements. Continued stanagemert involvenent ar.d support, and dedicated efforts from your staff to ider.tify .. and promptly correct problems, are necessary for you to continue to meet i y this difficult challenge. l l Sincerely, 4
- f ,
mes ky r 3 xecutive Director
-for Operations cc: See rext page )
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D_. l 4 i L.'- ___.__;-_._m_ v
or csc D /s 'e UNITED STATES
! " s ,, 7 ,' NUCLEAR REGULATORY COMMISSION
*- r -t WAsHIN G ton, D. C. 205s5 7, 1; February 3,1992 a -[/f No.
\ '..Docket
.+ '! 50-275
) Docket ho. 50-323 Mr. Richard A. Clarke Chair 4 nan of the Board and Chief Executive Officer 3 Pacific Gas and Electric Cou4pany 77 Beale Street San Francisco, California 94105
Dear Mr. Cische:
On January 14 and 15,1992, NRC Senior gianagers met to evaluate the nuclear safety performence of operating reactors, fuel f acilities, and other u.aterials licensees. The hRC conducts this meeting semiannually to deter.nine if the safety perfora.ance of the various licensees exhibits sufficient weaknesses to warrant increased NRC attention. In addition, at this meeting, senior managers , identify specific plants that have deu.onstrated
- level of safety performance
- that deserves formal NRC recognition. At the January 1992 Senior Managenent Meetirig, the Diablo Canyon nuclear power plant was identified as having achiev-ed a hign level of safety performance and met criteria for recognition of its perfora.an ce.
Ir. identifying such plants, senior managers perform an evaluation of performance 8 in snany areas including operational saf ety, self-assessment, problem resolution, and plant manage.nent organization and oversight. NRC senior saanagement recognizes that s..anagewnt involveuent in all phases of plant operation, the dedicated and knowledgeable staff that supports plant activities, and the couanitoent to safety throughout the organization are 8 necessary to achieve the level of performance demonstrated by the Diablo Canyon nuclear power plant. We consend you and your staff for achieving a high level of safety perfor4aance. Your achievenent is the result of dedicated efforts from your staf f anc is a positive exaa. pie to the industry. The greatest challenge that you now face is to maintain this level of perform-4 ance and not to rest on past achievenients. Continued manages.ent involveit.ent and support, and decicated efforts froni your staff to identify and promptly correct problems, are necessary for you to continue to meet this difficult challenge. Sincerely,
/
.nes M.
xecutive Director
- for Operations cc: See next page 9
4 .. O i EXHIBIT 20, i SALP REPORT
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O , O O O O O t O k O O
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j ff \, UNITED STATES i l 7 NUCLEAR REGULATORY COMMISSION g ,E REGION V 1450 MAR 1A1.ANE o WALNUT CREEK CAUFORNIA 94596-5368 ; g
. February 12, 1993 ;
Docket Nos. 50-275 and 50-323 i Pacific Gas and Electric Company ' y Nuclear Power Generation, B14A 17 Beale Street, Room 1451 ' P. O. Box 770000 , San Francisco, California 94177 Attention: Mr. G M. Rueger, Senior Vice President and General Manager ) Nuclear Power Generation Business Unit ,
Subject:
Systematic Assessment of Licensee Performance (SALP) ! Report Nos. 50-275/92-34 and 50-323/92-34 i The NRC's Systematic Assessment of Licensee Performance (SALP) Board has com- , 3 pleted its periodic evaluation of the performance of your Diablo Canyon Nuclear ; Plant for the period July 1,1991 through December 31, 1992. An Overview is provided as Section II.A of the enclosed Initial SALP Report. The performance of Diablo Canyon was evaluated in the functional areas of Plant l' Operations, Radiological Controls, Maintenance / Surveillance, Emergency g Preparedness, Security, Engineering / Technical Support, and Safety Assessment / ' Quality Verification. The criteria used in conducting this assessment and the SALP Board's evaluation of vour performance in these functional areas are outlined in NRC Manual Chapter 0516, " Systematic Assessment of Licensee , Performance," dated September 28, 1990. ! i 3 Overall, the SALP Board found the performance of licensed activities at Diablo Canyon to be superior. The Security functional area was assessed by the SALP r Board to be Category 2, improving, with all other functional areas evaluated Category 1. Based on the Board's assessment, we wish to recognizc the overall ' performance of your management and staff in exhibiting an attitude clearly-directed toward safe facility operation. 'l e' . Based upon discussions with your staff, a management meeting to discuss the results of the SALP Board's assessment has been scheduled for February 25, i 1993. Arrangements for this meeting will be discussed further with your staff in the near future, e- In that no functional area was assessed as Category 3, a written response to the enclosed initial SALP report is.not required. _ However, you may submit , comments on the enclosed report, if desired, within'30 days after the February 25 meeting. In accordance with Section 2.790 of the NRC's " Rules of Practice", Part 2, p Title 10, Code of Federal Regulations, a copy of this letter and the enclosed Initial SALP report will be placed in the NRC's Public Document Room. ; i
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n , l Should you have any questions concerning the SALP report, we will be pleased to discuss them with you. Since _ly, --. ff l 2 hn B. Martin Regional Administrator O
Enclosure:
Initial SALP Report Nos. 50-275/92-34 50-3.'3/92-34 cc: w/ enclosure: J. A. Sexton, PG&E
.O J. D. Townsend, Vice President / Plant Manager, PG&E C. W. Warner, Esq., Attorney O. A. Taggart, Director, Quality Support, PG&E B. Thomas, News Services, PG&E T. L. Grebel, Regulatory Compliance Supervisor, PG&E
~O State of California (Gordon K. Van Vleck) Bob Hendrix, County Administrator Sandra Silver INP0 1
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I J j U. S. NUCLEAR REGULATORY COMMISS10f1 REGION V SYSTEMATIC ASSESSMEf1T OF LICENSEE PERFORMANCE SALP BOARD REPORT Hos. 50-275/92-34 and 50-323/92-34 O PACIFIC GAS & ELECTRIC COMPANY g DIABLO CANYON POWER PLANT JULY I, 1991 THROUGH DECEMBER 31, 1992 D D e
i [) l 1 l-O i TABLE OF CONTENTS Pace O ! I. Introduction . . . . . . . . . . . . . . . . . . . . . . . . 1 l II. Summary of Results A. Overview. . . . . . . . . . . . . . . . . . . . . . . . 2 ()- B. Results of Board Assessment . . . . . . . . . . . . . . . 2 III. Performance Analysis A. Pl ant Operati on s . . . . . . . . . . . . . . . . . . . . 3 B. Radiological Control s . . . . . . . . . . . . . . . . . 5 () C. Maintenance / Surveillance. . . . . . . . . . . . . . . . 7 D. Emergency Preparedness. . . . . . . . . . . . . . . . . 9 E. Security. . . . . . . . . . . . . . . . . . . . . . . . 11 l F. Engineering / Technical Support . . . . . . . . . . . . . 13
- G. Safety Assessment / Quality Verification. . . . . . . . . 16
() IV. Supporting Data and Summaries A. Licensee Activities . . . . . . . . . . . . . . . . . . 19 , B. Inspection Activities . . . . . . . . . . . . . . . . . 20 C. Enforcement Activity ................. 21 D. Confirmatory Action Letters . . . . . . . . . . . . . . 21 () E. Licensee Event Reports. . . . . . . . . . . . . . . . . 21 l l () , t . r 3 +
O N I. INTRODUCTION The Systematic Assessment of Licensee Performance (SALP) is an integrated f C) HRC staff effort to collect available observations and data on a periodic basis and to evaluate licensee performance based on this information. The ' program is supplemental to normal regulatory processes used to ensure com-pliance with NRC rules and regulations. It is intended to be sufficiently i diagnostic to provide a rational basis for allocating NRC resources and to : provide meaningful feedback to licensee management regarding the NRC's l IC) assessment of their facility's performance in each functional area. 2 An NRC SALP Board, composed of the members listed below, met in the Region V office on January 21, 1993, to review abservations and data on the licensee's performance in accordance with NRC Manual Chapter 0516,
" Systematic Assessment of Licensee Performance."
'.O This report is the NRC's assessment of the licensee's safety performance at Diablo Canyon Power Plant for the period July 1,1991 through
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December 31, 1992. < The SALP Board meeting for Diablo Canyon was attended by: ,
- O Votina Members
~ K. Perkins, Director, Division of Reactor Safety and Projects, RV (SALP Board Chairman) 4 M. Virgilio, Assistant Director for Region IV & V Reactors, Division of C) Reactor Projects III, IV, V, NRR R. Scarano, Director, Division of Radiation Safety and Safeguards, RV L. Miller, Chief, Reactor Safety Branch, RV _ P. Johnson, Chief, Reactor Projects'Section 1, RV < S. Peterson, Project Manager, NRR M. Miller, Senior Resident Inspector, Diablo Canyon , Cs Other Attendees J. Reese, Chief, Facilities Radiological Protection Branch, RV R. Pate, Chief, Safeguards, Emergency Preparedness and Non-Power Reactor Branch, RV C) D. Kirsch, Technical Assistant, RV P. Morrill, Chief, Operations Section, RV W. Ang, Chief. Engineering Section, RV P. Narbut, Team Leader, RV D. Schuster, Safeguards Inspector, RV A. McQueen, Emergency Preparedness Analyst, RV i) C L. Norderhaug, Safeguards Inspector, RV L. Coblentz, Radiation Specialist, RV D. Corporandy, Project Inspector, RV C. Myers, Reactor Inspector, RV
;C)
J C)
O II.
SUMMARY
OF RESULTS A. Overview - O The licensee's overall performance level during this assessment period was good or superior in all areas. Examples of superior performance were demonstrated by relatively event-free operation, low occupational radiation exposure, awareness and training of personnel to minimize safety risks during outages, prompt and aggressive response to indications of O cracking in feedwater piping nozzles, and aggressive and well focused insight into performance weaknesses by the Onsite Safety Review Group. The strength; observed 'n the Operations, Radiological Controls, Engineering / Technical Support, Emergency Preparedness, and Safety Assess-ment / Quality Verification functional areas resulted in these areas being O rated as Category 1. The board noted in the functional area of tiaintenance/ Surveillance that early in the SALP period there were a few problems involving prompt problem identification and resolution, and engineering involvement in maintenance issues. The board concluded, however, that the overall performance was superior based on strong corrective actions and very high quality performance throughout the o remainder of the period. While strengths were noted in the security area, security management did not appear to have conducted an adequately broad examination of their activities to assure a high standard of performance throughout the organization. The board discussed HRC-identified problems at length, e particularly in compzrison with the high level of performance seen in most of the security organization. While corrective actions were taken for specific problems 'hatified by the NRC, it appeared that the requirements of the security organization had not been implemented with a consistent level of assurance of quality. Although management appeared to have , corrected weaknesses noted during the previous SALP period, weak manage-D ment involvement in maintaining high quality in all security program areas detracted from otherwise su;'rior performance in this area. B. Results of Board Assessment Overall, the SALP Board found the performance of NRC licensed activities e to be very effective and directed toward safe operation of Diablo Canyon. i The SALP Board has made specific recommendations in most functional areas ; for licensee management consideration. The results of the Board's assess-ment of the licensee's performance in each functional area, along with the ; results from the previous period, are as follows: l
@ Rating Rating Last This Functional Area Period Trend Period Trend A. Plant Operations 1 1 P. Radiological Controls 1 1 0 C. Maintenance / Surveillance 2 1 D. Emergency Preparedness 2 1 E. Security 2 Improving 2 Improving 9
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_O ! l I F. Engineering / Technical 1 1 i Support i G. Safety Assessment / 1 1 1 O Quality Verification l III. PERFORMANCE ANALYSIS The following is the Board's assessment of the licensee's performance in each of the functional areas, along with the Board's conclusion for each 'O area and its recommendations with respect to licensee actions and management emphasis. __ A. flant Operations j i
- 1. Analysis j During the assessment period, the resident inspectors conducted frequent inspections involving observation of operations activities.
Some engineering section and project inspector inspections also evaluated operations activities. Review of operations activities ! accounted for about 34 percent of the total inspection effort. The last SALP assessment rated the licensee's performance in this area Category 1. Strengths were noted in relatively event-free operation, and in superior operator response to plant transients. Operations exhibited significant strength in conservative operational , decisions. The previous SALP Board also noted weaknesses in occa- !
- O sional lack of timeliness in identifying and resolving problems and !
in issuing operability determir ations. During this SALP assessment period, the licensee continued to show t superior performance in this area. Strengths were observed in the - general high quality of the Operations staff's performance, and in
- O relatively event-free and uncomplicated operations. Management involvement has been frequent and probing, assuring timeliness in ;
identifying and resolving problems and in making operability deter-minations. Operations management has set high performance standards i which have usually been met or exceeded. 10 Recovery from each event, regardless of cause, and the subsequent root cause investigation indicated significant strengths. cThis was ' due in part to a high level of skill and sense of ownership among the Operations staff, and to intensive management involvement at all levels of the organization.
- o During operations at power and during outages, the Operations staff showed strong awareness of overall plant safety system availability and the significance of evolutions relative' to the risk to the plant.
i This appeared to have been a direct result of aggressive management j connitment to plant safety and risk reductions. The licensee developed and implemented a comprehensive and effective outage plan .o that appropriately considered risk associated with plant shutdown evolutions. Operations staffing levels appeared to be appropriate, and operations staff qualifications were strong.
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J Other examples of significant strengths were as follows: Active Operations involvement with maintenance crews near sensitive equipment helped to avert events. Documentation of operability determinations was strong, timely and consistent. Also, a very low threshold was established for the level of equipment degradation which required an operability evaluation. D + Operations simulator training was challenging and effective, and critiques appeared to be appropriately critical and probing. Toward the end of the SALP period, Operations personnel were progressively more alert to anomalous plant conditions. For D example, an operator's observation and followup of a failed fastener resulted in identification and repair of a degraded neutral connector to a main transformer, potentially averting a plant trip. During this SALP period, two severity Level IV violations occurred in this area. One was a repeat violation, for operation in Modes 2 and D 3 with one of two reactor cavity sump wide range level channels inoperable. The other violation involved inadequate instructions to operators for avoiding excessive piping vibration on loss of speed control to the positive displacement charging pump. Neither had an impact on safe plant operation, and each of these instances was promptly corrected. During the first part of the SALP period, a few Licensee Event Reports (LERs) were issued as a result of personnel errors. Although this was not an unusually high rate, the concern was that it indi-cated an increasing trend. The personnel error rate was reduced D later in the assessment period as a result of strong management invol vement. Four isolated instances of minor weakness were observed, either in following procedures or in coordination with other groups. The most significant involved an inadvertent chemical spill, which generated 8 noxious fumes and prompted declaration of an Unusual Event. Another instance occurred as a result of unclear procedures, which allowed a condenser vacuum pump to be started before its seal water isolation valve was opened. This uitimately resulted in a reactor trip. In each of these cases, root cause evaluation and corrective actions were immediate and appeared appropriate. 8 In summary, the performance of Operations has been strong, and has continued to improve. Weaknesses have been minor, isolated and infrequent, and have been corrected promptly and appropriately.
- 2. Performance Ratinq Performance Assessment: Category 1 O
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- 3. Recomendations None b B. Radiolooical Controls
- 1. Analysis Radiological controls inspections during this SALP period found that !
y the licensee continued to be aggressive in assuring quality. Radio- I active effluents continued to decrease, and occupational dose was ;
. reduced in 1992 despite a demanding outage schedule. A continued l strength was the licensee's innovative approaches to improving :
measures for personnel radiation protection. Minor weaknesses ) identified, related to radiological porting and labeling..were solved l y rapidly and thoroughly. Regional inspectors examining this : functional area contributed approximately 5 percent of the total inspection effort during this assessment period. [ The licensee's radiological controls performance during the previous ! SALP period was rated Category 1. The previous SALP Board recom- l mended that management continue _to fully support site and corporate k initiatives for improving performance. The board also recommended ; added emphasis toward correcting minor weaknesses in controlling personnel contaminations, reducing the backlog of non-Technical i Specification radiation monitoring equipment needing calibration, and ; training dosimetry clerks and radwaste handlers. l ) During this assessment period, management continued to be proactive in assuring quality. The ALARA awareness program, established to j reward outstanding outage performance, continued to be an effective ' incentive toward meeting rigorous ALARA goals. The 1991 average i occupational dose per reactor was 273 person-rem, and for 1992 was l 214 person-rem. Liquid effluents continued to decrease. Gaseous )- effluents were also maintained at a small fraction of the Technical Specification limits. Management support was evident in the elaborate remote monitoring capabilities used to support steam generator shot peening and eddy i ) current testing during the IR5 outage. Use of this equipment ! significantly reduced both the dose received and the radiological risk involved in conducting several complicated, high-dose tasks at once. In addition, corporate involvement and support was evident in , continuing efforts associated with a major upgrade of radiation and i effluent monitoring equipment. ) The licensee's approach to resolving technica' issues was conserva- ! tive and timely, and demonstrated a clear unde. standing of the issues involved. -In September 1992, the licensee voluntarily made a presen-tation to members of the NRC Region V staff concerning the status of radiation monitoring system upgrades. Detailed alternate monitoring y methods had.been analyzed, for use during interim periods while sys-r tem upgrades were being performed, to ensure proper monitor ranges, efficiencies, and sensitivity to airborne radioactivity. Technical )
e improvements were observed in licensee programs for radwaste classi-fication, the process control program, and radiological environmental monitoring. Technically sound judgment was also in evidence in the
- licensee's radiological controls preparations for potential high-dose outage tasks, such as steam generator shot peening, steam generator eddy current inspection, and core barrel inspection.
Licensee management support of training was demonstrated by the extensive efforts made in mock-up training prior to the IR5 outage. The steam generator cock-up included a fully operational shot-peening 3 apparatus. One weakness was observed involving failure to thoroughly train eddy current testing personnel on the impact that shot peening would have on steem generator airborne radioactivity hazards. A Severity Level IV violation was cited for the resulting hazard. The licensee took prompt corrective action to resolve this weakness. 3 The licensee's other training practices continued to exhibit excellence. Training and qualification programs made a positive contribution to the understanding of radiological controls issues and adherence to procedures. Staff members were kept abreast of industry knowledge and development through extensive participation in offsite J, owners' group meetings, Electric Power Research Institute (EPRI) conferences, and other opportunities for offsite involvement. An improvement was noticed in the licensee's training of radwaste handlers. Training on the new 10 CFR 20 requirements also continued for appropriate personnel. 3 The licensee's site and corporate radiological controls and chemistry groups continued to be well staffed. Key positions were generally filled on a p,iority basis. Authorities and responsibilities, both in the chemistry and radiation protection organizations, were well defined, and resulted in clear communications both within the groups and with other site orgsnizations. 3 One voluntary Licensee Event Report (LER) was submitted relevant to radiological controls during this assessment period. The LER dealt with overexposures received by contract radiographers, due to personnel error by the radiographers while performing radiography on the licensee's site. Three Severity Level IV violations were , identifi;J in this functional area. Two resulted from inadequate J posting and labeling, and one involved the failure to implement procedures to control airborne radioactivity from steam generator work. tieither the violations nor the LER indicated a programmatic breakdown of the radiation protection program. The licensee's root cause analyses and correttive actions were prompt and were ,J effectively implemented.
- 2. Perforrance Ratina Performance Assessment: Category 1
- 3. Poard Reco endation e
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g . C. Maintenance / Surveillance
- 1. Analysis 0:
During the assessment period, the resident inspectors conducted frequent inspections which included observation of maintenance and surveillance activities. Engineering inspections also evaluated ' i maintenance and surveillance programs. Review of maintenance and surveillance activities accounted for about 10 percent of the total l O Diablo Canyon inspection effort. { The last SALP assessment rated the licensee's performance in this I area Category 2. Strengths were noted in the initiation of a program ! for trending safety equipment out-of-service time, and in the use of 4 probabilistic risk assessment to evaluate preventive maintenance O programs. Weaknesses were noted in a lack of management aggressive- : ness in dealing with problem areas; occasional failure to follow : procedures, resulting in safety significant events; and a tendency for personnel errors due to lack of self-verification. The licensee was encouraged to involve management in timely problem identification -, and root cause investigation, and to continue to support industry : initiatives. O During this assessment period, the licensee generally displayed improved performance in this area. Virtually trouble-free plant > operation evidenced a high quality of maintenance work in that no plant events and almost no equipment failures occurred as a result i O of improper maintenance. Strengths were observed in the general. high quality of maintenance and surveillance work. Additionally, a , high level of management involvement in scheduling and pianning ; maintenance and surveillance work maximized safety system avail-ability from a probabilistic risk standpoint, both at power and during outages. This resulted in a ' considerable benefit to plant O safety. Noteworthy strengths were as follows: ' Outaae Manaaement: The management of outages was marked by an overriding understanding and emphasis of the probabilistic risk O. of each job and evolution. Work crews and planners were trained i and aware of the safety significance of the jobs and systems on ' which they worked at every stage of the outage. Oualifications: The training and qualification program for Maintenance personnel was strong. Well maintained training ; O facilities and a dedicated training staff were significant i factors in good performance, as was the sense of ownership shown : by Maintenance personnel. ' Plant Safety: Maintenance personnel were trained and informed i regarding overall plant safety system availability and the ; O. significance of their individually assigned work relative to its risk to the plant. , i O .
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- prioritization of Work:~ Outstanding work items were well prioritized, with safety-significant issues given high priority.
The backlog of non-outage safety related work items was low. 'O . Root Cause Investications: The routine involvement and leadership shown by the Plant Maintenance staff in root cause investigations was a significant strength, as was the routine integration of the Maintenance, Operations, and Engineering staffs in maintenance and surveillance operations. O . Reduction of Personnel Errors: A relatively high number of personnel errors were observed at the beginning of the SALP period. Several of these errors resulted in conditions wh'ch prompted a Licensee Event Report or Non-conformance Report. This number was reduced by about half during the remainder of the period due to a high level of management involvement O throughout the organization. Response to Problems: Overall, the maintenance staff improved their response to problems by identifying, analyzing and cor-recting maintenance and surveillance problems promptly. This O represented an improvement over the last assessment period. Examples of this improvement were the identification and correc-tion of an incorrect reactor coolant system leakage surveil-lance; prompt, in-depth evaluation and compensatory action for problems with auxiliary feedwater pump steam admission valve FCV-95; and improvement of the clarity of some instrumentation O and control surveillances. Four Level IV violations were cited in this area, involving improper maintenance of containment fan cooler unit (CFCU) backdraft dampers, failure to perform a containment airlock surveillance, failure to identify inconsistencies in a pump vibration measurement procedure by O writing an action request, and improper rigging of a cask. In some cases, as illustrated by the inoperable containment fan cooler unit backdraft dampers, Engineering involvement should have been more timely. Improper maintenance of CFCU dampers was significant in that the dampers were not functional, and only after additional analysis did the licensee determine that the CFCUs had been operable despite O the improper maintenance. These concerns appear to have been iso-lated, although the CFCU issue was potentially significant'to safety. Other weaknesses were also observed. One example was the improper tightening of setscrews on some motor operated valve actuators, resulting in a common mode failure vulnerability. Additional, less O significant weaknesses were observed. Most were identified by the licensee immediately upon occurrence. Management involvement was effective, and identified problems were promptly and appropriately corrected. Most of these examples occurred early in the SALP period. Since that time, significant improvement has been noted.
- 2. Performance Ratino O
Performance Assessment: Category 1 0
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- 3. Board Recommendations The Board encourages continued intrusive Engineering involvement in i O maintenance and surveillance issues, and focused management '
involvement to ensure continued low levels of personnel errors. 3 D. Emeroency Preparedness j
- 1. Analysis i O
Two routine emergency preparedness (EP) inspections and two annual emergency exercise team inspections were conducted during this assessment period. Review of the EP program accounted for approxi- ' mately 6 percent of the Diablo Canyon inspection effort. A strength , identified during the current assessment period was in making timely O and appropriate classifications during most actual emergency events, exercises and drills. A weakness was r.oted regarding the making of protective action recommendations (PARS) to offsite agencies during the 1991 annual emergency exercise. Generally, licensee performance in the EP area appears to have improved over the assessment period. ; 'O The licensee's EP performance in the last SALP cycle was rated Category 2. The SALP board at that time indicahi several recommen-dations: that management ensure the establishment and implementation of an effective corrective action plan for drill and exercise ' findings; that licensee management evaluate the adequacy of classroom : training provided to emergency response personnel and ensure that ! O personnel are given an adequate number of opportunities to practice their assigned tasks during periodic drills; that the additional dose assessment training provided to Control Room personnel continue; that r the need to adhere to radiation protection procedures under simulated emergency conditions also be stressed during classroom training and drills; that administrative procedures be enhanced to ensure that O drills and exercises consistently meet emergency plan requirements; and that simulation of sample collection during drills and exercises ; be avoided to enhance realism and increase the training value. During the current assessment period, licensee management appeared actively involved in EP activities and demonstrated support by , o providing the necessary resources to the EP staff. Management took interest in correcting problems and responding to NRC findings which indicated a need for corrective action. During the assessment i period, the licensee worked closely with the state, local county governments, and FEMA in resolving issues in offsite preparedness planning. Each of the recommendations from the previous SALP Board o was addressed by the licensev during this assessment period. Cor- : rective actions were evaluated by the NRC during routine inspections and observation of the two annual exercises, and improvement was , noted in each area. Dose assessment and projection, in particular, were noted as strengths in response facilities during the 1991 and 1992 annual emergency exercises. O Licensee management's approach to the resolution of technical issues appeared generally timely and thorough. During the assessment O
1 period, the licensee significantly upgraded the emergency urning siren system. The new primary system was completed, tested and l turned over to the county with 100 percent activation in September
- O 1992. The upgrade provided several new capabilities such as an activation system which allows selective sounding of individual or groups of sirens as opposed to the entire system, and a siren feedback system which provides input to the county when a " runaway siren" sounds without intended activation.
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- O One EP exercise weakness was identified during the 1991 annual emergency exercise. The licensee's system for providing Protective i Action Recommendations (PARS) appeared excessively complicated and caused delay in the issuance of PARS. The system was not based solely on plant conditions as would be appropriate, but included coordination of PARS with offsite agencies. This delayed and O possibly biased the licensee's decision making. The appropriate emergency plan implementing procedure (EPIP) was revised to insure licensee independence in PAR decision making and was validated through training, drills and exercises. The system appeared to have l been effectively implemented during the 1992 annual exercise.
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- O There were no enforcement actions in the EP area during the assess-cent period. Notifications to the NRC and offsite agencies were i consistent with regulatory requirements. The licensee reported nine i unusual events to the NRC during the assessment period, including i three earthquakes detected at the site. The other events were a
! reactor coolant system (RCS) leak, a grass and brush fire near the lO site, a turbine stop valve failure, a sulfuric acid spill, radiation overexposure of two contractor employees, and a temporary loss of communications with the California Office of Emergency Services. All events appear to have been properly identified and analyzed in accordance with regulatory requirements. EP staffing was an apparent strength, and staff members appeared 9 conscientious toward accomplishment of their assigned duties. No significant changes occurred in the composition of the emergency response organization (ERO) during the :tssessment period. The licensee had a system to ensure that new ERO personnel were properly trained prior to assignment to emergency organization positions. O EP staff and emergency response positions were clearly identified; authorities and responsibilities appeared clearly defined; and key positions were filled as appropriate. Decision-making authority appeared properly delegated to ensure quick identification of and response to problems and changes. Emergency facilities continued to
- be appropriately maintained and appeared ready for rapid activation.
The licensee provided adequate levels of dedicated staff to implement
- O the programs and to interact appropriately with offsite agencies.
During the assessment period, the licensee implemented what appeared a to be a substantial change to the EP training program. Previously, the site and corporate headquarters had separate EP training programs i and responsibilities. The company-wide responsibility for EP
-O training management and accomplishment was shifted entirely to the site. A system was established to ensure that required training is 4
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conducted and that training due dates are not exceeded, by linking accomplishment of EP training requirements to unescorted access privileges. To supplement and reinforce routine annual training, a O program of monthly integrated drills was conducted.
- 2. Performance Ratina Performance Assessment: Category 1 O 3. Board Recommendations The licensee should strive to maintain a consistent level of management oversight to continue and improve on the program quality achieved during this assessment period.
O E. Security
- 1. Analysis During this SALP period, approximately 4 percent of direct inspection effort was applied to the licensee's physical security and fitness O for duty programs. In addition to region based _ inspections, the resident inspectors also monitored implementation of this program as part of their routine inspection activities.
The previous SALP report rated the licensee's performance Category 2, Improving, for Security. Primary weaknesses identified in that O report focused on personnel access control to vital areas and failures of compensatory security measures. These weaknesses were significantly reduced during the current assessment period. In the-previous SALP report, the Board encouraged the licensee- to resolve a longstanding weakness in the CCTV alarm assessment capability, initially identified during a 1986 Regulatory Effectiveness Review. O. A significant equipment upgrade to incorporate a video capture system was installed during the current assessment period and has-largely eliminated this weakness. Some minor limitations remain with the video capture system and the licensee is actively exploring further equipment and/or procedural improvements. O The licensee's performance in the areas of physical security and fitness for duty appeared, on the basis of inspections conducted, event reports, and other observations and analysis, to be good in all assessment areas. Both program strengths (vital area barriers and armed response) and weaknesses (effectiveness of the audit program and the number of pending requests for security equipment modifica-O tion or maintenance) have been noted during the assessment period. Principal strengths in the licensee's security and fitness for duty programs included control of access to vital areas, the use of roving patrols dedicated to armed response (carrying carbines or shotguns, as appropriate, as well as side arms), and the availability of O Employee Assistance Programs for contractor employees. A major program upgrade to establish a search train at the intake structure protected area was completed during the assessment period. 'O
! A principal program weakne~ss was noted concerning the number of cited i and non-cited violations that could have been identified and 1 corrected by a stronger audit program. This indicated a need for 1 0 increhsed management attention to upgrade the audit program and make ! it more effective. I i Although the licensee has been aware of a large backlog of action j requests for maintenance or modification to security equipment, i little progress was seen in addressing this concern. This was also
- O seen as a program weakness. Discounting action requests of an i administrative nature or otherwise having no direct effect on
! security activities, approximately 90 requests were identified as i being more than 90 days old, nearly half of which wure more than a j year old. A more effective audit program could have identified this { weakness. This further demonstrated the need for increased lO management attention. i l One Licensee Event Report dealing with safeguards matters (requiring i prompt reporting pursuant to 10 CFR 73.71) was issued and adequately i resolved during the SALP period. This report dealt with failure of I circuit boards in the alarm annunciation system, and prompted a full 1 0 replacement of the obsolescent components which is scheduled to be completed in the near future. However, technical issues discussed in two reports issued in January and December 1990, and dealing with backup power to communications equipment and vital equipment pro-tected by compensatory measures, respectively, remain to be resolved. O Enforcement actions during the assessment period included four viola-tions, which were resolved by appropriate corrective actions: one each related to access control at the main and intake structure protected areas, one violation related to protection of Safeguards Information, and one violation related to urinalysis testing of fitness for duty program personnel. Two weaknesses related to O fitness for duty and four non-cited violations dealing with vital area access control, communications, lighting, and protection of Safeguards Information were promptly corrected by the licensee. The licensee's loggable safeguards events were promptly and completely reviewed and reported as required. The root cause and O trend analyses of these events determined that most of the events were related to aging equipment scheduled for replacement by major hardware upgrades then underway. The frequency of occurrence has exhibited a decreasing trend as those projects have been completed. Licensee staffirg appeared effective in most areas, although the O identified long delays in resolving security related action requests may indicate a need for additional senior management support. Key positions have been identified and responsibilitie., are well defined. Decision making authority appears properly assigned to ensure prompt identification and response to program challenges. During the current assessment period, management implemented team development O workshops for all staff. This training showed significant promise in improving staff communications and cohesiveness. O
_ 13 _ O-The licensee's guard training and qualification program was well defined and implemented with dedicated resources. During this SALP period, the licensee initiated sophisticated contingency drills l ,O incorporating diversionary tactics and covert penetrations.
- 2. Performance Ratino Performance assessment: Category 2, Improving ;
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- 3. Board Recommendation
.O l Licensee management is encouraged to more effectively identify and address weak areas. More attention shoLld be given to improving the - effectiveness of the audit program and to reducing the number and age of outstanding maintenance requests.
- O F. Enoineerina/ Technical Support
- 1. Analysis During the assessment period, NRC regional and Headquarters inspectors conducted a total of twelve inspections. Two of these O inspections were team inspections which addressed motor operated valves and shutdown risk management. The other inspections involved facility modifications, design changes, inservice inspection and testing, erosion / corrosion monitoring, eddy current testing of steam generator tubes, and procurement of a new emergency diesel generator.
The resident and project inspectors also conducted inspections in U this area. Review of Engineering and Technical Support activities accounted for approximately 15 percent of the totcl Diablo Canyon inspection effort. The last SALP assessment rated the licensee's performance in this ,O functional area Category 1. Improvements were recognized in Engi-neering involvement in plant operations and modification work, design basis reviews, setpoint reverification, vendor interface, personnel qualification and training. A particular strength was found in the commercial grade dedication program. Some weaknesses were noted in incomplete technical work and untimely identification and resolution O of problems due to a weak sense of ownership of plant problems. The Board recommended that the licensee provide emphasis on early identification, effective engineering involvement, and timely and thorough correction of plant problems. The licensee was encouraged-to continue building a strong interface between corporate and plant engineering groups, with corporate engineering taking a leadership .O role in the resolution of plant problems. Continuation of inns.3tive l corporate enginearing training programs was specifically encouraged. l During this SALP assessment period, the licensee showed continued l , high quality performance in this functional area. Strengths were . ( observed in a generally aggressive and thorough engineering attitude
- O in resolving technical problems, an extensive erosion / corrosion conitoring program, eddy current testing of steam generator tubes, 7 assessment of probabilistic risks to the shutdown plant, and overall t
e I
1 O ~ I4 - i i i emineering invcivement in plant operational activities. The NRR l staff observed excellent quality in the technical content and presentation of licensee submittals, which included documents in o support of lice 1se amendment requests, corrective actions regarding operations and Licensee Event Reports, and responses to NRC bulletins and generic letters. Improvements were observed in timely problem identification, engineering involvement, and problem ownership. Minor weaknesses were noted related to procurement of the new l emergency diesel generator (EDG) and certain inservice inspection and ; o testing activities. l Engineering involvement in resolving safety issues was generally timely. The mosi significant exception was Engineering's assessment of problems with containment fan cooler unit (CFCU) backdraft dampers t in early 1992. Additionally, resolution of Regulatory Guide 1.97 ; O issues was delayed by inadequate tracking of engineering actions, but the licensee later identified this weakness and pursued resolution in ; an aggressive manner. Substantial improvement was displayed later in the SALP period in Engineering's timely resolution of CFCU damper - blade cracking. o Strong Engineering performance and initiative were evidenced in Engineering's evaluation of setscrew loosening on motor operated valve (MOV) actuators and the licensee's decision to examine the steam generator feedwater nozzles in response to problems observed at another facility. The feedwater nozzle examinations were extensive and used state-of-the-art techniques. The Engineering staff's O assessment of crack indications in both feedwater piping and in a ' safety injection tank penetration resulted in a conservative decision , to replace affected piping segments. ; Proactive Engineering involvement was observed in the development of an extensive erosion / corrosion monitoring program. Despite extensive ' O involvement with the industry in the development of predictive analy- . tical computer programs, poor correlation between the quantitative t predictions and measured wear rates had been experienced by the litersee. The licensee's program exhibited a defense-in-depth ! approach to compensate for recognized limitations in the state of the art. Although a program weakness in the measurement of pipe wall O thickness was noted, strong engineering ownership of the program compensated for this minor weakness. l Throughout the SALP review period, the licensee demonstrated an aggressive engineering attitude in technical problem resolution. For ' example, the licensee instituted a supplemental program that is the o first surveillance program in a U. S. commercially coerated reactor vessel to investigate the effect of annealing and reirradiation on its reactor vessel beltline materials. 3 Another example of aggressive engineering was resolution of the long-term seismic program. The licensee performed a detailed analysis to 0-demonstrate that adequate seismic margins exist for the structures and equipment which could be affected by increased ground motion in certain frequency ranges at the Diablo Canyon site. O
I , The licensee also developed and implemented an effective outage risk assessment plan which was found superior to other plants which were . inspected. The technical support provided for the outage risk I 3 assessment plan was excellent. l The engineering program developed for eddy current testing (ECT) of ! steam generator tubes was observed to be a high quality program incorporating current technology and industry guidance. However, a weakness was noted in that engineering guidelines for ECT data 3 analysis and defect acceptance criteria, although adequate, were not controlled through the use of formal plant procedures. Specific strengths noted in engineering activities were as follows:
- The motor operated valve (MOV) program was found to be 9-aggressive and conducted in a well integrated manner. A minor weakness was identified in the lack of timely determination of operability following testing, due to the complexity of the engineering evaluation required to evaluate the test data.
A strong safety perspective was evident in the development of 9' engineering programs to resolve emerging technical issues. The programs were implemented with priority on safety significance. Quality assurance involvement was evident in the implementation of engineering programs. 3
- Design change packages for the installation of a new emergency diesel generator were generally thorough and complete, although minor housekeeping and cleanliness deficiencies were observed.
Four Severity Level IV violations, one Level V violation and one non-cited violation were identified. The violations were minor in nature 3 and did not evidence programmatic breakdowns. The low number of engineers and lack of clear goals for the plant's System Engineering staff was a concern earlier in the SALP period. The licensee has since increased the staff and clarified the goals for this group, and some improvement has been observed. O The commercial grade dedication of the sixth emergency diesel generator (EDG) presented unique challenges to the licensee's engineering and procurement activities. The Region and NRR Vendor Branch identified weaknesses in the quality of the procurement and ! commercial grade dedication of the new emergency diesel generator. 3 However, the licensee's root cause investigation was candid and thorough. Also, although most problems encountered during testing of the sixth EDG were found to have been documented and resolved, the test program did not require formal documentation of problems. This weakness was promptly corrected after identification by the NRC. e Inservice inspection and testing activities were found to comply with approved programs. Observed deficiencies in personnel qualifications 3
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l l and procedural adherence indicated minor weaknesses in the inservice inspection program. ! j In conclusion, Engineering and Technical Support demonstrated high quality, with continued strong performance. Some weaknesses were . observed, but these were minor in that they appeared isolated, of low significance, and were promptly and appropriately corrected.
- 2. Performance Ratina j
) Performance Assessment: Category 1 l
- 3. Board Recommendation r
The Board recommends that licensee management provide continued i p support for the development and long term integration of proactive l l engineering programs. , G. Safety Assessment /0uality Verification l 1. Analysis j ) Evaluation of this area was based on both region-based and resident . inspections. Review of Safety Assessment / Quality Verification activities accounted for about 26 percent of the total Diablo Canyon i inspection effort. p The last SALP assessment rated the licensee's performance in this i area Category 1. Strengths were noted in the implementation of Event l Investigation Teams (EITs). Weaknesses were noted in resolving problems in a timely manner and in occasional lack of management i aggressiveness in dealing with problem areas. The licensee was : encouraged to provide core management involvement in timely problem ! J identification and root cause investigation, particularly in the area y of repeat problems. t During this SALP assessment period, the licensee generally showed l improved performance in this area. Management was more aggressive j and timely in dealing with problems than during the previous period, . and Safety Assessment / Quality Verification performance by line and D ' quality organizations showed continued improvement. A weakness was noted in the identification and correction of precursors of ! potentially significant problems. i l l- A significant strength was the aggressive implementation of programs i to improve control of safety system availability during operating and I shutdown modes. These programs were implemented at all levels of the l licensee's organization. Plant design changes were also implemented to reduce risk, resulting in improved safety performance and safety system availability.
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Audits performed by the Quality Assurance organization were generally good. Lack of intrusive involvement by Quality Assurance in problems i such as the improper maintenance of the containment fan cooler back-
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i! - O ~"~ a draft dampers was a weakness. As discussed in Section III.E, a need i i for more effective audits was also noted in the Security area. In , fact, a factor in several of the problems experienced during this '
- O SALP period was insufficient QA involvement. Some improvement was observed in the latter portion of the SALP period. Audits required by Technical Specifications were adequate and appropriate. l l Additional audits performed as Quality Assurance initiatives showed
' significant technical depth, and identified weaknesses in complex technical areas not typically reviewed by quality organizations. A i 10 noteworthy improvement in QA effectiveness was evidenced in the increased use of surveillances, which are brief audits in specific areas of concern. These audits have allowed rapid focus of QA oversight in problem areas, which resulted in more timely management attention, root cause evaluation, and corrective action. , O Safety groups continued to be very strong in safety focus and depth of technical assessment. The Onsite Safety Review Group identified problems consistent with issues of higher safety significance. fianagement support of this group was adequate. The fiuclear Safety Oversight Committee improved during this assessment period as a result of focus on higher level concerns, and the addition of non-O licensee members. The Plant Safety Review Committee continued to be very strong, providing significant safety insight and conservative
- decision making.
fluclear Operations Support (f;05), which is not by charter a quality oversight group, performed several reviews and audits during this ! O assessment period which were instrumental in identifying and ; correcting problems in interfaces between licensee organizations. ' These 140S reviews and audits resulted in several improvements in the overall implementation of plant safety functions. During this assessment period the NRR staff reviewed a large number ! O of safety analyses performed by the licensee. The licensee's : submittals demonstrated a clear understanding of safety issues and a conservative approach to technical problem resolution. The submit- ; tais for license amendment requests were technically icequate and generally complete. Also, several of the licensee's submittals contained probabilistic risk assessment (PRA) analyses which were of l O high quality (the PPA technique requires considerable effort by the ' licensee, and when properly used, adds to the basis for approving ; proposed changes). The licensee's replies to NRC generic letters and ' bulletins were also timely, responsive and of generally high quality. l Throughout the SALP review period, the licensee consistently and ! O systematically addressed operability concerns in an aggressive nanner, and made appropriately conservative decisions until each concern was resolved. Licensee management kept the f1RC well informed of initial concerns as well as their followup plans for resolution. An increased number of personnel errors were observed in several O functional areas at the beginning of the SALP period. This number ; was reduced by about half during the remainder of the period as a i O ' i
J result of an effective human performance enhancement program and aggressive management involvement at all levels of the organization. s The licensee's program for assessing industry events was strong. A few vulnerabilities were identified and corrected promptly. Several programs were enhanced as a result of implementation of lessons learned from the industry. While the licensee typically has been aggressive in problem resolu-tion, there have been isolated examples of insufficient aggressive-ness in pursuing safety issues. For example, based on the review of licensing submittals requesting relief regarding pumps and valves, the licensee's a proach to the resolution of inservice testing program issues required amplifying information and in some cases were not technically justified. U Five Severity 1.evel IV violations were cited in this area, one for failure to correct reverse rotation of containment fan cooler units, and the others for failure to correct repeated problems of a lower safety significance in various functional areas. " A few weaknesses were observed. For example, in three cases, pre-cursors of plant problems occurred without being identified as such. An example of one of these three instances was an unplanned turbine speed-u" event, corrected by operators, which had two precursors which were not identified and corrected. In each case, the more significant problem occurred because the precursor had not been
' adequately addressed. Another weakness was that the guidelines used to trend root causes of problems were imprecise, in that root causes of problems could be assigned to more than one area.
In conclusion, the performance of the licensee's line organization is very streng in the assessment of safety and assurance of quality.
'g Independent safety groups, although already strong, showed additional strength during this assessment period. The Quality assurance organization's performance was not as strong as the line organiza-tions, but was above an adequate level.
- 2. Perforrance Ratina Performance Assessment: Category 1
- 3. Board Recommendations The Bcard recommends continued management involvement in Safety
, Assessment / Quality Verification activities, and strongly encourages prompt identification of problems, timely corrective action, 4 effective Quality Assurance audits and prevention of repeat problems. l ,9 l i O I
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IV. SUPPORTif4G DATA AfJD SUMMARIES A. Licensee Activities O Unit 1 Diablo Canyon. Unit I entered the assessment period at full power and operated nominally at full power during the SALP period, with occasional
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brief power curtailments for maintenance and testing activities, except as ; follows: ; On July 5,1991 an unplanned start of engineered safety features (ESF) , equipment occurred when a licensed operator inadvertently actuated the l wrong solid state protection system test switch. The control room operators promptly returned all actuated equipment to normal status, t O On March 6,1992 a plant trip occurred due to the loss of main feedwater pump 1-1. The cause of the trip was traced to a faulty fusible link in an i inverter. Unit I was restarted on March 9, 1992 after a new inverter was installed for feedwater pump 1-1 and fusible links for feedwater pump 1-2 i were inspected. Unit I reached 100% power on March 10, 1992. ' O On April 25, 1992, while conducting maintenance on main feedwater pump 1-1, vacuum in the condenser was lost, causing the main turbine and , reactor to trip. The primary cause was attributed to inadequate instruc- l tions, which allowed a vacuum pump to be started before its seal water , isolation valve was opened. Also, condenser vacuum pump suction line ' check valve Cf4C-1-747 was observed to leak excessively when the condenser O vacuum pump suction valve was opened. Unit I was restarted on April 27, , 1992 after evaluation of the event and correction of the cause of the i trip. Full oower was reached on April 28, 1992. i On July 24, 1992, after observing excessive flow noise from main turbine j O governor valve number 4, the licensee closed the valve, resulting in Unit 1 operating at 98% power. On Ser'e..ber 17, 1992, Unit I shut down for a scheduled 63-day refueling ' outage. The shutdown was complicated by spurious reopening of a main : turbine stop valve and two governor valves, which caused the turbine to O accelerate from 1100 RPM to 1870 RPM. A few hours later, during the cooldown, reactor coolant system (RCS) pressure rose above the 350 PSI setpoint. The system responded as expected, with the power operated relief valve opening and relieving pressure. On flovember 9,1992, Unit I completed its fifth refueling outage. The ! unit reached 100% power on fiovember 11, 1992. O On December 23, 1992, an operator observed fragments of a fastening device on the ground. Followup investigation revealeda ' partially melted neutral -
, line connector on a main transformer. The licensee curtailed power to 10% -
and separated from the grid, fixed the connector and inspected all similar i O connectors, and returned to 100% power. ; i O
4 Unit 2 i Diablo Canyon Unit 2 entered the assessment period at full power and ! i operated nominally at full power during the SALP period, with occasional !
- O brief power curtailments for maintenance and testing activities, except as
- follows-i !
j On August 31, 1991 Unit 2 shut down for its fourth refueling outage about ; 2 nin days early because of an unisolable leak in the charging system. The ~ ,O leak had been increasing since its discovery on August 13, 1991. The leak had not yet reached its Technical Specification limit at the time of
- shutdown. Unit 2's shutdown marked 482 d ys of continuous operation at power, a new world record.
4 On October 20, 1991, Unit 2 achieved criticality, marking the shortest j refueling outage in Diablo Canyon history. Full power was reached on : 50 October 31, 1991. l
- t On February 16, 1992, during a power curtailment to 50% for condenser cleaning, the Unit 2 reactor experienced an exaggerated quadrant power .
! tilt ratio as a result of slightly different efficiencies of the secondary ! loops. The licensee decreased power below 50%, the level below which the
- O quadrant power tilt action statement does not apply. The licensee also j entered the action statement, as a conservative measure. During curtail-4 4
ment operations, the power tilt decreased due to xenon burnup, and did no'. l recur until the following curtailment below 50% power for condenser clean- t ing on March 14, 1992. During this occurrence, the licensee repeated the earlier process, and entered the action statement. Upon power ascension, ; O the power tilt decreased to normal, and the licensee exited the action statement. On March 23, 1992, Unit 2 was shut down to investigate a failed turbine ! stop valve. The licensee disassembled the failed stop valve and found O that the nut nich secures the valve disk to the swing arm had disengaged, ! allowing the disk to separate and partially block main steam lead number 2. Unit 2 was returned to power on March 28, 1992 after the failed turbine stop valve was repaired and the other three Unit 2 turbine stop , valves were verified to be properly assembled. ' B. Inspection Activities ' O Fifty routine and special inspections were conducted during this assessment period (July 1991 through December 1992), as listed below.
- 1. Inspection Data
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O Inspection reports: 91-20, 91-22, 91-24 through 91-27, 91-29, 91-31, 91-32, 91-34 through 91-41, and 92-01 through 92-33. Six of these l reports documented management meetings and one documented an ' enforcement conference. O O. l
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- 2. Special Inspection Summary Special inspections included the following:
O 91-39 October 21 - tiovember 29, 1991: A review of the licensee's i Generic Letter 89-10 program for safety related motor operated valves 2-09 March 10 - March 17, 1992: Review of the licensee's e procurement activities for the 6th emergency diesel generator 92-17 March 17 - April 27, 1992: l'eview of the' licensee's maintenance and inspection activities for the containment fan cooler units (CFCUs), as well as licensee operability e assessments for the CFCUs 92-201 August 24 - October 30, 1992: Shutdown Risk Team Inspection C. Enforcement Activity O Inspections during this period identified 19 cited vio'iations. Of these, 18 were Geser ity Level IV and I was Severity Level V. lio deviations were identified during this period. D. Confirmatory Action Letters 3 fione. E. Licensee Event Report s Unit 1 LERs 3 Unit 1 issued 40 LERs during this reporting period. The LERs were 83-39, 91-011 through 91-021, and 92-001 through 92-028. LERs 91-021, 92-003, 92-006, 92-010, 92-015, 92-016, and 92-022 were voluntary. Unit 2 LERs 3 Unit 2 issued 18 LERs during this reporting period. The LERs were 91-001 through 91-012 and 92-001 through 92-006. LERs 91-002 and 91-008 were voluntary. O O 3}}