ML19282A661
| ML19282A661 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 02/27/1979 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| P-79050, NUDOCS 7903050322 | |
| Download: ML19282A661 (17) | |
Text
-
pubne service conipany e cenomde P. O. Box 361, Platteville, Colorado 80651 February 27, 1979 Fort S t.
Vrain Unit No. 1 P-79050 Mr. Karl V. Seyfrit, Direc to r Nuclear Regulatory Commission Region IV Office of Inspection and Enforcement 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76012 REF:
Facility Operating License No. DP R-34 Docket No. 50-267
Dear Mr. Seyfrit:
Enclosed please find two copies of the Report of Changes, Tests, and Experiments Not Requiring Prior Commission Approval in accordance with Part 50.59(b) of Title 10, Code of Federal Regulations, for the period of January 1,1978, through December 31, 1978.
If you have any questions concerning this report, please contact this o ffice.
Very truly yours,
- W M
Don Warembourg Manager, Nuclear Production DW/alk cc: Director, I & E 9i
\\
7903050 3AR
PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION REPORT OF CHANGES, TESTS, AND EXPERIMENTS NOT REQUIRING PRIOR COMMISSION APPROVAL PURSUANT TO 10CFR50.59(A)
January 1,1978, through December 31, 1978 e
I
4 TABLE OF CONTENTS Page Introduction - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
3 1.0 Public Service Company Change No tices - - - - - - - - - - - - - - - -
5 2.0 General Atomic Corr any Field Change Notices - - - - - - - - - - - - -
6 7
3.0 Sargent and Lundy Field Change Notices 8
4.0 General Atomic Company Requests for Test 12 5.0 Public Service Company Tests 13 6.0 Table of Abbreviations 7.0 System Number Identification 15 5
e l
s E
. INTRODUCTION This report is submitted to comply with the requirements of Part 50.59(b) of Vrain Nuclear Title 10, Code of Federal Regulations as they apply to Fort St.
Generating Station, Unit No.1.
It includes the period of January 1,1978, through December 31, 1978.
Some definitions of major terms used in this report may also be helpful:
Change Notice - Documents modification work proposed and installed by Public Service Company.
Field Change Notice - Documents modification work proposed and installed by contractor.
.".T" Tests - Documents tests proposed and conducted by Public Service Company.
"RT" Tests - Documents tests proposed and conducted as requested by con-tractor.
the Safety Evaluation for the changes and tests listed, is In this report, summarized. The terminology used in these summaries, is defined as follows:
Safety-Related Facilities - Those plant systems, structures, equipment, and components which include the following features:
a) Seismic Class I per the Final Safety Analysis Report, b) Safe Shutdown as described in the Final Safety Analysis Report.
c) Engineered Safeguards as described in the Final Safety Analysis Re-port.
d) Portions of radioactive waste systems whose f ailure would result in release of radioactive gas or liquid in excess of maximum permissible concentration.
Safety Significant Change a) Changes to the f acility, systems, components, or structures described in the Safety Analysis Report, that may affect the capability to pre-vent or mitigate the consequences of accident situations that could endanger the health and safety of the public or could result in ex-posures to plant personnel in excess of occupational limits, b) Changes in nuclear safety related systems which involve the addition, deletion, or rework of components, structures, equipment, or systems such that the original design intent is changed.
5, c) Any change that would result in an unreviewed safety question or re-quire a Technical Specification change.
A 2
. Unreviewed Safety Question - Any plant modification or activity is deemed to involve an unreviewed safety question when:
a) Tne probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Final Safety Analysis Report may be increased.
b) The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report, may be created.
c) The margin of safety as defined in the basis for any Technical Speci-fication, is reduced.
To reduce the size of this report, many repetitive terms have been abbreviated.
The reader is referred to Section 6.0 for this listing. Certain systems are identified by their generic numbers; Section 7.0 contains a list of System Number Identifications used.
3 3
.C
. 1.0 PUBLIC SERVICE COMPANY CHANGE NOTICES Change, Test, or Experiment System /
Description of Change, Summary of Safety Number Componen t Test, or Experiment Evaluation CN 30 Process Added sag le and test Not safety significant.
Radiation connections to RT-6212 Monitors piping.
CN 216 N/A Added a modular office Not safety significant.
complex to plant site.
CN 343 Circulator Removed modification Not safety significant.
Trip Mod-made by CN 218 and re-ule o f turned module to orig-Plant Pro-inal configuration, tective System CN 387 Plant Pro-Added test connections Not safety significant.
tective premanently to Plant System Protective System re-lays.
CN 403 Motor Removed shaf t driven Not safety significant.
Driven lube oil pump and in-Boiler Feed stalled DC motor driven Pump lube oil pump to backup AC driven pump.
CN 445 River Tem-Replaced entire system Not safety related.
perature with new design.
CN 464 Evaporation Added new evaporation Not safety related.
Pond pond for demineralizer regeneration effluent.
CN 526 Fire Hose Changed all fire hose Not safety significant.
Connections connections to national standard thread.
1 E
s E
. 2.0 GENERAL ATOMIC COMPANY FIELD CHANGE NOTICES Change, Test, or Experiment System /
Description of Change, Summary of Safety Number Componen t Test, or Experiment Evaluation GA FCN 3651 Helium Added booster pumps to Safety significant.
Circulators increase available cir-Not an unreviewed culator drive pressure safety question, when operating on firewater.
GA FCN 3790 Helium Added a two inch line Safety significant.
Purifica-and valve between the Not an unreviewed tion Sys-purification system safety question.
tem outlet and the reactor building exhaust duct for rapid primary coolant depressuriza-tion.
GA FCN 3898 Helium Added pressure gauges Not safety significant Purifica-on rapid depressuriza-tion Sys-tion vent line for tem monitoring primary coolant depressuriza-tion.
GA FCN 4089 480 Volt Added motor starters Safety significant.
Electrical to provide power to No t an unreviewed Distribu-new firewater booster safety question.
tion pumps for helium cir-culator drive.
GA FCN 4127 Rehea t Revised control system Not safety significant Steam Tem-to reduce reheat steam perature temperature at power Control levels less than 50%.
5 2
3 8
. 3.0 SARGENT AND LUNDY FIELD CHANGE NOTICES Change, Test, or Experiment System /
Description of Change, Summary of Safety Number Component Test, or Experiment Evalua tion S&L Reactor Added structural steel Not safety significant.
FCN 729 Building for support of the new Structure f acility for loading new fuel into the fuel handling machine.
E
~
1
.E O
. 4.0 GENERAL A'IDMIC COMPANY REQUESTS FOR TESTS Change, Test, or Experiment System /
Description of Change, Summary of Safety Number Component Test, or Experiment Evaluation RT-442 Hydraulic Determine the number Not safety significant.
Power Sys-o f tub es that can b e tem plugged in the hydrau-lic oil coolers with-out degrading required heat removal capacity.
RT-45 8 Reactor Determine by test if Not safety significant.
Plant Ven-modification of pipe tilation cavity recirculation fan support has re-duced vibration.
RT-469 Reacto r Investigate the cause N't safety significant.
Co re of core temperature fluctuations.
RT-474 Control Rod Determine by test the Not safety significant.
Drives temperature of control rod drives during rise to power program.
RT-475 Test Instru-Determine the optimum Not safety significant.
mentation location of vibration measuring instruments for core temperature fluctuation investiga-tion.
RT-476 Reactor Re-Determine by test the Not safety significant.
gion Outlet time constant of the Th ermo-region outlet tempera-couples ture thermocouples.
RT-478 Reactor Re-Determine by test the Not safety significant.
gion Outlet time response of the Thermo-core and the region couples outlet thermocouples to changes in orifice valve position and control rod position.
RT-4 80 Reacto r Induce and evaluate Not safety significant.
Core core temperature c
fluctuations.
. GENERAL ATOMIC COMPANY REQUESTS FOR TESTS (continued) 4.0 Change, Tes t,
or Experiment System /
Description of Change, Summary of Safety Numb er Component Test, or Experiment Evaluation RT-481 Prestressed Determine by test the Not safety significant.
Concrete temperature response Reactor of secondary closure Vessel plates and missile shield to changes of reactor power and flow.
RT-482 Reactor Determine by test the Not safety significant.
Core effect of changes in cold helium tempera-ture on core tempera-ture fluctuations.
RT-483 Reac to r Determine by test if Not safety significant.
Core core temperature flue-tuations can be induced by regulating rod mo-tion.
RT-484 PCRV Liner Determine by data re-Not safety significant.
Cooling cording the overall System heat load of the liner cooling system as a function of power.
RT-487 Reactor Determine by test the Not safety significant.
Core threshold power level to induce core tempera-ture fluctuations at various plant condi-tions.
RT-488 Control Determine by test the Not safety significant.
Rod Drives ef fect on drive inter-nal temperatures of changes in control rod drive purge flow.
RT-489 Secondary Determine by test the Not safety significant.
Coolant minimum water flow at System which the cold reheat attemperators may be opera ted.
E C
N I
. 4.0 GENERAL AIDMIC COMPANY REOUESTS FOR TESTS (continued)
Change, Test, or Experiment System /
Description of Change, Summary of Safety Numb er Component Test, or Experiment Evaluation RT-490 Helium Flow Determine by test the Not safety related.
Orifice operating character-Valves ist ics of specific orifice valves.
RT-491 Reacto r Determine by test if Not safety significant.
Core an increase in core pressure drop (AP) has an influence on the power level at which reactor temperature fluctuations occur.
RT-49 2 Secondary Determine by test the Not safety significant.
Coolant operating characteris-System tics of the motor driven boiler feed pump when operating in parallel with a turbine driven boiler feed pump.
RT-493 Main Tur-Evaluate the transfer Not safety significant.
bine Ex-speed of a turbine traction driven boiler feed Steam pump from extraction steam to auxiliary boiler steam during load runback.
RT-494 Helium Flow Determine by test the Not safety significant.
Orifice time constant of ori-fice valves as a func-tion of primary coolant flow.
RT-503 Reactor Determine by test the Not safety significant.
Core transfer function of the regulating rod at high power levels,
s c
I h.
. 4.0 GENERAL ATOMIC COMPANY REQUESTS FOR TESTS (continued)
Change, Test, or Experiment System /
Description of Change, Sumary of Safety Number Component Test, or Experiment Evaluation RT-506 Modified Perform preliminary Not safety significant.
Control Rod checkout of special in-Drives strumentation asso-ciated with modified control rod drives.
RT-516 Reactor Determine by data col-Not safety significant.
Vess el lection the structural response of the PCRV during pressurization from 0 psig to 650 psia.
c 1
.b
~
. 5.0 PUBLIC SERVICE COMPANY TESTS Change, Test, or Experiment System /
Description of Change, Summary of Safety Numb er Componen t Tes t, or Experiment Evaluation T-70A Helium Determine by test the Safety significant.
Circulators effect on circulator Not an unreviewed penetration leak rate safety question.
of changes in helium inlet temperature.
T-79 Helium Determine by test per-Not safety significant.
Circulator formance characteris-Auxiliaries tics of bearing watc.r pumps with one and two helium circulators being supplied.
T-87 Plant Determine the feasi-Not safety significant.
Elec trical bility of f ast trans-Distribu-fer of the 4 KV tion station service buses.
T-10 2 Service Determine by test the No t safety significant.
Water shutoff head of the System service water supply pumps.
T-103 Helium Determine adequacy of Not safety significant.
C 4.rculato r present settings for Auxiliaries proportional band and reset on controller for main drain of helium circula tors.
T-110 Helium Verify operability of Not safety significant.
Circulator bearing water accumu-Auxiliaries la to r.
e c
(
3 0.
. /
6.0 TABLE OF ABBREVIATIONS ACM Alternate Cooling Method ANSI American National Standards Institute ASCO Automatic Switch Company ASTM American Society for Testing Material C
Compressor CFM Cubic Feet / Minute CN Change Notice (PSC)
CO Carbon Dioxide 2
E Exchanger (Heat)
F Filter FCN Field Change Notice (non-PSC initiated change)
FE Flow Element
(
FIS Flow Indicator / Switch FSAR Final Safety Analysis Report GAC General Atomic Company HSV Hand Solenoid Valve HV Hand Valve HVAC Heating, Ventilating, and Air Conditioning IACM Interim Alternate Cooling Method K
Engine (Diesel or Gasoline)
L Line LCV Level Control Valve N
Nitrogen (_ Gas) 2 NRC Nuclear Regulatory Commission E
P Pump
~
A I
. 6.0 TABLE OF ABBREVIATIONS (continued)
PCRV Prestressed Concrete Reactor Vessel PDV Pressure Differential Valve PPS Plant Protective System PS Pressure Switch PSC Public Service Company of Colorado PSI Pounds / Square Inch PV Pressure Valve R
Refueling Region (when followed by a number)
RIS Radiation Indicator / Switch RT Request for Test (General Atomic Company)
S&L Sargent and Lundy T
Tank, Special Test (PSC)
TIG Tungsten Inert Gas TT Temperature Transmitter V
Valve e
3 i
E
. 7.0 SYSTEM NUMBER IDENTIFICATION TABLE 11 Reactor Vessel and Internal Components 12 Control Rods and Drives 13 Fuel Handling Equipment 14 Fucl S torage 15 Fuel Shipping Equipment 16 Auxiliary Equipment 17 Reflecto r 18 Fuel 21 Primary Coolant System (Helium Circulators and Auxiliaries) 22 Secondary Coolant System (Steam Generators) 23 Helium Purification System 24 Helium Storage System 25 Liquid Nitrogen System 31 Feedwater and Condensate 32 Feedwater Hester Vents and Drains 33 Water Treatment 41 Circulating Water System 42 Service Water System 44 Domestic Water System 45 Fire Protection System 46 Reactor Plant Cooling Water System 47 Purification Cooling Water System 48 Alternate Cooling Method (Interim and Permanent) 51 Turbine Generator and Auxiliaries 52 Turbine Steam s
i
. 7.0 SYSTEM NUMBER IDENTIFICATION TABLE (continued) 53 Extraction Steam 54 Turbine Lube Oil Purification 55 Turbine Vents and Drains 61 Decomtamination System 62 Radioactive Liquid Waste System 63 Radioactive Gas Waste System 72 Reactor Building (Vents and Drains) 73 Reactor Plant Ventilation System 75 Turbine Building (Vents and Drains, HVAC) 82 Instrument and Service Air 83 Communication System 84 Auxiliary Boiler and Heating System 91 Hydraulic Power 92 Electrical Power 93 Controls and instrumentation 98 Hydraulic Piping Snubbers 5
2 k
I