ML20087N273

Rept of Changes,Tests & Experiments Not Requiring Prior Commission Approval,Per 10CFR50.59(a),1983

ML20087N273
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/31/1983
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Jay Collins NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
P-84091, NUDOCS 8404030288
Download: ML20087N273 (18)

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PUBLIC SERVICE COMPANY OF COLO' RAD 01-

- FORT ST. VRAIN NUCLEAR'GENERATI G ST TION e,

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REPORT OF CHANGES, TESTS, AND EXPERIMENTS NOT REQUIRING PRIOR COMMISSION APPROVAL PURSUANT TO 10CFR50.59(a) c:

January 1,-1983, through December 31, 1983 i

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Section' Title

- Page Introduction.........................................;............

3 1.0 Public Service Company Change Notices (CN):....'..............

5 2.0. Public Service Company Tests (T-Tests)......:................ 10 3.0 GA Technologies Requests for Tests (RT-Tests)............................................ 13 4.0 System Operating Procedures..........................

....... 14 5.0 Table of. Abbreviations...................................... 15 6.0 System Number Identification Table.......................... 17 T

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, INTRODUCTION This report is submitted-to comply with the: requirements of Part 50.59(b) of Title 10, Code of Federal _ Regulations as they apply to Fort St. Vrain Nuclear Generating Station,. Unit No.1..

It includes the period of January:1, 1983, through December 31, 1983.

Some definitions of major terms used in'this report which may be helpful:

Change Notice - A document containing installation, inspection and testing requirements, design background information, and design document updating requirements which specify the design control requirements applicable to a

plant modification and authorizes changes to "as-built" plant design documentation.

"T" Tests - Tests proposed and conducted by Public Service Company of Colorado.

"RT" Tests - Tests proposed by GA Technologies and conducted by Public Service Company of Colorado.

In this report, the safety evaluation for the changes, tests, and experiments is~ summarized. The terminology used in these summaries is defined as follows:

Safety Related Items Those plant systems, structures, equipment, and components which are identified in the FSAR, and as detailed and supplemented by applicable piping and instrument (P & I) diagrams, IB and IC diagrams, and SR-6-2 and SR-6-8 lists to include the following:

a)

Class I per the Updated FSAR, Tables 1.4-1 and 1.4-3.

b)

Safe shutdown components per the Updated FSAR, Tables 1.4-2 and 1.4-3.

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J Safety Significant Change,

Changes to the

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systesse : components, or l

structures as described in the~FSAR that suy do any one of the following:

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Affect their capability to prevent or mitigate the consequences of accidents ldesc'ibed11n the FSAR.

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Could result in exposures (to fplant personnel in excess of occupational limits.

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Changes in the safety related systems which involve the addition, deletion, or repair of components, -structures, equipment, or systems such that the original design intent is changed (i.e.,

changes in redundancy,= performance characteristics, separation, circuitry logic, control, margins of safety, safe shutdown, accident analysis, or any change that would result in an unreviewed safety question or require a Technical Specification change).

Unreviewed Safety Question Any plant modification or activity that is deemed to involve an unreviewed safety question as defined in 10CFR50.59 (a)(2), in which:

a)

The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR may be increased; or b)

The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR may be created; or-c)

The margin of safety as defined in the basis for any Technical Specification is reduced.

To reduce the size of this report, many repetitive terms have been abbreviated.

The reader is referred to Section 5.0 for an index of the abbreviations used.

Certain systems are identified by their generic Fort St.

Vrain system numbers; Section 6.0 contains a list of system number identifications used.

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' 5-1.0 PUBLIC SERVICE COMPANY CHANGE NOTICES.

All CN's will be described in'the following1 order:

First - CN number.

Second - system / component number.

Third - description of the change.

Fourth - summary of safety evaluation.

CN-572 System 70/ General Structures This CN updated several documents to reflect as-built conditions of the revolving door at the entrance to Fort St.

Vrain.

The revolving door was installed in 1974, however, documents were not updated to show the new door.

Figure 1.2-18 of the FSAR is affected by this modification update.

Since this CN is only a formal mechanism for updating documents for an "as-built" modification, it does not involve an unreviewed safety question, nor is it a safety significant issue.

CN-1169 System 48/ Alternate Cooling Method i

This modification relocated the starting batteries for the Alternate Cooling Method (ACM) diesel generator to a heated building.

Prior to this modification,-the starting batteries for the ACM l

diesel generator were located in the shell of the l

diesel generator itself.

This location required the installation of heat tape at the beginning of each winter l

season. The relocation of the starting batteries to an adjacent building provided the necessary temperature and ventilation control to ensure the electrical start system was continuously in an operable condition. The addition does not increase the probability of any accidents analyzed in the FSAR. This CN is not safety significant and does not involve an unreviewed safety question.

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CN-1326 System 93/ Meterological.Instrumentatiod'

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This CN and it's associated revisions-installedia1610 meter tower and various instrumentation to measureLand. analyze.neterological conditions adjacent to the ' Fort ;St..AVraino site.

The installation was in response to NUREG-0654.

The meterological instrumentation was'installedito provide data on meterological conditions for plant opertions and to3 determine stability catagories per the station's RERP. The probability of occurrence of an accident or malfunction has~not been~ increased.

This CN does not involve an unreviewed safety question, nor is it safety significant.

CN-1394 Systems 79 and 90/ Technical Support Center and Station Computers This modification provided computer system upgrades for the station's Emergency Response Facilities.

The computer system upgrade incorporated new system monitoring points and recording capabilities to provide necessary information to the Technical Support Center as required per NUREG-0696. This change does not involve an unreviewed safety questien, nor is it safety significant.

CN-1401 System 93/ Radiation Detection Instrumentation l

i This modification installed a high range area radiation monitor on the east wall of the refueling floor.

1 The high range monitor replaced an existing low range monitor.

This modification was the result of a NRC commitment.

This change does not involve an unreviewed safety question, nor is it safety significant.

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CN-1433

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Systems 62 and 72/ Radioactive ' Liquid Wasth7Systemi I

This CN incorporated minorl eqEipment/compone$tJchanges and control circuit revisions to the radioactive;0 liquid.; waste and the reactor building sump' discharge; systems.-

-The modification installed a: three-way1 ball; val e at-the junction of the radioactive liquid wastec system' J(62) fand the reactor building sump discharge ~ (72)? lines. (This: prevents.a-simultaneous release from both: systems; whichricould. invalidate the calculated release -concentrations.

The change'.also installed a new automated proportional sampler infthe? discharge.

line of the reactor building sump.

JAlso. installed, was a permanent connection to the firewater system for. the purpose of i

flushing the radioactive liquid waste system. -This'CN and it's associated revisions affected the operating procedures for this system significantly.

The affected System Operating Procedure is addressed later in this report.

Original-design intent of the system remained unchanged, and the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR was not increased. This change is not safety significant and does not involve an unreviewed safety question.

CN-1436 System 11 This change to the steam generator penetration interspace pressure control system allows the interspaces to be operated at a pressure slightly greater than cold reheat steam pressure, but less than primary coolant (PCRV) pressure.

This modification was installed due to known leakage. The reduced operating pressure maintains this leakage at a minimum.

As this change did not affect the capability to prevent or mitigate -the consequences of accidents described in the FSAR, this change is 'not considered to involve an unreviewed safety question.

As the modification did, however, change the operating procedure and assoicated Technical Specification, the change is considered to be safety significant.

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CN-1460 m

System 70/ Building 1 10

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This CN provided the structural design lfor the installation of a new building on site.

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This change is not safety'significant, and the: change did'not increase the probability of an

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or_ _ malfunction previously discussed in the FSAR This change doe,s;not-involve

- an unreviewed safety question.

- CN-1461 System 70/ Building 10 This CN provided the mechanical design.for the installation of a new building on site.

This change is not safety significant, and did not increase the probability of an accident or malfunction previously discussed in the FSAR. This change did not involve an unreviewed safety question.

CN-1590 Systems 46 and 47/PCRV and Purification Cooling Water Systems 2

This modification allowed the use of a system 46 water chiller to supply cooling water to the helium purification cooler in the helium purification trains.

The modification allowed the physical connection between the chiller unit and the " front-end" coolers of both helium purification trains.

However, the actual connection was only made on "A" helium purification train. The use of the chiller unit allowed cooling water to be sent to the cooler which enhanced the removal rate of moisture from the primary coolant

- (helium).

As this modification did not change the basic functions of the helium purification system but rather facilitated the operation of the trains and provided for enhanced moisture removal from the primary

coolant, the modification did not create any new failure modes nor reduce any capabilities to deal with those existing.

Therefore, this change is' not safety significant and does not involve an unreviewed safety question.

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Systems 47 and.63/ Purification; Cooling Waterfand1 Radioactive Gas Waste Systems This CN modified the cooling ~ water piping N N 11um purification cooler, E-2302 to allow venting the cooling waterjsystem:to -.the gas waste system.

a result' of.'obtAin ng WprimaryTcoolant This modification was

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(helium) to purific? tion' coolingL water.b1_eak kin Othe-cooler.

Prior to this installation,-the leak created a gas build-up;in the cooling water sy: tem surge tanks.'. This build-up": eventually lead to gas binding of the system " pumps and' coolers. 1The installation of the vent path directly to the gas; wasteJ system allows continued operation of the cooler. until' it~t-can be replaced / repaired (scheduled for Refueling Outage.in early 1984).

As the design and operation of.the-components'did not change, this CN was not safety significant. The. installation of the vent path did not increase the probability of an accident or malfunction and this change does not involve an unreviewed safety question.

CN-1741-Systems 17 and 18/ Neutron Startup Source This modification prepared a neutron startup source for Region 22, Column 6, Layer 4 of the reactor core.

The addition of the fourth startup source into the Fort St.

Vrain core is required to provide adequate reactor control indication during startup operations.

This addition to the reactor core will take place during the next refueling outage, scheduled for early 1984. This modification does not increase the probability of an accident or malfunction. This change i s not safety significant, nor does it involve an unreviewed safety question.

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2.0 PUBLIC SERVICE COMPfilY TESTS (T-TESTS)1' ~

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This test determined. the.Jmagnitude-'ofi the flow resistences through the core support floor to the individual?CSF; vent,: lines (for both concrete and' column vents).

The test did not affect plant safety or operation. iThe test did not modify any system and was performed consistent with. current plant operating procedures. Testing.was performed at.less than full density and with the CSF vent system pressure at or; below-PCRV-pressure.

It did not involve any equipment needed-for safe shutdown. This test is not considered safety 1significant, nor does it involve an unreviewed safety question.

T-204 System 11/ Core Support Floor To identify any helium ingress from the PCRV to the CSF, and if that ingress can be terminated by varying the AP betkeen the PCRV and the CSF.

Although the test required exceeding full density, the reactor was shutdown and pressure and temperature were within operational limits.

The test was considered not safety significant and the probability of an accident or malfunction previously discussed in the FSAR was not increased. This test did not involve an unreviewed safety question.

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,n Determines 'the? magnitude E,off;the? flow 1 resistances;through.the-core support floor to thelindividual CSF vent. flines "(for ooth concrete and column. vents)(same as T-203).

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not modify any system and was performed: consistent;with ~ current plant operating procedures. Testing.was.performedfatiless>than

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full density and with the CSF vent system pressure;atf orf below-PCRV pressure.

It did not. involve-any' equipment needed for safe shutdown. This test is not considered safety. significant,, nor does it involve an unreviewed safety' question.: '

T-206 System 11/Fike RKB Rupture Discs *

• Brand Name To test for adequate leak-tightness of non-seal welded Fike RKB Rupture Disc Assemblies.

The seal weld, originally a shop weld, is to be ground away to gain access for the rupture disc examination required by Technical Specification SR 5.2.1.

It is not practical to perform the seal weld in place. Deletion of the seal weld would allow the performance of the examination without having to cut piping welds to remove the disc. The test was not performed on installed equipment, therefore it had no effect on plant safety.

If the test results show sufficient pressure boundary protection-without the use of the seal welds, a Change Notice will be issued to authorize the removal of the seal welds on the rupture disc assemblies in service.

This test was not safety significant nor an unreviewed safety question.

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System 62/RT-6212 and RT-6213 Di- '

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This test compared 0 sensitivity: values'obtain$d'on/RT-6212and-RT-6213 using radiciodine solution and a-serieslof point sources t

to establish a relationship. This;will.' provide:an easier and more efficient method 1 off following} Jthe: idetectors'.

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associated circuitry) performance.

1 The test was performed by station Radiochemistry personnel:in~

order to establish a' performance criteria base for thel radiation detectors in the radioactive : liquid Ewaste system.

This establishment of criteria will. enable _the> station :to iverify:

proper performance-of thes'e detectors, which ' subsequently ensures safe operation of the radioactive liquid waste" system.'

No releases were performed during the performance of this test.

The test was not considered safety significant, nor did it involve an unreviewed safety question.

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3.0 GA TECHNOLOGIES REQUESTS:FOR: TESTS TRT-TESTS) y t.9

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RT-485 System 12/ Control.R'od DriveIMechanisms' This

" request-for. test"i'was~-:initiallyLsubaittediby.GA Technologies :in. 1978.

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testLfealled:fforj installing-temperature : sensors on:: selected KCROMsf for Dthel-purpo'se Lof recording and analyzing the' temperatures. experienced by. the CRDMs during various plant operating conditions. ? Originally,.

.the CRDMs in core regions '4, 5, 34, 35 and 36 were equipped with the sensors.

Following: the -continuing-data acquisition,;the results were studied and the -. recommendation : 'of;, installing d

sensors on all the CRDMs, whenever they are pulled,;was made.

Public Service Company of Colorado has-followed-up ton -this recommendation -with the production of.: a Change. Notice to authorize the installation of temperature sensors.on the remaining control rod drive mechanisms whenever time and plant conditions allow. The installation and monitoring of: the~ CRDM temperatures will ensure that the mechanisms are operating properly. The test has been completed and was not considered safety significant, nor did it pose an unreviewed safety question.

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4.0 SYSTEM OPERATING PROCEDURES?

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Duetoseveral. minor _ equipment / component:andoherating' technique:

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changes of the Radioactive: Liquid Waste System,J Jthecassociated system operating procedure was extensively: revised. c Following-the changes madeE by CN-1433 -(see ;Sectiony,1.0),H the system operating procedure-for the radioactive liquid waste-system was completely reformatted and revised-to: reflectEthexequipment,:

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piping and operational changes.

The completed, procedure was reviewed for accuracy, completeness, and safe : operation.

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procedure was found not to be safety significant, nor did it.

involve an unreviewed safety question. -

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5.0 TABLE OF ABBREVIATIONS l --

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ACM

! Alternate _; Cooling; Method. ]

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ANSI American NationalLStandards Institut'e ASCO AutomaticLSwitchCompa'ny' ASTM American Society for Testing :and _ Materials; 2

C Compressor.

CFM Cubic Feet / Minute CN Change Notice' (Public[ Service. Company)

CO, Carbon Dioxide CRDM Control Rod Drive Mechanism CSF-Core Support Floor E

Exchange (Heat)

F Filter FCN Field Change Notice (Non-Public Service Company Initiated Change)

FE Flow Element FES Final Environmental Statement FIS Flow Indicator / Switch FSAR Final Safety Analysis Report GAT GA Technologies HSV Hand Solenoid Valve HV Hand Valve

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HVAC Heating,'Nentilating,andAir Conditioning i

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Engine.(Diesel orfGasbline).-

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Level ~ Control Valve N

~ Nitrogen (Gas) 2 NRC Nuclear Regulatory Commission'-

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' Pump

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PCRV Prestressed. Concrete Reactor Vesse1-PDIS Pressure Differential Indicating-Switch PDT Pressure Differential Transmitter 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)

RERP Radiological Emergency Response Plan RIS Radiation Indicator / Switch RT Request for Test (GA Technologies)

S&L Sargent and Lundy S0P System Operating Procedure T

Tank, Special Test (Public Service Company)

TIG Tungsten Inert Gas TT Temperature Transmitter V

Valve

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.6.0 SYSTEM NUMBER IDENTIFICATION: TABLE-:e

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2 Plant Site;.

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Reactor 1. Vessel;anUInternAl-Components.

12 Cohtrol. Rods land Drivest ~

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x.. s 14 Fuel Storage 15 Fuel Shipping Equipment.

16 Auxiliary Equipment 17 Reflector 18 Fuel' 21 Primary Coolant System (Helium Circulators and Auxiliaries) i l

22 Secondary Cooling System (Steam Generators) 23 Helium Purification System r

24 Helium Storage System 25 Liquid Nitrogen System 29 Gas Charging Facility 31 Feedwater and Condensate l-32 Feedwater Heater Vents and Drains r

33 Water Treatment 41~

' Circulating Water System j

42-Service Water System L

44 Domestic Water System i

45 Fire Protection System l

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W.. i8 Y Reactor Plantp Coolinglater Systim <

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?47 Purification.CoolingLWaterLSystem LAlternateCoblingiMethodj 48 51 LTurbin'e Generator and Aux 11'iar'ies5 52-

[TurbineLSteam

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-Extractionisteam-54

-Turbine' Lube 011: Purification' 55 Turbine Vents'and Drains.

61 Decontamination System

• 4 62 Radioacti_ve Liquid Waste: System:

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63 Radioactive Gas Waste System 70 Structures - General I

72 Reactor Building (Vents and Drains)

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73 Reactor Plant Ventilation System 75 Turbine Building (Vents and Drains, HVAC) 78 Security System 79 Technical Support Building 82 Instrument and Service Air 83 Communication System 84 Auxiliary Boiler and Heating System 90 Computer Systems l

91-Hydraulic Power 92' Electrical Power 93 Controls and Instrumentation 98 Hydraulic Piping Snubbers L

99 Miscellaneous i

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t public service company of C o11stra d e 16805 WCR 19 1/2, Platteville, Colorado 80651 20 March 23, 1984 Fort St. Vrain Unit #1 P-84091 Mr. John T. Collins, Regional Administrator Region IV g pgg Nuclear Regulatory Commission 611 Ryan Plaza Drive

~-g Suite 1000 Arlington, TX 76011

REFERENCE:

Facility Operating License No. DPR-34 Docket No. 50-267

Dear Mr. Collins:

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,.1983, through December 31, 1983.

If you have any questions concerning this report, please contact this office.

Very truly yours, 1

i 0A4 Don Warembourg Manager, Nuclear Pro ction l

DW/djo Enclosure i

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