ML18045A450
| ML18045A450 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 07/28/1980 |
| From: | Nishimura M EG&G, INC. |
| To: | Scholl R Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-01, TASK-3-1, TASK-RR ESD-6899, NUDOCS 8008060297 | |
| Download: ML18045A450 (10) | |
Text
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.c:;,~ EGc.G i,.*
E.nergy Measurements Group
- San Ramon Operations 2801 OLD CROW CANYON ROAD. SAN RAMQN. CA *,TEL. (415) 837-5381
- MAIL: BOX 204. SAN RAMON. CA 94583 28 July
- 1980
- ESD# 6899 Mr. R. F. Scholl, Jr; US Nuclear Regulatory Commission Division of Operating Reactors SEP Branch 7920 Norfolk Avenue Bethesda, MD 20014 -
SUBJECT:
PALISADES NUCLEAR POWER PLANT' (DOCKET NO. 50-255)
Dear Ray:
Attached is a rough draft. report on the Palisade~ plant detailing the NRC safety topic III-1 (Classification of Structures,. Components, and Systems),
Please review the report.and forward your comments and recommendations to me..
Sincerely, M.W.NISHIMURA GROUP.. LEADER MWN/ss cc:
EG&G D.
- Laudenbacl). *
. B. Mayn LLNL NRC M. Dittmore :.
D. Allison G. St, Leger-Birter f
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SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY TOPIC III-1 ASSOCIATED WITH THE ELECTRICAL, INSTRUMENTATION, AND CONTROL PORTION OF THE CLASSIFICATION OF STRUCTURES, COMPONENTSr AND SYSTEMS FOR THE PALISADES NUCLEAR POWER PLANT By M. w. Nishimura*
July 1980 ROUGH DRAFT
- EG&G, Inc., Energy Measurements Group, San Ramon Operations
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TABLE OF CONTENTS
- 1.
- 2.
REVIEW GUIDELINES.
- 3.
COMP1LATION OF IDENTIFIED SYSTEMS 3.1 Engineered Safety Feature Systems 3.2 Reactor Protection System.
3.3 Additional Systems REFERENCES.
'1 * **
APPENDIX A NRC SAFETY TOPICS RELATED TO THIS REPORT 1
2 3
3 3
4 5
6 l
Jl 1.J SYSTEMATIC EVALUATION PROGRAM REVIEW OF NRC SAFETY TOPIC II I-1 ASSOCIATED WITH THE ELECTRICAL, INSTRUMENTATION, AND CONTROL PORTION OF THE CLASSIFICATION OF ST~UCTURES, COMPONENTS, AND SYSTEMS FOR THE.PALISADES NUCLEAR POWER PLANT M. W. Nishimura EG&G, Inc., Energy Measurements Group San Ramon Operations
- 1.
INTRODUCTION
- Some of the SEP plant structures, systems, and components may not be designed to with stand the effects of. a safe shutdown earthquake and remain functional.
In some cases, systems and components important to
. safety may not be designed, fabricated, erected, and tested to quality standards commensurate with their safety function.
Compilation of the major systems required for design basis events (DBE) and for safe shutdown of the p 1 ant wi 11 be submitted in support of NRG Safety Topic III-12 *. This safety topic addresses whether the major systems identified meets current quality standards. ;).
- 2.
REVIEW GUIDELINES The objective of this review is to identify only the major elec-trical, instrumentation and control systems (EICS) required for DBE and for safe shutdown of the plant.
This identification is to be performed when reviewing each NRC safety* topic.
A detailed search will not be made to identify all required systems for DBE and for safe shutdown.
- 3.
COMPILATION OF IDENTIFIED SYSTEMS 3.1 ENGINEERED SAFETY FEATURE SYSTEMS The following engineered safety feature systems are required for DBE and safe shutdown:
- 1.
Safety injection system
- a.
High-pressure safety injection pumps
- b.
Low-pressure safety injection pumps
- c.
Safety injection tank
- 2.
Containment spray system
- 3.
Containment air coolers
- 4.
Iodine removal system
- 5.
Containment venting charcoal filter
- 6.
Electric hydrogen recombi ner system 3.2 REACTOR PROTECTION SYSTEMS The following reactor protection systems are required for DBE and safe shutdown:
- 1.
Power range safety channels
- 2.
Wide-range logarithmic neutron monitors
- 3.
Reactor cool ant fl ow
- 4.
Thermal margin/low pressurizer pressure
- 5.
High-pressurizer pressure
- 6.
Steam generator level
- 7.
Steam generator pressure ~I
- 8.
- 9.
Containment pressure Loss of load
- 10.
Protection system logic units
- 11.
Manual trips 3.3 ADDITIONAL SYSTEMS cu*e In addition to the ESF and RPS, the f()llowing systems"required for DBE and safe shutdown:
- 1.
Offsite power system
- 2.
- 3.
Safety i nj ecti on and ref.uel i ng water tank
- 4.
Containment sump
- 5.
Control room systems
- 6.
Compressed air system
- 7.
Engineered safeguards local panel-auxiliary building
- 8.
Service water system
- 9.
Component cooling system
- 10.
Auxi 1 i ary feedwater system
- 11.
Heating, ventilating and air conditioning systems.
REFERENCES
- 1.
Consumer Power Company, Palisades Final Safety Analysis Report, filmed June 1978.
- 2.
EG&G, San Ramon Division, technical evaluation reports on NRC safety topics:
a)
Safety Topic VI-7.A.3, "ECCS Actuation." July 1980 b)
Safety Topic VI-10.A, "Testing of RTS and ESF, Including Response Ti me,
11 J u 1 y 198 0
- c)
Safety Topic VII-LA, "Isolation of RPS from Non-safety Systems, Including Qualification of Isolation Devices," July 1980.
d)
Safety Topic VI-7.B, "ESF Switchover from Injection to Recircula-tion Mode, 11 July 1980.
e)
Safety Topic VII-2, "ESF System Control Logic and D_esign, 11 July 1980.
APPENDIX A NRC SAFET~ TOPICS RELATED TO THIS REPORT
- 1.
Safety Topic VI-7.A.3
- 2.
Safety Topic VI-10.A
- 3.
Safety Topic VII-1.A
- 4.
Safety Topic VI-7.B
- 5.
Safety Topic VII-2
- 6.
Safety Topic III-2
- 7.
Safety Topic I I I-3
- 8.
Safety Topic III-4
- 9.
Safety Topic III-5
- 10.
Safety Topic III-6 "ECCS Actuation."
"Testing of RTS and ESF, Including Response Time.
11 "Isolation of RPS from Non-safety Systems, Including Qualification of Isolation Devices."
"ESF.Switchover from Injection to Recir-culation Mode.
"ESF System Control Logic and Design."
"Wind and Tornado Loadings."
11 Hydrodynamic Loads."
a)
Effects of high water level on structures b)
Structural and other consequences (e.g.,
flooding of safety-related equipment in basements) of failure of underdrain systems c)
Inservice inspection of water control structures.
11Missile Generation and Protection."
a)
Internally-generated missiles d)
Site proximity missiles (including aircraft).
"Evaluation of Pipe Breaks."
a)
Effects of pipe break on structures, systems and components inside con-tainment b)
Pipe break outside containment.
"Seismic Design Considerations."
./... 1:,;
- 11.
Safety Topic III-7
- 12.
Safety Topic III-8
- 13.
Safety Topic III-9
- 14.
Safety Topic III-10
- 15.
Safety Topic III-11
- 16.
Safety Topic III-12 Cathy #4/#12/CEB/amr 11Category *I Structures Integrity.
11 a)
Inservice inspection, including pre-
~tressed concrete containments with either grouted or grouted tendons b)
Design codes, design criteria, load combinations, and reactor cavity design criteria c)
Delamination of prestressed concrete containment structures d)
Containment structural integrity tests.
11Reactor Vessel Internals Integrity.
11 a)
Loose parts monitoring and core barrel vibration monito~ing b)
Control rod drive mechanism integrity c)
Irradiation damage, use of sensitized steel and fatigue resistance d)
Core supports and fuel integrity.
11 Support Integrity.
11 "Pumps and Valves Integrity.
11 a)
Thermal-overload protection for motors of motor-operated valves b)
Pump flywheel integrity c) Surveillance requirementss on BWR recirculation pumps and dischrage valves.
"Component* Integrity."
"Environmental Qualification of Safety Related Equipment." i i
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