ML20247G984
| ML20247G984 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 01/22/1989 |
| From: | Brey H PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| References | |
| P-89255, NUDOCS 8907280244 | |
| Download: ML20247G984 (32) | |
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1 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 i
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l July 21, 1989 Fort St. Vrain
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Unit No. 1 j
P-89255 1
i U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Docket No. 50-267
SUBJECT:
10 CFR 50.59 ANNUAL REPORT SUBMITTAL
REFERENCE:
Facility Operating License No. DPR-34 Gentlemen:
This letter transmits the Annual Report of Changes, Tests, and Experiments affecting the Fort St. Vrain Nuclear Generating Station pursuant to Part 50.59(b) of Title 10 Code of Federal Regulations.
This report covers the period of January 23, 1988 through January 22, 1989.
If you have any questions concerning this report, please contact Mr.
M. H. Holmes at (303) 480-6960.
Very truly yours, n-H. L. Brey, Manager l
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. 40 Nuclear Licensing and
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Resource Management 18 gg HLB /GMK/km
W Attachment Z8
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Regional Administrator, Region IV l p-ATTN: Mr. T. F. Westerman 13d o o:
Chief, Projects Section B l ~ $8cc
'l Mr. Robert Farrell Senior Resident Inspector i
Fort St. Vrain
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PUBLIC SERVICE COMPANY'0F COLORADO FORT SAINT-VRAIN NUCLEAR GENERATING STATION ANNUAL REPORT OF CHANGES, TESTS, AND EXPERIMENTS NOT REQUIRING PRIOR COMMISSION APPROVAL PURSUANT TO 10 CFR 50.59 January 23, 1988 through January 22, 1989
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o, TABLE OF CONTENTS Section Title Page INTRODUCTION........................................................
2 1.0 Change Notices (CN)......................................
5 2.0 Document Change Notices (DCN)............................ 17 3.0 Setpoint Change Reports (SCR)..................
......... 18 4.0 Speci al Te st s (T-Te s t s).................................. 19.
5.0 Procedures............................................. 29 1.
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INTRODUCTION This report is. submitted to. comply with the requirements of Part 50.59(b) 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 23, 1988 through January 22, 1989.
The following defines certain activities contained in this report:
Change Notice (CN) - 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.
Document Change Notice (DCN)
A document which aut.horizes a change to design documents. As a minimum, it contains a design
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input statement,- a. design analysis statement, a document update list and the document update information.
Setpoint Change Report (SCR)
A document which authorizes setpoint changes which do not constitute an alteration to the design of the affected equipment.
T-Tests - Special tests proposed and conducted by Public Service Company of Colorado.
The following is a list of abbreviations used in this report:
EQ - Environmental Qualification FSAR - Final Safety Analysis Report HELB - High Energy Line Break LCO - Limiting Condition for Operation P&I - Piping and Instrument Drawing PCRV - Prestressed Concrete Reactor Vessel PPS - Plant Protective System RERP - Radiological Emergency Response Plan SLRDIS - Steam Line Rupture Detection and Isolation System S0P - System Operating Procedure - _ _ __-_ -_ _ _ _
The following' defines terms used in safety evaluation summaries contained in this report:
Safety Related Those plant systems, structures, equipment and components.which are identified by the FSAR and as detailed and supplemented by applicable P&I-drawings, "IB" and "IC" diagrams, "E" schematic diagrams, the Cable Tab, SR-6-2 and SR-6-8 lists to include the following:
a)
Class I per the FSAR, Table 1.4-1 b)
Safe shutdown components per the FSAR,. Table 1.4-2 c)
Alternate Cooling Method (ACM) equipment d)
Interface circuits (IC) within the EQ Program Safety Significant Changes to the facility, system, components, or structures as described in the FSAR that may do any one of the following:
a) affect their capability to prevent or mitigate the consequences of accidents described in the FSAR; b) could result in exposures to plant personnel in excess of occupational limits.
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, ci rcuitry logic, control, margins of safety, safe shutdown, accident-analysis or any change that would result in an unreviewed safety question or requirr; a Technical Specification change).
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Unreviewed Safety Question Any plant modification or activity that is deemed to involve an unreviewed safety question as defined in 10 CFR 50.59:
a) if the probability of oc:urrence or the consequences of an
.'E accident or malfunction of equipment important' to safety previously evaluated in the FSAR may be-increased; or b) if a possibility for an accident or malfunction of a different type-than any evaluated previously. in the FSAR may
.be created; or c) if the margin of safety as defined in the basis for any Technical Specification is reduced. L _ =____ - __ ________ _ - _ -. _.
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1.0 CHANGE NOTICES (CN)
CN-1255D System 92/ Accessory Electrical Equipment The original CN was reported to the NRC in P-84224, dated July 20, 1984, and authorized installation of. independent instrument power transformers as backups. for inverter / static. transfer switches 1A and 18, which provide power to non-interruptible buses 1A and 18.
Also, inverters IA and IB were upgraded to 25
.KVA.
Reissue D revised the Safety Evaluation to identify changes to FSAR Table I.4-1.
FSAR Table I.4-1 has been revised to reflect this change. This activity was classified safety related and safety significant based on the original CN, but did not' involve an unreviewed safety question.
CN-1395 System 93/ Controls and Instrumentation This CN authorized modifications to Main Control Board I-04 to accommodate a Safety Parameter Display System (SPDS).
This CN provided for:
rearrangement of I-04 layout, a seismically qualified support for the SPDS monitor, installation of wiring and labeling, and removal 'of non-required instrumentation and display information from I-04.
Modifications to I-04 have not compromised the Control Board's design function or that of the installed instrumentation. Adequate circuit protection exists for the monitor while the data logger power supply is adequate to handle the additional load.
Either equivalent instrumentation exists in the control room for the removed instruments, or the removed instruments are not required for safe plant operation.
No redundancy considerations are affected.
FSAR Sections 3.6.7, 7.2.2.2, 7.2.3.2, and 7.4.1 have been revised to reflect this modification.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question. - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
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CN-1878 System /Various This CN authorized changes resulting from the Control' Room Design Review which satisfies, in part, requirements of NUREG-0737.
Specific changes involved relocation'of certain system parameter recorders from the Main Control : Boards.
With one exception, these recorders remain located in the Control Room.
One recorder, used to monitor helium system component temperatures',
was relocated to the refueling floor.
However, parameter indications which are required for post-accident monitoring per Reg. Guide 1.97 have been retained in the Control Room.
A multipair-thermocouple cable was also added by this CN. Recorder power supplies remain unchanged and recorder relocation reduces Main Control Board congestion.
FSAR Figures 8.2-15 and 8.2-20 have been revised to reflect this modification. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
CN-1895 System 93/ Controls and Instrumentation j
This CN authorized changes resulting from the Control Room Design Review which satisfies, in part, requirements' of NUREG-0737.
Main. Control Board I-10,.which contains instrumentation for the Plant Protective System (PPS), has been modified through functional grouping and demarcation, relabeling, and incorporation of standard conventions (i.e.
color coding and 1
switch positions).
This CN also corrected documents for hand switch labeling discrepancies.
The enhancements of this CN provide easier information processing by the operators and support personnel, thereby enhancing margins of safety in the operational activities of the plant.
FSAR. Figures 7.1-14 and 7.1-15 have been revised to reflect the
. modifications of this CN. This activity was classified safety related, but was not safety significar,t and did not involve an unreviewed safety question.
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. System 25/ Nitrogen System This CN authorized installation of instrumentation to automatically maintain levels in T-2501 (liquid nitrogen storage tank)'and T-2502 (liquid nitrogen surge tank). No specific level was stipulated for control by this CN except that T-2501 level will be maintained above the low level alarm setpoint (> 650 gallons), as required by the Technical Specifications and FSAR for reactor power operation.
FSAR Section 9.6.4 has been revised to' delete a statement that the liquid nitrogen storage tank is normally maintained at least 1000. gallons.
FSAR Table I.4-1 has been revised to indicate T-2501 as the liquid nitrogen storage tank instead of T-2502. This.
activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
CN-1996A System 92/ Accessory Electrical Equipment The original CN was reported to the NRC in P-86454, dated July 22, 1986, and authorized installation of fused disconnect safety switches in the Building 10 circuit breaker test facility. Cable
- 3201 was installed as part of.this modification.
Reissue A authorized this cable to be renumbered as #3254, since #3201 was discovered to have been previously used elsewhere.
FSAR Figure 8.2-9 has been revised to reflect this change. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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System 21/ Primary Coolant System
.This CN authorized installation of a third Turbine Water Removal Pump (designated P-2103SX). This pump was added'to provide the preper separation between the Turbine Water Removal Pumps as required by Appendix R, and was sized for removal of water resultant from running one helium circulator on its water turbine drive supplied with condensate. This third pump was not intended to meet Technical Specification requirements, but is relied on to function after a fire in a non-congested cable area.
Its platform was designed not to fail during a seismic event.
The margin of safety regarding the operation of a Turbine Water Removal. Pump after a fire has been increased due to the addition of this third pump and its separation from the other two pumps.
FSAR Sections 4.2.2.3.3, 4.2.2.3.5, Figures 1.2-10 and'4.2-12 have been revised to reflect this modification.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
CN-2109
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System /Various This CN authorized removal or modification of alarms associated with the annunciator system that were redundant, of a nuisance type, or an indication of equipment status. This was done for human factors considerations.
This CN satisfies, in part, requirements of NUREG 0737.
Removal of unnecessary alarms enhances operating staff attentiveness to valid information requiring their assessment and corrective actions.
FSAR Sections 9.7.3.1, 9.11.4, Figures 7.1-15 and 9.7-1 have been revised to reflect these modifications.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question. -
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l System 12/ Control Rods and Drives
.This CN added new documentation, design drawings, and new design specification 12-D-15 for a new control rod design. This new rod design was intended to replace the old design gradually during l
refuelings. The new rod. design includes material. changes to
-Inconel and reduces.the chances for stress corrosion cracking.
The new design will result in a decrease in the total amount of boron in the control rods by 1.2%.
However, this remains in compliance with the shutdown margins.and rod worths specified in the FSAR and Technical Specifications. The ability of the rods I
to be inserted will not change.
FSAR Sections 3.5.3.1, 3.8.1.2, D.1.3.3.2, D.3.3.1, and Figure 3.8-11 have been revised and Figures 3.8-13 and D 3-55 have been added to. incorporate information on the new control rod design.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed. safety question'.
CN-2175 System 70/ Structures - General System 78/ Security System System 83/ Communication System This CN authorized modifications to the Communications Room.
This room was enlarged to allow for additional equipment.
Modification to this facility has not compromised any previous FSAR analyses. No new accident modes were created and no margins of safety were affected by this activity.
FSAR Figure 1.2-18 has been revised to reflect this modification.
This activity was classified safety related, but was' not safety significant and did not involve an unreviewed safety question.
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CN-2176, 2176D System 21/ Primary Coolant System System 22/ Secondary Coolant System System 93/ Controls and Instrumentation This CN authorized installation of a Steam Line Rupture Detection and Isolation System (SLRDIS). This system utilizes sensing from four temperature rate-of-rise detectors in the Turbine Building and four detectors in the Reactor Building to automatically isolate selected high energy line breaks in the secondary coolant system. This system uses 2-out-of-4 transmission logic for actuation. The system is intended to preclude, through automatic action, a harsh environment which could render safety related electr' cal equipment inoperable prior to operator action. Due to the autcmatic isolations upon actuation of the SLRDIS, the l
probability-of occurrence of an interruption of forced circulation cooling has been increased.
SLRDIS interface into the Circulator Trip Logic will override existing design features and cause automatic shutdown of both primary and secondary coolant loops, and inhibit the auto-start of the water turbine 3
drives for a loss of forced circulation cooling. This means of overriding the two loop shutdown inhibit reduces the margin of safety for assuring continued forced circulation cooling.
However, accident consequences are not increased as restoration of forced cooling by the operators within the analyzed time of the FSAR is not precluded by SLRDIS operational design. No single failure / malfunction can result in the actuation of the SLRDIS safety function er preclude the safety function from actuating.
Technical Specifications LC0 4.4.1 and SR 5.4.1 were revised in Amendment #50 to reflect this modification.
Various FSAR sections, tables, and figures of Sections 1, 4, 7, 8, 10, 12, 14, and Appendix C have been revised to reflect this modification.
This activity was classified safety related and safety significant, and was originally determined to involve an unreviewed safety question.
Following receipt of Amendment #50 to the Technical Specifications and the NRC's Safety Evaluation Report, this activity was no longer determined to involve an unreviewed safety question.
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CN-2238 i
System 93/ Controls and Instrumentation This CN authorized removal of the Steam Pipe Rupture Detection System (SPRDS) safety. function 'from the PPS.
This system consisted of detectors and circuitry to initiate a loop shutdown upon the detection of a steam line rupture.. This system is no
' longer needed due to -the installation of the SLRDIS (See CN-2176). No FSAR analyses are affected and no margins of safety are decreased due' to removal of the SPRDS since the SLRDIS.
installation performs the preferred functions.
Technical Specifications LC0 4.4.1 and SR 5.4.1 were revised in Amendment #50 to reflect this modification.
Various FSAR sections, tables, and figures of Sections 1, 4, 7, 8, 10,-12, 14, and Appendix C have been revised to reflect this modification.
This activity was classified safety related and safety significant, but did not involve an unreviewed safety question.
CN-22578 System 22/ Secondary Coolant System This CN authorized documentation updates to the System 22 Safety Related Data Base, Essential Cable List and associated documents.
Essential parent components were identified and their related essential subcomponents and cables were added to respective parent component data fields.
Separation / segregation requirements have been maintained for essential cabling.
FSAR Figure 7.1-16 has been revised to incorporate additional safety related instruments. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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CN-2376 System 21/ Primary Coolant System 1
This CN authorized modification of the helium circulator s
L steam / water drain automatic backpressure control circuitry.
Steam / water drain backpressure is now controlled with feedback ~
from existing circulator lower bearing housing temperature sensing elenents.
This modification allows finer automatic i
control of the steam / water drain backpressure. This modification does not affect main drain to steam / water drain pressure differential circuitry, which provides PPS input, and does not affect' manual operation.
Circulator protection features remain unchanged.
FSAR Section 4.2.2.3.5 and Figure 4.2-12 have been revised to reflect this modification. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
CN-2393 System 63/ Radioactive Gas Waste System System 82/ Instrument and Service Air This CN determined the acceptability of discrepancies between the as-built configuration and the indicated configuration on P&I 63-2 of the Gas Waste System relative to Service Air interface. 'Two untagged valves were tagged and documents updated.
No physical modifications to the, plant resulted from this activity and documents reflect the as-built configuration.
This activity enhanced proper operation and maintenance.
FSAR Figure 11.1-2 has been revised to reflect the as-bwlt configuration. This activity was not safety related or safety significant, and did not involve an unreviewed safety question. i
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Cable #560 powers safety related instrument rack I-9343 from Interruptible Bus #3 distribution L
panel.
Required separation for redundancy and physical locations has been met. Additional tray loading is within design limits and the new cable was determined to be acceptable for its intended service and environment.
FSAR Figure 8.2-20 has been revised to reflect this modification.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
CN-2668 System 21/ Primary Coolant System This CN justified use of KALREZ elastomer (KALREZ 4079) as an alternate material to the installed metallic helium circulator primary seals, and justified a new shutdown seal due to existing vibrational problems.
KALREZ 4079 will perform as good or better as KALREZ 1045, which was previously evaluated by General Atonncs (GA) as an acceptable alternate to the installed metallic seal.
The new shutdown seal has the same critical dimensions, critical clearances, and is made'of the same material as the installed seal.
The new seals provide'the same functions as the existing seals. No new accident modes were created and no margins of safety decreased.
FSAR Sections 5.8.2.5.5 and E.23.3 have been revised to discuss the use of either elastomer or metallic seals on the primary closure of the helium circulators. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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..y CN-2672 System 92/ Accessory Electrical Equipment This-CN authorized replacement of Exide lead-calcium Station
-Batteries 1A, IB, and 1C due to their deterioration.
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batteries were replaced with Exide lead-antimony batteries. A l
review of the new batte ry load profiles, margins, and other design considerations indicated their acceptability.
Replacement of degraded batteries with new batteries of relative design equivalence did not decrease any margin of safety.
Technical Specifications LC0 4.6.1 and FSAR Section 8.2.3.4 have been revised to reflect this change.
Based on the corrected Safety Evaluation of CN Reissue C, this activity was classified safety related and safety significant, but did not involv6 an unreviewed safety question.
CN-2698 System 91/ Hydraulic Power System This CN authorized installation of equipment to the hydraulic power units for. performance monitoring. This equipment included flow meters permanently added to the system. Pressure gauges and flexible hoses will be connected to the system when needed for monitoring.
The increased system pressure drop due to the flow meter addition is negligible, and meter support meets seismic criteria for pressure boundaries.
This modification will allow system monitoring for troubleshooting purposes and therefore enhances system availability.
FSAR Section 9.11.3 and Figure 9.11-1 have been revised.to reflect this modification. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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l System 52/ Turbine Steam System System 84/ Auxiliary Boiler System i
This CN authorized rearrangement of the Backup Auxiliary Boile*
instrumentation, replacement of'the superheater, increasing the capacity of check valve V-52168, and other miscellaneous enhancements. This modification rejuvenated the Backup Auxiliary l
Boiler capacity to its originally designed capacity of 115,000 i
lb/hr. The Backup Auxiliary Boiler is not relied on for any i
analyzed accident.
No new accident modes have been created by this activity.
FSAR Sections 7.3.10.3.1 and 13.2.6 have been revised to reflect this modification and to standardize " backup auxiliary boiler" terminology.
This activity was not safety related or safety significant, and did not involve an unreviewed safety question.
CN-2765 System 31/Feedwater and Condensate Systems This CN authorized a change to the control circuitry for HV-31207, which is a 4 inch discharge bypass valve for motor driven Boiler Feed Pump P-3102. The valve will now automatically shift to manual control when HV-3109 (10 inch discharge valve for P-3102) fully opens.
This precludes the circuitry from futilely.
attempting to control differential pressure across HV-31207 when HV-3109 is fully open, and precludes the operators from having to manually shift control modes during this condition.
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modification does not hinder operator control of the valves. The function and operability of the pump and valves remain the same, and no new failure modes are introduced.
FSAR Section 10.2.3.1 has been revised to reflect this modification. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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l CN-2tg System 91/ Hydraulic Power System l
This CN' authorized remova'l of cap-side relief valves from HV-l 2223, HV-2224, HV-2253, HV-2.292, and HV-2293..These reliefs were determined not to be needed and the valve operators are not adversely affected by their removal.
Hydraulic oil flow capability, valve stroke time, system process flow, and isolation capabilities remain unchanged.
No margin of safety is affected 1
by tnis activity.
FSAR Figure 9.11-2 has been revised to reflect this modification.
This. activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
C1 2842, 2842A, 28428, 2842C System 93/ Controls and Instrumentation This CN authorized modification of the SLRDIS program to add a fixed temperature trip.and to adjust recorder start temperature.
TFe additional trip setpoint neither causes any new single ft.ilure-points nor compromises the system detection. The margin of safety for assuring the detection of steam leaks is enhanced l
by the addition of a new detection point at which protective action is automatically initiated,-
as required for the environmental qualification of safe shutdown systems.
FSAR Sections 1.4 6, 7.3.10, 14.5.1.1 and Technical Specifications LCO 4.4.1 and SR 5.4.1 have been revised to reflect this modification. This activity was classified safety related and safety significant, but did not involve an unreviewed i
safety question.
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L 2.0 00CUMENT CHANGE NOTICES (DCN)
.i DCN-74 l
l System 91/ Hydraulic Power System This DCN authorized document changes to delete relief valve Vr91694 from the cap side of HV-2254. Deletion of this relief valve does not affect the operation of the hydraulic system, HV-2254, or the Secondary Coolant System.
FSAR Figure 9.11-2 has been revised to reflect this change. This ace.ivity was classified safety related, but was not safety significant and did not involve an unreviewed safety questien.
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DCN-116 System /Various This DCN authorized the update of documentation associated with setpoint changes made in 1988 which satisfies, in part, a commitment ~ made to the NRC in P-86543 to maintain setpoint documentation. This activity included previously approved SCR's with' individual safety evaluations. No new safety concerns were created by this activity.
FSAR Section 9.6.6 has been revised to indicate that a low level alarm will be received prior to the liquid nitrogen storage tank reaching 650 gallons.
This DCN activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
DCN-119 System 13/ Fuel Handling Equipment This DCN authorized document changes to reflect the manufacturer's new part number for Reactor Isolation Valve motor drives.
Since the manufacturer's design change involved only the motor frame size, the motor can still perform the required design function; and the ability of the Reactor Isolation Valves to provide a pressure boundary will remain unchanged.
FSAR Figure 9.1-6 has been revised to reflect this change. This l
activity was classified safety related, but was not safety i
significant and did not involve an unreviewad safety question.
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3.0 SETp0 INT CHANGE REPORTS (SCR)
SCR 88-075 System 25/ Nitrogen System This SCR authorized setpoint changes to increase the pressure at which the Moisture Monitor (MM) liquid nitrogen supply dewars are controlled.
These changes were due to inadequate flow of nitrogen cooling for. proper temperature control of the MM mirrors.
This activity enhances proper MM operation, and does not exceed any-system design parameters or affect any FSAR analyses.
FSAR Section 7.3.2.2.2 has been revised to reflect this activity.
l This activity was classified safety related, but was not-safety significant and did not involve an unreviewed safety question.
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e 4.0 SPECIAL' TESTS (T-TESTS) i T-216 System 11/ Prestressed Concrete Reactor Vessel.
Purpose:
To determine the magnitude of difference between the indicated core power from power range detectors and that from a heat balance (Decalibration Factor) for Cycle 4 operation, and
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confirm that the calculated. decalibration factors used in determining the PPS setpoints for.' Cycle 4 are adequate.
Decalibration of the six PPS nuclear detectors occurs due to control. rod motion and its shielding effects.
The performance of this test involved taking data during i
ascension to power after moving certain control rod grotps and allowing power to stabilize. The comparison between the neasured.
and calculated decalibration data for. control rod group i
withdrawal is consistently as good as that for previous cycles and supports the analytical methods.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-381 System 93/ Controls and Instrumentation
Purpose:
Attempt to determine resolution of electrical noise problems in Instrument Panel I-10.
Electrical noise spikes have resulted in erroneous high circulator speed signals.
Arc suppressors were installed and tested.
Testing indicated that the arc suppressors were extremely effective in eliminating pulse test transients. A Change Notice i
made the suppressor installation permanent.
The reactor was critical throughout the duration of this test. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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l Th382 System 22/ Secondary Coolant System
Purpose:
To verify system response to the installation of zener
-diodes across the remote setpoint input resistors of the_ cold reheat.attemperation flow controllers prior to permanent installation of the diodes. The intended installation 're;ults from modification to the circuitry that provides feedwater system runback following a hot reheat steam scram, and an associated i
Ger.eral Atomics analysis report on a continuous rod withdrawal accident.
l The :ener diode installation limits controller input voltage to correspond to the maximum attemperation flow desired. No adverse t
system ' responses resulted and a'
Change Notice made the diode 1
installation permanent.
This installation and test did not decrease any margin of safety.
The reactor was critical throughout the duration of this test.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-383 System 63/ Radioactive Gas Waste System
Purpose:
To simulate Core Support Floor (CSF) vent flow during a plant scram and determine if the vent system can handle the flow without lifting relief valve V-6389. This test was conducted following maintenance on the CSF vent system aimed at reducing j
flow restrictions in the system.
CSF vent flow was redirected to the Gas Waste Vacuum Tank so that any vent flow through the Reactor Building stack would be from V-l 6389.
The test concluded that with V-6389 set to relieve at 10 l
p.s.i.g. instead of 5 p.s.i.g., and with the CSF vent pressure controller set at 95 p.s.i.g.
during periods of high flow, instead of 60 p.s.i.g.,
that the CSF vent system functions i
properly and does not lift V-6389.
This test involved planned l
releases of radioactive gaseous waste via paths previously analyzed in the FSAR.
Based 'on the recent primary coolant samples, no release would result in exceeding the Maximum Permissible Concentration, and the allowable release rate would not be exceeded.
The reactor was shutdown throughout the duration of this test.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question, l
- n-T-387 System 11/ Prestressed Concrete Reactor Vessel System 22/ Secondary Coolant System
Purpose:
To verify' a leak path between the Steam Generator Interspace and Cold Reheat Steam Piping on the Loop 2 Steam Generator, and to determine if a leak exists between the Interspace and Reactor Building atmosphere.
Steam Generator Interspace pressure was raised at intervals from 476 p.s.i.g.
to 645 p.s.i.g.,
resulting in a simultaneous increase in helium concentration at the outlet of Steam Jet Air Ejector. The test concluded the presence of a leak to the reheat piping and purified helium leakage from the interspace to the Reactor Building atmosphere.
Response to the test included locating and repairing interspace to atmosphere leaks. The margin of safety as defined in the Technical. Specifications for interspace. leakage was confirmed by this test. The reactor was critical throughout the duration of this test. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-389 System 31/Feedwater and Condensato Systems System 92/ Accessory Electrical Equipment
Purpose:
To test Boiler Feed Pump 1B DC Lube Oil Pump (P-3108SX) motor starting / running current and voltage to accurately update the load profile for recently replaced Station Battery 1B. Based on name plate and previous data, this pump is a major contributor to the Battery 1B load profile.
Conduct of the test revealed differences between actual data and I
name plate data. Worst case values, as measured during the test, were used in generating the load profile. During this test, the DC Lube Oil Pump and associated system were operated as designed, and no new failure modes were introduced.
The reactor was shutdown throughout the duration of this test. This activity was i
not safety related or safety significant, and did not involve an unreviewed safety question.
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4-T-390-System 31/Feedwater and Condensate Systems System 92/ Accessory Electrical Equipment-
Purpose:
To test Boiler Feed Pump IC DC Lube Oil Pump (P-31095X) motor starting / running current and voltage to accurately update
- the load profile for recently replaced Station Battery 18.
Based-on name plate and previous data, this pump is a major contributor to the Battery IB load profile.
Data collected during the test resulted in an overall reduction of the load profile. During this test, the DC Lube.0il. Pump and associated system were operated as designed,.and no new failure modes were introduced. The reactor was shutdown throughout the duration of this test. This activity was not safety related.or safety significant, and did not involve an unreviewed safety question.
T-391 System 51/ Turbine Generator and Auxiliaries System 92/ Accessory Electrical Equipment
Purpose:
To test Main Turbine DC Hydrogen Seal Oil Pump (P-5105X) motor starting / running current-and voltage to accurately update the load profile for recently replaced Station Battery 18. Based on name plate and previous data, this pump is a major contributor to the Battery 18 load profile.
Data collected during the test resulted in an overall reduction of the load profile. During this test, the DC Seal Oil Pump and associated system were operated as designed, and no new failure modes were introduced. The reactor was shutdown throughout the duration of this test. This activity was not safety related or safety significant, and did not involve an unreviewed safety question.
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T-397 System 75/ Turbine Building System 92/ Accessory Electrical Equipment
Purpose:
To veri fy that ventilation is adequate to remove hydrogen generated during the charging of Station Battery 1C.
Following the installation of Station Battery IC, this test monitored hydrogen concentration in the battery's room during a freshening charge.
The maximum concentration measured was approximately 0.2% hydrogen, which is well below the maximum 1.0%
limit.
Since the freshening charge results in a higher. voltage and current (and thus more hydrogen production) than the more frequent equalizing charges, this test concluded that the Battery IC room ventilation is adequate.
During.this
- test, the batteries, chargers, and ventilation system were operated as designed. Also, the test and charge would have been terminated upon reaching a hydrogen concentration of 1%. The reactor was shutdown throughout the duration of this test. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-398 System 45/ Fire Protection System System 92/ Accessory Electrical Equipment
Purpose:
To verify that the emergency lighting system, as enhanced by. CN-2002, will provide adequate illumination to operate / start-up various manual emergency equipment required for safe shutdown cooling of the reactor plant upon loss of normal lighting / electrical power.
Affected area lighting was deenergized and the adequacy of emergency area illumination was checked. The test concluded with certain deficiencies noted for correction and retest.
Announcements were made prior to deenergizing the lighting in each area.
Personnel were instructed to carry flashlights and instructed in their proper use during the test.
Provisions were made to terminate the test at any time plant conditions warranted.
The reactor was shutdown throughout the duration of this test.
This activity was not safety related or safety significant, and did not involve an unreviewed safety question.
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a T-401 System 92/ Accessory Electrical Equipment
Purpose:
To conduct an eight hour test discharge of Station Battery 1A, following installation, due to low specific gravity and a single low cell voltage. The test discharge was followed by a charge to return 125'4 of the amp-hours removed during the discharge, per the manufacturer's recommendations.
No improvements in specific gravity or the low cell voltage were noticed following the eight hour test discharge.
However, a subsequent service test discharge indicated the battery was adequate to meet the required load profile.
The reactor was shutdown throughout the duration of this test. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-405 System 92/ Accessory Electrical Equipment
Purpose:
To perform a post installation four hour service test discharge of Station Battery 1A.
This test was terminated eight minutes following its start due to improper functioning of the Battery Discharge Test Unit.
See T-406 for the performance of the service test discharge. The reactor was shutdown-throughout the duration of this test.
Technical Specification Surveillance Procedure SR 5.6.2c-Al was used as a guide for the conduct of this test. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-406 System 92/ Accessory Electrical Equipment
Purpose:
To perform a post installation four hour service test discharge of Station Battery 1A.
Station Battery 1A successfully passed the service test discharge. The results of this test meet all requirements of Technical Specification Surveillance Procedure SR 5.6.2c-A1. The reactor was shutdown throughout the duration of this test.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question. )
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-a T-408 System 92/ Accessory Electrical. Equipment
Purpose:
To verify the ability of Battery Charger ID to supply power to a DC Bus without a battery connected to the bus (battery eliminatormode).
A jumper was installed on Battery Charger'1D as required by the manufacturer for operating it in the battery eliminator mode.
The charger was placed on DC Bus IB while Station Battery 18 and Battery. Charger IB were removed from service. Equipment powered' from Bus IB was started to test the battery charger's response to the transient. Voltage initially dropped as a result of the transient, as expected, due to no battery available as a system buffer; This voltage dip was of sufficient length.to cau,e the Instrument Inverter IB.to transfer to an alternate AC power source.
The inverter was manually transferred back to DC power and operated ' satisfactorily.
The test resulted in Battery Charger _10 operating satisfactorily -in the battery eliminator mode.
The test also concluded that the DC Buses can be alternately powered while maintaining independence from each other, meeting Technical Specification requirements. 'The reactor was shutdown throughout the duration of this test. This activity was' classified safety related, but was not safety significant'ano did not involve an unreviewed safety question.
T-409 System 21/ Primary Coolant System
Purpose:
To determine the operational adequacy of newly installed P-2103SX (Turbine Water Removal Pump 1C).
Operational testing concluded that P-2103SX could pump to the condenser if downstream control valve (PCV-31213) setpoint does not exceed 10 p.s.1.g.
This pump cannot pump to the deaerator.
Test results were satisfactory.
The reactor was shutdown throughout the duration of this test.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question. i
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T-411 System'92/ Accessory Electrical Equipment
Purpose:
To evaluate voltage and amperage response of DC Bus 1B when being powered by Battery Charger 10 (battery eliminator mode) following modification of.this battery charger by a change notice to permanently install the required jumper described under T-408 above.
Battery Charger ID was placed on DC Bus 18 while Station Battery 1B and Battery Charger 18 were removed from service.
Equipment powered from Bus IB was started to test the battery charger's response to the transient.
Voltage initially dropped as a result of the transient, as expected, due to no battery available as a system buffer. On.this occasion, Instrument Inverter IB. output frequency dropped to 56 Hertz and the normal DC power indicating light extinguished momentarily. However, the transient was of a short duration and the inverter did not complete power transfer..
The test concluded that with a larger load on the DC bus and/or a larger starting load the transient duration would be longer and the inverter would probably shift power sources.
However, this is not of consequence and the. test concludes operation of Battery Charger 10 as a power source to a DC bus is satisfactory.
The reactor was shutdown throughout the duration of this test. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-412 System 46/ Reactor Plant Cooling Water System
Purpose:
To determine which Core Support Floor (CSF) cooling water. tube in Subheader #238 was leaking.
Us ng recently installed individual CSF cooling tube isolation va ves and test-connections, and temporary test gauges,. cooling t'aes were ir,dividually isolated and monitored to determine which t;be(s) increased to PCRV pressure.
Cooling tube F8T10 was found to be the leaking tube. During this test, the potential existed for primary coolant to leak into the Reactnr Building.
Due to the length of shutdown, primary coolant activity was less than the Maximum Permissible Concentration by a factor of ten.
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reactor was shutdown throughout the duration of this test. Also, the Reactor Building ventilation and exhaust syrtem was maintained in operation.
Therefore, expected primary coolant i
j leakage to the Reactor Building was appropriately processed and i
did not exceed any defined limits. This activity was classified safety related, but was not safety significant and did not l
l-ir,,21ve an unreviewed safety question.
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T-413 System 48/ Alternate Cooling Method (ACM)
Purpose:
To provide an additional freshening charge to cells #1 through #4 of the ACM Diesel Starting Battery due to low specific gravity.
Following the freshening charge of the newly installed ACM Diesel Starting Battery, the specific gravity of cells #1 through #4 was lower than expected.
Per vendor recommendation, a portable battery charger was used to provide an additional freshening charge to cells #1 through #4 for approximately two weeks.
The final specific gravity was above the minimum acceptable value.
The ACM Battery remained in service and the reactor was shutdown i
throughout the duration of this charge.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-415 System 22/ Secondary Coolant System
Purpose:
To isolate the Economizer-Evaporator-Superheater (EES) section of either Steam Generator and monitor pressure trends to determine if a Steam Generator tube leak exists.
Loop 2 Steam Generator EES section was isolated from the feedwater and main steam systems.
PCRV pressure was held relatively constant at about 175 p.s.1.a.
The solid EES section was connected to a pressure indicator and monitored for approximately three days. During this period, the EES pressura i
remained between 64 and 68 p.s.i.g., and did not approach PCRV pressure.
This test provided enough information to conclude that no tube leak exists.
Loop 1 EES was not tested.
During the test, systems required for safe shutdown cooling remained capable i
of providing their safety function.
The reactor was shutdown throughout the duration of this test.
This activity was classificd safety related, but was not safety significant and did not involve an unreviewed safety question..
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T-416 System /Various i
Purpose:
During performance of Technical Specification l
Surveillance Procedure SR 5.6.le-1.5Y,
" Loss of Offsite i
Electrical Power," six components (not affected by clearance 1
procedures) failed to auto start during the sequenced return to operation.
T-416 was generated to determine operability of the auto-start control circuits for the affected plant components.
Electrical jumpers were installed to simulate the sequenced loading of the electrical buses follow og Emergency Diesel 1
Generator start-up/ loading. Operation of the drum sequencer and the auto-start relays were visually observed.
Test sequences were completed satisfactorily with no failures noted.
(See Licensee Event Report 88-015 for additional information.)
Equipment operational status was not affected during this test.
The reactor was shutdown throughout the duration of the test.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
T-417
. System 48/ Alternate Cooling Method (ACM)
Purpose:
To provide an additional freshening charge to cell #14 of the ACM Diesel Starting Battery due to low voltage.
Several weeks following the freshening charge of the newly installed ACM Diesel Starting Battery, the float voltage of cell
- 14 was lower than minimum acceptable.
Per vendor recommendation, a portable battery charger was used to provide an additional freshening charge to the cell for seven days. The final voltage for cell #14 was above the minimum acceptable value, following the charge. The ACM Battery remained in service and the reactor was shutdown throughout the duration of this charge. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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5.0 PROCEDURES S0P 25-01, Issue 14 System 25/ Nitrogen System
" Nitrogen System" operating procedure was revised, in part, to include procedural coverage for level controllers (LC-2524 and LC-2507) which were permanently installed by CN-1932.
These controllers accommodate level control for T-2501 and T-2502 (See CN-1932 in Section 1.0 of this report). This procedure stipulates a normal control range for the controllers and meets the minimum 650 gallon requirement for T-2501 per the Technical Specifications.
FSAR Section 9.6.4 has been revised to reflect this change. This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
50P 63-01, Issue 28 System 63/ Radioactive Gas Waste System
" Radioactive Gas Waste System" operating procedure was revised, in part, to provide for placing both Core Support Floor (CSF) vent filters in service during periods of high flow, and to increase CSF pressure control from 60 p.s.i.g. to a maximum of 95 p.s.i.g.
at the Shift Supervisor's discretion.
The FSAR discusses the acceptability of controlling CSF pressure up to 100 p.s.i.g.
without resulting in structural deformation. Placing both CSF filters in service during certain periods of high flow, with permission of the Shift Supervisor, does not increase the probability of equipment malfunction.
FSAR Sections 5.9.2.4, 11.1.1.3, 11.1.2.3, 11.1.3.4, and 11.1.4 have been revised to reflect these changes.
This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.
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RERP Section 9, Issue 9 System /None RERP Section 9, " Recovery", describes administrative requirements involved with recovery from a RERP event. The organization and specific recovery responsibilities are now given in RERP Section 9.
FSAR Sections 12.3.2, 12.3.3, and 12.3.4 have been updated based on this Issue.
This activity was not safety related or safety significant, and did not involve an unreviewed safety question.
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