ML20046B358
| ML20046B358 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 07/14/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20046B351 | List: |
| References | |
| NUDOCS 9308040111 | |
| Download: ML20046B358 (12) | |
Text
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9 UNITED STATES NUCLEAR REGULATORY COMMISSION gv,8 WASHINGTON D C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.161
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TO FACILITY OPERATING LICENSE NO. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET NO. 50-213 P
1.0 INTRODUCTION
By letter dated January 29, 1993, as supplemented April 30, 1993, the Connecticut Yankee Atomic Power Company (CYAPC0/the licensee)' submitted a request for changes to the Haddam Neck Plant Technical Specifications (TS).
The requested changes would modify the TS to reflect modifications made to modernize the steam generator feedwater control system.
2.0 EVALUATION CYAPC0 requested changes to the TS relating to the Feedwater (FW) control system design modification using a digital control system. The design modification replaces the existing obsolete Hagan Feedwater control system with a Foxboro SPEC 200 MICR0 system and replaces other instrument-loop components including transmitters, indicators, recorders, bi-stables for instrumentation loops measuring steam Generator (SG) feedwater flow, SG steam flow, SG pressure, SG narrow range level, and steam line break flow. The a
modification also adds two vital inverters, and revises the logic arrangement of the steam flow / feed flow mismatch, and SG overfill protection circuits.
In addition to the above design changes, the modification includes all required TS changes reflecting the above described hardware and logic circuit changes.
The proposed modification will increase system reliability, add redundancy and provide the on-line testing capability for the reactor trip instrument functions.
2.1 Existino System Confiouration The FW system is comprised of two motor driven FW pumps, FW heaters, and SG headers one for each SG.
Each header consists of an automatically controlled main FW regulating valve, a motor operated FW isolation valve, a manual isolation valve and a check valve. A bypass line is provided around each set of main FW regulating valves and contains a FW regulating bypass valve.
Each bypass valve has the capability of 10 percent of full power flow.
I 9308040111 930714 PDR ADOCK 05000213 P
b The Haddam Neck Plant has four SGs, and each is provided with the SG Water Level Control System (SGWLCS), which is a non-safety system. The purpose of SGWLCS is to maintain the SG water level at the desired setpoint during all phases of normal power operation, by modulating the main feedwater regulating valve.
Level in each SG is controlled by a separate and independent SGWLCS channel. The SGWLCS is a three-element control system, and uses input signals from steam flow, FW flow and SG level transmitters of each SG to develop a flow demand signal to position the main FW regulating valve to the desired position. The FW flow to the SG is controlled using steam flow as a demand signal for FW flow, and a trimming action is achieved by using the SG level signal which is used by the controller to establish the level setpoint. The steam header pressure transmitter signal is used to compensate the steam flow signal for density changes.
The existing design uses Hagan transmitters to monitor steam flow and FW flow.
At present, SG level is monitored using a narrow range SG level Hagan Powermag Differential Pressure Transducer.
The FW demand signal from the three-element controller is transmitted through an auto / manual (A/M) flow control station to a Fisher electro-pneumatic (E/P) converter, which is mounted at the FW control valve.
3.0 PROPOSED MODIFICATIONS 3.1 FW Control and SG Level Control Desian Modification The proposed design modification upgrades the FW control system by replacing the existing analog controllers with Foxboro SPEC 200 MICR0 digital controllers, and also replaces existing obsolete instrumentation system used for SG level measurement and FW flow regulation with new instrumentation.
In addition, this modification revises the logic configuration of the reactor trip circuits associated with steam flow / feed flow mismatch, SG low level and steam line break flow trips, changes allowable and setpoint values of the process variable for SG overfill protection, and changes the automatic closure logic of the FW regulating valve.
The proposed changes are described as follows:
(1) SG Feedwater Flow: The existing eight non-redundant flow transmitters would be replaced by eight redundant post accident monitoring (PAM) Category 1, Class 1E transmitters. Also, the existing loop circuitry including the indicator and recorder would be replaced.
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. t (2) SG Steam Flow: The existing four flow transmitters would be replaced by eight redundant Class IE transmitters. Also, the existing loop circuitry including the indicator and recorder would be replaced.
(3) SG Pressure: The existing four containment mounted SG pressure transmitters would be replaced by four PAM Category 1, Class 1E transmitters.
Also, the existing loop circuitry including the indicator and recorder would be replaced.
(4) SG Narrow Ranae level: The existing four transmitters would be replaced by twelve (three channels per SG) redundant PAM Category 1, Class IE transmitters. Also, the existing loop circuitry including the indicator and recorder would be replaced.
(5) Steam Line Break Flow: The existing four transmitters would be replaced by four PAM Category 1, Class 1E transmitters. Also, the existing loop circuitry including the indicator and recorder would be replaced.
(6) Reactor Trio Circuit Loaic:
a.
The existing reactor trip circuit for steam flow / feed flow mismatch would be replaced by a one out of two coincidence logic using solid-state equipment and a suitable field interface.
b.
The existing reactor trip circuit for SG narrow range low-level trip would be replaced by a two out of three coincidence logic using solid-state equipment and a suitable field interface.
c.
The existing reactor trip circuit for steam line break flow would be 3
replaced by solid state equipment and a suitable field interface.
d.
On-line test capability would be provided for the above circuits.
(7) SG Overfill Protection loaic:
a.
The existing logic using a one out of two SG high level signal from the wide-range level instrument would be replaced with a two out of three high level signals from the narrow-range level instrument
- loops, b.
The value for the Engineered Safety Features Actuation System (ESFAS) trip setpoint for FW isolation on SG high water level would be changed from 5 69% span of the wide-range level instrument to S 74% span of the narrow-range level instrument.
The allowable t
value for this trip setpoint would be changed from 5 72% span of the wide-range level instrument to s 80% span of the narrow-range level instrument.
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The modified circuit for SG overfill protection would provide automatic closing of the FW regulating valve, regardless of the mode of operation, as well as automatic closure of the FW isolation motor operated valve.
l (8) Vital Inverter: The existing Vital inverter A and B would be replaced with two Class IE 5 KVA inverters and backup transformers.
(9) FW Bypass Valves: The modified design would provide automatic control for i
the FW bypass valves to control SG level at lower power lev ls, and would replace control board indicators for bypass flow.
-(10) level Controller: The three-element level controller would be replaced by a Foxboro SPEC 200 MICR0 digital control system with related field l'
interfaces and isolation devices.
(11) Recorders:
The modification would add non-safety related recorders for recording steam / feed flow and SG narrow range level.
(12) Auto / Manual Stations: The modification would = replace the existing Auto / Manual (A/M) control stations for the main FW regulating valve and for the auxiliary FW regulating valve controls.
l 3.2 Technical Soecification (TS) Chanaes:
I The proposed modifications described in Section 4.1 above needed a revision to the existing TS Sections and Tables, for reactor trip system (RTS). instrumentation,
-i surveillance requirements and bases, and ESFAS instrumentation trip setpoints, surveillance requirements and bases. These proposed TS changes would revise TS Tables 3.3-1, 4.3-1, 3.3-2, 3.3-3, 4.3-2 and Bases 3/4.3.1 and 3/4.3.2.
l 4.0 EVALUATION j
The staff reviewed the licensee's documentation submittal relating to the proposed modifications, and additional information submitted by the licensee in response to staff's request for additional information to clarify several design details.
4.1 Diaital System Desion and Implementation j
The proposed design. utilizes a microprocessor based SPEC 200 MICR0 digital l
control system (DCS) manufactured by the Foxboro Company. The DCS receives i
inputs from instruments including transmitters monitoring; FW flow, steam flow, SG level, steam pressure, and other signals and generate various outputs including; flow control valve demand signal, alarms, indications, and t
modulates the FW regulating valve such that water level in SG is maintained at l
its preset level during all modes of normal plant operation.
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l i 4.1.1 The digital controllers are normally more accurate, faster and more flexible when compared to its conventional analog counterparts but the accuracy of the any digital system depends upon proper interfacing of peripheral hardware, accuracy of software design and implementation, adequate protection of hardware from noise & radio frequency interference (RFI).
In addition, the reliability of the digital system depends on adequate control on the microprocessor software, and adequate protection against possible failure i
modes including common cause and common mode failures.
The staff evaluated t
all these attributes for the proposed DCS.
A.
Verification and Validation (V&V) of software-Accuracy of the software and its proper implementation is assured through the V&V process.
Verification during design process is basically determining whether or not the product of each phase of system develoment fulfills all
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the requirements imposed by the previous phase, and validation is testing of l
the integrated software to determine that it will perform the designed functions correctly.
USNRC Regulatory Guide 1.152,1985 " criteria for pro-grammable digital computer system software in safety-related systems of nuclear power plants" which endorses the standard ANSI /IEEE-ANS-7.4.3.2,1982
" Application criteria for programmable digital computer systems in safety i
systems of nuclear power generating stations" establishes requirements and guidelines for the V&V process.
The acceptability of Foxboro SPEC 200 MICR0 equipment usage in safety systems of nuclear power generating stations including software V&V has been i
documented in Foxboro Test reports Q0AAE01, Q0AAE03, Q0AAE04 and Q0AAE05.
These reports have been reviewed and approved by the staff for tuo previous RPS modifications at the Haddam Neck Plant. The licensee confirmed that the control modules used for this modification are similar to those used in past for the RPS system modification at the Haddam Neck Plant. The licensee further stated that the configuration of the SPEC 200 MICR0 is controlled by an i
Instrumentation and Controls Department Instruction, ICDI-29, " SPEC 200 MICR0 Configuration Change Control," and two station Corrective ~ Maintenance Procedures, CMP 8.2-76.1, " Downloading of SPEC 200 MICR0 Configurator," and CMP 8.2-76.2, " Uploading of SPEC 200 MICR0 Configurator." These procedures apply to all configuration changes made to the SPEC 200 MICR0 equipment installed in all plant systems at the Haddam Neck Plant. - Since. the control algorithm software modules are similar to pre-approved modules the staff i
accepts that the V&V performed is acceptable for this modification and that the licensee has adequately followed their in-house procedures.
i B.
Protection against Electro-Magnetic interference (EMI), Radio frequency interference (RFI) and ambient temperature parameters specific to the location of installation:
Conducted and radiated EMI and RFI could introduce undesirable characteristics and responses in the digital control-system and could lead to malfunctions of the FW control system.
In addition, large variations in other environmental conditions could lead to hardware failures.
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The licensee stated that the equipment for this modification, which has been purchased QA Category I, Class IE, is identical to the Foxboro digital equipment used for RPS I and RPS 11 modifications, and is located in the control room which is considered a mild environment.
Equipment utilized for RPS I and RPS II modifications was qualified by the Foxboro Company for all attributes of the control room environment including RFI, EMI, highest ambient l
temperatures and humidity.
EMI/RFI testing was performed on location by the licensee subsequent to installation of the RPS I and RPS 11 modifications, and demonstrated that the environment of the installed equipment was enveloped by Foxboro's noise test.
In addition, successful operation of the digital equipment from the RPS modifications during the last 3 years provides a high i
degree of confidence in EMI/RFI immunity of the SPEC 200 MICR0 equipment.
In response to the staff's request for additional information, the licensee informed the staff that at present an independent contractor is evaluating test results of the qualification tests of the Foxboro Company to verify that i
the values of test attributes used during tests either meet or exceed the equipment location environment. The contractor would also evaluate and verify if the test results of previous on-location tests could be applied to the proposed modification.
Results of this evaluation will indicate if any additional tests are required. This evaluation and required corrective action (s) would be completed prior to declaring the modification operational.
If corrective actions are deemed necessary by this evaluation then they would be reviewed by the staff in the future. The staff finds this commitment acceptable.
C.
Software control:
Programmable software could be intentionally or unintentionally altered for various reasons. After the V&V has been performed, if software alterations are not controlled properly, unnoticed errors could be introduced which could cause the controlled system to malfunction.
At the Haddam Neck Plant, the software that has undergone Foxboro's V&V program are contained in a programmable read only memory (PROM). The configuration of the SPEC 200 MICR0 is controlled by the Instrumentation and Controls Department Instruction, ICDI-29, " SPEC 200 MICR0 Configuration Change Control," with two station Corrective Maintenance Procedures, CMP 8.2-76.1,
" Downloading of SPEC 200 MICR0 Configurator," and CMP 8.2-76.2, " Uploading of SPEC 200 MICR0 Configurator." These procedures apply to all configuration changes made to SPEC 200 MICR0 equipment installed in plant systems. After release cf the system to operations, any configuration changes required would be controlled administratively by the design change process. The staff finds t
this acceptable.
D.
Failure modes:
This modification has not introduced any new failure modes not evaluated in past. The modification changes the existing hard wired analog control for FW regulating valves with a digital programmable control and replaces obsolete instruments of related instrument loops, including transmitters, indicating meters and recorders with new equipment.
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3 The licensee has analyzed the design modification to verify that the modified system would perform all safety functions required for a design basis event in the presence of:
(a) any single detectable failure within the safety systems concurrent with all identifiable but non-detectable failures; (b) all failures caused by the single failure; (c) all failures and spurious system actions which cause or are caused by the design basis event requiring the safety function. The evaluation demonstrated that the design change would enhance the adherence to the single failure criterion by; (I) increasing channel and train independence and redundancy, (II) providing design and test capabilities to eliminate identified non-detectable failures, (III) reducing potential cascaded and design basis event failures, (IV) decreasing potential for certain common mode failures through the use of qualified Clas's lE components.
The failure of the digital controller and/or any of its peripheral equipment could lead to the FW regulating valve failing either as-is, or at a full open position or at a full closed position.
The licensee evaluated these failure modes and after staff review, we agree that there are no concerns and find the consequences of these failure modes acceptable.
The licensee stated that the SPEC 200 MICR0 cards are designed _ to fail safe.
L All analog outputs fail to zero and all logic outputs fail to the trip condition.
Failure of the output to zero would close the FW valve and would generate a trip. The operators would receive an alarm of this situation and would evaluate the status accordingly.
Similarly, a failure in the SG overfill protection function would be processed and annunciated as one channel of FW isolation.
Each SPEC 200 MICR0 control card includes an LED indicator of card failure. Additionally, monthly surveillance testing will be performed in accordance with plant TS.
4.1.2 Environment Qualification:
i The safety system equipment including field transmitters, main centrol board indicators and vital inverters, that are used in this design change have been procured as QA Category 1, Class IE and qualified in accordance with IEEE Std. 323-1974. All new cables were procured as QA Category 1 and satisfy the flame retardant requirements of IEEE 383. New recorders used for recording steam / feed flow and SG narrow range level will be procured as non-safety related items. These recorders are not designated to provide indication per requirements of Regulatory Guide (RG) 1.97.
New auto / manual stations for the main and auxiliary FW regulating valve control will also be procured as non-safety related items.
4.1.3 Design Modification Implementation:
The staff reviewed modification design documents to verify that:
A.
All supports are seismically qualified.
B.
All possibilities of installing the Category II equipment over the Category I equipment have been considered and corrected.
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, C. Alarms or other means to detect all possible failures have been provided.
D. Analyses to address all possible single failures and or common mode failures has been performed and protection against these failures have been incorporated.
E. All controlled copies of tha control room reference documents have been revised to show the modified configuration.
F F. Adequacy of the Class IE batteries to support the new inverter has been verified from the battery sizing calculations. That the Class IE DC distribution system calculations have been evaluated for changed values of short circuit currents, voltage drops, coordination, protective gear settings and for equipment rating considerations.
G. The heating, ventilation, air conditioning (HVAC) calculations for all areas have been revised where new equipment is to be installed and have been reviewed for additional heat loading.
H. The effect of this modification on the fire protection system was evaluated.
J. The affected non-Class IE AC distribution system calculations were reviewed for changed values of short circuit currents, voltage drop, coordination, protective gear settings and for equipment rating consid-erations.
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l K. The station operators have been trained for following aspects of the modification: operation, surveillance, testing, maintenance, emergency operating procedures (EOP), and any design imposed operating limitations imposed on existing systems / components.
j L. All new installations for instruments, cables, digital control cabinets and racks, field interfaces and peripherals, Class IE/non-Class IE isolation devices, instrument tubing, panel wiring have been reviewed to ensure that new installations do not in any way adversely affect the ability of other safety systems and components in their proximity.
M. Proper isolation between Class 1E and non Class IE circuits and between redundant Class IE circuits and components has been provided for signal and for power connections.
N. Equipment qualification for all equipment has been verified for the effects of high energy line breaks (HELBs), flooding, and saturated steam environment if applicable.
4.2 Technical Specifications (TS) Chanaes (1) Table 3.3 Functional Unit 9 (Steam Generator Water Level-low Coincident With Steam /Feedwater Flow Mismatch)
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-g-Proposed chanaes: The " TOTAL NO OF CHANNELS" is being changed from 1 SG level to 3 SG level and 1 steam /feedwater flow mismatch in each SG to 2 i
steam /feedwater flow mismatch in each SG.
The " CHANNELS TO TRIP" is being changed from 1 to 2/SG level coincident with 1 steam /feedwater flow mismatch in the same loop. The " MINIMUM CHANNELS OPERABLE" for SG level is being changed from 1 to 2/SG. ACTION 5 is being deleted. ACTION 6 is being invoked.
1 Evaluation: The above described changes support the installation of three narrow-range level, two steam flow and two FW flow transmitters per SG as i
opposed to the existing design of one narrow-range level, one steam flow, and one FW flow transmitter per SG, and are acceptable to the staff.
ACTION 5 is replaced by ACTION 6.
ACTION 6 requires the inoperable channel be placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and, allows the inoperable channel to be bypassed for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, if the number of OPERABLE channels is one less than the TOTAL NO 0F CHANNELS.
Extending the time to put the inoperable channel in tripped condition from I hour to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is acceptable to the staff.
This extension has been previously approved as described by the RTS SER for Westinghouse WCAP-10271 issued by the NRC on February 21, 1985.
The time for the inoperable channel to be in a bypassed condition has been extended from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this situation 2/2 channels are available for trip provided the channel being surveillance tested is put in a tripped condition (this guarantees 2/2 channels available for trip).
Therefore, changing the bypass time from 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is acceptable to the staff.
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- 12) Table 4.3 Functional Unit 8 (Steam Flow--High) and Functional Unit 9 (Steam Generator Water Level--Low Coincident with Steam /Feedwater Flow Mismatch).
Proposed chanaes: The frequency of ANALOG CHANNEL OPERATIONAL TEST for these Functional Units is being changed from every refueling outage to every 6 weeks.
1 Evaluation: The revised frequency is conservative and reflects an on-line testing capability of the modified design.
This is acceptable to the staff.
(3) Table 3.3 Functional Unit 6.a (Feedwater Isolation:
Wide Range Steam Generator Water Level--High).
Proposed chanaes:
Feedwater Isolation on SG water level high signal would be generated from the narrow-range SG level transmitters. The TOTAL NO. OF CHANNELS are being changed from 2/SG in each operating loop to 3/SG in each operating loop. The CHANNELS TO ACTUATE are being changed from 1/SG in each operating loop to 2/SG in each operating loop. Table notations (c) and (d) are no longer applicable to Functional Unit 6.a.
Table notation (c) is being deleted as not applicable to any functional Unit. All references to 6.a are being removed from Table notation (d). Change ACTION 26 to ACTION 24.
. Evaluation: The above described changes support the installation of three narrow-range level transmitters and use their signal for the feedwater isolation function.
In a condition with the number of OPERABLE channels one less than the required MINIMUM CHANNELS OPERABLE requirement, tha existing ACTION 26 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore the inoperable channel to OPERABLE status or reduce THERMAL POWER to below 10% within the following I hour.
In the revised TS, ACTION 26 would be replaced by ACTION 24.
In a condition with the number of OPERABLE channels one less than the required MINIMUM CHANNELS OPERABLE requirement, ACTION 24 requires that the inoperable channei be placed in the tripped condition within I hour and, if the requirement for the minimum operable channels is met, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of other channels per Specification i
4.3.1.1.
In the existing TS, TOTAL NO 0F CHANNELS are 2, CHANNELS TO ACTUATE are 1 and MINIMUM CHANNELS OPERABLE are 2.
Therefore, if I out of 2 MINIMUM CHANNELS OPERABLE becomes inoperable, the actuation logic is reduced to 1/1 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In the revised TS, TOTAL NO OF CHANNELS are 3, CHANNELS TO ACTUATE i
are i and MINIMUM CHANNELS OPERABLE are 2.
If I out cf 3 MINIMUM CHANNELS OPERABLE becomes inoperable, actuation logic for this Ft;NCTIONAL UNIT is reduced to 2/2 for I hour and after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, when the inoperable channel is placed in the " tripped" status, the actuation logic becomes 1/2.
In this situation if another channel is placed in a bypassed state, the actuation logic becomes 1/1 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Tt Js, time for the actuation logic to be in 1/1 state (i.e., without meeting the single failure criterion) is reduced from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Again, since the total number of channels are increased, i
flexibility is improved. Therefore, replacing ACTION 26 by ACTION 24 is 1
acceptable to the staff.
1 (4) Table 3,3 Functional Unit 6.a Feedwater Isolation (Steam Generator Water Level--High).
j Proposed chances:
The ESFAS instrumentation trip setpoint for FW isolation on l
SG high water level is being changed from 5 69 percent of wide-range 1
instrument span to 5 74 percent of the narrow-range instrument span. The l
allowable value is being changed from s 72 percent of wide-range instrument span to 5 80 percent of the narrow-range instrument span.
Evaluation: The new values are conservative with respect to overfill i
protection and will allow plant operations without spurious overfill protection.
This is acceptable to the staff.
(5) Table 4.3 Functional Unit 6.a Feedwater Isolation (Steam Generator Water Level--High).
Proposed chances: The ESFAS instrumentation surveillance requirements for FW isolation on SG high water level TRIP ACTUATING DEVICE OPERATIONAL TEST is being changed from N.A.(not applicable) to R (refueling) frequency.
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. Evaluation: The new circuitry allows for a complete logic test to be performed. Therefore, the TS change is acceptable to the staff.
(6) Bases for 3/4.3 (Instrumentation), 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION Proposed chances:
The following paragraph is added: "The Feedwater Isolation circuit is considered to be OPERABLE provided either the Feedwater Regulating Valve or the Feedwater Isolation MOV is available for automatic closure."
Evaluation:
The above addition supports the design modification. This is acceptable to the staff.
4.3 Mechanical Failure of Main Feedwater Reaulatina Valve As part of the Sys'ematic Evaluation Program (SEP) and Integrated Safety Assessment Program (ISAP) reviews it was determined that the excess feedwater event is initiated by a failure of the main feedwater regulating valve. This failure can be initiated by either an instrument or mechanical failure of the main feedwater regulating valves.
Instrument failure was resolved by the closure of USI A-47, " Safety Implications of Control Systems." The mechanical failure was proposed to be closed as part of SEP Topic IV-2, " Reactivity Control Systems, Including Functional Design and Protection Against Single Failure," and Integrated Safety Assessment Program Topic 1.54, " Safety Implications of Control Systems." The ISAP program review determined that since there is a need to promptly isolate the feedwater line given a high steam generator level condition and a mechanical failure of the main feedwater regulating valve, the feedwater isolation signal will also be sent to the feedwater isolation valve.
Even though this valve closes more slowly than the feedwater regulating valve, initiating the automatic closure of the isolation valve on a high steam generator level will improve safety as it will isolate main feedwater if the main feedwater regulating valve were to mechanically fail open.
Based on the licensee's proposed modification, the staff considers SEP Topic IV-2 and ISAP Topic 1.54 complete.
4.4 Correction of TS Paaes 3/4 3-15 In telephone discussions with the licensee on July 14, 1993, it became apparent that TS Table 3.3-2 function Unit 3.a.1 and 3.a.2 did not have the correct table notations. Amendment 149, which was issued on February 24, 1992, did not incorporate changes made to TS page 3/4 3-15 in Amendments 141 and 148 which were issued July 3, 1991, and January 17, 1992, respectively.
The corrections have been made and are included with the attached TS pages.
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3.0 STATE CONSULTATION
i In accordance with the Commission's regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The State i
official had no comments.
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4.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32, and 51.35, an env rn imental assessment and finding of no significant impact have been prepared and published in the Federal Reaister on June 4, 1993 (58 FR 31760). Accordingly, based upon the environmental assessment, the staff has determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.
5.0 CONCLUSION
Based on the above evaluation, the staff found the instrumentation and control design modification of the FW control system and the proposed technical specification changes to be acceptable.
In addition, the Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the l
Commission's regulations, and (3) the issuance of the amendment will not be I
inimical to the common defense and security or to the health and safety of the I
public.
Principal Contributor:
S.V. Athavale Date:
July 14, 1993
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