ML17223B229
| ML17223B229 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/16/1991 |
| From: | Sager D FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| L-91-198, NUDOCS 9107230353 | |
| Download: ML17223B229 (47) | |
Text
, ALCELERATED DIS UTION DEMONSTRA ON SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9107230353 DOC.DATE: 91/07/16 NOTARIZED: NO DOCKET g FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power & Light Co. 05000335 AUTH. NAME AUTHOR AFFILIATION SAGER,D.A. Florida Power & Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk) R
SUBJECT:
Forwards description & summary of safety evaluation of plant changes/mods made, reportable per 10CFR50.59.Rept includes TITLE:
NOTES OR plant changes/mods completed between 900123-910122.
DISTRIBUTION CODE: AOOID COPIES RECEIVED:LTR Submittal: General Distribution i ENCL l SIZE: 0 D RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID 'CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 D NORRIS,J 2 2 INTERNAL: ACRS 6 6 NRR/DET/ECMB 7D 1 1 D NRR/DET/ESGB 1 1 NRR/DOEA/OTS B1 1 1 1 NRR/DST 8E2 1 1 NRR/DST/SELB 7E 1 1 NRR/DST/SICB8H7 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 OC 1 0 OGC/HDS3 1 0 G~~% 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 D
D D
NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED: LTTR 24 ENCL 22
P.O. Box 128, . Pierce, FL 34954-0128 JUL $ 6 $ 99t L-91-198 10 CFR 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:
Re: St. Lucie Unit 1 Docket No. 50-335 Re ort of 10 CFR 50.59 Plant Chan es Pursuant to 10 CFR 50.59 (b)(2), the enclosed report contains a brief description and summary of the safety evaluation of Plant Changes/Modifications (PCMs) which were made, and are reportable, pursuant to 10 CFR 50.59. Included with the brief description of each PCM is a summary of the safety evaluation completed by Florida Power & Light Company for that PCM. This report includes PCMs completed between January 23, 1990, and January 22, 1991, and correlates with the information included in Revision 10 of the Updated Final Safety Analysis Report submitted under separate cover.
Should there be any questions on this information, please contact us ~
Very truly yours,
~if D. A. er Vice P sident St. Lu e Plant DAS/JJB/kw Enclosure cc: Stewart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie Plant II, USNRC DAS/PSL N471-91 9i07230353 9107l6 PDR ADOCK 05000335 p'0 p PDR c>rior APL Group company
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P C/M 089-186 Rev. 0-3 ABSTRACT The modifications associated with this change provide for the upgrade of the Main Steam Isolation Valves (MSIV), their control and test circuits. An upgraded pneumatic control scheme for the MSIV actuators and an additional air accumulator for each MSIV is also provided. The MSIV test circuits and Control Room wiring are modified to make the safety related control circuits independent from the test circuits for each MSIV. The implementation of this modification did not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval because the modifications provide for increased reliability via upgraded components and circuit design while maintaining the existing design requirements and basis. The systems functional requirements and capabilities remain the same as the original design.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
This EP does not involve an unreviewed safety question and the following are the bases for this justification:
i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. This modification will provide increased MSIV reliability and will increase the fatigue life of various internal valve components. This modification has been designed to the original requirements of USAS B31.7, 1969 Edition. The probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated is not increased since the new control system, is more reliable and meets all existing design basis. The consequences of an accident or malfunction have not increased since the control'system and MSIV will perform on demand to mitigate any accident as previously analyzed.
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P C/M 089-186 Rev. 0-3 SAFETY EVALUATION (Continued) ii) As a result an accident of this modification, there is no possibility for or malfunction of a different type than any previously evaluated because the modification does not modify the intended operation or test requirements of the MSIV. The new control system will perform the same function as the existing system. Therefore, the functional capability remains the same.
iii) This modification does not reduce the defined in the basis for any Technical margin of safety as Specification. MSIV operability as defined by Technical Specification 3.7.1.5 requires the MSIV to close within 6 seconds in modes 1, 2, 3.
This modification meets this requirement. Therefore, the basis for the Technical Specification remains the same. There is no change in the functional requirements of the MSIV.
Technical Specification Surveillance 4.7.1.5 shall be performed in accordance with procedures to verify the functional requirements.
The implementation of this PCM does not require a change to the plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or a change to the Technical Specifications, and prior commission approval for the implementation of this PCM is not required.
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P C/M 088-187 Rev. 2 ABSTRACT The modifications associated with this change provide for additional instrumentation and indication changes for the remote reactor vessel indication system. Changes provided by this P C/M revision do not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval as they provide improved display and alarm functions for monitoring reactor vessel water level during refueling operations.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident. or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
This EP does not involve an unreviewed safety question, and the following are bases for this justification:
i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased since this modification provides a means whereby an accurate Reactor Vessel water level can be readily determined during refueling. During power operation, this system is isolated from the RCS. The portions of this modification within the normal RCS pressure boundary have been designed to the original requirements of the RCS pressure boundary.
As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated because the modification provides double isolation valving which will isolate the system from the RCS during power operation.
iii) This modification does not reduce the margin of safety as defined in the basis for any Technical Specification because it neither changes the design parameter of the RCS nor does change the RCS design flow or functional requirements.
it The implementation of this PCM does not require a change to the plant Technical Specification.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or a change to the Technical Specifications, and prior commission approval for the implementation of this PCM is not required.
P C/M 040-188 Rev. 0-1 ABSTRACT The modifications associated with this change provided for an upgrade to the Westinghouse generator exciter system to prevent potential unit trips and correct problems encountered during exciter startup and operation. Changes provided by this modification do not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since the generator excitation system performs no safety related function and is not part of the basis for any Technical Specification. After implementation the generator excitation system is better able to respond to changes in the electrical system grid.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) consequences if the probability of occurrence or the of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications included in this Engineering Package do not involve an unreviewed safety question because:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased by this modification because it does not affect or change the availability, redundancy, capacity, or function of any equipment required to prevent or to mitigate the effects of an accident.
There is no possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report since no new failure modes are introduced, as stated in Section 2.1.7 of this EP.
(iii) The margin of safety as defined in the bases for any Technical Specification is not reduced since this modification does not degrade the Generator Excitation System and the Generator Excitation System does not form the bases of any Technical Specification.
The implementation of this EP does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior NRC approval for the implementation of this PCM is not required.
P C/M 062-188 Rev. 0 ABSTRACT The modifications associated with this change provide for the introduction of nitrogen gas into the main condenser hotwells. The purpose is to increase the ability to remove non-condensable gases and reduce the dissolved oxygen level in the condensate system.
Provision has also been made to inject hydrazine during shutdowns via valved connections. Changes provided by this modification do
.not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since the systems involved perform no safety related function and do not form the basis for any Technical Specification. The modification serves to improve condenser efficiency and minimize corrosion in the feedwater system.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increasedg or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. Neither the nitrogen gas system nor the condenser air evacuation system are used in any safety analysis for accidents or malfunction of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety.
ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. The components involved in this modification have no safety related function and no changes have been made to the operational design of the systems impacted by this EP.
iii) The margin of safety as Specification is not defined in the bases for any Technical affected 'by this PCM, since the components involved in this modification are not included in the bases of any Technical Specification.
The implementation of this PCM does not require a change to the Plant Technical Specification.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.
I P C/M 047-189 Rev. 0 ABSTRACT The modifications associated with this change provide for the deletion or modification of particular small thermal movement snubbers. Stress Analysis calculations were performed to document that the modifications and deletions could be performed and that the resultant pipe stresses are within the code allowables. The analysis shows that the resultant pipe configurations meet the original design requirements. Changes provided by this modification do not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since the resultant configuration was shown by analysis to not be a change to the original design requirements. Technical Specifications do not require any change because the original design and code requirements continue to be met.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) consequences if the probability of occurrence or the of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
This EP does not involve an unreviewed safety question, and the following are the bases for this conclusion:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased because 1~ there is no modification performed to the piping and all supports/restraints, except those for which snubbers have been deleted, have remained functionally identical to their existing design configurations.
- 3. piping stresses for the modified condition have remained within the stress limits allowed in USAS B31.7 Code 1969 Edition for safety class II piping. Other applicable considerations have remained within acceptable limits.
4 ~ the existing and modified support/restraints have been demonstrated to be adequate for revised stress analysis loads in accordance with the applicable codes & criteria.
P C/M 047-189 Rev. 0 SAFETY EVALUATION (Continued)
(ii) As a for result of this modification, there is no possibility an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report
'ecause the piping stresses for the modified condition have remained within the stress limits allowed in USAS B31.7 Code 1969 Edition for safety class II piping. Other applicable considerations have remained within acceptable limits.
- 2. the existing and modified supports/restraints have been demonstrated to be adequate for revised stress analysis loads in accordance with the applicable codes & criteria.
(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification because it neither changes the design parameter of the systems (Reactor Coolant Gas Vent or Safety Injection) nor does it change the design flow or functional requirements of the systems. This modification does not affect the integrity of the pressure boundary of the piping.
The implementation of this PCM does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or a change to the Technical Specifications, and prior NRC approval for the implementation of this PCM is not required.
P C/M 102-189 Rev. 0-2 ABSTRACT The modifications associated with this change provide for a diverse means of removing power from the Control Element Assembly Drive Motors (CEADM) in response to the requirements of the ATWS Rule, 10CFR 50.62. The modification uses electrically isolated input signals from the pressurizer pressure instrument loop to provide a signal to open new load contactors in the output circuit of the motor generator sets which power the CEADMs. These load contactors are in series with the existing output breakers so that the operation of either accomplishes the required result. These additions do not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since they are an additional and diverse means of creating a shutdown signal and removal of power to the CEADMs. Isolators and the series contactor provide assurance that the original design shutdown methods are undisturbed and therefore no changes to any Technical Specification are required.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or, malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications included in this Engineering Package do not involve an unreviewed safety question because:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. This is confirmed by the following:
To ensure that the existing Reactor Protection System (RPS) is fully capable of functioning as designed to protect against accidents and malfunctions of equipment important to safety that have been evaluated and documented in the FSAR and form the design basis for the licensing commitments, the NRC has issued specific guidance for electrical independence for the RPS and the Diverse Scram System (DSS).
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P C/M 102-189 Rev. 0-2 SAFETY EVALUATION (Continued)
Electrical independence is required from sensor output to the final actuation device at which point non-safety circuits must be isolated from safety related circuits.
To achieve this, the DSS is designed so that inputs to the DSS from the sensors will be electrically isolated from the Reactor Protection System by isolation devices (I/I converters) installed in RTGB-106. The DSS logic circuits will be contained in the ESFAS cabinets which are electrically independent from the Reactor Protection System. The DSS outputs will be directed to the CEA drive MG set load contactors, which are not safety-related nor associated with the Reactor Protection System trip functions. DSS outputs will be isolated non-safety to prevent adverse electrical interactions between the actuation devices and the safety related portion of the DSS installed in the ESFAS cabinet.
Physical separation from the RPS is not required, but the separation criteria applied to the existing protection system must not be violated. Since the DSS is not maintained in the same cabinets as the RPS, separation criteria are not affected.
Consolidated Controls Corporation has designed the DSS such that it will not adversely interact with ESFAS, with which is shares the same cabinets. It interfaces with the ESFAS in only three areas: 1) Power supplies The DSS is separately fused so that a DSS fault cannot propagate through the power supplies to any ESFAS functional loops. 2) ATI The DSS is optically isolated from the ATI to ensure independence 'from other ESFAS functional loops. 3) Input signal (pressurizer pressure) The DSS is designed so that single failure at the bistable input will not affect the ESFAS functions tha't use the same input. Since the DSS is designed to meet or exceed the requirements of the existing ESFAS design, the DSS will not cause adverse interactions through the interfaces with ESFAS.
No change has been made to the input parameters or the ability of the RPS and ESFAS to perform safe shutdown functions based on these parameters. Therefore, the probability of occurrence -or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased.
P C/M 102-189 Rev. 0-2 SAFETY EVALUATION (Continued)
There is no possibility for an accident or malfunction of a different type than any previously evaluated. This is confirmed by the following:
Design of the DSS does not introduce new components that could violate the RCS pressure boundary, release radioactive material to the environment, or damage the reactor fuel. It does not change the method or ability of the RPS or ESFAS to detect system parameters and perform required safety functions based on the values of those parameters. Malfunction of the DSS cannot cause an event other than a reactor trip. The new DSS will reduce the risk of an over-pressure condition due to an ATWS event and the subsequent stressing of the RCS in excess of ASME Level C pressure of 3200 psia (Reference 6.29, Section 1.3). Therefore, the DSS does not increase the possibility for an accident or malfunction of a different type than any previously evaluated, but actually reduces it.
(iii) This modification does not reduce the margin of safety as defined in the basis for any technical specification.
This is confirmed by the following:
This modification does not adversely affect equipment whose operation is defined by the Plant Technical Specifications and does not affect the operation of the RPS in any manner. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.
Implementation of this EP does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question nor a change to the Plant Technical Specifications. NRC approval has been obtained for the conceptual design with the Safety Evaluation dated September 6, 1989. Prior NRC approval for the implementation of this PCM is not required.
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P C/M 131-189 Rev. 0-1 ABSTRACT The modifications associated with this change provide for the upgrade of two existing and the installation of two new pressure switches for the Emergency Diesel Generator air receivers. These switches control the starting and stopping of air compressors which charge the receivers. The two new switches are installed on the two air receivers which were previously not monitored. All four switches are qualified to meet the design requirements of the system. These additions and upgrades do not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since all switches are able to meet original design requirements and the new switches provide additional monitoring to keep all receivers at the required pressure. There are no changes required to the Technical Specifications since the original configuration has been upgraded and expanded to continue to meet the design requirements and operation of the air system.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications included in this Engineering Package (EP) do not involve an unreviewed safety question because:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased by this modification because it does not reduce the availability, redundancy, capacity, or function of any equipment required to mitigate the effects of an accident. The pressure switches involved in this EP are not required to operate during or after a design basis earthquake. The pressure switches are required to maintain the readiness of the diesel generators during the normal operation of the plant. The new pressure switches and their associated tubing connections to the air start receivers are designed to meet seismic category I requirements to assure the maintenance of the pressure boundary of the Diesel Generator air starting systems during and after a seismic event. This modification will increase the reliability of the diesel generator air starting system by assuring that all air receivers are always fully charged and in a state of readiness when called upon to perform their safety function.
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P C/M 131-189 Rev. 0-1 SAFETY EVALUATION (Continued)
(ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated since no changes have been made to the operation of the diesel generator air start system. This modification does not affect any other systems which perform or monitor a safety related function.
(iii) This modification does not change the margin of safety as defined in the basis for any Technical Specification since the pressure switches being replaced do not form the bases of any Technical Specifications.
Implementation of this Nuclear Safety Related PCM does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question nor a change to the Plant Technical Specification; thus, prior NRC approval for the implementation of this PCM is not required.
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P C/M 152-189 Rev. 0 ABSTRACT The modifications associated with this change provide for the installation and modification of relaying and control equipment associated with protection against certain inadvertent non-synchronized connections of the turbine generator to the power grid. Modifications to the unit main transformer cooling system circuitry were also performed. These additions and modifications do not create an unreviewed safety question per 10CFR.50.59 or require prior NRC approval since the changes are to the generator protection system which serves no safety function and does not form the basis of any Technical Specification.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications included in the Engineering Package (EP) do not involve an unreviewed safety question because:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased by these relaying modifications as enhance the generator protection. The possible it will failure of this equipment will not prevent safety related equipment from performing their intended functions.
Therefore, the implementation of these modifications cannot increase the probability of occurrence or the consequences of an accident or malfunction of equipment.
The possibility of an accident or malfunction of equipment of a different type than any evaluated previously is not created. The equipment added/modified by this EP is not required during an accident condition nor will it prevent safety related equipment from performing their functions. This modification does not affect any safety related equipment.
(iii) The margin of safety as defined in the bases for any Technical Specification is not reduced by this modification. The equipment added/modified by this EP does not form basis of any Technical Specifications.
PC/M 152-189 Rev. 0 SAFETY- EVALUATXON (Continued)
The implementation of this PC/M does not require a change to plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question or a change in the Technical Specifications and prior NRC approval for the implementation of.
this PC/M is not required.
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P C/M 160-189 Rev. 0 ABSTRACT The modifications associated with this change provide for the installation and connection of sound powered communications between the St Lucie Unit 1 Control Room and the Technical Support Center.
This system replaces the existing system with a permanent system requiring no external power. This modification does not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since this communications system serves no safety function.
and does not form the basis of any Technical Specification.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications included in this Engineering Package (EP) do not include an unreviewed safety question because:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. The installation of a sound powered communication system between the Control Room and the Technical Support Center does not create any new safety concerns previously not evaluated in the FSAR. The communication system is not safety related and all the new equipment has been seismically mounted/supported precluding any effects on safety systems.
As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated. This modification does not change the operation of any safety related equipment, and there is no introduction of any new failure mode for safety related equipment.
(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification.
The margin of safety provided by the Technical Specification is not affected as the equipment modified (sound powered jack stations) does not form basis of any Technical Specifications. Therefore, no change to the Technical Specification is required.
P C/M 160-189 Rev. 0 SAFETY EVALUATION (Continued)
The foregoing constitutes, per 10 CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior NRC approval for the implementation of this PC/M is not required.
P C/M 250-189 Rev. 0
'ABSTRACT The modifications associated with this change provide for the installation, as required, of steam generator tube plugs which meet the design requirements for the Reactor Coolant System pressure boundary. The use of tube plugs does not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since tube plugging is an analyzed activity and failures are enveloped by the steam generator tube rupture even't. The number of allowable plugged tubes is a basis for Technical Specifications but since the total number of plugged tubes is less than that assumed in the analysis, no change to any Technical Specification is required.
10 CFR 50.59 allows a change to a nuclear facility without prior NRC approval if an unreviewed safety question does not exist and if changes to Technical Specifications are not involved. The following arguments demonstrate that an unreviewed.safety question does not exist relative to this modification:
i) The probability of occurrence of a design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased since this modification does not decrease the design margin of the RCS pressure boundary (the tube plugs meet or exceed all design requirements for ASME Section III, Class 1 components).
The consequences of a previously postulated design basis accident or malfunction of equipment important to safety previously evaluated in the FSAR are not made more severe for the same, reasons given above and since no existing accident mitigation equipment or systems are altered by this modification.
The possibility of an accident of a different type than previously addressed in the FSAR does not exist since no new systems or equipment are introduced by this modification.
Failure of a tube plug would be no more severe than a steam generator tube rupture, a previously evaluated condition.
Therefore, no new accidents are created.
iii) The margin of safety as defined in the basis for any technical specification is not reduced since the total number of tubes plugged in the steam generators following this modification is less than assumed in the Cycle Ten Reload Analysis.
Since the above arguments demonstrate that an unreviewed safety question does not exist, and since a revision to the Technical Specifications is not required, the addition of the Combustion Engineering tube plugs to the Unit 1 steam generators does not require prior NRC approval.
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P C/M 319-189 Rev. 0 ABSTRACT The modifications associated with this change provide for upgrades to the Reactor Coolant Pump (RCP) motor and oil supply system by changes to covers and screens and oil system instrumentation. All changes have been designed with seismic considerations but since the RCP motors and their oil systems serve no safety function, these changes do not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval. The RCP motors and oil systems do not form the basis for any Technical Specification.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
This Engineering Package (EP) does not involve an unreviewed safety question, and the following are bases for this conclusion:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. Since this modification is an improvement of existing non-safety related equipment and added equipment is seismically mounted so as not to impact Safety Related equipment, to plant it will have safety.
no adverse effect on equipment vital ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. Added equipment is seismically mounted so as not to impact Safety Related equipment. Other modifications are for the RCP motors which have no effect on the consequences of an accident, nor are used to mitigate an accident.
iii) The margin of safety as defined in the bases Specification is not affected by this PCM.
for any Technical The modifications to the RCP motor do not compromise integrity or reliability, and the operating characteristics of the motor are not changed.
P C/M 319-189 Rev. 0 SAFETY EVALUATION (Continued)
The implementation of this -PCM does not require a change to the Plant Technical Specification.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.
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P C/M 336-189 Rev. 0-1 ABSTRACT The modifications associated with this change are to piping, instrumentation and control circuits which relate to boric acid systems which require changes due to the reduction in boric acid concentration. Technical Specifications have been revised and all required safety evaluations and analyses are in place to support the lower concentration. The lower concentration allows for the removal of piping relief valves and heat trace, changes in temperature controls and alarms and a change to level alarms. A local temperature indicator was installed near boric acid piping to monitor ambient temperatures. The changes introduced by this modification=-do not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval because the equipment affected is no longer required in order for the remaining components to perform their required functions. Any required Technical Specification changes were performed during the original analysis effort and the removed or changed equipment does not require further changes.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications have been evaluated under 10CFR 50.59 and been determined that this modification does not it involve has an unreviewed safety question. The following are the bases for this conclusion.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased. Utilization of the BAMT heaters is not essential due to the reduction of the boric acid concentration. The addition of a local air temperature indicator is done as a convenience and will have no impact on the plant. The piping and support modifications are for the removal of equipment which is no longer necessary. These changes do not affect the design requirements of any other instruments or equipment in the Boric Acid Makeup System. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety has not increased.
lg P C/M 336-189 Rev. 0-1 SAFETY EVALUATION (Continued)
(ii) The possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report is not created because components involved with this modification introduce no new type of accidents and cannot cause malfunctions of any other safety related equipment.
(iii) The margin of safety as defined in the bases for any Technical Specification is not reduced since this modification does not affect the bases of any Technical Specification.
The implementation of this PC/M does not require a change to the plant Technical Specifications, nor does Safety Question.
it create an Unreviewed The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an Unreviewed Safety Question and prior Nuclear Regulatory Commission approval for the implementation of this PC/M is not required.
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P C/M 342-189 Rev. 0 ABSTRACT The modifications associated with this change provide for finer voltage regulation and control for the Control Element Assembly Motor Generator Sets. The voltage supplied to the Control Element Assemblies remains the same with a better degree of load sharing between motor generators operating in parallel. This improvement in voltage regulation does not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval because the motor generator sets and their voltage regulation system do not perform a safety function and do not form the basis of any Technical Specification.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety..question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications have been evaluated under 10CFR 50.59 and been determined that this EP does not involve an unreviewed it has safety question. The following are the bases for this conclusion.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report is not increased. The MG sets are not safety related and do not mitigate or monitor accident conditions or interface with safety related systems.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report is not created. Components involved with this modification introduce a new type of accident (i.e., new rheostat opens), however these components cannot cause malfunctions of any safety related equipment.
(iii) The margin of safety as defined in the bases for any Technical Specification is not reduced since these minor wiring modifications do not affect any Technical Specification, nor does the CEAMG system form the bases of any Technical Specification bases.
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P C/M 342-189 Rev. 0 SAFETY EVALUATION (Continued)
The implementation of this PC/M does not require a change to the plant technical specifications, nor does it create an unreviewed safety question.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Nuclear Regulatory Commission approval for the implementation of this PC/M is not required.
P C/M 348-189 Rev. 0 ABSTRACT The modification associated with this change provides for the permanent removal of the startup neutron sources from the St Lucie Unit 1 core. Irradiated fuel reloaded from previous cycles is to be used as the neutron source for monitoring subcritical multiplication during all future reload operations. The removed startup neutron sources are to be stored in the spent fuel pool.
This change does not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since the required function is being maintained by another method and the sources are being stored in an approved location. The existing startup neutron sources do not form the basis for any Technical Specifications.
Title 10 of the Code of Federal Regulations Section 50.59 (Reference 6.3) states that the licensee may make changes to the facility as described in the FSAR without prior Nuclear Regulatory Commission approval unless the proposed change involves a change to the technical specifications or an unreviewed safety question. A proposed change involves an unreviewed safety question if:
i) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is increased, or ii) the possibility than for an accident or malfunction of a different evaluated previously in the safety analysis type any report may be created, or iii) the margins of safety as defined in the basis for any technical specification is reduced.
The modification proposed in this engineering package neither involves a change to the plant technical specifications nor an unreviewed safety question. Plant operation utilizing irradiated fuel as a neutron source is not a safety concern. The basis for this conclusion are addressed below.
- i. 1) increased probability of a malfunction/accident The removal of the startup neutron sources and their replacement with irradiated fuel does not increase the probability of a malfunction or accident.
P C/M 348-189 Rev. 0 SAFETY EVALUATION (Continued)
The fuel assemblies that are to reside in the source locations for a given operating cycle may not meet the criteria given in Reference 6.2 required for a fuel assembly to be used as a neutron source during refueling. These criteria define a minimum accumulated exposure which a fuel assembly must receive in order to produce a source strength large enough for positive indication on the ex-core neutron detectors.
Therefore, temporary placement of irradiated fuel assemblies into the source locations (2 nearest the ex-core wide range detectors) may be required to produce a count rate off the ex-core detectors that is sufficient for monitoring subcritical multiplication. When enough fuel assemblies are placed in the core to produce a sufficient count. rate off the ex-core neutron detectors, the temporary fuel assemblies may be removed from the source locations and replaced with the fuel
, assemblies, that are to reside in the source location during the operating cycle. The additional transfer of at most two fuel assemblies does not increase the probability of an accident or a malfunction by a measurable quantity.
The removal of the startup sources is needed since the source is becoming damaged and it is becoming difficult to transfer the startup sources from one fuel assembly to another during fuel reload operations. Due to wear of the upper end fitting, it is difficult to latch the source with the source lifting tool. The probability of a malfunction of the source handling tool is reduced since the sources are to be permanently removed from the core and latching of the sources with the lifting tool will no longer be required.
1~ 2) increased consequences of a malfunction/accident The removal of the startup neutron sources and their replacement with irradiated fuel does not increase the consequences of a malfunction or accident.
The consequences of malfunction of the fuel handling tool are not increased (FSAR section 15.4.3 fuel handling accident).
new type of malfunction or accident No new types of accidents or malfunctions exist as a result of removing the startup neutron sources and utilizing irradiated fuel as the startup neutron source.
ill~ This modification does not reduce the margin of safety in the basis for any Technical Specification.
The neutron count rate on the ex-core neutron detector will not decrease due to this modification.
P C/M 348-189 Rev. 0 SAFETY EVALUATION (Continued)
The implementation of this PC/M does not require a change to or impact the plant Technical Specifications.
As per the requirements of 10CFR 50.59, this change does not involve an unreviewed safety question or a change to the Technical Specifications, therefore, prior NRC approval for the implementation of this PC/M is not required.
P C/M 053-190 Rev. 0 ABSTRACT The modification associated with this change provides for the removal of two 1 inch drain lines located downstream of the Main Steam Isolation Valves. Required drainage can be performed by other drain lines. This change does not create an unreviewed safety question per 10CFR 50.59 or require prior NRC approval since these drain lines do not perform any safety function and do not form the basis for any Technical Specification.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) consequences if the probability of occurrence or the of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modification included in this Engineering Package (EP) does not involve an unreviewed safety question, and the following are the bases for this conclusion:
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The 1-MS-49 Main Steam drain lines are not used in any safety analysis for accidents or malfunction of equipment and as such are Non-Safety Related and will have no effect on equipment vital to plant safety.
The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not created. The components involved in this modification have no safety related function and no changes have been made to the operational design of the systems impacted by this EP.
iii) The margin of safety as defined in the bases for any Technical Specification is not affected by this PCM, since the components involved in this modification are not included in the bases of any Technical Specification.
The implementation of this PCM does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.
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I r5 P C/M 243-189 Rev. 0 ABSTRACT The modification associated with this change provides for a means for an operator to regain manual control of- the Steam Generator Blowdown Containment Isolation Valves after a Containment Isolation Signal or High Radiation signal has been received. The override of the isolation signal requires two operator manual actions to the switch modified by this change. The purpose is to allow sampling and blowdown of the Steam Generators during recovery from certain design basis events. This change does not create an unreviewed safety question per the requirements of 10CFR 50.59 and does not require prior NRC approval since there has been no change to the systems capability to respond to an isolation signal. There is no change required to the Technical Specifications since there was no change to the valves or their circuitry which affects the original isolation design. A new isolation signal will cause the valves to close regardless of prior switch operation.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
The modifications included in this Engineering Package do not involve an unreviewed safety question because:
The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased since the use of the override capability will be utilized only in an off-normal/emergency operating condition and strict adherence to the governing procedure will require alignment of the blowdown system to receive process fluid with potentially high activity levels. The system's normal operation is unaffected by this modification.
(ii) The possibility for an accident or malfunction of a different type than any previously evaluated in the Safety Analysis Report is not created because components involved with this modification introduce no new type of accidents and cannot cause malfunctions of any safety related equipment. No changes have been made to the normal operational design of any. control circuits or associated systems which are important to safety.
P C/M 243-189 Rev. 0 SAFETY EVALUATION (Continued)
(iii) This modification does not change the margin of safety as defined in the basis for any Technical Specification since the steam generator blowdown system is not part of any Technical Specification bases. The CIS portion of the circuit will still function in accordance with the Technical Specification.
Implementation of this Nuclear Safety Related PCM does not require a change to the Plant Technical Specifications.
The foregoing constitutes, 10CFR 50.59(b), the written safety evaluation which provides the basis that this change does not involve an unreviewed safety question nor a change to the Plant Technical Specifications; thus, prior NRC approval for the Implementation of this PCM is not required.
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