ML20064C228

Annual Operating Rept Jan-Dec 1993

ML20064C228
Person / Time
Site:	Cooper
Issue date:	12/31/1993
From:	Horn G NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD940132, NUDOCS 9403090214
Download: ML20064C228 (29)

Text

I 4

1 G"S y: r 1sternone sc2>e2,B3ei,M 68321 P.O. BOX LL NE m

s c

Ay Nebraska Publ.ic Power Distr. t=.. _. _ _ _. _ _ _. - _ _ _ _. = = = _ "

ic

=*==2 as s

=== = ===__ _

y NSD940132 March 1,1994 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

Subject:

Annual Operating Report Cooper Nuclear Station NRC Docket No. 50-298, DPR-46 In accordance with Paragraph 6.5.1.C of the Cooper Nuclear Station Technical Specifications, the Nebraska Public Power District, hereby submits the Cooper Nuclear Station Annual Operating Report for the period of January 1,1993, through December 31,1993.

We are enclosing one signed original for your use and, in accordance with 10 CFR 50.4 are transmitting one copy to the NRC Regional Office, and one copy to the NRC Resident Inspector for Cooper Nuclear Station.

Should you have any questions or comments regarding this report, please contact me.

Sine rely,

/ r,_ ~

t i

CrR. IIorn Vidii ' resident Nuclear GRll/tja:93an-rpt.ltr Attachment cc:

NRC Regional Office Region IV NRC Resident Inspector Cooper Nuclear Station REIRS Project Manager (w/ Personnel Man-Rem Report only)

Office of NRR - USNRC Washington, DC l

,.. l,.b

/

t 6

x 8

l 9403090214 931231 PDR ADOCK 05000298 S

PDR

,,c $

qrmne ygp^ O Ypb ^ynemrewnwvWw=P"Tm'"'"7 "

$$ymen; hila $3r CBMCIDOrd ' O.,A *' *f~$5gA.

SUlb ran g *Ee" nam adadu M M "'" "

n COOPER NUCLEAR STATION BROWNVILLE, NEBRASKA ANNUAL OPERATING REPORT JANUARY 1,1993 THROUGH DECEMBER 31,1993 USNRC DOCKET 50-298

I TABLE OF CONTENTS SECTION PAGE I.

PERFORMANCE CHARACTERISTIC 1

i Fuel Performance 2

MSV and MSRV Failures and Challenges 2

II.

FACILITY CHANGES, TESTS, OR EXPERIMENTS REPORTABLE UNDER 10CFR50.59 3

Reportable Special Procedures /Special Test Procedures 4

Reportable Design Changes 9

Reportable Activities (Setpoint, Procedure Changes) 21 Ill.

PERSONNEL AND MAN-REM EXPOSURE 25 By Work and Job Function 26 i

s 2

m-F b

e 1

i l.

PERFORMANCE CHARACTERISTICS i

1

FUEL PERFORMANCE Cycle 15 operation continued from January 1.1993, through March 4,1993. The unit was shut down on March 5,1993 to begin Reload 15 (RE15) refueling / maintenance outage. Cycle 16 was started on July 21,1993, with the reactor mode switch being placed to RUN on July 23,1993. Due to site area flooding, the unit was shut down on July, 23, 1993. The unit was restarted on July 29,1993, and continued operation until December 14,1993. The unit experienced an automatic scram due to reactor feed pump control failure. The unit restarted on December 19,1993, and continued operation through December 31,1993.

Cycle 15/16 off-gas activity continued at essentially steady state levels with reactor coolant dose equivalent I-131 equilibrium values and off-gas release rates maintained well within the limits specified by the CNS Technical Specifications. Compansons of actual control rod densities predicted by computer program calculations at various core exposures indicated no reactivity anomalies of 1% or greater.

1 MSV AND MSRV FAILURES AND CHALLENGES (Ref ' NUREG-0737, Action Item II.k.3 3)

There were no failures or challenges to the Safety Valves during 1993.

During 1993, there was one challenge to the MSRVs, this includes all startups and shutdowns during 1993. MSRV-F inadvertently opened at 1224 on December 28,1993, during performance of Surveillance Procedure 6.1.12, " ADS Reactor Pressure Permissive Calibration and Function / Functional and Logic Tests (Reactor in RUN)". The incident was attnbuted to personnel error. The valve actuation was satisfactory.

There were no Failures to the Safety Relief Valves.

L 1

2 l

1 II.

FACILITY CHANGES, TESTS, OR EXPERIMENTS REPORTABLE UNDER 10CFR50.59 l'

l 3

, RdFMRTABLE SPECIAL PROCEDURES (SPs)I SPECIAL TEST PROCEDURES (STPa)

SP 90-180_

TITLE:

Fire Seals Visual Inspection DESCRIPTION: This Special Procedure provided detailed instructions necessary to verify the integrity of the Dow Corning RTV silicone foam fire penetration seals. This was accomplished by performing a direct visual inspection of readily visible fire penetration seals or fire bairier assemblies in the fire area barriers which consisted of Dow Corning RTV silicone foam with or without damming boards.

SAFETY ANALYSIS:

The fire barriers perform no safety-related function nor support the operation of safety equipment. The purpose of the barrier is to provide a passive method of fire separation from prevalent fire hazards within the plant and/or separation between redundant safety-related components in the event of a fire. This Special Procedure (SP) defined the method for performing penetration sealinspection and documented the results This SP did not authorize any deviation from approved CNS procedures and did not authorize any equipment alterations.

SP 91-038 and STP 91-038 TITLE:

Torus Coating Inspection and Repair

'ESCRIPTION. This Special Procedure (SP)in conjunction with Special Test Procedure (STP)91-038 provided for the inspection and repair, as needed, of the coating system on various underwater portions in four bays of the Primary Containment Suppression Pool (Torus).

SAFETY ANALYSIS:

The work performed by this SP and STP was performed during a refueling outage while the plant was in a cold shutdown condition and primary containment integrity was not required.

SGPAl Procedures OCP-10-1-CNS-7068, " Underwater Coating inspection", QCP-10-2-CNS-7068, " Underwater Coating Repair", and WP-1-CNS-7068, " Diving Operations in Radiological Contaminated Areas" was used to control this work and to minimize potential personnel dangers. The probability of occurrence or the consequences of an accident or malfunction of equipment important to nuclear safety previously evaluated in the USAR were not increased.

No operational aspects of any safe shutdown system were affected or changed by this SP and STP.

SP 91-091 TITLE:

Service Water Booster Pump (SWBP) Motor A Removal and Replacement DE 3CRIPTION: This Special Procedure authorized RHR Service Water Booster Pump (SWBP) Motor A removal and rigging. Additionally, SWBP motor inspection, maintenance as necessary, re-installation, and post maintenance testing of the SWBP motor were covered by the SP.

ANALYSIS:

The removal and replacement of SWBP motor A was performed while the reactor was in a cold shutdown condition, during which, the B Service Water loop was maintained in an operable status. Acceptance testing was performed in accordance with CNS Surveillance Procedure 6.3.20.1, "RHR SW Booster Pump Flow Test and Valve Operability," to ensure IST and Technical Specification operabilLy criteria were satisfied after the motor was re-installed. This j

SP did not create an accident or malfunction of a different type, decrease the margin of safety, j

nor did it allow operations of plant systems that were not previously evaluated in the USAR.

j SP 93-057 TITLE:

RHR Shutdown Cooling / Fuel Pool Cooling Parallel Operation DESCRIPTION: The purpose of this Special Procedure was to verify that temporary spent fuel pool cooling can be maintained by lowering the Reactor Cavity weirs 1" below the elevation of the fuel pool weirs with the RHR system in Shutdown Cooling operation and the Fuel Pool Cooling (FPC) system circulating pumps operating. The SP also venfied that, if required, temporary spent fuel pool cooling can be maintained by realigning the FPC system to discharge to the Reactor Cavity through existing diffuser piping and utilizing the existing RHR system to FPC inter-tie to supply cooled water to the pool.

4

.- ~

SAFETY

' ANA' LYSIS:

The RHR B loop heat exchanger has sufficient capacity to maintain the fuel pool temperature below USAR limits when the reactor is shutdown per General Electric (GE) specification i

22A1472. In addition, RHR flow through the Class llS FPC system piping did not result in an increase in the probability of occurrence of a pipe break since the FPC system design pressure is greater than the pressure resulting from a RHR system diverted flow of approximately 1000 l

gpm. Should a seismic event occur during the time RHR flow is diverted through the Class llS piping, a pipe break may occur resulting in approximately 1000 gpm being released to the Reactor Building until the operator who was stationed at the inter-tie valve closes the valve.

Manual closure of the valve was conservatively estimated to require less than 1 minute.

Therefore, a maximum of 1000 gpm would be lost during the seismic eventm not enough to result in uncovering of fuel er in reducing the fuel pool level below Technical Specification limit cf 8.5 feet for radiation shielding protection. In addition, this volume of water is well within the capacity of the Reactor Building sump.

The ability of the RHR' system to maintain the fuel pool temperature while in concurrent i

SDC/FPC was verified. Furthermore, Fuel Pool temperature was continuously monitored in the Control Room and the FPC trouble alarm would alert operators in a timely manner of any problems in maintaining fuel pool level. Additionally, the RHR A loop and the Core Spray system were available for assisting in vessel inventory control if required during the -

performance of this SP. Therefore, this SP did not increase the probability of accident occurrence; create a previously unidentified accident; reduce the margin of safety as defined in the Technical Specifications or violate existing CNS Technical Specifications.

STP 88-003 TITLE:

GE10 Lead Test Assemblies DESCRIPTION: This Special Test Procedure addressed four (4) Lead Test Assemblies (LTAs) designed and manufactured by GE forloading in the Cooper Nuclear Station core as part of reload 11. The LTAs contained design features of the GE10 fuel _ assembly design. The purpose of the LTAs is to confirm the expected performance of the GE10 features with respect to both mechanical and nuclear characteristics.

SAFETY ANALYSIS:

The LTAs were interchangeable with other fuel contained in the Cooper core, from both a mechanical and nuclear standpoint. The LTAs required no changes to the manner in which the reactor is operated because they were placed in non limiting core locations. The LTAs met all the limits prescribed in NEDE-24011-P-A for the applicable operating conditions, including accident conditions. Additionally, analyses were performed confirming the applicability of the Technical Specification limits to the LTAs therefore, assuring that Technical Specification limits..

for the fuel would be met. Thus, the LTAs did not create the possibility of a new or different kind of accident, and the margin of safety was not reduced.

STP 88-005 Amendment 3 TITLE:

Specimen Irradiated Capsule Program (Installation / Removal)

DESCRIPTION: The purpose of this Special Test Procedure (STP) was to address the installation and removal of a specimen capsule containing structural materials that may be utilized in future fuel designs -

i by GE.

The intent of this program is to allow specimen irradiation and to provide data to demonstrate the improved resistance to corrosion for zirconium alloys, swelling of hafnium, and -

to irradiation-assisted stress corrosion cracking for special stainless steels.

SAFETY ANALYSIS:

The irradiation capsule was installed into the existing neutron source holder position in the l

core. The only impacted system associated with this STP was the reactor core. There is no significant difference between capsule and original neutron source holder and there was no impact on lattice physics calculations. Additionally, analysis performed showed that the specimen would not escape the capsule so no new type of accident was created. This STP.

did not change the plant facility or procedures as described in the USAR or the Technical Specifications. All safety aspects were reviewed, and it was determined that there was no possibility of an accident or malfunction of equipment important to safety resulting from performance of this test procedure.

5.

~__

' STP 88-006 Tl.TLE:

Pilot Production Zircaloy Fuel Channels DESCRIPTION: The purpose of this Special Test Procedure was to permit the use of pilot production zircaloy-2 and zircaloy-4 fuel channels manufactured by GE in Cooper Nuclear Station fuel bundles in order to gain in-core operating experience and data on the improved corrosion resistant of the materials and processes used in the fabrication of these channels.

SAFETY ANALYSIS:

The probability of occurrence of an accident or malfunction of equipment important to safety as evaluated in the USAR was not increased as a result of installing the pilot production -

channels, since these channels are virtually identical to the current production channels supplied by GE. The pilot production channels are interchangeable with the other channels in terms of nuclear and mechanical performance. The worst case scenario of catastrophic failure of a fuel channel, which could have resulted in the channel breaking into sections, damaging the attached fuel assembly, or restricting flow to other fuel bundles is conservatively bounded -

by the Loss-of-Coolant Accident postulated in the USAR.

STP 90-064 TITLE:

Condenser Protective Tube Shields DESCRIPTION. The purpose of this Special Test Procedure was to install six condenser protector shields on tubes located in the area adjacent to the main steam bypass sparger. The shields were installed during the 1991 outage and were removed during the 1993.

SAFETf ANALYSIS:

The installation of the six condenser protector tube shields did not affect the operation of the condensers. The intent of the tube protector shields was to reduce wear on the condenser tubes due to high energy steam, thereby reducing condenser tube leaks. This STP did not make any functional changes in system operation, components, or equipment. No possibility of an accident or malfunction of a different type than previously evaluated in the USAR or Technical Specifications was created as a result of this change.

STP 90-197 TITLE:

Drywell Sand Cushion Drain Vacuum Test DESCRIPTION: The purpose of this Special Test Procedure was to prove Drywell Sand Cushion Drain Operability by performing " Vacuum Tests" on the Drywell Sand Cushion Drain lines. This test i

was performed as part of the District's action plan for dispositioning NRC (IN) 86-99, and (GL) 87-05 which were issued to alert licensees about potential corrosion of the drywell liner due to trapped water in the sand cushion area.

SAFETY ANALYSIS:

This STP gathered information to be used by engineering to evaluate whether the Sand Cushion Drain Lines were open. The sand cushion drain lines are used to remove moisture from contact with the outer drywell steel liner, reducing its potential for corrosion. This STP did -

not alter the capabilities of the Sand Cushion' Drain Lines, nor did it change the function of the -

affected components, the margin of safety was not reduced, and all. accident analyses documented in the USAR remained bounding.

STP 90-352A and Amendments 1 & 2 -

TITLE:

Motor Operated Valve (MOV) Design and Switch Testing DESCRIPTION: The purpose of this Special Test Procedure was to perform in-situ testing of certain Motor Operated Valves (MOVs) to obtain data which will be used to provide assurance that they will function when subjected to design basb conditions. This STP and Amendments were implemented as part of the District's MOV Program Propet formed to respond to the NRC's Generic Letter 89-10 and Supplements. Amendment 1 to this STP performed in-situ testing on Core Spray Valve CS-MOV-MO5A, and Amendment 2 to this STP performed in-situ testing on Core Spray Valve CS-MOV-MO5B.

6 a

m m

SAFETY

' ANALYSIS:

All scenario test conditions performed by this STP were.within the original design basis conditions. This STP provided verification of the design and motor-operated switch settings of.

i specific valves using normal system operation, design basis accident operating scenarios, and

'l off normal system / plant configurations; Plant equipment was operated in accordance with applicable CNS approved procedures, Technical Specifications and this STP. The equipment tested was retumed to the as found position or the position required by the Operations Department after completion of the testing of each valve. This STP was performed when those portions of the affected systems were not required to perform their safety function as governed by the limiting conditions of operation of the Technical Specifications.. The testing performed by this STP did not in any way degrade the Cooper Nuclear Station with respect to personnel, equipment, or nuclear safety. Performance of this STP did not present an unreviewed safety question as defined in 10CFR50.59.

1 STP 92-016 TITLE:

Quarterly Local Leak Rate Testing on Primary Containment Purge and Vent Valves DESCRIPTION: The Purpose of this Special Test Procedure was to provide guidance for performing quar 1erly leak rate tests on the primary containment purge and vent isolation valves. The valves are normally leak tested once per operating cycle per 10CFR50 Appendix J.

However,- NRC -

concems have prompted increased frequency of the tests to verify the adequacy of the valves' new resilient seats. These quarterly tests will be performed until it is demonstrated that the valve seats do not exhibit degradation due to aging.

SAFETY ANALYSIS:

Since the tests described in this Special Test Procedure were performed with both the inboard and outboard valves of each purge and vent penetration in the closed position, the quarterly leak tests did not degrade the Primary Containment isolation capability of the subject valves.

This STP did not require abnormal operation of any plant systems or procedures, did not alter any plant equipment or have any effect overall plant safety.

STP 92-034 TITLE:

DC Motor Performance Test DESCRIPTION. The purpose of this Special Test Procedure was to investigate DC motor-performance characteristics as a function of voltage. This was accomplished by the use of a Lab Volt dynamometer and DC motor modules to collect various motor parameters as a function of supply voltage and motor load. The main elements of this STP were calibration of the dynamometer, configuration of the motor windings, and collection of performance data.

SAFETY ANALYSIS:

This STP gathered engineering data to be used in characterizing DC motor performance as a function of voltage.' The testing was completed on spare DC motors in the lab on a workbench and did not require the use or operation of;any plant system or component and therefore did.

not affect any plant system or component.

STP 92-092 TITLE:

HPCI-CV-17CV Open Flow Test Verification i

DESCRIPTION: The purpose of this Special Test Procedure was to determine if flow testing of HPCI-CV-17CV, in the HPCI pump minimum flow line is achievable with the use of a clamp-on flow meter. The flow measurement was attempted during the performance of Surveillance Procedure (SP) 6.3.3.1. This STP provided instructions to defeat an interlock for HPCI-MOV-MO25 closure during-the performance of SP 6.3.3.1.

SAFETY ANALYSIS:

This STP did not affect any aspect of the HPCI system operation or design. Even though the -

HPCI minimum flow was not isolated for a short period of time during this test, instructions existed and a operator was standing by to restore isolation upon receipt of an automatic initiation signal. This STP documented actions to be taken if an automatic initiation signal for HPCI was received. This STP did not prevent HPCI initiation. Compensatory measures were put in place and operators were standing by to compensate fc. the automatic actions which were defeated during the performance of this test. The defeated interlock was immediately put back into service at the completion of this test.. HPCI initiation and isolation functions as i

required by the Technical Specifications were maintained.

7

~.-

,. STP 93-036 TITLE:

Elevated Release Point Flow Measurement DESCRIPTION: The purpose of this Special Test Procedure was to determine the optimum rotameter insertion depth by obtaining the flow profile of the gases being expelled out of the Elevated Release Point (ERP) and converting them to a standard. Other flow indication equipment was also' monitored to determine their accuracy.

SAFETY ANALYSIS:

The system design parameters for both the Off Gas system and the Elevated Release Point were not exceeded with the performance of this STP..These systems are non-safety related -

and the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated h the USAR was not increased. All margins of safety as defined by Technical Specifications, USAR, and plant procedures were maintained.

STP 93-077 TITLE:

Service Water Pump Room Fan Test DESCRIPTION: This Special Test Procedure established the operability of the Service Water Pump Room (SWPR) Halon 1301 system during normal plant operating conditions, and at a time when four 2" diameter schedule 40 pipes were open to the area below the SWPR while the plant was in a cold shutdown condition. To complete this STP a door fan test was completed on the SWPR to assure that it would take longer than 10 minutes for the Halon 1301 concentrations to fall below 5% after a full discharge (per NFPA 12A 1992 edition). Halon 1301 was not discharged into the room as part of this STP.

SAFETY ANALYSIS:

This STP measured the retention time to maintain a 5% concentration of Halon after a full discharge and did not involve any physical changes to plant safety related equipment. The SWPR temperature was monitored and controlled at all times during this STP, and controls for terminating the STP were in place, if the SWPR temperature reached limiting temperatures for essential equipment. The administrative limits were established to ensure that margins existed before any degradation to the motors took place and original design document limits were not approached or exceeded during the STP. Therefore, this STP did not constitute an unreviewed safety question.

STP 93-086 TITLE:

Off Gas Dilution Fan Flow Test DESCRIPTION: The purpose of this Special Test Procedure was to obtain and graph information that would accurately represent the flow for each Off Gas (OG) fan. The test was performed with one OG fan running with no other flow inputs to the Elevated Release Point. A graph of the dilution fan flow as a function of recorded differential pressure will then be made from the data collected.

SAFETY ANALYSIS:

Dilution fan flow is only necessary to provide a redundant approximation of Elevated Release Point flow rates if normal indication is lost. The systems and components involved in this test were not safety related. Additionally, since system design pararneters for the OG system were not exceeded, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the USAR was not created. All margins of safety as defined by Technical Specifications, USAR, and plant procedures were maintained.

8

REPORTABLE DESIGN CHANGES DC 86-005 TITLE:

Fuel Pool Skimmer Surge Tank Level Gauge DESCRIPTION: The purpose of this Design Change was to install a local level indicator gauge to provide operators indication of the water level in the fuel pool skimmer surge tanks.

SAFETY ANALYSIS:

Failure of the level gauge or the tubing can in no way cause draining of the fuel pool because the skimmer surge tank water inlet is from the fuel pool overflow into the tank. This modification did not make any functional changes in system operation, components, or equipment. No possibility of an accident or malfunction of a different type than previously evaluated in the USAR or Technical Specifications was created as a result of this change.

DC's 87-01SMF TITLE:

CNS Annunciator Upgrade Project Panels 9-4, 9-5, B, C, J, & S DESCRIPTION: This Design Change (DC) was a continuation of the Detailed Control Room Design Review (DCRDR) Annunciation Upgrade Project. This Project replaced the existing non-essential (not safety related) Cooper Nuclear Station (CNS) Control Room Annunciator System with an upgraded non-essential Annunciator system which incorporates reflash and sequential events recording capabilities and is designed to meet Human Factors Engineering (HFE) guidelines.

4 This DC accomplished the replacement of the existing Panalarm Control Room Windowboxes in the following Control Room Panels (9-4,9-5, B, C, J. & S) with new windowboxes, Control Room Supervisor's CRT, alarm printers, panel CRTs, and the necessary connections.

SAFETY ANALYSIS.

This DC improved the annunciator system performance and reliability as well as resolved Human Factors Engineering deficiencies. This DC did not affect the actual systems (alarm inputs) that the affected panels monitor. The performance and reliability of the systems which provide input to the Annunciator system were not changed b/ this DC, and the Plant was in a cold shutdown condition while the work was implemented. The Annunciator system is a non-essential system and its operation, although desirable, is not necessary to obtain safe shutdown of the plant. A failure of any portion, or the entire Annunciator system will not jeopardize the j

plant safe shutdown capabilities.

j l

Although the Annunciator System interfaces with safety systems, neither an Annunciator System failure or an interface failure will jeopardize the plant safety system capabilities. This DC did not modify the function of any safety system. The upgraded Annunciator System only utilizes passive monitoring of existing instrun.cntation contacts. In addition, all applicable Technical Specifications were adherbd to during implementation of this Design Change.

Therefore, these modifications did not change the existing accident analyzes for Cooper Nuclear Station, nor the probability or consequences of an accident as analyzed in the USAR.

No reduction in the margin of safety was involved with implementation of this DC.

DC 88-302B TITLE:

Pipe Support Mod:fications DESCRIPTION: The purpose of the Design Change was to modify, remove, and install pipe supports, these modifications ensured that the member design stresses for the supports remain below code allowables. These modifications were the result of an upgraded pipe stress and hammer '

analysis.

SAFETY ANALYSIS:

All modifications performed under this DC were performed with either proper restraint in place, approved engineering analysis which documented that the system remained operable, or during a cold shutdown condition when the system was not required to be operable. Implementation of this DC fully code qualified existing piping supports in accordance with the USAR. This DC did not in any way degrade any safety related equipment or components, this DC met or exceeded these quality standards established by the original installation of the interfacing systems, all seismic design are within full code compliance. Therefore, the ability of affected systems to perform their safety function was unchanged.

9

DC 89-138

]

TITLE:

Instrument Air (IA) Isolation Valves for Fire Protection (FP) Accumulators j

i DESCRIPTION: This Design Change installed a 1/2" manual isolation valve on each of the two Instrument Air (IA) lines supplying the Fire Protection (FP) System 9 and System 10 accumulators. The new valves will allow maintenance work on the accumulators without isolating a portion of the lA system, which also supplies components required for operation of the Reactor Feed Pumps.

SAFETY ANALYSIS:

This Design Change was completed during cold shutdown because the Reactor Feed Pumps are not required for operation during cold shutdown. A roving fire watch was implemented during this modification while the FP system 9 and system 10 were out of service. Installation of the manual isolation valves did not degrade the air flow to any components in the IA system i

or change any operating characteristics of components. The addition of these passive components did not impact any fail-safe functions of the lA system. Therefore, this DC did not create an unreviewed safety question nor have any adverse effect on nuclear safety.

DC 89-2528 j

TITLE:

Trim Modification to RHR Valve RHR-MOV-MO34A 1

DESCRIPTION: The purpose of this Design Change was to install anti-cavitation trim in valve RHR-MOV--

MO34A. This valve is the RHR Suppression Pool Cooling "A" Loop Throttle Valve. The trim modification consisted of replacing the existing disc and seat ring. The modification required complete disassembly of the valve, machining out the old seat ring, welding in a new seat ring, and valve reassembly.

SAFETY ANALYSIS:

System performance was not adversely affected by this Design Change. The intent of-installation of the new valve trim was to eliminate wall thinning of the RHR Valve due to cavitation. The new tnm did add flow resistance. Flow testing was conducted after installation to verify that the minimum flow requirements would be achieved for Suppression Pool Cooling.

The function and operation of this valve did not change. Therefore, this activity did not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR.

DC 89-252E 1

TITLE:

Trim Modification to RHR Valve RHR-MOV-MO27A DESCRIPTION: The purpose of this Design Change was to install anti-cavitation trim in valve RHR-MOV-MO27A, and replace the existing RHR pump restriction onfices with new restriction orifices in the pump discharge lines. In addition, the existing Limitorque Model SMB-4 operator on the RHR-MO27A Valve was replaced with a new Limitorque SB-4 operator. The trim modification 3

consisted of replacing the existing disc, complete disassembly of the valve, machining out the old seat ring, welding in a new seat ring, and valve reassembly.

SAFETY ANALYSIS:

System performance was not adversely affected by this Design Change. The intent of.

installation of the new valve trim was to eliminate wall thinning of the RHR Valve due to cevitation. A flow test was conducted after installation to verify the required flows can be achieved for LPCI Injection. The function and operation of this valve did not change. The..

limitorque operator was replaced because it was an obsolete model, the new model has an i

improved design which incorporates a spring pack to compensate for expansion and contraction as well as inertia effects. Therefore, this activity did not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR.

DC 89-286C TITLE:

Phase 111 Performance / Reliability Monitoring Instrumentation DESCRIPTION: This Design Change installed instruments to monitor the performance and reliability of safety related and non safety related components. The safety related components were located in l

the Diesel Generator Diesel Oil System, and the Residual Heat Removal System. The non safety related components were located in the Circulating Water System, and the Plant Management Information System.

These instruments were included in the Performance /ReliabiUty Monitoring Program implemented at CNS. The installation of these 10 4

1

_. _. ~ _,

~

-- ~..

instruments will allow for the collection of data required for in-depth analysis of system components and their performance. This analysis will further ensure the reliability of the safety systems by providing valuable trending data.

SAFETY ANALYSIS:

No personnel, equipment, or nuclear safety concerns existed with the implementation of this Design Change (DC). The components installed per this DC are passive components (local indicators and sensors) and do not provide any controlling functions, and all separation requirements were maintained. The addition of these instruments does not affect the operational characteristics of any of the safety related or non-essential components / systems.

Since there was no change in any system components or operating characteristics, the effect on overall plant safety was not changed.

DC 90-023 TITLE:

Fuel Oil Boiler "B" Removal DESCRIPTION: The purpose of this Design Change was to remove Station Heating Boiler "B" to provide floor space for equipment storage and the installation of the new Turbine Equipment Cooling (TEC) filter demineralizer skid. Heating Boiler "B" is no longer required for heating and cooling purposes due to the installation of the electric boilers that were installed in 1982.

SAFETY ANALYSIS:

All systems, subsystems, and components removed by this modification were nonessential and did not perform any safety-related functions. As such, this modification did not increase the '

possibility of an accident occurrence, create the possibility of a previously evaluated accident occurrence, or decrease the margin of safety as defined in the Technical Specifications.

DC 90-036 Testable Gaskets (CS, RHR, Torus Drainline) and Pressure Taps on the Torus Drainline -

TITLE:

This Design Change authorized the replacement of gasketed, flanged piping joints, located in the discharge piping of various RHR and CS relief valves which discharge into the Suppression -

Pool, with new joints utilizing a testable, double o-ring scheme. Two additional joints in the Torus drainline were also converted to a testable scheme. These new joints will endie Type B Local Leakrate testing (LLRT) to be performed, thereby, verifying seal integrity.

SAFETY ANALYSIS:

This DC did not increase the probability of occurrence of an accident previously evaluated in the USAR because, the affected systems are unchanged in terms of performance, margin of safety, and physical configuration. No new failure modes or effects of systems or components were created. There were no changes to system operating characteristics, nor changes to any setpoint, and no new possible accidents or equipment malfu?ctions were created.

DC 90-085 TITLE:

REC Filter Demineralizer Skid Addition DESCRIPTION: The purpose of the Design Change (DC) was to provide improved monitoring and control of i

Reactor Equipment Cooling (REC) water chemistry. This was accomplished by adding a filter demineralizer skid to the REC system through the Non-Critical Service Header. Implementation of this DC was required to meet the comrritments made by the Distnct in response to Generic Letter 89-13.

SAFETY ANALYSIS:

This DC improved plant safety and reliability by providing improved chemistry control to the REC System Additional chemistry control reduces corrosion rates.to the REC piping, associated components, and heat exchangers. No functional changs were rnade to the affected systems, and all design criteria in accordance with applicable governing codes, standards, and practices were maintained. All margins of safety as defined by the basis of the Technical Specifications, USAR and plant procedures were maintained.

i DC 90-174A and Amendment 1 TITLE:

RHR Service Water Booster Pumps and Gland Water Modifications:

i DESCRIPTION: This Design Change replaced the existing Essential Gland Water injection System forthe RHR Service Water Booster Pumps (SWBPs) with a simplified essential system to improve system

)

reliability and reduce maintenance requirements. The new simplified essential system uses i

11

river water from the discharge of the RHR SWBPs and modified the pumps and mechanical seals to enable the direct use of river water without filtration equipment..The Amendment physically removed the old RHR Gland Water injection System including the gland water -

pumps, cyclone separators, transfer tanks, strainers, eductors, valves, instrumentation, and

-associated piping.

SAFETY ANALYSIS:

The replacement of the essential RHR SWBP Gland Water injection System will not reduce the performance or reliability of the RHR SWBPs nor did the modifications affect their ability to perform the safety function as described in the USAR.

Additionally, the pump / seat modifications were evaluated by Byron-Jackson /Durametallic (who also performed an analysis concerning the flush requirements of the mechanical seals), and a third party review was-performed by Stone & Webster to ensure no safety concerns were created. Modification to the pumps and piping did not reduce the performance of the pumps nor did it affect their ability to perform their safety function. This modification did not increase the probability of an accident occurrence, create the possibility of a previously evaluated accident occurrence, or decrease the margin of safety as defined in the basis for any Technical Specification.

DC 90-174B TITLE:

Service Water Pump Bearing and Shaft Modification DESCRIPTION: This Design Change replaced the bronze bearings, sleeved shafts in the four (4) Service Water (SW) Pumps SW-P-1A/B/C/D with Cutiess rubber bearings and pump shafts overlaid with hardfacing. These modifications allow the use of direct SW gland injection and thus the 1

removal of the old SW gland injection system.

SAFETY ANALYSIS:

This DC did not change the redundancy, decrease the reliability nor degrade pump performance. The replacement of the bearings, sleeved shafts was performed one pump at a time, and a 30 day test / trial period using direct SW gland injection to Service Water Pump B was performed (with acceptable results) to verify performance of the rubber bearings and hard-faced shafts. Implementation of this DC did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.

DC 90-174B Amendment 1 TITLE:

Service Water Pump Gland System Replacement and Miscellaneous Electrical Modifications DESCRIPTION: The purpose of this Amendment was to replace the essential Service Water Pump Gland Water Injection System with a simplified essential system. This Amendment also changed the time dial settings on the Service Water (SW) Pump Motor protective relays and installed fuses on the primary side of the 480/240-120 volt transformer located in the SW Pump Room.

SAFETY ANALYSIS:

The replacement of the essential SW Gland Water Injection System did not reduce the performance of the SW Pumps nor did the modifications affect their ability to perform the safety function as described in the USAR. The gland water piping modifications did not reduce the performance of the pumps nor did it affect their ability to perform their safety function. The change in the time dial settings for the relays provided for proper protection from excessive motor start time or motor stall. The changes in the time dial settings ensure that the SW system will continue to perform as analyzed. The installation of the fuses on the primary side of the transformer will protect the transformers from damage during a secondary short circuit-current. As such, this modification did not increase the probability of an accident occurrence, create the possibility of a new accident, or decrease the margin of safety as defined in the.

basis for any Technical Specifications DC 90-183 TITLE:

TEC Filter Demineralizer Skid Addition DESCRIPTION: The purpose of the Design Change (DC) was to provide improved monitoring and control of Turbine Equipment Cooling (TEC) water chemistry. This was accomplished by adding a filter demineralizer skid to the TEC system through the Non-Cntical Service Water Header. The control switch for the TEC surge tank level control was also modified to enhance operations of the TEC system.

12

(

SAFETY ANALYSIS:

This DC improved reliability by providing improved chemistry control to the TEC System.

Additional chemistry control reduces corrosion rates to the TEC piping, associated components, and heat exchangers. No functional changes were made to the affected systems, and all design criteria in accordance with applicable governing codes, standards, and practices were maintained. All margins of safety as defined by the basis of the Technical Specifications, USAR and plant procedures were maintained.

i DC 90-22G and Amendment 2 TITLE:

RMV-RM-1 and RMV-RM-4 Replacement DESCRIPTION: This Design Change and Amendment replaced the existing Ventilation Radiation Monitors for the Control Building Intake (RMV-RM-1) and the Drywell Vent (RMV-RM-4). The new monitors will perform tne same function as the old vent monitors. Additionally, this DC and Amendment installed 4 new, air operated containment isolation valves in the suction and return lines for the RMV-RM-4 monitor.

SAFETY ANALYSIS:

The isolation scheme chosen to implement this DC and Amendment conforms to Class B isolation of containment as described in the USAR, to ensure primary containment integrity is maintained. Two normally open, reverse acting isolation valves in series outside of containment were used for this containment isolation scheme, with the inboard isolation valve (nearest containment) and the outboard isolation receiving divisionally separate, Group Two isolation signals form the Primary Containment isolation System. This -DC did not change the performance or the capability of the two Radiation Monitors. Therefore, the margin of safety has not been decreased.

DC 90-278 TITLE:

Permanent Power to Turbine Deck Disconnect Switches DESCRIPTION: This Design Change provided a permanent source of power to the seven Turbine Deck Disconnect Switches. Only the Electrical System was affected by this modification.

SAFETY ANALYSIS:

The portion of the electrical system supplying power to the Turbine Deck Disconnect Switches is part of the 12.5 kV system which is non-essential and does not provide any safety function as described in the USAR.This DC did not require abnormal operation of any plant systems or procedures, and did not introduce any plant equipment alteration. Therefore, the effect on overall plant safety was not changed.

DC 90-320 TITLE:

Hydrogen / Oxygen Analyzer System Modifications DESCRIPTION: The Primary Containment Hydrogen / Oxygen Analyzer System was installed by Design Change (DC)87-036. This DC implemented modifications and enhancements to the Reg. Guide 1.97 H O primary containment analyzers to better assure system operation during normal and 2 3 design basis accident events.

SAFETY ANALYSIS:

Equipment and nuclear safety were not compromised by this DC. Implementation of this DC did not affect the functional capabilities of the Division i or ll Primary Containment monitors.

The system operation during design basis events remained as specified in the USAR. The Containment Atmosphere Monitoring System was tested per IEEE 323-1974, meets all Regulatory Guide 1.100 requirements and will function properly during all postulated design basis events in which it must function. The primary containment oxygen concentration continues to be measured and recorded twice a week and is maintained below 4% during normal operation, as per Technical Specifications.

DC 90-326 TITLE:

TEC/ REC to Plant Air Compressors A & B DESCRIPTION: The purpose of this Design Change was to decrease the possibility of inadvertent cross connection of the Turbine Equipment Cooling (TEC) and Reactor Equipment Cooling (REC) systems at Plant Air Compressors A, B, and C. This DC also allowed the plant air compressors to be restarted sooner following a high temperature trip by reducing cool down time. The 13.

~.

9 coolant piping at Compressors A and B was rerouted similar to the piping at Compressor C, which increases efficiency and reduces the discharge temperatures and decreases the number of high temperature trips.

SAFETY ANALYSIS:

This modification decreased compressor air discharge temperatures, improved compressor efficiencies, and reduced cool down time after high temperature trips. Also, the four isolation valves for each compressor will be controlled by one switch. accessible at the control panel.

The need to operate four manual valves to achieve a single objective is now reduced to one operation. This DC also relocated the compressor cooling water and oilindicators to a more accessible rack. No functional changes were made to any of the systems affected by this modification, and all other design criteria were maintained in accordance with applicable codes and standards. All previous accident analyses as documented in the USAR remain bounding, and no unreviewed safety question was created.

DC 90-331 TITLE:

Rod Sequence Control System (RSCS) Removal DESCRIPTION: This Design Change removed the existing Rod Sequence Control System (RSCS) and the Group Notch subsystem of RSCS, and all associated relays, pressure transmitters, power supplies, and annunciators. This DC also lowered the Rod Worth Minimizer (RWM) Techr,ical Specification low power setpoint, from 20% to 10% of rated thermal power. This is the setpoint below which the RWM is required to operable.

SAFEW ANALYSIS:

The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR were not significantly affected by this Design Change. RSCS and/or RWM were not designed to prevent a Rod Drop Accident (RDA), but instead impose restrictions on rod movement to mitigate the consequences of a postulated RDA. The probability of a RDA is a function of the Control Rod Drive System (CRDS), and is not dependent on the RSCS or RWM. New analyses have shown that when above 10% reactor power, a RDA which exceeds the 280 cal /gm limit will not occur. Therefore removal of the RSCS and reducing the RWM low power setpoint from 220% to 210% does not involve a significant increase in the probability or consequence of an accident previously evaluated. This design was previously reviewed and found acceptable by the NRC in Cooper Nuclear Station License Amendment 156, dated December 22,1992.

DC 90-352 and Amendments TITLE:

MOV Actuator, Thermal Overload Protection, and RWCU Valve Modifications DESCRIPTION: The purpose of this Design Change was to replace Reactor Water Cleanup valve RWCU-MOV-MO18, including LLRT connection valves; replace RWCU-MOV-MO15 valve stem and stem nut; replace torque switches on fourteen Motor Operated Valve Actuators: install spring pack modifications on seventy valves for hydraulic lock concerns; replace RHR-MOV-MO18 actuator motor and associated overload heaters; and replace RHR-MOT-MO57 overload heaters.

SAFEW ANALYSIS:

Implementation of this DC and Amendments was performed when those portions of the affected systems were not required to perform their safety functions, as governed by the Limiting Conditions of Operation as defined in the plant Technical Specifications. The testing and verification sections of this DC ensured that the affected components would adequately perform their safety and design functions. All margins of safety as defined in any basis for Technical Specifications, USAR, and plant procedures were maintained. Plant equipment was operated in accordance with applicable CNS approved procedures, and Technical Specifications. This i

DC and Amendments did not degrade Cooper Nuclear Station with respect to personnel, l

equipment, or nuclear safety, nor did this DC and Amendments present an unreviewed safety j

question as defined in 10CFR50.59.

DC 90-367 and Amendment 1 TITLE:

RMP-RM-351 Sample Line Upgrade DESCRIPTION This Design Change replaced the Sampling Pump SW-P-SWS and sample piping associated H

. with Radiation Monitoring Process RMP-RM-351 in the Service Water System toimprove flow and reduce maintenance. The pump was replaced with a larger capacity pump, and the existing carbon steel piping was replaced with stainless steel tubing. The Amendment allowed for the replacement of the thermal flow switch with a magnetic flow switch.

14

SAFETY-ANA' LYSIS:

This DC modified non-essential components in ron safety-related portions of plant systems,.

and did not introduce any failure modes that could affect essential plant oparations or nuclear safety. The installation of the non-intrusive magnetic flow switch improved plant monitoring by eliminating the clogging problem associated with the thermal flow switch. This modification had i

no affect on the performance of this system. The Radiation Monitoring Process serves only a monitoring function as described in the USAR and failure will not affect any safety-related i

system, equipment, or components. This DC ar,d Amendment did not affect the probability of an accident occurrence, create a previously unidentified accident, or reduce the margin of safety as defined in the basis of any Technical Specification.

DC 91-001 Amendment 1 TITLE:

Torus to Drywell Vacuum Breaker Disc Position Indication Replacement DESCRIPTION: This Design Change Amendment authorized the reduction of the Torus to Drywell vacuum breakers setpoint lower limit from 0.18 psid to 0.1 psid. This was performed to ensure that the vacuum breakers would continue to perform their safety function without exceeding allowable leak rates.

SAFETY ANALYSIS:

This Design Change and Amendment were performed during a refueling outage when primary containment integrity was not required. This change did not negatively affect the pressure relieving function of the vacuum breakers and maintained the required leak-tightness of the valves. This was verified by the acceptance testing outlined in DC 91-001. Testing during the outage of the vacuum breakers verified that the valves maintained their leak-tightness. In addition, the manufacturer has stated that the opening setpoint of the valves should be approximately 0.1 psid by design. This Amendment did not affect the original plant design, it only ensured the leak-tightness of the vacuum breakers, thereby maintaining the operation and reliability of the original design. The potential for bypass leakage from the drywell to the torus was not increased during a LOCA. The pressure relieving capacity of the vacuum breakers was not decreased by the lower setpoint. The upper setpoint limit of the vacuum breakers t

remained in conformance with the Technical Specifications limit of 0.5 psid. Reduction of the lower setpoint limit to 0.1 psid was in conformance with the vendor's recommendation. The setpoint change did not degrade CNS with regard to personnel, equipment, or nuclear safety.

DC 91-031 TITLE:

Smoke / Fire Dampers for Control Room Ventilation DESCRIPTION: This Design Change installed four fire / smoke dampers in the Control Room HVAC System ductwork to prevent the flow of smoke and hot gases into the Control Room in the event of a fire in the Cable Spreading Room. Two fire / smoke dampers were installed side-by-side in the Control Room HVAC internal supply ductwork and, similarly, two dampers side-by-side in the control Room HVAC return ductwork.

SAFETY ANALYSIS:

The fire / smoke dampers were connected to existing smoke detector FP-SD-1001, which is -

located in the Control Room HVAC System return duct in the Cable Spreading Room < This smoke detector shuts down the Control Room HVAC supply fans upon smoke detection in the Cable Spreading Room. The dampers installed were. seismically qualified, and when closed i

will provide a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating with Class 11 leakage in accordance with U.L Standard 555S.

The dampers are activated by one of three methods, a signal from a smoke detector,. heat, or manually. Installation of the dampers will prevent the flow of smoke and hot gases into the Control Room. This DC did not reduce the effectiveness of the Control Room HVAC system-or interfere with the safety function of the Emergency bypass system. Implementation of this design change did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.

DC 91-041 TITLE:

Torus Hard Pipe Vent DESCRIPTION: The purpose of this Design Change was to install a Torus Hard Pipe Vent (THPV) consistent with Generic Letter (GL) 89-16, and the BWR Owners Group hardened vent general design criteria. The THPV will be used to relieve pressure in primary containment during a loss of long 15

~

- term decay heat removal capability (TW Sequence). The DC installed a new 10" line connecting the Torus Purge / Vent supply line to the two Standby Gas Treatment discharge lines which provide the flow path required by GL 89-16 from primary containment to the elevated release point.

SAFETY ANALYSIS:

This modification did not affect the capability, performance, or the reliability of the Primary Containment System. The THPV tied into the Suppression Chamber purge inlet ductwork outside of the outboard containment isolation valve. A containment. isolation signal override was added to the containment isolation system. This did not adversely affect containment -

j isolation since two actions are required of the operator to override the containment isolation signal including two keylock switches to override the isolation signal, and operation of valve control switches. The modification also did not affect the performance and reliability of the Standby Gas System. The THPV tied into the SGT discharge piping in the discharge cross-tie.

i Isolation of the SGT discharge is maintained using rupture disc which is a passive component.

')

Radiation monitoring was provided to alert control room operators of radioactive releases during venting. The THPV system performs no function during normal plant operation, it is designed to provide pressure relief of the primary containment to reduce vulnerability to severe accident challenges during the loss of long term decay heat removal capability. The margin of safety was not reduced nor was the possibility of an accident or malfunction created or increased by implementation of this Design Change.

I DC 91-068 TITLE:

Radiological Support System DESCRIPTION: This Design Change installed the equipment associated with the Radiological Support System (RSS), which included the installation of fiber optic communication link between the RSS, the Whole Body Counting System, IBM Mainframe, and the Administration, Engineering, and HP/ Chem networks.

The RRS will support the Health Physics Department in the implementation of the changes resulting from the revision of 10CFR20.

SAFETY ANALYSIS:

The installation of the Radiological Support System does not interface with any equipment important to safety, nor does any of this equipment perform a safety-related function. No safety design basis or functional requirements of any systems were affected. Therefore, this modification did not change the existing safety analysis for Cooper Nuclear Station, nor_-

increase the prcbability or consequences of an accident as analyzed in the CNS USAR.

DC 91-088 TITLE:

lWn Steam Line Radiation Monitor Scram and Group 1 Function Removal DESCRIPTION: This Design Change (DC) removed the automatic scram signal and the Group 1 isolation signal i

from the Main Steam Line Radiation Monitors (MSLRMs). With implementation of the this DC the MSLRMs will initiate a Group 7 (Isolation of recirculation loop sample lines), close mechanical vacuum pump inlet and outlet valves, and prevent or stop the condenser mechanical vacuum pumps from running. The MSLRMs will initiate these actions at 3 times the normal radiation level to ensure the isolation of the direct release path to the ERP tower.

Additionally, the Steam Jet Air Ejector radiation monitor's alarm setpoint was reduced to 1.5 times normal radiation levels.

SAFETY ANALYSIS:

The Main Steam Line Radiation Monitors (MSLRMs) provide an early indication of gross fuel failure. However, inadvertent scrams and group isolations attributed to MSLRMs for Boiling Water Reactors have imposed transients on the reactor vessel and have actuated safety systems unnecessarily. The NRC approved GE report NEDO-31400 " Safety Evaluation for i

Eliminating the Boiling Water Reactor Main Steam Line isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor", which concluded that the automatic scram and closure of the MSIVs may be removed from the MSLRMs. This report indicates an insignificant increase in the reactivity control failure frequency as a result of removing the MSLRM scram function _which is offset by a reduction in transient initiating events caused by spurious scrams from the MSLRMs resulting in an approximate 0.3% reduction in core damage i

frequency. This represents an overall improvement in safety. This design was previously' reviewed and found acceptable by the NRC in Cooper Nuclear Station License Amendment 158 dated March 2,1993.

4 l

16

DC 91-121 A '

TITLE:

69 kV Capacitor Bank Installation DESCRIPTION: This Design Change installed a 69 kV,5.4MVAR capacitor bank on the 69 kV line in support of the CNS Emergency Transformer. This capacitor bank will increase the availability of the 69 kV line by providing a means to boost the voltage on the Emergency Transformer should the 69 kV line voltage experience load induced voltage reductions.

SAFETY ANALYSIS:

The capacitor bank was fused on the 69 kV line and each capacitor cell was individually fused.

The capacitor bank has no defined safety function and is physically located away from all equipment important to safety. The capacitor bank is not connected with other sources of offsite power to the station. Additionally, partial or complete loss of the capacitor bank will not impact the operation or availability of the 69 kV line. This design change did not degrade any safety related equipment and/or components and therefore, the ability of these systems to perform their safety function remains unchanged.

DC 91-1218 and Amendment 1 TITLE:

Emergency Transformer Replacement DESCRIPTION: This Design Change was to replace the Emergency Transformer with a new Load Tap Changer Transformer however, during testing of this new transformer a short was discovered in one of its windings and it could not be utilized. Therefore, the Amendment reinstalled the original emergency transformer to take the place of the other transformer. This Amendment also raised the setpoint of the emergency transformer auto closure permissive relays.

SAFETY ANALYSIS:

The setpoint for the emergency transformer auto closure permissive relays was required to be l

revised to meet the line conditions for the original emergency transformer. The revised setpoint for these voltage permissive relays will still ensure adequate voltage is available for worst case LOCA loading on the emergency transformer, The capabilities and performance of the 69 kV offsite power supply and the emergency transformer remain as specified in the USAR. Use of the original transformer in conjunction with the new setpoint of the auto closure permissive relays will not degrade its performance and capability to carry loads on the emergency AC buses as described in the USAR. No possibility of an accident or malfunction of a different type than previously evaluated in the USAR or Technical Specifications was created as a result of i

this change.

1 ESC 92-029 Amendment 1 TITLE:

IRM/SRM Drytube and Detector Replacement I

DESCRIPTION: This Equipment Specification Change (ESC) authorized the replacement of 12 IRM/SRM I

drytubes, shuttle tubes detectors and other associated equipment.

SAFETY ANALYSIS:

The replacement drytubes, shuttle tubes, detectors and other associated equipment were identical in form, fit, and function to the existing equipmant Additionally, all equipment replaced j

with this ESC maintained equal or better material standards in regards to the reactor pressure -

D boundary. Furthermore, if any drytube were to fail, the resultant leak would not be more severe-than that analyzed for design basis LOCA event. All accidents analyses remain bounding and no safety concerns were identified with implementation of this ESC.

DC 92-031 TITLE:

Reload 15 / Cycle 16 Operation DESCRIPTION: The purpose of this Design Change was to address the Cycle 15 core reload design and safety analysis. This DC documented the acceptability of the reconfiguration of the core that resulted from the replacement of depleted fuel assemblies with new fuel assemblies. This change also reviewed the results of the safety analysis performed by General Electric for Cooper Nuclear Station Cycle 16 core reload design.

SAFETY ANALYSIS:

The Cycle 16 reload design was reviewed and analyzed using methodologies described in NEDE-24011-P-A (latest approved version). The analysis for the specified abnormal operational transients and design basis accidents of Section XIV of the CNS USAR remains bounding for the Cycle 16 reload. Therefore, by operating the plant in accordance with the' nuclear safety 17

~

l 1

operational requirements as specified in Technical Specifications, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety was not increased, nor was the margin of safety as defined in the basis of any Technical Specifications reduced. No physical changes to plant safety systems were implemented by this Design Change and all safety systems will continue to be operated in their normal (as designed) configuration.

DC 92-097 and Amendment 1 TITLE:

Thermo-Lag Fire Barrier Replacement DESCRIPTION. The purpose of this Design Change and Amendment was to replace all of the Thermo-Lag 330-1 Fire Barrier Material, manufactured by Thermal Sciences Inc. at CNS. The Design Change l

1 replaced the Thermo-Lag in the Control Building Basement with a concrete one hour barrier.

The Amendment replaced the Thermo-Lag on two radiant energy shields located in the Cable Spreading Room, and one Plume impingement shield located in the Cable Expansion Room with Promat-H Board, manufactured by Eternit Inc.

ANALYSIS:

This Design Change and Amendment did not affect the operation of any system. The I

replacement of the Thermo-Lag material with concrete and Promat ensures CNS fire protection capabihties as discussed in the fire hazards analysis. There is no performance or operational modes associated with this material other than being a fire barrior.. This design change and amendment resulted in no physical modifications to components, systems or equipment. No possibihty of an accident or malfunction of a different type than previously evaluated in the USAR or Technical Specifications was created as a result of this change.

DC 92-145 TITLE:

Control Room HV Emergency Bypass and inlet Valve DESCRIPTION: This Design Change documented the work performed by SORC approved MWR 92-1578, which removed pneumatic relays, and poppet valves and replaced them with an essential 4-way solenoid valve. The solenoid valve performs the function of controlling the Control Room Emergency Bypass System inlet valve previously performed by the removed components.

SAFETY ANALYSIS:

This modification simphfied the system, but still maintained the original design requirements.

The functional characteristics of the replaced components is duplicated by the new components, The performance of the new 4-way solenoid does not impact any safety system

'l nor impacts any system that is relied upon to perform a safety-related function. This DC did not introduce any failure modes that could affect plant operations - or nuclear safety, i

implementation of this design change did not increase the probabikty of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.

ESC 92-163 i

TITLE:

Control Rod Flange Housing Bolt Material Change DESCRIPTION: This Equipment Specification Change (ESC) documented the replacement of the control rod drive unit flange housing bolts with improved material. The installed bolts were identified in GE SIL 483 as being susceptible to IGSCC.

SAFETY ANALYSIS; The new bolting material is similar to the original material but, improved because it is more resistant to IGSCC. The design requirements of the CRD housing bolts is that they must have a tensile strength of 125,000 psi. The new CRD bolting material has a minimum tensile strength of 165,000 psi and therefore, meet this requirement. Form, fit, and function of the bolts have not changed and operability was not degraded. The safety design basis and functional requirements of the closure bolting material were not affected, since the critical charactenstics of the replacement material meet the requirements of the original closure bolting material.

DC 93-034 1

TITLE:

HV and RW AOV Air Supply Upgrade DESCRIPTION. This Design Change (DC) documents SORC approved Maintenance Work Requests (MWRs) 93-2580 and 93-2896, which resolved closure problems associated with Air Operated Valves 18

-(AOVs) located in the Heating & Ventilation (HV), and Radwaste (RW) systems.

The modifications implemented by this DC consisted of the replacement of the pilot operated solenoid valves with direct acting solenoid valves and the installation of air filters at the solenoid valve cylinder and supply ports SAFETY ANALYSIS:

The implementation of this Design Change was performed while the plant was in a cold shutdown condition. This DC did not change the original design basis of the air operated valves or affect the safety function of the affected systems. This Design Change did not alter the capabilities of the HV or RW air operated valve, nor did it change any functions of the affected components during operation. The margin of safety was not reduced nor was the possibility of an accident or malfunction created orincreased by the implementation of this DC.

DC 93-057 dnd Amendment 1 TITLE:

Service Water (SW) and Reactor Equipment Cooling (REC) System Modifications DESCRIPTION: This Design Change and Amendment established divisional alignment within and between the Service Water (SW) and the Reactor Equipment Cooling (REC) Systems and provided for complete electrical Division I and Division ll isolation of the REC non-critical loads to assure that required SW and REC cooling are available to the critical loads designated to respond to a Design Basis Accident DBA LOCA/ LOOP with loss of one emergency diesel generator.

SAFETY ANALYSIS:

The safety function of SW and REC Systems remain unaltered. Divisional separation and divisional alignment of the REC and SW System assure adequate cooling to critical station loads without operator entry into the Reactor Building to perform manual operations. Even though the REC System is not mechanically redundant, adequate cooling to critical station loads will be available via the SW-REC inter-tie. No new accident type or equipment malfunction was introduced by the changes implemented by the Design Change. A single motor operator malfunction or operator error which would result in closure of a motor operated valve, will not, by itself, render the SW or REC incapable of performing their safety function.

The changes made in this Design Change and Amendment assure adequate SW and REC J

cooling flow to critical loads will be provided in response to postulated Design Basis Accidents.

DC 93-058 TITLE:

125/250 Volt DC Battery Charger Modification DESCRIPTION This Design Change authorized the permanent disabling of the load sharing circuitry of the 125 volt DC Battery Chargers 1A,1B, and 1C, and the 250 volt DC Battery Chargers 1A,18, and 1C. This change also authorized removal of temperature switches from the battery chargers.

SAFETY ANALYSIS:

This Design Change did not create the possibility of an accident or malfunction other than any evaluated in the USAR. This modification disabled the load sharing circuit and removed temperature switches from the 125/250 volt battery chargers, neither of which are necessary -

for the battery chargers to perform their safety function. The_ temperature switches are no longer required since the essential HVAC for the Control Building will maintain the temperature in the room below the temperature which could potentially damage the chargers. The load.

sharing circuit was designed to balance load between chargers which are supplying the same'

)

bus. This is not the configuration at CNS aad therefore it was disabled. The 125/250 volt j

battery chargers are still capable of performing their safety function as described in the USAR.

DC 93-062 and Amendment 1 TITLE:

Hatch Plug Removal, Northwest and Southwest Quads DESCRIPTION: This Design Change (DC) and Amendment removed the existing hatch plugs in the Northwest and Southwest Quads in the Reactor Building. The openings created by the removal of the j

hatch plugs from the 881'-9" and 9031-6" elevations were covered by steel floor grating. Dams

~

were installed around the perimeter of the openings on elevation 903'-6". These dam's will prevent water from internal flooding from cascading down the openings. These changes will allow natural convection cooling for the Reactor Building Northwest and Southwest Quads containing the RHR pumps to ensure their operation during certain design basis events.

SAFETY ANALYSIS:

This modification has no negative effect on the performance or reliability of quads in their safety function of providing separation for the redundant trains'of the RHR System. The removal of the hatches does not affect Appendix R compliance, since no credit is taken for the quad 19

= ~

interior walls or ceiling as fire barriers and there are already unsealed penetrations in the

+

ceilings where the hatches are to be removed. Further, this modification does not affect the basis for the Appendix R exemptions granted for the quads or the 903'-6" elevation of the Reactor Building. Further, this modification has no negative impact on the performance or reliability of the quads under post accident environmental conditions. Internal flooding is also not a concern because a flood dam of the appropriate height > 'as installed around the open hatch on the 903'-6" elevation of the Reactor Building.

DC 93-064 TITLE:

Standby Gas Treatment (SGT) System Modifications DESCRIPTION: The purpose of this Design Change (DC)was to document SORC approved Maintenance Work Requests (MWRs) 93-1172 and 93-1814, and modifications needed to bring the Standby Gas Treatment (SGT) System into compliance with all design requirements and ensure its ability to perform in all operational modes. This consisted of disconnecting the instrument air supply to the SGT room supply damper PC-AD-(AD-R-1C) as well as documenting the removal of the valve internals of check valve SGT-CV-10CV. These modifications permanently positioned the damper open and made the damper a passive component, thereby eliminating a possible single failure vulnerability. MWR93-1172, replaced the essential solenoid operated valves (SOV) with higher pressure-rated essential SOVs.

SAFETY ANALYSIS:

By disarming the damper and verifying the removal of the check valve internals assurance was obtained that the flow path for both Standby Gas Treatment (SGT) trains would remain open, which is the fail safe position of the system. There was no need for the check valve or damper to isolate the reactor building exhaust plenum from SGT inlet piping. If high radiation in the reactor building should initiate a Group 6 isolation, the open pathway established by.this modification ensures that the SGT would function as required. The replacement of the SOVs with higher pressure-rated models ensured that the failure of the non-essential pressure regulator would not over pressurize a SOV in a manner that would prevent the SGT system from operating. This DC established the conditions necessary for either SGT train to maintain a negative pressure on the Reactor Building and process the effluents through either train in the event of an accident. This DC did not introduce any new failure modes for the SGT system, while it removed a single failure vulnerability. Also, this DC did not cause any negative safety effects on the operation _of the Reactor Building Ventilation System, Secondary Containment System, or SGT System.

DC 93-076 TITLE:

Cold Reference Leg Continuous Backfill Modification DESCRIPTION: The purpose of this Design change (DC) was to authorize modifications that wnu!d enable a continuous backfill of water from the Control Rod Drive Hydraulic System (CRDH) to the cold reference legs (condensing chambers) in the Nuclear Boiler Instrumentation System (NBI). This DC documents SORC approval of Maintenance Work Requests (MWRs) 92-2259 arid 93-3023 which implemented the addition of taps and isolation valves in the NBl and CRDH systems, and included the installation of tubing, flow control valves (manual needle), and flow instrumentation forthe affected systems. These modifications when implemented and placed into service would prevent the buildup of saturated water in the cold reference leg piping and thereby, eliminate the possibility of water level errors during vessel depressurization events caused by degassing.

SAFETY ANALYSIS:

Dual isolation valves were installed on the CRDH tie-in side and the NBl tie-in side, to assure pressure boundary isolation during installation of the modification.

This DC included.

acceptance testing of the CRD backfill modification at low and high reactor pressure testing to answer any nuclear safety issues involved with the continuous operation of the backfill system.

The error on existing equipment caused by the continuous backfill of the reference legs was analyzed in NEDC 93-136, and determined negligible when compared to the potential errors attributed to the degassing phenomena in the cold reference legs. -No functional changes were -

made to the affected systems, and all other design criteria in accordance with_ applicable governing codes, standards, and practices were maintained. All margins of safety as defined by the basis for any Technical Specification, USAR, and plant procedures were maintained.

Implementation of this DC completes the District actions given in response to NRC Bulletin 93-03 and Generic Letter (GL) 92-04.

20

REPORTABLE ACTIVITIES M)NR 93&1204 for EWR 93-021 TITLE:

Installation of Loop Seal in Rupture Seal Drain Line DESCRIPTION: This SORC approved Maintenance Work Request (MWR) for Engineering Work Request (EWR)93-021, modified the 10" rupture seal drain line in the Fuel Pool Cooling piping. A loop seal was installed in this drain line to provide a barrier between Secondary Containment and the Radwaste Building. The Pddition of the loop seal prevents the possibility of an open leak path between the Radwaste Building and the bellows seal. A Design Change will be written at a later date to more fully document this modification.

SAFETY ANALYSIS:

This Maintenance Work Request enhanced plant design by separating the Reactor Building from the Radwaste Building HVAC systems, therefore, ensuring design basis building pressure relationships are maintained during all operating modes. The modification enhanced the Secondary Containment system and operation by implementing a positive means of sealing the Fuel Pool cooling Reactor Building /Radwaste Building penetration. Installation of the loop seal eliminated a potential leak path of Seccndary Containment therefore, the performance and reliability of any plant system were not degraded.

MWR 93-2260 for EWR 93-075 TITLE:

Test Connections for RHR-MOV-MO57 and RHR-MOV-MO67 DESCRIPTION: This SORC approved Maintenance Work Request (MWR) for Engineering Work Request (EWR)93-075. authorized the installation of a 4" isolation valve and 3/4" test connections on the piping for motor operated valves RHR-MOV-MO57 and RHR-MOV-MO67 These valves are downstream of the RHR pumps which circulate reactor coolant during LOCA conditions, and-have direct contact with primary containment. The modifications performed by this MWR will allow for leak tests to be performed on MO57 and MO67 to quantify leakage to the Radwaste Surge Tank.

ANALYSIS:

Installation of the test connections and the isolation valves enhances testing capabilities to quantify leakage through RHR and RW systems interface and does not affect the safety capabilities of either system. Therefore, the probability of occurrence or the consequences of 1

an accident previously evaluated in the USAR was not increased. The test connections and j

isolation valves were installed to the same quality standards as the utilized during original construction. No accident or malfunction of a different type is postulated as a result of this modification. The test connections and isolation valves did not change the perforrnance or function of the system, therefore, the margin of safety was not reduced.

MWR 93-2712 for EWR 93-091 TITLE:

Capping and Isolation of Smaller Bore Service Water Lines DESCRIPTION: This SORC approved Maintenance Work Request (MWR) for Engineering Work Request (EWR)93-091, authorized the installation of four essential valves to the Service Water (SW) sample supply piping from the REC and RHR heat exchangers, added two essential isolation valves to the SW sample return piping, replaced portions of the carbon steel SW sample retum lines j

with stainless steel piping, and capped and removed a portion of the SW Booster pump i

minimum flow lines, which were previously capped and were dead legs.

j SAFETY ANALYSIS:

This MWR installed passive components whose safety functions are only to' maintain the SW system pressure boundary. The installed components are in compliance with the design requirements of the system, and meet or exceed the original quality standards specified for the SW system. The existing failure modes and effects of the SW system remain applicable. All margins of safety defined by the basis of any Technical Specifications, USAR and plant procedures were maintained. Therefore, this MWR did not create an unreviewed safety question or have an adverse effect on nuclear safety.

21

_. _ _. _ _~ _

PTM 92-007 TlTLE:

Jumpering of HV-AOV-271AV DESCRIPTION: This Plant Temporary Modification (PTM) installed a mechanical jumper to bypass all pneumatic controls associated with HV-AOV-271 AV, of the Control Room Emergency Bypass System such that it would be maintained in the open position at all times. By maintaining the AOV in the open position the Control Room Emergency Bypass System can remain operable.

SAFET(

ANALYSIS:

Bypassing the controls for HV-AOV-271AV does not affect the operation of any other component. The valve will be in its fail safe position and all other components will be capable of performing their safety functions. Additionally, the filtering function of the emergency bypass system was not changed nor were the design bypass flow requirements altered.

PTM 93-033 TITLE:

Replacement of REC-Fl-475B with Spool Piece DESCRIPTION: The Purpose of this Plant Temporary Modification (PTM) was to replace flow indicator REC-FI-475B with a spool piece. This replacement was required because the flow glass broke and no repairs were available. This PTM will either become a permanent modification with an appropriate design chenge or the flow glass will be repaired when an additional sight glass becomes available.

SAFETY e

ANALYSIS:

This PTM did not create the possibility of an accident or malfunction than any evaluated in the USAR. The spool piece installed had flanges and gaskets duplicating the flow indicator installation, and the spool piece has comparable flow characteristics as the flow indicator. The spool piece was manufactured from material consistent with the system piping material.

PTM 93 037 TITLE:

Replacement of REC-FI-467B with Spool Piece j

DESCRIPTION: The Purpose of this Plant Temporary Modification (PTM) was to replace flow indicator REC-F1-467B with a spool piece. This replacement was required because the flow glass broke and no l

repairs were available. This PTM will either become a permanent modification with an appropriate design change or the flow glass will be repaired when an additional sight glass becomes available.

SAFETY ANALYSIS:

This PTM did not create the possibility of an accident or malfunction other than any evaluated in the USAR. The spool piece installed had flanges and gaskets duplicating the flow indicator installation, and the spool piece has comparable flow characteristics as the flow indicator. The spool piece was manufactured from material consistent with the system piping material.

PTM 93-049 TITLE:

Main Condenser System, Condensate Chemistry DESCRIPTION: The purpose of the Plant Temporary Modification (PTM) was to provide a method for injecting oxygen into the feedwater system in order to increase dissolved oxygen concentration level.

This PTM installed a means to add oxygen to the condensate system at the condensate pump combined discharge header.

SAFETY

- ANALYSIS:

Oxygen concentrations in the condensate system do not affect any USAR accident analyses.

Over injection of oxygen would result in increased corrosion which is monitoted by chemistry (metals). This PTM did not require abnormal operation of any plant systems or procedures, and did not introduce any plant equipment alteration.

Procedure Chance Notice (PCN) 2.2.77 (Revision 34)

TITLE:

Turbine Generator DESCRIPTION. Added subsection 14 to Section V " Limitation" to the procedure to provide guidance in the adjustment of the turbine pressure setpoint during normal full power operation and end of cycle (EOC) coastdown. This pressure setpoint during EOC coastdown may be adjusted to maintain reactor dome pressure.

22

SAFETY ANALYSIS:

An evaluation (GE report G-HPO-2-052), was performed for plant operation with constant reactor dome pressure at a full power value during EOC coastdown by increasing the pressure control setpoint. This report concluded that transient and accident conditions for this new plant operation during EOC coastdown are bounded by the normal reload analyses and the bounding design basis analyses reported in the CNS USAR. Additionally, the turbine vendor ensured that the turbine inlet pressure was within range of pressure variation allowable for the steam turbine performance consideration.

Procedure Chance Notice (PCN) 10.23 (Revision 7)

TITLE:

New Fuel Inspection, Channeling and Control Blade Inspection DESCRIPTION: This Procedure Change Notice was issued to address an extremely remote possibikty for inadvertently establishing critical conditions in new fuel storage racks as described in GE Service information Letter (SIL) 152.

SAFETY ANALYSIS:

The PCN included additional administrative controls which were implemented to address the concerns of SIL No.152, regarding usage of new fuel storage racks. This PCN involved no technical or operational aspects that directly affect normal station operation. This procedure change did not require abnormal operation of any plant systems or procedures, and did not introduce any plant equipment alteration.

New Procedure 6 310 8.1 (Revision Oj, TITLE:

Secondary Containment Valve Operability Test DESCRIPTION: This new procedure was written to allow increased closure times of the Reactor Building isolation Dampers. This procedure will document the time required to close the dampers to ensure it is consistent with the specified times in the USAR.

SAFETY ANALYSIS:

The effect of an increase in the time required for isolation of the reactor building following an accident was investigated and verified by GE. The conclusion, by GE was that the potential increase in the radiological consequences of an accident as presented in the USAR was minimal (remained well within 10CFR requirements and the CNS licensing basis) and that the accident analyses found in the USAR remain bounding for the implementation of the new procedure.

Other Activities TITLE:

QA Program Audit Requirements DESCRIPTION: This activity documented that future QA Program Audits will be performed per the frequencies stipulated in the CNS Technical Specifications, and the QA Program Policy Document which references Reg. Guide 1.33 (revision 2). This will allow QA Program Audits to be performed at least once per 24 months with the exception of Training, EP, Security, Fitness for Duty, Surveillance testing, and Fire Protection which will be performed at least once per year, and Corrective Action which will be performed every six months.

SAFETY ANALYSIS:

The changing of the audit frequency is an administrative change. Changing the audit frequency has no operational impact on any safety systems, structures, components or any equipment that are described in the USAR. The changing of audit frequency meets or exceeds all currently approved Technical Specifications and Reg. Guide 1.33 requirements. This is an administrative change only.

Other Activities TITLE:

CNS Emergency Procedure Guidelines DESCRIPTION: This evaluation was conducted to ensure that Revision 2 of the CNS Emergency Procedure Guideline (EPG) would be consistent with Revision 4 of the Boiling Water Reactor Owners Group (BWROG) EPG and does not conflict with either the design basis or operation of CNS as described in the USAR and other licensing documents.

SAFETY ANALYSIS:

The NRC reviewed and approved the BWROG EPG, including steps to accomplish actions beyond the design basis of the plant, for use in developing plant specific EPGs. The NRC 23

review and approval of the BWROG EPG is the basis for the CNS EPG, and EOPs. Since the revised EPG does not modify the operation or design basis of the plant, the possibility of accic'ents or malfunctions of different types than those listed in the USAR is not increased.

Additionally, since the EOPs are intended for use during plant conditions beyond those analyzed or required to be included in the plant licensed design basis, the margin of safety is not reduced.

_Other Activities TITLE:

RHR Service Water Booster Pump Room Design Temperatures DESCRIPTION. The purpose of this evaluation was to document and specify a temperature limit of 131"F for

]

the RHR Service Water Booster Pump (SWBP) Room. The existing ventilation systems for the RHR SWBP Room are non-essential, and cannot be relied upon to maintain the temperatures currently specified in the USAR under worst case design conditions. During an accident or abnormal conditions, temperatures in the SWBP Room could increase to 131*F without this ventilation. This evaluation showed that temperatures up to 131*F in the SWBP room are acceptable.

SAFETY ANALYSIS:

The performance and reliability of the RHR Service Water Booster Pump System was not affected by this change. Evaluation of equipment operability limits were performed and showed that the equipment required to operate these systems can withstand the temperature limit of -

131*F. There were no physical changes to the plant and the temperature limit does not adversely affect any existing essential equipment, and all existing systems, subsystems, structures, and components remain capable of performing their intended safety function. The only change is the temperature limit for the SWBP Room listed in the USAR, all accident analyses remain the same, and no new accidents or malfunctions were created.

Other Activities TITLE:

Nuclear Power Group (NPG) Organizational Changes DESCRIPTION. This activity documented the NPG organizational changes that were modified by the creation of the Vice-President Nuclear and Senior Nuclear Division Manager of Safety Assessment position.

Additionally, the NPG organization was modified by shifting the reporting responsibilities of some individuals, and various position title changes.

SAFETY ANALYSIS:

Changes in the NPG organization structure do not affect the design or operation of any plant system, structure, or component described in the USAR accident analysis. The organization changes that took place affect only the title of positions and reporting responsibilities. All persons filling the new positions were qualified to perform the assigned tasks and respons!bilities. These changes are considered to be an administrative change to the NPG organization which does not affect the performance of the organization to effectively respond

+

to plant transients or emergencies.

For additional information reference CNS Proposed Technical Specification Change Nos.121 and 123 submitted to the NRC on August 23,1993, and September 28,1993, respectively.

Other Activities TITLE:

Setpoint Change Requests (SCRs)03-063, and 93-064 DESCRIPTION. Setpoint Change Requests (SCRs)93-063, and 93-064, changed the existing setpoints of relays RHR-REL-K70A and RHR-REL-K708 respectively. The setpoint log shows that these relays were set at 0.0 seconds however, the minimum setting of the relays is 0 2 seconds. The -

Technical Specification limit for the relays is s 0.5 seconds, the SCRs documented the setpoint for these time delay relays as s 0.25 seconds the setpoint was calculated in accordance with -

Engineering Procedure 3.26.3 "Instrumeni Setpoint and Channel Error Calculation Methodology".

SAFETY ANALYSIS:

This change ensured that the margin of safety as defined in the basis for any Technical Specifications is met under postulated LOCA conditions. The time delay setpoint is maintained such that the start of RHR pumps 1A and 1D occurs prior to the start of RHR pumps 1B and 4

4 1 C.

This ensures that proper diesel generator loading is not compromised. The setpoint change does not alter the function of the relays and maintains the setpoint within the limits of the Technical Specifications Therefore, all accident analyses as described in the USAR remain bounding.

l 24

r i

l i

1 f

i f

i Ill.

PERSONNEL AND MAN-REM EXPOSURE i

1 l

1 25 r

I PE%S0WEL AND MAN REM BY WCCK LND J01 TUNCTIG!

Number of Personnel

(> 100 mrem)

Total Man-Rem Station Utility Contractor &

Station Utility Contractor &

Work and Job Fmetion Enployees Enployeen others Encloyees Enployees Others Pf ACTOR Of1 Rail 0NS & SUPV.

Maintenance Personnel 8

0 6

0.278 0.000 0.553 Operating Personnel 43 0

0 14.285 0.000 0.000 Health Physics Personnel 21 0

18 3.403 0.000 4.146 Supervisory Personnel 4

0 2

0.598 0.000 0.148 Engineering Personnel 23 0

4 7.696 0.000 0.637 RGITINE MAINTlNANCE Maintenance Personnel 76 0

337 62.120 0.000 170.812 Operating Personnel 5

0 0

0.032 0.000 0.000 Health Physics Personnel 30 0

62 27.803 0.000 32.941 Supervisory Personnel 4

9 18 1.398 5.833 5.332 Engineering Fersonnel 7

21 11 0.190 10.490 4.360 SPECIAL MAINTEt:NR Maintenance Pc snnel 0

0 19 0.000 0.000 5.040 operating Personnel 1

0 0

0.002 0.000 0.000 Health Physics Personnel 6

0 0

1.602 0.000 0.000

$ merviaory Peraonnel 1

0 0

0.029 0.000 0.000 Engineering Personnel 0

0 0

0.000 0.000 0.000 WASTE PROCESSINJ Maintenance Personnel 2

0 1

0.017 0.000 0.003 Operating Personnel 4

0 0

2.258 0.000 0.000 Health Physics Personnel 4

0 2

1.512 0.000 0.056 Supervisory Persomel 0

0 0

0.000 0.000 0.000 Engineering Persomel 0

0 0

0.000 0.000 0.000 ROtXL1NG Maintenance Personnel 0

0 5

0.000 0.000 0.367 Operating Persomet 15 0

0 0.381 0.000 0.000 Health Physics Persomel 0

0 0

0.000 0.000 0.000 Supervisory Persomel 0

0 0

0.000 0.000 0.000 Engineering Personnel 2

0 0

0.081 0.000 0.000 IN_SFRVICE ]NSPECTION Maintenance Personnel 0

0 26 0.000 0.000 11.100 Operating Personnel 1

0 0

0.013 0.000 0.000 Health Physics Persomet 0

0 0

0.000 0.000 0.000 Supervisory Persomel 0

0 0

0.000 0.000 0.000 Engineering Personnel 4

0 0

0.694 0.000 0.000 TOTAL Maintenance Personnel 76 0

371 62.415 0.000 187.875 Operating Persomel 47 0

0 16.971 0.000 0.000-Health Physics Personnel 31 0

62 34.320 0.000 37.143 Supervisory Persomet 7

9 18 2.025 5.833 5.480 Engineering Personnel 23 21 14 8.661 10.490 4.997 CRAD TOTALS 184 30 465 124.392 16.323 235.495 1

26 l.I