ML17308A491
| ML17308A491 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/02/1990 |
| From: | Sager D FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| L-90-110, NUDOCS 9004200672 | |
| Download: ML17308A491 (114) | |
Text
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ACCESSION NBR:9004200672 DOC.DATE: 90/04/02 NOTARIZED: NO 'DOCKET FACIL:50-389 St. Lucie Plant, Unit 2, Florida Power a Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION SAGERiD.A. Florida Power & Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards description a summary of safety evaluations of plant changes/mods reportable per 10CFR50.59.
DISTRIBUTION CODE: ZE47D COPIES RECEIVED:LTR l ENCL Q SIZE:
TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made H out Approv NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 0 PD2-2 PD 5 5 NORRIS p J 1 0 INTERNAL: ACRS 6 6 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 NRR/DLPQ/LHFB11 1 1 NRR OEAB11 1 1 NUDOCS-ABSTRACT 1 1 LE 02 1 1 RGN2 FILE 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 R
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P.O. Box14000, Juno Beach, FL 33408-0420
'APRIV 0 2 1990 L-90-110 10 CFR 50.59 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:
Re: St. Lucie Unit 2 Docket No. 50-389 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 October 7, 1988, and October 6, 1989, and correlates with the information included in Revision 5 of the Updated Final Safety Analysis Report submitted under separate cover.
Very truly yours, D. A. ger Site i e President St. L 'e Plant DAS/EJW/gp Enclosure cc: Stewart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie Plant II, USNRC 9004200672 900402 PDR ADOCK 05000389 P PNU an FPL Group company 2<97
ENCLOSURE ST LUCIE UNIT REPORT OF PLANT CHANGES HADE PURSUANT TO 10 CFR 50.59 OCTOBER 7 I 1988 TO OCTOBER 6 I 1989 50-389 ST. T.UCIE PLANT $2 FP&UD BEPORT OF PLY CHANGES MADE PURSUANT TO 10 CFR 50.59 Oct. 7, 1988 to Oct 6, 1989 w/1tr dtd 4/2/90 $P9004200672
1IT CHANGE/MOD REPORTABLE PUREU TO 10CFR50.59 FOR ST LUCIE UNIT 2 FSAR AMENDMENT 5 SUPPLEMENT TITLE 182-285 0-1 PROTECTIVE COATINGS REPAIR AND/OR REPL IN RCB 048-286 0 LOW PRESSURE TURBINE UPGRADE 051-286 0-2 INSTRUMENT AIR UPGRADE 075-286 2 HEATER DRAIN PUMP MECHANICAL SEAL DEMINERALIZED WATER SUPPLY 121-286 0-1 HPS SECURITY LIGHTS AND EDG LOADING 135-286 0 RCP SEAL INJECTION INTERFERENCE 003-287 0 BECHMAN WASTE GAS OXYGEN ANALYZER REPL 052-287 1 CONDENSATE POLISHER TIE-INS 057-287 0 'DG LOCKOUT RELAY MOD 060-287 0 BORIC ACID FLOW XMTR REPL'ACEMENT 077-287 2-3 ERDADS/SAS UPGRADE 129-287 0-1 SIT & CONT FAN COOLER INSTRUMENT UPGRADE 140-287 0 CONDENSATE POLISHER CROSS-TIE 154-287 0 RPS EXCORE SAFETY CHANNEL MODIFICATION 068-288' POWER FEED TO MH261 SUMP PUMP 090-288 0-2 PRESSURIZER NOZZLE REPLACEMENT 091-288 0-2 RCS HOT LEG NOZZLES REPLACEMENT 144>>288 0 480V PZR HTR XFMR 2A3 & 2B3 REPL 147-288 0 RCP 2A1 SNUBBER MOD 189-288 0 PROT COATINGS REP/REPL-RCB 268-288 0 AFW MOV-09-13, 14, BREAKERS 274-288 0 FWRV MODIFICATION 275-288 0 MOV THERMAL OVERLOAD MOD 281-288 0 RX CAVITY SEAL RILE WATER FILL MOD
P T CHANGE/MOD REPORTABLE PURSU TO 10CFR50.59 FOR ST LUCIE UNIT 2 FSAR AMENDMENT 5 NUMBER SUPPLEMENT 1ITLE 283-288 0-2 BORIC ACID CONCENTRATION REDUCTION 284-288 . 0 INST CHANGES HUMAN FACTORS CONCERNS 297-288 0 ELIM OF SM THERM SNUBBER 367-288 0 MFIV MAINTENANCE LIFT BEAMS 389-288 0-1 FUEL ROD DESIGN-REGION G 404-288 0-1 PIPE WHIP RESTRAINT REMOVE 035-289 0 KW PIPING 6 RESTRAINTS OF FIS 21-9B 044-289 0 STM GEN TUBE PLUGWE DESIGN 171-985 0 HYPOCHLORITE SYS I&C ENHANCEMENTS 111-986 1 SIMULATOR BLDG PIPING TIE-INS
CM 182-285 ge 1 of 1 PROTECTIVE COATINGS REPAIR AND/OR REPL IN RCB ABSTRACT This engineering package covers the maintenance of Service Level I protective coatings on concrete and steel surfaces inside the Reactor Containment Building.
This project is classified as nuclear safety related and does not constitute an unreviewed safety question.
Supplement No. 1 This supplement includes Revision 1 to Specification No. PSL-C-182-285.2. The specification revision is noted on pages 3, 0 and 5 and addresses the removal of blisters in the concrete topcoat without recoating the concrete surfacer under the blisters. The safety evaluation has not been affected since the surfacer material has been tested without a topcoat and has passed the design basis accident conditions including radiation.
SAFETY EVALUATION The safety related aspect of the. protective coatings used inside the Reactor Containment Building is that they must not peel off, become lodged in the containment sump and render it inoperable.
The Unit 2 FSAR addresses this concern and includes the requirement that the coatings be DBA tested to withstand accident conditions without delaminating or peeling off the applicable surfaces.
The coatings specified for this project are DBA tested and their composition, functional and testing requirements are addressed in the Unit 2 FSAR..-For these reasons the probability of occurence or consequences of a design basis accident or malfunction of equipment important to the safety of the plant has not been increased. In addition, there will continue to be no possibility of an accident or malfunction different than those previously evaluated in the Unit 2 FSAR. Finally, the margin of safety as defined in the plant technical specifications has not been reduced. It is therefore concluded that the repair and/or replacement of protectiv~ coatings on surfaces inside the Reactor Containment Building as outlined in this package does not pose an unreviewed safety question pursuant to 10 CFR 50.59.
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P M 048-286 elof I LOW PRESSURE TURBINE UPGRADE ABSTRACT This Engineering Package provides documentation of the design enhancement performed by Westinghouse, the Original Equipment Manufacturer (OEM), during the refurbishment of the PSL Unit 1 low pressure turbine rotors The upgrade; heavy disc keyplate design (HD/KP), was developed by Westinghouse to mitigate the consequences of a gen8ric problem of stress corrosion cracking (SCC) plaguing a majority of the nuclear LP turbine rotors in service.
This modification is classified as quality related based on the potential for missile generation. However, it does not constitute and unreviewed safety question.
SAFETY EVALUATION The low pressure turbine upgrade (heavy disc keyplate "esign) was developed by Westinghouse in an effort to mitigate the stress corrosion cracking problem of nuclear low pressure turbine discs.
The Westinghouse HD/KP design provides the following benefits over the original design in regards to the SCC problem.
Lower Operating Stresses Lower Yield Strength Materials Elimination of Keyways from Disc Bores Longer Inspection Interval Lower Probability of Disc Rupture The Westinghouse design enhancement has been utilized by several utilities, initially at PSL I in 1933, when the original spare rotors were procured to the HD/KP design. It was shown at that time that the new design provides for lower yield strength disc material, and tougher, more crack resistant discs.
The HD/KP design change reduces the probability of L.P. disc generated missiles by an order of magnitude, (Ref. 6.3). The Westinghouse Model; MSTC-I-P. "criteria for low pressure nuclear turbine Disc inspection," used to calculate disc inspection intervals has been approved by the NRC, (REF 7.0). Since the yield strength of discs is discussed in the FSAR (Section 10.2),
the physical changes in the L.P. rotor design are subject to review pursuant to 10 CFR 50.59. The use of the refurbished (HD/KP) L.P, rotc@ at St. Lucie 2 is an acceptable change to the St. Lucie Plant unit 2 facility in that it does not involve a change in the technical specifications incorporated in the license or an unreviewed safety question as defined at 10 CFR 50.59(a)(2) because:
- 1) The probability of occurrence of turbine missile "eneration has been reduced and the consequences of a failure as previously evaluated in the safety analy:is report has not increased.
- 2) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report has not been created by replacement of the LP rotors.
- 3) No St. Lucie Plant unit 2 technical specification nor the margin of safety as defined in the basis for any technical specification is affected by this change to the plant.
CM 051-286 age 1 of 2 INSTRUMENT AIR UPGRADE This Engineering Package (EP) is for the installation of 2 nev air compressors> 2 aev desiccant air dryers (including prefilter and afterfilter packages) aad removal of the existing desiccant air dryer and afterfilter package which do not have sufficient capacity to accommodate the nev compressors. For normal operation one of the two aev air compressnrs aad one aev air dryer vill operate and, the other vill serve as a standby. The existing compressors vill remain as backups ia the case of loss of offsite pover, since only these compressors are capable of being loaded oa the diesel generator.
This EP is classified as Non-Safety Related since the instrument air compressors and associated equipment perform no safety function. The safety evaluation has determined that this RP does not constitute an unrevieved safety questioa and implementation of the EP does not require a change to the Plant Technical Specifications. Therefore, pxior NRC notification for implementation of this EP is not, required.
This EP has no impact on plant safety and opexatioa.
Su lement 1 The supplement I provides revised design for the piping and piping supports/restraints to install additional flexible hoses to minimize piping vibration. It also provides additional design details. for the instrument air control. tubing, tagging of intercooler temperature gauges, cable splice detail and revised foundation details for the compressor skid.
The original results of evaluation as stated in the safety evaluatioa remain uachaagedo Su lement 2 Mals Engineering Package {EP) revision 2 provides revised design basis/analysis, operation and maintenance guide1incs and design for:
installatioa of check valves in the compressor discharge lines, replacement of compressor tubiag with stainless steel tubing, replacement of all pressure and temperature indicators with standard indicating scs1cs, replacement and autoaaticm of the dxycx prcfiltcr drain ~m> .installation of .stai331css stce1 socket waded pipiagfor the dryer potage sad imp2ovad'removability of mftcrcxlolcr't'utooatXg drain valve It also provides,:far replaccaeat'f,thecal overload relays for compressor motors and revised design for individual control air lines to the cempressors'nloaders and pressure switches to prevent compressor hunting.
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'Ihc original results of evaluation as stated in thc safety evaluation remain mchaagedi
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SAFETY EVALUATION PCM 051-286 page 2 of 2 With respect to Tit 10 of the Code of Federal R oa, Part 50.59,
. a proposed change shall be deemed to fnvolve an unrevfewed safety question; (f) if the probability of occurrence or the coasequences of an accident or malfunction of equipmeat important to safety previously evaluated fn 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 ia the safety analysis report may be created; or (ifi) ff the margin of safety as defined fn the bases for any Technical Specification is reduced.
This EP fs for the addition of two 100Z capacity new compressorsf two new desiccant afr dryers and removal of the existing low capac1ty desiccant dryer and afterfflter package.
Failure of the instrument air compressors and compoaents resulting in loss of IA and consequent effects as stated in the FSAR Subsection 9.3.1.3 have been reviewed. This modification does not add any new failure modes for the safety related afr operated valves. This modff1cation is therefore classified=as nonnuclear Safety Quslfty Group D aad nones 1E.
The increase in IA requirements fxom 155 SCFH to 400 SCFN aad pressure from 90-100 psfg to 110-115 psig is based on FPL studies fox the requiremeat of the IA.
Based on the above description, the modff1cation included fn this EP fs considered to be non-safety related This EP does not involve an uarevfewed safety quest1oa, and the following are bases for this justification:
(1) The'probability of occurrence or the consequences of aa accideat or malfunctfoa of equipment important to safety
,previously evaluated fa the safety aaalysfs xepoxt fs aot increased. The instrument air system compressors aad associated equfpment are not used directly ia any safety analysis for accidents or malfunction of equipment and ae such are non-safety related and will have no effect on equipment vital to plant safety.
(11) The possibility for an accident or malfunction of a different type than any evaluated previously ia the safety analysis report 1s aot created. The components involved in this modification have no safety related function and no changes have been made to the normal operational design of the system with the compressoxe2C and 2D ia operation. In this mode the IA compreseors 2A aad 2B discharge valves V18109 snd V18118 are closed to prevent IA leakage vfa these compressors.
Similarly, whenevex the IA compreseors 2A and 2B are required to operate, valve V18660 is closed to pxevent IA leakage via comnressors 2C aad 2D.
(111) The margin of safety as defined in the bases fox any Technical Specf fication 1 s not affected by thf e PQf, since the components involved in thfe modification are aot included in .
the bases of any Techafcal Specification.
The implementation of this P(H does not xequfre a change to the plant Technical Specification.
The foregoing constitutes, per 10CFR50-59 (b), the written safety evaluation which provided the bases that this chaage does aot involve an unrevfewed safety question or a change to the Plant Technical Specifications aad prior Commission approval for the implementation of the P(M fs not requfred.
M 075-286 ge 1 of 2 HEATER DRAIN PUMP MECHANICAL SEAL DEMINERALIZED PATER SUPPLY This design package provides the required engineering for adding permanent piping from the demfneralfzed water system to the Unit 2 heater drain pumps'echanical seals. The piping will make available to the seals the necessary backup flushing water meeting the appropriate chemistry requirements. The backup water source is requfred during initial plant startup whenever the pumps sit idle.
8ased on the failure modes analysis and 10CFR50.59 review, thfs modification does not impact any safety related equipment and fs not relied upon for any accident prevention or mitigation. Thus, ft does not constitute an unreviewed safety questfon and is correctly classified as non-nuclear safety related. ?mplementatfon of this modification, therefore, does not z"equfre prior NRC approval.
This package revision provides valve drawings for valves added by this PC/M and modifies the expiration date to reflect the correct format. The scope of work specified by this Engineering Package has not been affected by this revision. The
. safety classification and the safety evaluation as stated is correct and fs not impacted.
u lement This Supplement provides revisions to certain drawings to correct improperly identified generic valve tag numbers. No other changes are addressed. The safety classification does not change, the safety evaluation remains valid and the Technical Speciffcatfons are not affected as a result of this revision.
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PCS 075.-286 page 2 of 2 SAFETY EVALUATION The Unit 2 Heater Drain Pumps are located in a Non-Nuclear Safety Related system and as such are not required to function during any existing analyzed accident scenario. Therefore, modifications to these pumps affect only Non-Nuclear Safety Rehted, Quality Group D equipment.
Based on the failure mode analysis, faflure of the demineralized water supply piping could result only in failure of the heater drain pumps. Since the piping and components are located remote from any safety related equfpment or components, failure of this equipment will not inhibit operation of any safety related equipment or components.
Based on the above evaluation and informatfon supplied in the desfgn analysis it can be demonstrated that an unreviewed safety question as defined in 10CFR50.59 does not exht.
o The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report has not been increased.
Sfnce tMs desfgn change does not alter or affect equfpment used to mitigate accidents, the probability of occurrence of analyzed accidents remains unchanged..
o The possfbflfty of an accfdent or malfunctfon of a dffferent type than any evaluated previously in the safety analysis report has not been created.
There fs no new failure mode introduced by this change that has not been evaluated previously fn the PSAR.
o The margin of safety as defined in the basis for any Technical Specifications has not been reduced.
This change has no affect on any existing Technical Specifications.
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HPS CURITY LIGHTS AND EDG LOADING a y f2 ABSTRIL CZ This Engineering Package (EP) covers the replacement of all outdoor security lighting fixtures along the plant perimeter and within plant protected area to reduce the high maintenance cost as well as to enhance the security system illumination level.
The lighting fixtures being replaced are connected to Normal Emergency Lighting Panels LP2-2A1, LP2-2A2, LP2-281 & LP2-2B2 which are loaded on Emergency Diesel Generators 2A and 2B during certain accident scenarios. As such, this package is classified as safety related.
A review of the changes to be implemented by this PC/M was performed against the requirements of 10 CFR'50.59 as indicated in section 3.0 of this EP. As a result, the replacement of the present lighting fixtures and the increased lighting load incurred by the new fixtures do not constitute an unreviewed safety question, will not affect plant safety and its operation and will not require a change to the Technical Specification. Therefore prior Commission approval is not required for implementation of this PC/M.
SUPPLEMENT 1 This supplement provides for the replacement of the fixed light poles on the west bank of the intake canal with new hinged poles. The function of the light poles (to hold luminaires) is not changed by this modification. The new poles and pole foundations have been evaluated and found to be able to withstand 120 MPH wind velocity loading as required by South Florida Building Code, Section 2306. As such the original safety evaluation is not affected by this supplement. There are no unreviewed safety questions per 10CHt50.59 and no change to the plant Technical Specification is required. Therefore prior Commission approval is not required for implementation of this supplement.
SAFETY EVAWATION Pith 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 probabQ.ity 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 modification replaces the existing 150M Appleton HPS fixtures with General Electric 250M HPS fixtures for the outdoor security lighting to decrease plant maintenance of these fixtures. These new lighting fixtures are installed with top and side visors minimizing the skyward brightness as required by St Lucie Unit 2 conditions of Certification for environmental protection to minimize turtle disorientation.. These fixtures are part of the outdoor security plan which is not required to monitor or mitigate the consequences of an accident. As part of the normal emergency lighting system the outdoor security lighting system is powered from the Emergency Diesel Generators. This assures that in the unlikely event of a loss of offsite power, adequate plant lighting is avai1able for the plant and personnel safety. Since these lights are connected to the Diesel Generators, this package is considered safety related.
PCM 121-286 page 2 of 2 In order to insure that the increased fixture wattage w111 not impact the Diesel Generator load block value specified in UPSAR Amendment 4 Table 8.3.2 an evaluation of the 1mpact of the replacement fixtures was performed (Ebasco Study PLO'108-44.5000, attachment 7.3). This study conc1uded that the increased lighting load incurred by the new fixtures will not increase the 125KW nominal load specified in the Diesel Generator loading schedule.
Supplement 1 of the modification replaces light poles on the west bank of the intake canal with new lowerable hinged poles. The function of the poles (to hold luminaires) is not changed by this modificat1on. The hinged design allows convenient servic1ng of luminaires at shoulder height. The pole and pole foundations were analyzed for the new 11ght fixtures and found to be able to withstand wind loading of 120 MPH as required by South Florida Building Code, Section 2306.
Based on the preceeding, the following conclusions can be made:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the UFSAR Amendment 4 is not increased, since the modification in outdoor security lighting systems enhances the illumination levels. The added lighting load does not increase the present UPSAR Amendment 4 load block value of 125KW for the Emergency Diesel Generator loading.
(ii) As a result of this modification there is no poss1bility for an accident or malfunction of a different type than any previously evaluated because the modification simply replaces the existing lighting fixtures, the nominal load specified in UPSAR Amendment 4 Table 8.3-2 has not been exceeded and no additional changes are required. The existing poles along w1th the replacement poles and pole foundations on which these new fixtures are mounted, have been analyzed for the added lighting fixture load and found to be able to withstand the 120 mph wind loads as required by the South Plorida Building Code Section 2306.
(iii) This modification does not reduce the margin of safety as def1ned in the bases for any Technical Specif1cation. In
" fact, this modification will increase the margin of safety since it will enhance the existing lighting system by providing additional lumens.
The implementation of this PC/M does not require a change to the plant technical specification.
The foregoing constitutes, per 10 CFR 50.59 (b), the written safety evaluation which provides the bases that th1s change does not involve an unreviewed safety question and prior Commission approval for the implementation of this PC/M is not required.
CM 135-286 age 1 of 2 RCP SEAL INJECTlON ENTERPERENCE ABSTRACT This Engineering Package (EP) provides design for the installation of flanges to permit removal of the RCP seal infection piping inside the RCP motor support structure. The supports/restraints for this piping have also been redesigned to be removable. Piping'nd supports/restraints are uniquely identified to facilitate their reinstallation.
Removable piping and supports/restraints have been provided to eliminate piping interference with the seal removal trolley and to minimize hazard to personnel performing work on the RCP seal cartridge inside the RCP motor support structure.
This EP is classified as nuclear safety'related since it modifies safety related piping. The safety evaluation has shown that this EP does not constitute any unreviewed safety questions, nor does it require a technical specification change. Therefore, prior NRC approval is not required for implementation of this PCM.
This EP has no adverse impact on nuclear plant safety and operation.
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PCM 135-'286 page 2 of 2 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 modification installs flanges to provide removal of the RCP seal in)ection piping inside the RCP motor support structure. The associated supports/restraints are also removable. Removability of piping and supports/restraints is provided to eliminate piping interference with the seal removal trolley and to minimize hazard to personnel performing work on the RCP seal cartridges inside the RCP motor support structure.
The modification included in this Engineering Package is considered to be safety related and does not involve an unreviewed safety question because:
(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 the addition of flanges in the RCP seal infection system are designed and stress analyzed in accordance with the applicable design codes and regulatory requirements for Safety Class 2 systems.
(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis is not created since no changes have been made to the operational design of the RCP seal infection system.
(iii) This modification does not change the margin of safety as defined in the bases for any technical specification since the installation of flanges is a piping change which is designed
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to the same codes used for original plant construction.
The operation of the RCP Seal in)ection system is neither required for plant safe shutdown nor for mitigating consequences of an accident. The system is installed to protect plant investment The implementation of this PCM does not require a change to the plant Technical Specification.
The foregoing constitutes, per 10CFR50.59(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 PCS is not required.
CM 003-287 gelof 2 BECHMAN WASTE CAS SYS OXYGEN ANALYZER REPL ABSTRACT In order to increase the effectiveness of the oxygen analyzers for continuous monitoring of the oxygen levels within the Waste Decay Tanks as required by the plant technical specifications, the existing oxygen analyzers will be replaced with more reliable analyzers. The replacement analyzers provide an additional advantage in that the sensing element is suitable for sampling oxygen in either a liquid or gaseous sample environment.
The inherent design features of the replacement analyzers will include the design and operational criteria for sample monitoring and installation in potentially hazardous locations, therefore, this design shall be considered as Quality Related.
The implementation of this PCM will have no impact on plant safety or plant operation.
A review of the changes to be implemented by this PCM was performed against the requirements of 10CFR50.59. As indicated in Section 3.0 of this Engineering Package (EP), this PCM does not involve an unreviewed safety question, nor does it require a revision to the technical specification; therefore, prior Commission approval is not required for implementation of this PCM.
SAFETY EVALUATION PCM 003-2b I page 2 of 2 With respect to Tit 0 of the Code of Federal Regu ions, 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 if in the safety analysis report may be created, or (iii) 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 since the Oxygen Analyzers are used for frequent monitoring of oxygen concentrations in the waste decay tanks and as described in PSL-2 FSAR Subsection 11.3.2 this system's function is not essential for the safety of the plant. The replacement of Oxygen Analyzers will provide control improvements to maintain the Waste Gas Analysis System functional with significant reduction in system maintenance and component replacements. Since this equipment is required by the technical specification to continuously monitor oxygen concentrations in the Waste Gas Decay Tanks, this EP is considered to be Quality Related.
ii) The possibility of an accident or malfunction of a different type other than any evaluated previously in the safety analysis report is not created since:
a This installation is in accordance with the Code of Federal Regulation 10 CFR 50.49 and no impact is incurred by this installation.
b The new equipment mountings and added components have been analyzed in accordance with the specification for the Design Fabrication and Erection of Structural Steel for Building, and it has been determined that the stresses with the new equipment are less than the panel stresses with the original equipment.
c This installation is in accordance with the Code of Federal Regulation 10 CFR 50.49 and has been determined to have no impact on the Environmental Qualification criteria since the equipment does not monitor or mitigate the event causing the harsh environment.
d The GWMS Oxygen Analyzers are neither required for safe shutdown nor for mitigating the consequences of an accident.
iii) The margin of safety as defined Specifications is not affected in the bases for any Technical by this EP since the components involved in this modification are not included in the bases of any Technical Specification.
The implementation of this PC/M does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10 CFR50.59(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 PCM is not required.
CM 052-287 age 1 of 2 CONDENSATE POLISHER TIE-INS This Engineering Package (EP) is for the installation of the 24 inch tie-in piping and valves required for the future connection of the Condensate Polisher System (CPS) to the Unit 2 Condensate System. It also includes the installation of the b~ss flow control valve required for operating the CPS using Unit 2 condensate and the instailation of a connection to the Unit 2 condensate storage tank for providing the capability of using Unit 2 condensate for backwashing the condensate polishers.
This EP is classified non-safety related since the portions of the Condensate System and Condensate Storage Tank piping where this modification will be implemented do not perform any safety function.
The safety evaluation has determined that this EP does not constitute an unreviewed safety question and implementatioa of the EP does not require a change to the Plant Technical Specification. Therefore, prior NRC notification for implementing this EP is not required.
This EP has no impact on plant safety and operation.
SUPPLEMENT 1 This supplement is to include reinstallation of the bypass flow control valve and a change in location of chemical in)ection points.
results of the evaluation as stated in the safety evaluation The'riginal remain unchanged.
This supplement has no impact on plant safety and operation.
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PCM 052-281 page 2 of 2 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulation, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question; (1) aa accident or if the probability of occurrence or the consequences of malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or (11) possibility for aa accident or malfunction of a d1fferent type than aay if e evaluated previously in the safety analysis report may be created; or (iii)'if the margin of safety as defined ia the bases for any Technical Specification is reduced.
This Engineering Package (EP) ie for the installation of the 24 inch piping aad valves required for the future connection of the Condensate Polisher System (CPS) to the Unit 2 Condensate System. It also includes the installat1on of the by~ass flow control valve required for operating the CPS using Unit 2 condensate and the 1nstallation of a connection to the Unit 2 condensate tank for providing the capability of using Unit 2 condensate for backwashing the condensate polishers.
In addition, the chemical ingecti,on points have been relocated.
Neither the portions of the chemical infection system nor the portions of the .Condensate System, Condensate Storage Tank piping and the CPS that this modification will be implementing perform any safety funct1on or interact with safety related equipment; therefore this package is classified as nonnuclear safety related.
Based on the above description, the modification included in this Engineering Package (EP) is considered to be aon-safety related. This EP does not involve an unreviewed safety. question, aad the following are bases for this gustificati.on!
The probability of occurrence or the consequences of aa accident or malfunction of equ1pment important to safety previously evaluated in the safety analysis report is not increased. The Condensate System and the CPS are not used in any safety analysis for accidents or malfunction of equipment and ae such are aoa-safety related and will have no effect on equipment vital to plant safety.
(11) 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 beea made to the operational design of the system.
(iii) The margin of safety as defined in the bases for any Technical Specif1cation 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 th1s PCM does aot require a change to the plant Technical Specificatione.
The foregoing constitutes, per 10CFR 50.59 (b), the written safety evaluation which provided the bases that this change does not involve an unreviewed safety. question aad prior Commission approval for the implemeatatioa of the PCM is not required.
M 057-287 age 1 of 2 EDG LOCKOUT RELAY MOD This Engineering Package (EP) modifies circuits and components in the Diesel Generator 2A and 2B Control Panels to provide improvements as follows:
- 1) Install tripping diode to allow the trip circuit failure relay to drop out in the event of an open circuit in the lockout relay coil.
- 2) Add a Diesel Generator breaker contact to, and modify wiring associated with, the voltage balance relay to eliminate potential damage to the relay resulting from contact chattering.
This EP is classified as Nuclear SafetyWelated since it provides for modification to Nuclear SafetyWelated Class lE equipment. The Safety Evaluation has shown that the implementation of this EP does not constitute an unreviewed safety question nor would implementation affect Plant Technical Specifications. Thus, Commission approval is not required prior to implementation.
This EP has no impact on plant safety or operation.
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PCH 057-287 page 2 of 2 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change sha11 be deemed to involve an unreviewed safety question: (1) if the probability of occurrence or the consequences of an accident or malfunctI.on of equipment important to safety prev1ously evaluated in the Safety Analysis Report may be increased; or (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report may.be created; or (111) if the margin of safety as defined In the basis for any technIcal specification is reduced.
The modificat1ons included in this Engineering Package do not involve an unreviewed safety question because:
- 1) The probability of occurrence or the consequences of an accident or malfunct1on of equipment important to safety previously evaluated are not increased s1nce th1s EP provides for increased reliability of the Emergency Diesel Generator trip circuit failure alarm indication through improved lockout relay circuit design, and for increased reliability of the Voltage Balance relay'by reduction of contact wear during lockout cixcuit trip testing.
- 11) There is no possibility for an accident or malfunction of a different type than any previously evaluated. This EP does not, modify the intended operation or test requirements of the lockout circuit. The addition of the diode does not 1ntroduce the potential for new accidents because the diode is in a circuit location in which no anticipated mode of diode failure will affect lockout'relay operation under conditions when the Diesel Generators are required to shut down the plant or to mitigate the consequences of an accident. The addition of a set of breaker contacts and the wiring modification to the voltage balance xelay do not alter its intended operation or create a new failuxe mode.
111) This modification does not change the margin of safety as def1ned in the basis for any Technical Specification, since no anticipated mode of diode failure will affect lockout relay operation under conditions when the diesel generators are required to shut down the plant or to mitigate the consquences of an accident.
The implementation of th1s change does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR50.59(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 implementati,on of this PCM is not required.
~CM 060-287
~age 1 of 3 BORIC ACID PLOW XMTR REPLACEMENT This Engineering Package involves .the replacement of boric acid makeup flow element.FE-2210Y and transmitter FT-2210Y. The existing Controlotron model 241N ultrasonic instrument has proven to be insufficiently accurate to allow the plant operators to control boric acid makeup flow rates to yield the desired Chemical and Volume Control System (CVCS) concentrations.
Furthermore, the Controlotron ultrasonic instrument has been the source of continuous maintenance problems which are compounded by the unavailability of detailed manufacturer's drawings. Additionally, the fact that the Controlotron flow element is strapped onto the outside of the makeup line, has resulted in misalignment which, yields invalid data.
The replacement instrument is a Fischer & Porter Series 10D1477. This device consists of an in-line, magnetic type flow element and a remotely mounted transmitter which is similar to the instrumentation used for this function at St Lucie Unit 1. The new flow element and transmitter will enable the plant operatoxs to accurately control the CVCS boric acid concentration and will eliminate the maintenance difficulties.
The fact that the new flow element is an in-line device requires a section of the boric acid makeup line to be modified. These modifications will be conducted on a quality group "D" section of pipe downstream of valve FCV-2210Y- The modifications included are: 1) Removal of the heat tracing on this section of pipe. 2) Removal of a section of the pipe. 3) installation of flanges to interface with the new flow element. The removal of the heat tracing is to be coordinated with the implementation of PCM 283-288, "Boric Acid'oncentration Reduction." PCM 283-288 provides for the deenergization of the boric acid system heat tracing which must be implemented before the heat tracing removal required by this PCM can be implemented.
Implementation of this PCM is contingent on the implementation of PCM 283-288, "Boric Acid Concentration Reduction".
Although boric acid makeup flow element FE-2210Y and transmitter FT-2210Y axe classified as Non"Nuclear Safety Related, the safety classification of this PCM is Quality Related since the piping modified to allow installation of the in-line flow element must be seismically analyzed and since this instrumentation assists in controlling the reactivity of the core. The safety evaluation of this EP has determined that this PCM does not constitute an unreviewed safety question and does not xequire'a change in the Plant Technical Specifications. This PCM can be implemented without prior NRC approval.
This PCM has no impact on plant safety or operation.
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PCH 060-2S7 page 2 of 3 SAPETY 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'reviously evaluated in the Safety Analysis Report may be increased; or (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be cx'eated; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.
This EP documents the replacement of flow element and transmitter FE-2210Y and FT-2210Y, respectively. The eristing equipment has been shown to be inaccuxate and has also been a continuous maintenance problem. The existing flow element is a ultrasonic type device which is stx'apped onto the surface of a qua1.ity group "B" section of the box'ic acid make-up line The new flow element is a magnetic type device which is installed in-line. To reduce the impact. on the piping, the new element is being installed downstream of PCV-2210Y in a quality group "D" section of the makeup line. This section of the make-up line does not serve any safety function.
This package is classified as guality Related because the piping which is being modified to interface with the boric acid flow element must be seismically re-analyzed and because this equipment assists in controlling reactivity of the core.
This EP has no effect on the design basis as provided in St Lucie Unit 2 PSAR Section 9.3. However, St Lucie Unit 2 PSAR Subsection 9.3.4 and Pigure 9.3-5b have been modified to reflect the new location and instrument type of the replacement flow element.
The modifications included in this Engineexing Package do not involve an unreviewed safety question because:
(i) The probability of occurxence or the consequences of an accident or malfunction of equipment important to safety pxeviously evaluated is not increased since the equipment and piping being modified'neither directly nor indirectly serve any function required to mitigate the effects of an accident or to bring the plant to safe shutdown.
PCM 060-287 page 3 of 3 (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 operational design of any control
.circuits or associated systems which are important to safety.
(iii) This modification does not change the margin of safety as defined in the basis for any technical specification since the new flow e1ement and transmitter perform a nonnuclear safety related function. Further, neither the flow element and transmitter being replaced nor the flow path containing the section of pipe being modified are included in the bases of any technical specification Due to the fact that the EP does not involve any cables, essential to safe reactor shutdown or systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 10CFR50 Appendix "R" fire protection requirements. Therefore, the proposed design of this package is in compliance with the applicable codes and St Lucie Unit 2 FSAR requirements for fire protection equipment.
Implementation of this ~lity Related PCM does not require a change to the Plant Technical Specifications.
Implementation of this PCM is contingent upon the implementation of heat tracing deenergization for the affected piping under PCM 283-288, "Boric Acid Concentration Reduction". Implementation of PCM 283-288 constitutes a HOLD POINT RELEASE for this EP.
The foregoing .constitutes, per 10CPR50.59 (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.
M 077-287 page 1 of 3 ERDADS/SAS UPGRADE This Engineering Package (EP) provides for modifications in Control Room equipment to upgrade the Emergency Response Data Acquisition and Display System (ERDADS), which is also known as the Safety Assessment System (SAS) and includes Safety Parameter Display System (SPDS) equipment. This EP will improve the performance and d1splay capabilities of the exist1ng system and will include new display CRTs and keyboards and a new color hardcopier.
The Engineering Package is classif1ed as Quality Related since the SAS system is a computer based data processing and display system which assists Control Room personnel in evaluating the safety status of the plant and since the modifications in the Control Room involve inst'.ation of equipment in RTGB-204. Implementation of this PCM does not involve an unrev1ewed safety question or a change to .the Plant Technical Specifications. It can be implemented without prior Commission approval.
Implementation of the PCM will not affect the safety or operation of the plant.
SUPPLEMENT l This EP revision provides for modifications in the Control Room in preparation for implementing an, upgrade to the ERDADS/SAS equipment Included in this work are installation of conduit and cable, relocation of existing ERDADS/SAS equipment, and 1nstallation of mounting hardware to allow future installation of ERDADS/SAS equipment.
The Engineering Package is classif1ed as Quality Related s1nce SAS is a computer based data processing and display system which assists Control Room personnel in evaluating the safety status of the plant. Implementation of this PCM does not involve an unreviewed safety question or a change to the Plant Technical Specifications. It can be implemented without prior Commission approval.
Implementation of the PCM will not affect the safety or operation of the plant.
SUPPLEMENT 2 This EP revision modifies the design to replace the CRTs, video generators, and supporting components which were originally specified 1n Supplement 0 and Supplement l due to hardware incompatibility problems.
remains the same The overall design The Engineering Package is classified as Quality Related since ERDADS/SAS is a computer based data processing and display system which assists Control Room personnel in evaluating the safety status of the plant. Implementation of this PCM does not involve an unreviewed safety question or a change to the Plant Technical Specifications. It can be implemented without prior Commission approval.
Implementation of the PCM will not affect the safety. or operation of the plant.
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PCM 077-287 page 2 of 3 ABSTRACT (Continued)
SUPPLEMENT 3 This EP revS. sion modifies the design to add an input to the SAS fox valve FCV-25-36, Continuous Hydrogen Purge Containment Isolation Valve and incorporate the results of the Human Factors Engineering review performed on the Safety Parameter Display System (SPDS) and the Non-SPDS portion of the Emexgency Response Data Acquisition and Display System (ERDADS).
With the exception of the modifications to provide a new ERDADS input from valve FCV-25-36, all work to be conducted under this EP is considered Quality Related. The modifications required to provide the ERDADS input from FCV-25-36 are considered Nuclear Safety Related since the limit switch used to originate the new 'input is powered by a safety related source. While this modification involves safety related equipment, the new ERDADS input signal is not considered safety related since ERDADS/SAS assists the operatoxs in evaluating the safety status of the plant but serves.no Nuclear Safety Related function. Furthermore, isolation is provided between the safety related power supply and ERDADS/SAS via SAS Isolation Cabinet 5 (SB). Since this Revision involves safety related equipment, the safety classification of this EP has been upgraded to Nuclear Safety Related. As a result, the original Safety Evaluation has been revised. The Safety Evaluation still concludes, however, that implementation of this PCM does not involve an unreviewed safety question or a change to the Plant Technical Specifications. It can be implemented without prior Commission approval.
Implementation of the PCM will not affect the safety or operation of the plant.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a pxoposed 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 incxeased- or (ii) if the 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 is xeduced.
defined in the basis for any technical specification The modifications included in this Engineering Package do not involve an unreviewed safety question because:
i) The probability'f occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated are not increased si'nce the existing input isolation of the ERDADS/SAS equipment will not be modified and will maintain the same level of protection for safety-related equipment.
PCM 077-?87 page 3 of 3 Also, an evaluation of the modifications required to provide the valve position input to ERDADS/SAS from FCV-25-36, Continuous Hydrogen Purge Containment Isolation Valve, concludes that the existing input isolation of the ERDADS/SAS equipment will protect the valve control circuitry in the event of ERDADS/SAS fai1ure and that neither the operational design nor the function of the valve control circuitry has been changed. Thus, these modifications will not increase the probability of occurrence or the consequences of an accident ox malfunction of equipment important to safety previously evaluated.
There is no possibility for an accident or malfunction of a different type than any previously evaluated since no new safety related functions are cxeated by this EP and since no changes have been made to the operational design of any control circuits or associated systems Furthermore, the ERDADS/SAS system is,provided with input isolation devices to prevent ERDADS/SAS failure from affecting safety related components and systems This modification does not change the margin of safety as defined in the basis for any Technical Specification, since no equipment instaLLed or modified by this EP affects any paxameter referenced in 'the Technical Specifications.
All work to be conducted under this Engineering Package, with the exception of the modifications associated with valve FCV-25-36, is considered quality related. However, this Engineering Package is classified as Nuclear Safety Related since it involves modification to safety related cixcuitry associated'with FCV-25-36, Continuous Hydrogen Purge Containment Isolation Valve, to provide an input signal to ERDADS/SAS-The Human Factors Engineering evaluating of the SPDS portion of the ERDADS system found seventy four (74) HEDs. All four (4) Priority 1 HEDs have been corrected. Therefore, the HEDs found through this Human Factors Engineering xeview do not affect plant safety.
This EP has no effect on cables or components necessary for safe shutdown of the plant, or on equipment on the EssentiaL Equipment List. Changes to equipment and structures involving 10CFR50 Appendiz "R" fire protection xequirements hive been addxessed. (See Attachment 7-1) Thus, the proposed design is in compliance with applicable
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requirements for fire protection.
The implementation of this change does not xequire a change to the Plant Technical Specifications.
The foregoing constitutes, per LOCFR50.59(b), the written safety evaluation which provides the bases that this changes does not involve an unreviewed safety question and prior Commission approval for the implementatg~gf this PCH is not required.
CM 129-287 age 1 of 3 SIT & CONT PAN COOLER INSTRUMENT UPGRADE ABSTRACT This Engineerfag Package addressee level, temperature, and flov fastrumeatatfon upgrades for the Safety Infection Tanks (SIT), Componeat Cooling Water System and Containment Pan Coolers.
The Safety Ingectfon Taake are part of the Safety Infection System vhich automatically discharges borated water'into the Reactor Coolant System on depreeeurieatioa of RCS which occurs as a result of a Lose of Coolant Accident (LOCA). The level iaetrumentatfon being upgraded measures the Safety Infection Tank vater level and provides indicatioa at the RTGB.
The Contaiameat Pan Coolers sre paxt of the Containment Cooling System which provides the means of Containmeat heat removal during normal'perations aad subsequent to a LOCA. The tempexature detecting elements (thermocouples) located at the inlet and outlet of the Containment Pan Coolers, used to meaeuxe the duct air temperature are being upgxaded. The flov instrumentatioa which detects lov Component Cooling Water flov through the Containmeat Pan Coolexs and provides local indication and remote annunciation 1s being credited as part of the Contaiameat Heat Remova1 System to satisfy USNRC Reg Guide 1.97.
These instruments currently are designated as Non-Nuclear Safety Related. This effort vill upgrade selected fnstrumentatfon, associated electrical circuit loops and structural support to Nuclear Safety Related, fn accordance with USNRC Regulatory Guide 1.97, Rev 3, Category 2, Type D Variable.
Th1s guide deffnee a method acceptable to the NRC for complying with the regulations to provide instrumentation to monitor plant variables and systems duriag and following sn accideat. Specifically, Type D vaxiables are defined as those variables that provide 1nformstion to indicate the operation of fndividual safety systems and other systems important to safety. Category 2 pxovides for design and qualification criteria whfch is less strfagent thea Category 1 in that it does aot include seismic qualification, redundancy or contfauoue displays and requfres only a high~eliabilfty povex source.
Based oa the usage of these fnstrumente to monitor safety related equipmeat, this EP is classified as Nuclear Safety Related.
The safety evaluation of this package has shown that the implementation of this PCM does aot constitute an unreviewed safety question and prior NRC appxoval for its implementation ie aot, requixed.
This EP has ao impact on plant safety and opexation or Plant Technical Specifications.
Su lemeat 1 Revision 1 to th1s Engineering Package provides for the xelease of hold points regarding work to be performed inside the Containment Building. Eng1neering end design of electrical conduit and box supports necessary to complete the implementation of this PCM is detailed herein-The original Safety Evaluation and Technical Specifications are not affected, amended or changed as a result of this supplemeat.
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PCM 129-287, page 2 of 3 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, ox (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 modification is for the upgrade of the Safety Infection System and Containment Cooling System instrumentation in accordance with USNRC Regulatory Guide 1.97, Rev 3, Category 2, Type D Variables.
This modification upgrade will provide more reliable and qualified instrumentation loop to detect and monitor Safety In5ection System and Containment Fan Heat Removal System operation during and following an accident. No modification is required fox the Component Cooling Water flow indicating switches, since these switches are-Clsss IE devices and qualified to 'meet the above mentioned cxiteria Hence, this EP is considered Nuclear Safety Related. Since this modification replaces existing monitoring instrumentation with qus1ified devices or qualifies the 'existing instrumentation and involves no other modifications to safety related equipment, the degree of protection provided to nucleax safety related equipment is unchanged. The probability of malfunction of equipment important to safety previously evaluated in the FSAR remains unchanged. The consequences of malfunction of equipment important to safety previously evaluated in the FSAR are unchanged. The possibility of malfunctions of a different type than those analyzed in the FSAR is not created.
The modifications included in this Engineering Package do not involve an unreviewed safety question because of the following xeasons:
(i) The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously'valuated in the Safety Analysis Report will not be increased by this modification because the existing monitoring equipment availability, redundancy, capacity, or function requixed to mitigate the effects of an accident is not, affected and is in fact upgraded per USNRC Reg. Guide 1.97
PCM 129-2S7 page 3 of 3 (ii) The possibility for an accident or malfunction of a different type than any evaluated previously ia the Safety hnslyeis Report will aot be created by this modificatioa because replaciag and/or qualifying the monitoring instrumentation is in accordance with the criteria of USNRC Regulatory Guide I 91. In addition, this new equipment ie seismically and environmentally qualified to withstand the normal aad accident conditions anticipated in the areas that they are iastalled.
(iii) The margin of safety as defined in the bases for any technical specification is aot reduced since this modification addresses flow indicatiag switches and installs qualified thermocouples and level transmitters which will enhance the moaitoriag of the Coatainment Heat Removal System and the Safety In)ection System duriag aad followiag an accideat and does not alter the bases of the technical specificatioas associated. with these systems.
The implementation of this EP does not require a chaage to the Plant Technical Specificatioas, nor does question.
it create an unreviewed safety The foregoing constitutes, per 10CPR50.59(b), the written safety evaluation which provides the bases that this chaage does aot involve an uareviewed safety questioa and prior HRC approval for the implementation of this PCM is not required.
M 140-287 age 1 of 2 CONDENSATE POLISHER CROSS-TIE ABSTRACT The Condansata Polfeher fs currently used during Vndt 1 etartup for cleaning the fluid in the condensate system as part of the steam generator protection program. This modification in con)unction with PCMs 052-287, 053-987 and 054-187 will provide the necessary tie-ins so the U'n'it 1 Condensate Polisher may be utilized for either the Unit 1 or Unit 2 Condensate System.
This fourth modification will provide the electrical tie-ins and instrumentation and controls modifications to all'ow operation of the Condensate Polisher for Unit 2 Service.
This EP is classified +mitty Related because, although the Condensate System 'and Condensate Polisher affected by this modification do not perform any safety function, this modification - adds seismically supported conduit in the RAB and modifies a seismically qualified panel in the Unit 2 Control Room (PACB-2). The safety evaluation has determined that this EP does not constitute an unreviewed safety question and implementation of the EP does not require a change to the Plant Technical Specification. Therefore, prior NRC notification for implementing this EP is not required.
This EP has no impact on plant safety and operation'AFETY EVALUATION With respect to Title 10 of the Code of Federal ReguLation, Part 50.59, a proposed change shaLl be deemed to involve an unreviewed safety question; (i) if the probability of occurrence or the conse-quences 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 createdp or (iii) if the margin of safety as defined in the bases for any Technical Specification is reduced.
This modification wiU. provide the controls instrumentation, the electrical tie-ins and instrumentation and controls modifications to allow operaton of the Condensate Polisher for Unit 2 Service We Unit 2 condensate bypass valve will automatically be driven fully open upon any one of'he following, thereby preventing a unit trip-
- 1) loss of Condensate Polisher Panel power;
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- 2) misalignment of the Unit 1 and Unit 2 input/output isolation valves or;
- 3) high inlet/outlet differential pressure across the Condensate Polisher;
- 4) improper unit selector switch alignment;
- 5) loss of ac power and or control signal to the Unit 2 Bypass Valve Control Cabinet..
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PCM 140-287 page 2 of 2 diss, snnunedstoOeill ldght/sound dn both the Polisher 'Contxol Cabinet.
t 2 Control Roon and on the Condensate Cloeiag of the Unit 2 Condensate Polisher bypass valve wiU. be achieved through manual control only. Furthermore, an interlock is provided to immediately disconnect the closing circuitry from the bypass valve and drive the valve fully open should one or more of the above stated conditions occur.
The portions of the Condensate System that this modification will be affecting do not perform any safety function or interact with safety relate'd equipment; however, since this EP also involves the addition of controls, indication and alarms to the Plant Auxiliary Control Board (PACB-2), which ie a seismically qualified control panel in the Control Room, and installs seismically supported conduit in the RAB>
it is classified anality Related.
The new components that will be added to PACB-2 have been designed and will be instaU.ed to the same requirements as existing components ia the Coatrol Room This addition of components to 'with PACB-2 have been reviewed and considered acceptable and in compliance the seismic requirements applicable to the Control Room. Requirements for Electxical separation of Class 1E circuits from Non Class 1E circuits have been maintained Based on the above description, the modification included in this EP is considered to be quality related. This EP does not involve an unreviewed safety question, and the following are bases for this Justification:
s i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety hnalysis Report is not iacxeaeed. The portions of the condensate system, where this modification will be implemented, and the Condensate Polisher System are not used in any safety analysis for accidents or malfunctioa of equipment and as such are non-safety related and will have no effect on equipment vital to plant safety The additioa of the contxols, indications and alarms to the PACB-2 have beea reviewed aad considered to be acceptable and in compliance with the seismic xequirements applicable to the Control Room. ELectrical separation requirements have been met.
- ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report ie 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 system iii) The maxgin of safety ae defined in the bases for any Technical
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Specificatioa 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 Specificatione.
The foregoing constitutee, per 10CFR50.59(b), the written safety evaluation which provided 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.
M 154-287 ge 1 of 3 RPS EXCORE SAFETY CHANNEL MODIFICATION This engineering package covers modifications to the Excore Linear Power Safety Channels at St. Lucie Unit 2. The major feature of this package is the modificatidn of two of the feedback loops on the linear power subchannel inputs with feedback loops of higher gain. This modification is necessary to compensate for the lower values of leakage flux at the excore detectors that resulted from the current St. Lucie Unit 2 fuel management program. Two new gain ranges are being incorporated to give plant personnel the option of selecting an amplifier gain that is most appropriate for use with the exhibited leakage flux values.
Because the excore linear power safety channels are safety related, all work covered by this engineering package is classified as Nuclear Safety Related.
Based on a 10CFR50.59 safety evaluation this modification does not affect plant safety or operation, nor does it involve any unreviewed safety question or require changes .to the plant technical specifications. As such, prior NRC approval is not required to implement this engineering package.
10 CFR 50.59 allows changes to a facility if an unreviewed safety questions does not exist and if a change to the Technical Specification is not required.
This change can be performed under 10 CFR 50.59 since it does not effect the technical specifications and since it meets the applicable criteria as defined below:
- 1. The change described herein does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
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Pdi~i l~4-2v/
page 2 of 3 This engineering pack ge has considered the safety related consequences of the modifications proposed to the excore linear power safety drawers. Because the excore linear power safety channels are safety related, all work covered by this engineering package is classified as Nuclear Safety Related.
All oF the modifications implemented by this package are confined to the linear .amp and sumer cards (U51-3) which are located inside the excore saFety channel drawers. The changes consist of modifying two of the three first stage feedback loops on the two linear subchannels,to accommodate the actual, and anticipated future flux levels as seen by the detectors. Section l0 describes the modification in detail.
There are no changes to any 'other equipment interfacing with the excore linear power safety channels, and specifically, there are no changes, with the exception of the aforementioned changes to the Reactor Protection System.
These changes were reviewed to determine the impact on the existing seismic and environmental qualification with no negative findings.
Based on the above, the modification is confined only to the excore drawer, and 'has no impact on existing analysis or design basis, the existing FSAR Failure Effects and Modes Analysis was reviewed and it was determined that the changes would have no negative impact on the system. Therefore, the probability of occurrence or the consequences of an accident of malfunction of equipment important to safety has not increased.
The change does not create the possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report.
There is no increase in the subchannels uncertainties as a result of this change, therefore the changes do not impact any of the current settings associated with the following:
PCM 154-287 page 3 of 3 Var 5ab'le High Power Level 20 Thermal Power Gener ation 30 High Low Power Density Thermal Margin/Low Pressure Zero Power Hade Bypass The changes have not resulted in any new functional circuitry being added to the equipment, therefore the procedures for performing maintenance and calibration for the drawer are essentially identical, and there is no impact on personnel performing maintenance on this equipment.
Based on the above, the change has no effect on existing setpoints, and system operation, and therefore, does not create the possibility for a malfunction of a different type than evaluated previously.
The change does not reduce the margin of safety as defined in the basis for any technical specification.
As stated above, this modification will not incrase the subchannel uncertainties, therefore, there is no impact on existing systems or setpoints.
The change does not result in an increase in the surveillance requirements for the excore linear power safety channels. In addition, operational parameters are not impacted by the changes of the Engineering Package, therefore, no ch'ange to the technical specifications are required.
Since the subchannel uncertainties and surveillance requirements are unchanged this change does not reduce the margin of safety as defined in the basis for any technical specification.
In conclusion, the change proposed in this Engineering Package is acceptable from the standpoint of nuclear safety as it does not involve an unreviewed safety question and does not change the Technical Specifications'. Therefore, prior NRC approval is not required to implement this procedure.
CM 068-288 age I of 2 POWER FEED TO MH261 SUMP PUMP This Engineering Package covers the modifications required to introduce a second, redundant fault current interrupting device (isolation device), into the power feed circuit of the sump pump located in MH 261, to meet the requirements of Regulatory Guide 1.75.
This Engineering Package will provide the engineeering and design details required to install a second isolation device (the existing cixcuit breaker will be the first device), consisting of fuses and their fuseholder, mounted in a box ad)scent to PP-203.
The sump pump is required to operate during a flood condition to drain the manhole. Although the sump pump is non-safety xelated, it is fed fxom a Class 1E power panel. Therefore, this package is classified as Nuclear Safety Related.
This EP does not constitute an unreviewed safety question since the modifications described above were reviewed in accordance with 10CFR50.59 and were determined to have no adverse impact on plant operations or safety related equipment. The implementation of this PCM does not require a change to the Plant Technical Specifications. Prior NRC approval for the implementation of this PCM is not required.
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PCM 068;288 SAFETY EVALUATEO page 2 of 2 With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a pxoposed 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 x'eport may be created; or (iii) if the margin of safety as .defined in the bases for any Technical Specification is reduced.
This Engineering Package provides the engineering and design details required to install a second, redundant fault current interx'upting device (isolation device), into the powex'eed circuit of the sump pump located in MH 261, tb meet the requirements of Regulatory Guide 1.75.
Although the sump pump is non-safety x'elated, it is fed from a Class 1E power panel. A non>>safety load which is considered important for operation and is connected to a Class 1E power. source, must be provided with two, Class 1E, fault current interrupting devices. The addition of fuses '(and fuseholdexs) to the power circuit of the sump pump, along with the existing circuit breaker, meets the requirements of Regulatory Guide 1.75. The fuseholder, and the box it is mounted in, will be seismically installed.
The implementation of this EP incxeases the availability of the Class lE electric system by preventing a malfunction in one section of a circuit (non-safety) from causing .an unacceptable influence in another section of the circuit (safety)<<'his EP has been classified as Nucleax'afety Rel'ated.
Based on the preceding, the following conclusions can be made-(i)'he pxobability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased. The installation of a second, redundant fault current interrupting device in the power feed circuit of MH 261 sump pump will prevent a malfunction in one section of the circuit (non-safety) from affecting another section of the circuit (safety).
(ii) As a result of this modification, thex'e is no possibility for an accident or malfunction of a different type than any previously evaluated. Although this modification does affect safety related equipment, there is no introduction of any new failure mode for the equipment.
(iii) This modification does not reduce the margin of safety as defined in the bases for any Technical Specification. The proposed design pxovides the modifications required to maintain the independence of electrical circuits The margin of safety provided by the Technical Specifications is preserved.
The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides the bases that this change does not involve a change to the Plant Technical Specifications or an unreviewed safety question and prior NRC approval for the implementation of this PCM is not required.
GM 090-288 age 1 of 3 PRESSURIZER NOZZLE REPLACEMENT Abstract This Engineering Package (EP) provides for the replacement of one (1) St.
Lucie Unit 2 Pressurizer Lower Level Instrument Nozzle (CE Pc. No. 684-107).
The Level Instrument Nozzle is made of a specific heat of material that is known to be susceptible to Intergranular Stress Corrosion Cracking (IGSCC).
The purpose of this'replacement is to preclude the chance of a future nozzle failur e.
This replacement is classified Nuclear Safety Related since it involves the of the Reactor reactor coolant pressure boundary and components which are part Protection System (RPS). This EP does not have any adverse impact on plant safety and operation.
A review of the changes to be implemented by this PCN was performed against the requirements of 10CFR50.59. As indicated in the Safety Evaluation (Sec-tion 3.0), this PN does not involve an unreviewed safety question, nor does it require a revision to the Plant Technical Specifications. This modification will have no effect on plant safety or operation. Prior Coanission approval is not required for the implementation of this PCN.
This EP revision incorporates minor drawing revisions, provides instructions for retermination of pressurizer heater cables, and incorporates minor corrections in the .word document.
These revisions do not require revision to the plant technical specifications, nor does it effect the Safety Evaluation provided in Section 3.0 of this EP. Therefore, pursuant to 10CFR 50.59 this modification can be implemented without prior commission approval.
These revisions do not have any adverse impact on plant safety and operation. \
This supplement incorporates revised drawings issued by Combustion Engineering for this prospect. The drawing revisions consisted of a change to one end of the sleeves (beveled to flat) due to changes in the seal weld configuration, the addition of'a drawing for the. instrument line restrictor.orifice, and other minor corrections.
This supplement does not require revision to the plant technical specifications, it it does not effect the Safety Evaluation provided in Section 3.0 of this EP and Therefore, does not have any adverse impact on plant safety or operation.
pursuant to 10CFR 50.59 this modification can be implemented without prior coaeission approval.
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0 PCM 090-288 page 2 of 3 Saf et Evaluati on Mith 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: (1) if the probability of occurrence or the conse-quences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report may be 1ncreased, or (11) if the possibility for an accident or malfunction of a different type than any evaluated previously 1n the safety analpeis report may be. created, or (111) 1f the margin of safety as defined in the bas1s for any technical spec1fication 1s reduced.
The modifications included in this Engineering Package do not involve an unreviewed safety question because:
(1) The probabil1ty of occurrence or the consequences of an accident or'malfunction of equipment important to safety previously evaluated is not increased s1nce the replacement pressurizer legal instrument nozzle meets the requirements of the FSAR. The replacement instrument nozzle does not alter any other equipment or components.
(11) There is no possibil1ty for an accident or malfunction of a different type than any previously evaluated since no changes have been made to the operational design of any other compo-nents or equipment.
Installation of the Pressurizer Level Instrument Nozzle is controlled by CE procedures. Meld1ng of the weld buildup pads, seal welds and partial penetration weld is to be in accordance with CE developed weld procedures. The Instrument Nozzle has been sub)ected to non-destructive examination (NDE) per the codes listed in Sections 2.3.5 of this EP. By adhering to these codes and standards 1n the 1mplementation of this PCN, there 1s no possibility for an accident or malfunction different than any previously evaluated 1nvolving the Primary Coolant Pressure Boundary. The replacement Instrument Nozzle has approximately the same weight as that be1ng replaced. Therefore the insignificant change in weight does not have any affect on the pressurizer support re-straints.
(111) This replacement/modification does change the design analysis for the pressurizer. This modification is analyzed in the Addenda Report No. CENC 1831 titled Addendum No. 2 to the Analytical Report for Florida Power and Light Company, St.
Lucie Unit 2, pressurizer. The nozzle penetration hole s1ze through the pressurizer was not increased to allow for the sleeve 1'nstallatfon.
The design limitations of the reactor coolant pressure boundary, as delineated in FSAR 5.1, are maintained with the implementation of this PCN.
v ~ Yrv V page 3'of 3 Since this EP affects equipment that is identified as Nuclear Safety Related (the Pressurizer Level Instrument Nozzle is ASME Class 1),
this package is considered Nuclear Safety Related.
A pressure test is required after this replacement modification, per ASME Section XI, Paragraph IMA-4400. A system pressure leak test will be performed to ascertain that the implementation of this PCN has met the requirement of no allowable leakage of reactor coolant.
This'EP has no effect on cables essential to safe reactor shutdown and alternate shutdown components. There are no other changes to equipment which involves 10CRF50 Appendix "R" fire protection (see .1). Thus, the proposed design of this package is in compliance with the applicable codes and FSAR requirements for fire protection equipment.
Implementation of this PCM does not require a change to the Plant Technical Specifications and may be implemented without prior Comnis-sion approval.
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 PCN is not required.
CX 091-288 ge j. of 3 RCS HOT LEG NOZZLES REPLACEMENT Abstract This Engineering Package (EP) provides for the replacement of five (5) St.
Lucie Unit 2 Reactor Coolant System (RCS) hot leg Resistance Temperature Detection (RTO) Instrument Nozzles (CE Pc. No. 772-104-R). The RTO Instrument Nozzles are made of a specific heat of material that is known to be susceptible to Intergranular Stress Corrosion Cracking (IGSCC). The purpose of this replacement is to preclude the chance of a future nozzle failure.
This replacemint is classified Nuclear Safety Related since it involves the reactor coolant pr essur ~ boundary and components which are part of the Reactor Protection .System (RPS). This EP does not have any adverse impact on plant safety and operation.
A review of the changes to be implemented by this PCN was performed against the requirements of 10CFR50.59. As indicated in the Safety Evaluation (Section 3.0), this PCN does not involve an unreviewed safety question, nor does it require a revision to the Plant Technical Specifications. This modification will have no effect on plant safety or operation. Prior Commission approval is not required for the implementation of this PCN.
This EP revision provides instructions for removal, reinstallation, and start-up testing for the RTD's affected by this EP, incorporates minor drawing changes and minor corrections in the word document.
These revisions do not require revision to the plant technical specifications, nor does it effect the Safety Evaluation provided in Section 3.0 of this EP. Therefore, pursuant to 10CFR 50.59, this modification can be implemented. without prior coaefssion approval.
These revisions do not have any adverse impact on plant safety and operation.
This supplement incorporates a revised drawing issued by Combustion Engineering for this pro5ect. The drawing revision consisted of a change to one end of the sleeves (beveled to flat) due to changes in the seal weld configuration.
This supplement does not require revision to the plant technical specifications, it.does not effect the Safety Evaluation provided in Section 3.0 of this EP and it does not have any adverse impact on plant safety or operation. Therefore, without prior pursuant to 10CFR 50.59 this modification can be implemented commission approval.
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PCM 091-288 page 2 of 3 Safet Eva 1uat1 on With respect to Title 10 of the Code of Federal Regulations, Part 5D.59, a proposed change shall be deemed to involve an unreviewed safety question: '(1) 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 (11) if the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analys1s report may be created, or (111) if the margin of safety as defined 1n the bas1s for any technical specification is reduced.
The modifications 1ncluded in this Engineer1ng Package do not involve an unr eviewed safety question because:
(1) The probability of occurrence or the consequences of an acc1dent or malfunction of equipment important to safety previously evaluated is not 1ncreased since the replacement RTD instrument nozzles meet the requirements of the FSAR.
The replacement RTD instrument nozzles do.not alter any other equipment or components.
(11)'here is no poss1bility for an accident or malfunction of a different type than any previously evaluated since no changes have been made to the operational des1gn of any other components or equipment.
Installation of the RTD Instrument Nozzles 1s controlled by CE procedures. Meld1ng of the weld buildup pads, seal welds and partial penetration weld is to be in accordance with CE developed weld procedures. The RTD/Instrument Nozzles have been subjected to non-destructive examination (NDE) per the codes listed in Section 2;3.5 of this EP., By .adher1ng to these codes and standards in the 1mplementation of this PCM, there is 'no possibility for an accident or malfunct1on different than any previously evaluated 1nvolving the Primary Coolant Pressure Boundary.'he replacement RTD/
Instrument Nozzles have approximately the same weight as those being replaced. Therefore the insign1ficant change in weight does not have any affect on the pipe support
'estrainti.
(111) 'This repl.acement/modification does change the design analysis for the RCS hot legs. This modificat1on 1s analyzed in the Addenda Report No. CENC 1832 titled Addendum No. 3 to the Analytical Report for Florida Power and Light Company, St.
Lucie Unit 2, Piping. The nozzle penetration holy size through the piping was increased to allow for the sleeve 1nstallat)on.
The design limitations of the reactor coolant pressure boundary, as delineated in FSAR 5.1, are ma1ntained with the implementation of this PCN.
C
0 PCM 091-288 page 3 of 3 Since.this EP affects equipment that is identified as Nuclear Safety Related (the RTD Instrument Nozzles are ASNE Class I), this package ls considered Nuclear Safety Related.
A press~re test ls required after the installation of the RTD, per ASIDE Section XI, Paragraph IMA-4400. An ln-service leak test will be performed to ascertain that the implementation of this PCN has met the requirement of no allowable leakage of reacto~ coolant.
The only effect this EP has on cables essential to safe reactor shutdown and alternate shutdown components is ln the disconnection of the existing RTDs and reconnection after completion of the installation of the replacement RTD Instrument Nozzles and RTD/
thermowell assemblies. There are no other changes to equipment which involves IOCRF50 Appendix "R" fire protection (see Attachment Y.I).
Thus, the proposed design of this package ls ln compliance with the applicable codes and FSAR requirements for fire protection equipment.
Implementation of this PCH does not require a change to the Plant Technical Specificatloos and may be implemented without prior Comalsslon approval.
The foregoing constitutes, per IOCFR 50.59 (b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question and prior Comnisslon approval for the implementation of this PCM iij not required.
144-288 age 1 of 4 480V PZR HTR XPK 2A3 6 2B3 REPL This Engineering Package provides for the replacement of the existing PCB filled pressurizer heater transformers with equivalent transformers of dry type construction the removal of the concrete curbs surrounding the transformers and the removal of an existIng wall between 480V Pressurizer Heater Transformers 2A3 and 2B3 for access purposes. The electrical equipment located on the wall will be removed and relocated on an exist1ng wall and a nearby column. The. conduit and cable associated with thIs equipment will be reworked and/or replaced as required.
The existing high energy missile barriers which surround the transformers will not be required for uee with the replacement transformers.
The transformers have 6, l/20 HP low energy fans installed under the core and coil assembly. These low energy devices could not generate a missile which could penetrate the 14 gage sheet steel transformer enclosure.
A concrete wall located between the transformers must be removed to permit transformer 2A3 to be removed. The waU. has electrical equipment which must be relocated and the conduit and cable associated with it must be reworked or replaced, as required.
Pressurizer heater transformers 2A3 and 2B3 perform nonnuclear safety related functIons. Because of their importance Ia plant operations they are fed from safety related busee, 4160V 2A3 and 2B3. This package will also relocate componeats that are located on a concrete wall that separate the pressurizer heater transformers. One of the componeats, Isolation Box 2952, is safety related, therefore, this engineering package is classified as safety related Results of the safety evaluation conclude that modifications presented by this Eagiaeeriag Package do not coastitute an unreviewed safety question, .do not require any changes to the Plant Technical SpecIfications and do not require prior NRC approva1 for the implementation of this PCM.
The Implementation of this PCM will aot have an adverse impact on plant safety or operations'
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PCS 144-288 page 2 of 4 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 diffegent 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 safety analysis focuses on the replacement of the pressurizer heater transformers and the removal of the high energy missile barrier surrounding pressurizer heater transformers 2A3 and 2B3. Transformers 2A3 and 2B3 are non-safety related devices which provide power to non-safety related 480V Switchgar Buses 2A3 and 2B3. The transformers are supplied electrical power from 4160V safety related buses 2A3 and 2B3. The transformers are located at elevation 43'n the cable spreading area in the RAB which contains safety related equipment and they are seismically mounted. This engineering package is classified nuclear safety related because an isolation box containing lE cables must be relocated. Transformers 2A3 and 2B3 are 750 KVA, 4160 to 480V, 3 phase PCB liquid filled. The missile barriers were utilized to satisfy Appendix 1, Section 5.1.3 to Regulatory Guide 1.75 as endorsed by IEEE 384-1981, Section 6.1.3.1.(1) where it is stated that cable spreading areas should not contain high~nergy equipment such as transformers. This is a condition aimed at liquid filled transformers which under certain fault conditions can build up very high pressure inside the transformer tank containing the liquid cooling/insulating medium. A study was made at the NRC's request in which it was postulated that 12 missiles could be generated by a transformer rupture. Therefore, missile barriers were designed which surrounded each of the transformers and were sufficient to prevent perforation by missiles The barrier satisfied the design bases by shielding the equipment in the room from possible missiles. This was in accordance with FPL commitment per letter L-82%79 dated October 29, 1982 to the NRC. The NRC stipulated that the FPL commitment was acceptable per NUREG 0843 Supplement 3 dated April 1983.
FPL submitted their design philosophy for the missile barrier to the NRC by letter L-84-44 dated February 27, 1984. The design was approved by the NRC The missile barrier exists because of a condition of licensing required to meet Reg. Guide 1.75. The condition of license was satisfied by erection of the missile barriers. The NRC has since removed this condition from the FPL Operating License, since FPL complied with the NRC.
At the present time, FPL is replacing the PCB liquid filled Pressurizer Heater Transformers. The dry type replacement transformers do not have any means for containing gases produced by any non~echanistic failure conditions. The high energy fault is dissipated in the form of heat
PCM 144-288 page 3 of 4 which in an extreme case will cause a meltdown of the transformer core and coil. However, the transformer enclosure is vented to obtain air movement for transformer cooling, therefore, there is no way to produce a pressure buildup which can generate a missile. Furthermore, the air cooled transformer does not have any appurtenances which could become missiles such as valves, handhole covers, gauges, etc, which are standard equipment for liquid cooled transformers. The dry type transformers will have 6 small (1/20 HP) fans inside the enclosure which will be individually cage enclosed. These are low energy devices which could not generate a missile which would penetrate its own wire guard and the transformer metal enc1osuxe. Consequently, there is no need for a missile barrier and the barrier's removal is compatible with the intent of the design bases.
The pressurizer heater transformers 2A3 and 2B3 are located on elevation 43 in the Unit 2 cable spreading area. These non-safety transformers are powexed from Class 1E 4.16 KV switchgear through qualified isolation devices.
Dry type transformers are not tank enclosed. The coils are silicone resin vacuum pressure impregnated to insure excellent protection against the tropical climate high humidity, and salt laden air, which is the design environment for this plant. The coils for each phase are mounted on the individual steel cores, which are then mounted together on a structural frame in an enclosure. The most probable failure mechanism is a phase to phase fault whexe the coil connections are made. Pailure like that would result in copper vaporization with copper splatter egected in the vicinity of the fault. The sheet metal of the enclosure is more than adequate to contain the hot splatter.
There is no possibility for pressure build up, with resultant generation of a missile as is the case with transformers whose cooling/insulating medium is a liquid.
The Nomex Conductor insulation is Class H rated for 220 Deg. C and highly fir'e resistant. The transformer coils are coated with eight mils of silicone sealant which is also Class H, 220 Deg. C tested to 250 Deg. C. If the fault generated sufficient heat in the conductor to cause the Nomex and the silicone resin to burn it would be unrealistic to assume that a temperature could be rapidly built up to a level which would cause the base metals (copper and steel) of the transformer to explode.
J The transformer protection would operate within a relay trip time instantaneous setting, long befoxe any catastrophic temperature could be reached. There is no evidence that a mi'ssile could be generated which would pierce the sheet metal housing of the transformer.
In the event of a turn to turn fault occuring there is a possibility that the arc which formed at the point of fault would cause the copper conductor to melt and splatter. This copper splatter would easily be contained by the 14 gage sheet metal transformer enclosure. Circuit breaker operation would remove the fault current stopping the copper melt.
PCizi 144-288 page 4 of 4 The equipment vendor has informed us that in fifty (50) years of manufacturing and operating dry type transformers they have never encountered a situation where a failed transformer caused a missile to be generated. Furthermore, during the transformer design development progxam the manufacturer's laboratory tested the transformers. The transformers where exposed to total destruction failure modes. We vere verbally advised that during these teste there never vas a missile geaexated. Furthermore, Attachment 7.10 presents vendors position on the subject.
Based on the above, the high energy missile barriers surrounding the pressurizer heater transformers may be removed because they serve no useful function with the dry type transformers. The standard manufacturer's enclosure vill contain any poss1ble products of an internal short circuit and this satisfies the Design Bases.
Based on the preceding, the folloviag conclusions can be made:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR vill not be incxeased because the existiag transformers ax'e being replaced on a one-formac basis by txansformers that are equivalent in function, capacity and electrical characteristics. Removal of the missile barrier vill not have any safety related implication. This action will not 1ncxease the likelihood of generating any missiles and therefore we have met the design bases.
(11) This modification does not change the operation of the non-safety x'elated pressurizer heater txaneformex'e aad switchgear.
Therefore, per the design bases there is no possibility that an accident non~echanieti,c failure or malfunction of a different type than any evaluated in the FSAR may be cxeated.
(iii) The replacement pressurizer heater transformers are equivalent in purpose and function to the existing transformers and perform no safety xelated functions. Thex'efoxe, in accordance with the design bases this m'odification does not xeduce the margin of safety ae defined in the bases for any Technical Specification.
The foregoing constitutes per 10CFR50.59('b) the written safety evaluation which provides the bases that this change does not involve any change to the safety evaluation nor change to the technical specifications and therefore prior NRC approval for the implementation of this PCM is not required.
CM 147-288 age 1 of 2 RCP 2A1 SNUBBER MOD ABSTRACX Engineering Package (EP) provides the engineering and design information
'his required to eliminate the existing snubber for S/R Mark f RC-26-R10A and replace it with a rigid restraint. The subject snubber with FPL Tag 0 2-034, Ebasco Mark 0 RC-26-RlOA, is located on RCP 2Al Mid Seal Cavity pressure indication line 0 I-3/4-RC-174.
An investigation into the cause of the Reactor Coolant Pump oil fire at St Lucie 0 2 on Nov. 28, 1987, revealed 'that a portion of the RCP oil collection system had been disassembled to facilitate inspection of the subject snubber.
The corrective action recommended in the Licensee Event Repox't for this fire, LER 87-005 dated Dec. 15, 1987, includes relocation of the snubber, which was physically interfering with the RCP 2Al oil collection pan. In lieu of this recommendation this EP eliminates the existing interfex'ence with the oil collection pan by removing the subject snubber and installing a rigid restraint in its place. Future disassembly of the oil collection pan for snubber inspection would, therefore, not be required.
The modification made in this EP involves the xeactor coolant system piping which is Safety Class 1, Quality Group A, and Seismic Category I; and hence the supports/restraints for this piping are classified as 'Nuclear Safety Related and Seismic Category I.
The safety evaluation has shown that this EP does not constitute an unreviewed safety question and prior NRC approval is not required fox'mplementation.
The implementation of this EP does not require a change to the Technical Specification and does not reduce the margin of safety for any Technical Specification.
The implementation of the EP will have no impact on plant safety or operation SAFETY EVALUATION Mith respect to Title 10 of the 'Code of Federal Regulations, Part P<<posed change shall be deemed to involve an unreviewed safety question: (i) if the probability of occuxrence or the consequences of an accident or malfunction of safety previously evaluated in the Safety Analysis equipment important to increased'r (ii) of a if Report may be a possibility for an accident or malfunction different type than any evaluated previously in the Safety Analysis Report may be created; or (iii) if the maxgin of safety as defined in the basis for any Technical Specification is reduced.
Thy modification included in this EP is to Mark 8 RC-26-R10A to a rigid restraint. change the snubber at S/R located on Reactor Coolant System piping, onThethesub5ect snubber is RCP 2A1 Mid Seal Cavity Pressure indicator line 8 I-3/4"-RC-174.
Based on the above description, the modification Engineering Package is considered safety related. included in this This involve an unreviewed safety question, and the following are the EP does not bases for this conclusion:
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PCM 147-288 page 2 of 2 (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 because:
There is no modification performed to the piping and all the supports/restraints, except Mark I RC-26-RlOA, have remained functionally identical to their existing configurations.
- 2. The replacement of snubber Mark 0 RC-26-RlOA with a rigid restraint, which eliminates the possibility of snubber malfunction, has been evaluated to ensure that:
- a. Piping stresses for the modified condition have remained within the stress limits allowed in the ASME> Section XlI Code.
- b. The existing and modified support/restraints have been demonstrated to be adequate for revised stress analysis loads in accordance with the applicable codes..
(ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any evaluated previously in a safety analysis report because the modification has no adverse impact on the piping stresses and the associated piping supporting system.
(iii) defined This modification does not reduce the margin of safety as in the bases for any Technical Specification because it neither changes the design parameter of the RCS nor does it the RCS design flow or functional requirements.
change This modification does not affect the pressure boundry integrity of the piping for the RCS.
The implementation of this PCM does not require a change to the Plant Technical Specification.
The foregoing constitutes, per LOCFR 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.
CM 189-288 age 1 of 2 PROT COATINGS REP/REPL-RCB This engineering package covers the maintenance of Service Level 1 protective coatings on concrete and steel surfaces inside the Reactor Containment Building. The coatings perform no. safety function and do not affect plant safety or operation, however, this engineering package is classified as Quality Related since the failure of the coatings could have an adverse affect on the containment sump and thus affect the ECCS. This modification does not constitute an unreviewed safety question, nor does any changes to the Plant Technical Specifications.
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CM 268-288 page 1 of 2 APW MOV-09-13, 14 BREAKERS This Engineering Package involves the addition of conduit entrance seals at Auxiliary Feedwatex Pump 2A and 2B discharge cross-tie valves MV-09-13 and MV&9-14 and associated local control stations to allow the Motor Control center circuit breakers associated with these valve operators to be "racked-in" during all modes of plant opexation. These cixcuit breakers are currently "rackedmut" during operating modes 1 through 3 to ensuxe that spurious operation (opening) of these valves does not occur as a result of harsh environment conditions associated with a Main Steam Line Break (MSLB)o This package includes an evaluation of the temperatures of MOV and local control station components, within the sealed enclosures, during. a MSLB to establish the envixonment experienced by these components as mild.
0 In addition, the heater circuits fox the MOV limit switch enclosures ax'e disconnected and spared as part of this package. The Design Analysis of the package (paragraph 2.1.1) includes reference to Limitorque documentation stating~hat energization of these heaters is not required for valve operation.
These..valves provide redundant cross-tie isolation between the discharge pipi'ng of motor driven Auxiliary Feedwater Pumps 2A and 2B. These valves axe normally closed and may be opened via manual action at the local control stations to provide Auxiliary Feedwater from one of the pumps to either or both Steam Generator(s) during fill operations following refueling.
The control and power circuits for MV-09-13 and MV-09-14 operators are Class lE and the conduit entxance seals added by this EP are to be included in the 10CFR50.49 'Equipment Qualification list For this reason, this package ~is classified as Nuclear Safety Related.
This EP does not constitute an unreviewed safety question since the modifications described wexe reviewed in accordance with 10CFR50.59 and were determined to have no adverse impact on plant operations or safety related equipment The implementation of this PC/M does not requixe a change to the Plant Technical Specifications.
change does not involve an unreviewed safety question and prior NRC approval for the implementation of this PC/M is not required.
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PCM 268-288 page 2 of 2 SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shaQ. be deemed to involve an unxeviewed safety question: (i) if the probability of occurx'ence 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 the possibility for an previously in accident or malfunction of a different type than any evaluated the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced.
This EP provides for the addition of conduit entrance seals for Auxiliary Feedwater System motor operated valves MV-09-13 and MV&9-14 and associated local control stations. These valves provide redundant isolation between the discharge lines of motor driven Auxiliary Feedwater pumps 2A and 2B. The addition of seals for this equipment, which is located in the Steam Trestle ares on the 19.5'levation, will allow the cixcuit breakexs for these valves to bh "racked-in" during plant operation. The circuit breakers are currently "racked out" during plant operation to prevent concurrent spurious operation (open'ihg) of both valves during a Design Basis Event requix'ing operation of the Auxilsry Feedwater System. The modifications included in this EP do not change the function or operation of the affected valves.
Since this modification involves the energization of safety related (Class 1E) circuits and the addition of environmentally qualified conduit entrance seals to be included in the 10CFR50.49 list, this modification is classified as Nuclear Safety Related. However, it does 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 Spurious opex'ation (opening) of either valve does not impact:the cause or consequences of any existing failure mode.
The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis ri port is not increased. These valves provide redundant isolation between AFW txains Spurious operation (opening) of either valve does not incxease the possibility of an accident ox malfunction on that train and, due to the valve redundancy, does not impact the operation of the opposite AFW train iii) The margin of safety, as defined in the bases for any Technical Specification, is not affected by this PCM since the components
'nvolved in this modification are not included in the bases of any Technical Specifications.
The implementation of this change does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR 50.49 (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.
CM 274-288 page 1 of 2 FWRV MODIFICATION St Lucie Unit 2 has two feedwater regulating valves (FCV-9011 and FCV-9021) controlling the amount of water fed from the feedwater pumps to the steam generators. These feedwater regulating valves are air operated contxol valves which need to remain in their last position (fail as-is) upon loss of air supply. This prevents undesired closure of the feedwater regulating valves. This modification removes pressure switches and solenoid valves used to isolate the regulating valve on loss of air supply A pneumatically actuated trip valve will be installed to isolate the regulating valve.
The equipment involved with this modification is nonnuclear safety related. However, based on the consequences of feedwater flow perturbations on the plant and the resulting unnecessary challenges to nuclear safety related systems, this Engineering Package is classified as Qality Related.
The safety evaluation of this package has shown that the implementation of this PQi does not constitute an unreviewed safety question and prior Commission approval is not required for implementation. This PCM has no impact on plant safety and operation or the Plant Technical Specifications.
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PCM 274-288 page 2 of 2 SAFETY EVALUATION With respect'o Title 10 of the Code of Federal Regulations, Part 50.59, a proposed change shall be deemed to involve an unreviewed safety question:
accident (i) or if the probability malfunction of of occurrence equipment important or the coxzequences of to safety previously an evaluated in the Safety Analysis Report may be incxeased, (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 equipment involved in this modification is non<<nuclear safety related as described in FSAR Section 7.7.1.1.4. However, considering the consequences of feedwater flow fluctuations and the resulting unnecessary challenges to safety related systems, this package is classified as Quality Related- These valves are not required for isolation of feedwater in the event of a main steam line break. These valves sole purpose is to regulate feedwater flow to the steam generators during opexation above 15 percent power.
This modification does not involve any equipment listed on the St Lucie Unit 2 Essential Equipment List, Drawing No. 2998-B&49, Rev. 0 and has no effect on safe reactor shutdown or alternate shutdown procedures.
10CFR50 Appendix "R" Fire Prote'ction requirements are affected by the temporary removal of a cable tray fire stop for cable pulling. This package requires the restoration of this cable tray fire stop. See Attachment 7.1, Section 2.1.2, and Section 9.3.
The modifications included in this Engineering Package do not involve an unreviewed safety question because of the following reasons:
(i) The probability of occurrence and the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report will not be increased by this modification because the installation of the backup air supply and the new trip valve does not affect the availability, capability, or function of any equipment required for the mitI.glutton of a design bases accident (DBA). The feedwater regulating valves have no .safety related function and are not required by any design bases for the mitigation of a DBE.
(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report will not be cx'eated by this modification because the components involved in this modification have no safety related function.
(iii) The margin of safety as defined in the bases for any Technical Specification is not reduced because the feedwater regulating valves do not form the basis of any Technical Specifications.
The implementation of this PCM does not require a change to the Plant Technical Speci'fications.
The foregoing constitutes, per 10CFR50.59 (b), the written safety evaluation which provides the basis that this change does net'nvo-':e an unreviewed safety question and prior Commission approval for the implementation of this PCM is not required.
CM 275-288 age 1 of 2 MOV THERMAL OVERLOAD MOD Motor .Operated Valve (MOV) MV-08-3 is the Auxiliary Feedwater (AFW) pump 2C Turbine Trip & Throttle Valve. Presently, MV-08-3 is provided w1th a manually operated thermal overload protection (TOL) bypass selector switch which bypasses the thermal overload protective trip devices and'onsequent MOV trip during AFW System automatic start. The TOLs are bypassed during normal plant operation and reinstated during MOV testing. With the TOL bypass selector switch in the Maintenance/Test mode, annunciation is provided in the Control Room in the event of TOL actuation. However, the TOL bypass selector switch when placed 1n the Normal/Bypass mode, 1nhibits the TOL trip annunciation in the Control Room-This Engineering Package (EP) includes engineering and design necessary to modify the circuit for MV-08-3, AFW Turbine Trip 6 Throttle Valve, such that Thermal Overload (TOL) trip annunciation will be provided in the control room regardless of the position of the TOL bypass selector switch; also, the annunciator window engraving will be modified to indicate TOL actuation.
This package is classified as Nuclear Safety Related since it involves modifications of control circuitry to the AFW System which is required for safe shutdown and performs a Nuclear Safety Related function. The modifications are of a type which do not affect the function, availability or capab1lity of the Nuclear Safety Related AFW System. The safety evaluation
~ has determined that this EP does not constitute an unreviewed safety question and does not require a change in the Plant Technical Specif1cation. Th1s PCM can be implemented without prior NRC approval. Therefore, this PCM has no impact on plant safety or operation.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part to involve an unreviewed 50.59, a proposed change shall be deemed safety question: (1) if the probability of occurrence or the to consequences of an accident or malfunction of equipment important safety previously evaluated in the Safety Analysis Report may be increased; or (ii) if the possibility for an acc1dent or malfunction of a different type than any evaluated previously inofthe Safety Analysis Report may be created; or (111) if the margin safety as defined in the basis for any technical specification is reduced.
(EP) do not The modifications included in this Engineering Package involve an unreviewed safety question because:
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0 PCM 275-288 page 2 of 2 (i) 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.
This modification is designed to provide Control Room operators with annunciation of a degraded condition of MV&8-3. This circuit modification does not modify the thermal overload
'ypass requirements as required by Regulatory Guide 1.106.
The modification to the circuitry for the AFR Turbine Trip Throttle Valve MV-08-3 does not affect or modify the function of the valve nor the availability, redundancy, capability, or function of any equipment required for the mitigation of a design bases event (DBE) or safe shutdown.
(ii) The possibility for an accident or malfunction of a different type than any evaluated previously in the St Lucie Unit 2 Final Safety Analysis Report are not increased by this modification because this circuit modification does not alter the functions of MV-08-3 nor the Thermal Overload Protection design as required by RG 1.106. Annunciator G-46 has been modified to provide additional information in the Control Room. Furthermore, the reliability of the annunciator G-46 has not been compromised by this change.
(iii) The margin of safety as defined in the bases of any technical specification is not reduced since it has no negative effects The availability and on safety related components or systems.
operation of the motor operated valve MV-08-3 as originally specified in the St Lucie Unit 2 Final Safety Analsis Report and Plant Technical Specifications is not being reduced.
Since this EP affects equipment that is identified as Nuclear Safety Related, it is considered Nuclear Safety Related.
Due to the fact that the EP does not involve any fire protection systems, fire rated assemblies or systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 10CFR50 Appendix "R" fire protection requirements. Therefore, the proposed design of this package is in compliance with the applicable codes and St Lucie Unit 2 FSAR requirements for fire protection equipment.
Implementation of this PCM does not require any change to the Plant Technical Specifications, nor does it create an unreviewed safety question.
The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety question, nor does implementation of Nuclear Safety Related PCM 275-288 require a change to Plant Technical Specifications. Therefore, prior NRC approval for the implementation of this P(M is not required
PCM 281-288 page 1 of 2 RX CAVITY SEAL RING MATER FILL MOD ABSTRACT An Engineezing Package is required to modify the Reactor Cavity Seal Ring (seal ring) so that it can be filled with water during reactor refueling. The water-filled ring girder will be more effective as a shield against radiation exposure. The water will be removed from the seal ring and discharged to the floor drains of the Reactor Building at the conclusion of the refueling operation.
The'odifications included in this EP include providing pipe plugs on each section of the Reactor Cavity Seal Ring toroid to permit filling and draining of water and providing 2 inch diameter holes through the seal plate to access the plugs. Any required structural modifications to the seal zing, its supports and lifting lugs are also 'included.
The Reactor Cavity Seal Ring is nonnuclear safety related and non-seismic, based on the St Lucie Unit 2 FSAR and the fact that its use is for refueling operations to allow filling the Reactor Cavity with water. This EP is classified Quality Related to obtain added assurance that the Reactor Cavity Seal Ring will not adversely affect the Reactor Vessel, since the seal ring is lifted over and installed over the open Reactor Vessel.
This modification does not involve an unreviewed safety question, has no effect on plant safety and operation, and does not involve a change to any plant Technical Specification, based on a 10CPR50.59 review. Therefore, prior NRC approval is not required for implementation of this Engineering Package.
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SAFETY EVALUATION
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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 the possibility for an accident or malfunction PCM 281-288 page 2 of 2 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 basis for any technical specification is reduced.
The modifications included in this EP are limited to the Reactor Cavity Seal Ring which is non-safety-related. The modifications consist of adding penetrations and pipe plugs to the seal ring to allo~ it to be filled while with the water.
seal This ring is water in provides place. The additional water will radiation shielding be removed at the conclusion of the outage. The seal ring lifting lugs are modified to meet the requirements of Subsection 3.8.4 of the St Lucie Unit 2 FSAR since the lifted load has increased. No safety-related equipment is affected. This modification is classified as Quality Related to obtain additional control during the engineering, design and installation to assure that the Reactor Cavity Seal Ring will perform its intended function and will not adversely affect the reactor vessel. Based on the above, the following provides the justification that an unreviewed safety question does not exist:
i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. The modifications included in this EP have been reviewed and it has been determined that they will not adversely affect safety-related equipment. Since the Reactor Cavity Seal Ring is not considered by the FSAR in determining the probability of accidents, possible types of accidents, or in the evaluation of consequences of accidents, it can be concluded that the probability of occurrence or the consequences of accidents previously addressed in the FSAR remain unchanged.
ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated. The seal ring will be filled with water only during outages when the plant is shut down. The structural members affected by this modification have been analyzed for the increased load and have been found to be acceptable. The sealing portion of the seal ring has not cha:~ed. Therefore, the possibility of an accident of a different type has not been created.
iii) This modification does not change the'argin of safety &s defined in the basis for any Technical Specification. The installation of the seal ring is not specifically addressed in the Technical Specifications. Furthermore, the sealing portion of the seal ring has not been changed and structural effects have been reviewed and found to be acceptable.
The implementation of this change does not require a change to the Plant Technical Specifications.
The foregoing constitutes, per 10CFR50.59(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.
CM 283-288 page 1 of 3 BORIC ACID CONCENTRATION REDUCTION This Engineering Package provides aa evaluation of the effect of reducing the boric acid concentration and temperature tn tbe Boric Acid Makeup System on the associated system tnstrumcntattoa. In addition, the demaergtstng of portions of the heat traciag system associated with the Boric Acid Makeup System will be addressed. These cvaluatioas include the suitability of the present taetrumeatatton, the ef fcct on tnstrumeatatioa quali ficatioa documentation, aad address changes tn the calibrat toa procedures where required Because portions of the Boric Acid Makeup System are safety related, all work covered by thts Engiaeeriag Package (EP) is classified as nuclear safety related.
Based on a 10CFR50.59 safety evaluation, the modifications identified ia this package do not affect pleat safety or operation, do not iavolve any unrevtewed safety queetioas, nor do they involve additional changes to Plant Technical Spcciftcattone. NRC approval is not required prtor to implementation of this Engineering Package.
SUPPLEMENT 1 Supplcmcat 1 to thts,PCM provides detailed informattoa on the modifications to the Boric Acid Makeup Tank Temperature Indicat tag Controllers, tbe installatioa of an RAB air temperature indicator, aad the de~rgtsatton of the beat tracing. The Essential Equipment List, ts being revised to reflect the d~nergisation of the tank heaters and beat tracing Additional Enviroaneatal ~liftcatton Documentation Packages associated with other PCMs in progress have beiag revised The modifications included in this supplement have beea addressed ia the revised Safety Evaluation.
SUPPLEMENT 2 Supplement 2 to tht,s PCM removes the boid potat for implementation as dispositioned via CRN-283-288-1071 ~ This hold point wae based on the proposed change to the Tcchnical Specifications. This change hae been approved by the NRC aad the hold point ie ao longer applicable. Tbe Safety Evaluation bss been revised to remove the bold point.
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PCM 283-288 page 2 of 3 SAF EXP EVALUATION This Engineering Package has considered the safety related consequences of reducing the boric acid concentration and temperature of the Boric Acid Makeup System on the instrumentation within the Boric Acid Makeup and Iodine Removal Systems. De~nergixing portions of the heat tracing systems associated with the Boric Acid Makeup System was also considered. Because portions of the Boric Acid Makeup System are safety related, all work covered by this Engineering Package is classified as Nuclear Safety Related.
With respect to Title 10 of the Code of Federal Regulations, 10 CFR 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 revisions included in this Engineering Package do not involve any unreviewed safety questions because of the following reasons:
The change described herein does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report The de-energized heat tracing is no longer required for system operation due to the lower temperature requirements. The de-energization of the heat tracing will reduce the loading of the emergency diesel generators during an accident. The Boric Acid Makeup Tank level instrumentation calibration has no impact on the instruments'unction, but is done solely to maintain instrument accuracy. Utilisation of the Boric Acid Makeup Tank Hester Controllers is not essential due to the reduction of the boric acid minimum temperature. The addition of a local air temperature indicator is done as a convenience, and will have no impact on the plant All other instrumentation within the Boric Acid Makeup and Iodine Removal Systems remains unchanged and does not require calibration or hardware changes. This change does. not affect the design requirements of any other instruments 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.
0 PCiaf 283 288 page 3 of 3 ii. The change does not create the possibility for an accident or malfunction of a differeat type than evaluated previously in the safety aaalyeis repoxt. The de-energieed beat txacing fs no longer required fox system opexation due to tbe lowex'empexature requirements The Wmnergieatioa of the beat tracing will reduce the loadfng of the emergency diesel generators during an accident. The Boric Acid Makeup Tank level instrumentation calibration has no impact oa the iastrumeats'unction, but . is done solely to maintain instrument accuracy. Utflfration of the Boric Acid Makeup Teak heater controllers is not essential due to the reduction of the boric acid minimum temperature. The addition of a local air temperature fadicator fs done as a convenience. These chaages do not create the possibility for an accident ox malfunction of a different type than previously evaluated.
iii. This modification does not, reduce the margin of safety as defined fa the bases for any technical epecificatioa. The instrument changes identified in this package will be done in order to conveniently verify the max'gine as defined in the revision to the Techaical Speciffcatione. The requfrement for heat tracing has been xemoved from the proposed Technical Specifications; therefore, the demnergisation of beat traciag bas no impact on the Technical Specifications.
The implemeatetion of this Eagineexiag Package does aot require any additioaal changes to the Plant Technical Specificatioas other than those pxovided duxiag the Borfc Acid Concentration Reduction Effort nor does ft create an unxeviewed safety question.
The foregoing constitutee, pex 10CPR50.59(b), tbe written safety evaluation which provides the bases that thfs change does not involve aa unrevfewed safety question, aad prior NRC approval fox'he implementatioa of this PCM is not required The implementation of this 'PCM does aot require a change to the Plant Techaical Specificatioas.
PCM 284-288 page 1 of 2 INST CHANGE HUMAN FACTORS CONCERNS This Engineering Package (EP) .includes engineering and design details necessary to implement instrumentation changes to resolve four (4) outstanding Human Engineering Discrepancy Reports (HEDs) against the St Lucie Unit 2 (PSL-2) control panels. This EP also includes modifications to provide consistency between St Lucie Unit 1 and 2 control panels. The control panels affected by this modification are the Control Room (remote) Hydx'ogen Analyzer Panels 2A h 2B and RTG Boards 201, 202 and 204 The four (4). outstanding HEDs are addressed by the three (3) modifications which constitute the "Removal of Unused Instruments" category. These modifications are (1) the removal of a pushbutton from RTGB-204, (2) the removal of five control element assembly group deviation lights and rearrangement of the remaining seven deviation lights on RTGB-204 and (3) the removal of the vibration and eccentricity meters and the four power drawers which had provided input to the meters from RTGB-201 The remaining modifications are being implemented to provide increased consistency between the Unit 1 6 2 control panels. The Addition of Status Indication" category involves the installation of local/remote control indi'cating lights on Control Room .(remote) Hydxogen Analyzer Panels 2A & 2B.
The "Device Layout Changes" category includes two modifications. One of these modification involves rearranging the legend lights for the diesel generator loading sequence to match the actual order of loading/actuation. The other modification in this category involves the addition of two nameplates and demarcation to separate and identify the reheater block valves and the warm up valves associated with low pxessuxe turbines 2A and 2B The last category "Labeling Changes" involves one modification which consists of the replacement of the nameplates fox the steam generator 2A and 2B atmospheric dump isolation valve control switches. This replacement is to provide a more complete description of the function of the valve.
This package is classified modifications to the Control as Room Nuclear Safety Related since it involves (remote) Hydrogen Analyzer Panels 2A and 2B and the RTG Boards which are classified Nuclear Safety Related. The modifications are of a type which do not affect the function, availability or capability of the Nuclear -Safety Related instrumentation. The safety evaluation has determined that this EP does not constitute an unreviewed safety question and does not require a change in the Plant Technical Specifications. This PCM can be implemented without prior NRC approval-This EP has no impact on plant safety or operation.
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page 2 of 2 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: (1) if the probability of occurrence or the consequences of an accident or malfunct1on of equipment important to safety previously evaluated in the Safety Analysis Report may be 1ncxeased; or (ii) if the possibility for an accident ox malfunction of a different type than any evaluated pxev1ously 1n the Safety Analysis Repoxt may be created; or if (iii) the marg1n of safety as defined in the basis for any technical specification is reduced.
The modifications included in this Engineering Package (EP) do not involve an unreviewed safety question because:
(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not 1ncreased by th1s modification because 1t does not affect or change the availabQ.ity, redundancy, capacity, or function of any equipment required to mitigate the effects of an
'accident.
There is no possibility for an accident or malfunction of a d1fferent type than any px'eviously evaluated since no changes have been made to the operational des1gn of any control circu1ts or associated systems.
(111) The margin of safety as defined in the bases of any technical specification is not reduced since the instrumentation changes provided by this package involve only the removal of unused devices, the rearrangement and 1nstallation of indicating lights',
and the addition and modification of nameplates and demarcation.
The Safety Related circuits which were modified have been analyzed, and it has been determined that there is no effect on the purpose, function or operation of the control circuits.
Since this EP affects equipment that is identified as Nuclear Safety Related, it is considered Nuclear Safety Related.
Due to the fact that the EP does not 1nvolve any fixe protection systems, fire rated assemblies or systems associated with achieving and maintaining safe shutdown conditions, this package has no impact on 10CFR50 Appendix "R" fire protection requirements. Therefore, the proposed design of this package is in compliance with the applicable codes and St Lucie Unit 2 FSAR requirements for fire protection equ1pment.
The foregoing constitutes, per 10CFR 50.59 (b), the wxitten safety evaluat1on which provides the bases that this change does not involve an unreviewed safety question nor does implementation of Nuclear Safety Related PCM 284-288 require a change to Plant Technical Specifications.
Thex'efore, prior NRC approval fox the 1mplementation of th1s PCM is not required.
M 297-288 age 1 of 2 EXIM OP SM THERM SNUBBER ABSTRACT This Engineering Package provides the engineering and design information for the deletion of ten (10) snubbers.
The six (6) snubbers for S/R Mark Nos. B-2501-6B (which has two snubbers],
CH-l-R2, CH-1-R5, CH-l-R8, and SI-2419-50A have been eliminated by replacing them with rigid struts or rigid restraints. The applicable stress analysis calculations (B-2501, CH-2085, and SI-2419) have been reanalyzed to include" the change from a snubber to a rigid restraint and to demonstrate that the piping stresses are within code allowables. The revised stress analysis is acceptable.
The four (4) snubbers for S/R Mark Nos. MS-32-R3B, MS-32-R4B, MS-33-R3B, and MS-33-R4B have been replaced with antenna-type rigid restraints without performing any reanalysis of stress calculations since the rigid restraints have been designed to be attached to the 34 inch diameter main run of the pipe. The new restraints provide an additional Y-restraint function, which consequently renders the eight (8) dead~eight spring supports, Mark Nos.
MS-32-Rl, MS-32-R2, MS-32-R3A, MS-32W4A and MS-33-R1, MS-33-R2, MS-33-R3A, MS-33-R4A unnecessary and thus they have been removed.
The affected supports/restraints have been reviewed and qualified for the revised piping loads in accordance with applicable codes and criteria.
The elimination of the 10 snubbers and their consequent removal from the Inservice Inspection Program will contribute significantly to the reduction of personnel time spent in Radiation Controlled Areas.
The modifications made in this EP involve the supports/restraints on Nuclear Safety Related piping, which is Safety Class 2 and 3 (ality Group B and C) and Seismic Category I. Hence this package which modified the supports/
restraints for the piping is classified Nuclear Safety Related and Seismic Category I.
The implementation of this EP does not require a change to the Technical Specification and does not reduce the margin of safety for .any Technical Specification.
The .safety evaluation has.shown that this EP does not constitute an unreviewed safety question and therefore prior NRC approval is not required for the implementation of this EP.
The implementatf vn of the EP will have no impact on plant safety or operation.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulations, Part 50.59, a roposed change shall be deemed to i volv en unr risque~
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 in the Safety of a different type than any evaluated previously Analysis Report may be created; or (iii) if the margin is of safety as defined in the basis for any Technical Specification reduced.
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PCM 297-,288 page 2 of 2 The modifications included in this EP are to replace the snubbers at.
S/R Mark Nos. B-2501-6B fwhich has two snubbersj, CH-l-R2, CH"1-R5, CH-l-R8, SI-2419-50A, MS-32-R3B, MS-32-R4B, MS-33-R3B, and MS-33-R4B with rigid restraints. Eight existing spring hangers for Mark Nos.
MS<<32-Rl, MS-32"R2, MS-32-R3A, MS-32-R4A and MS-33-Rl, MS-33-R2, MS-33-R3A, and MS-33-R4A have been deleted as the redesign of MS-32-Rl and MS-33-Rl provided this Y-function. The sub5ect snubbers are located on the Blowdown, Chemical 6 Volume Control, Containment Spray, Safety Infection, and Main Steam systems.
Since the modifications included in this Engineering Package involve supports/restraints for safety related piping systems, the package is classified as safety related. This EP does not involve an unreviewed safety question, and the following are the bases for this conclusion:
(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 because the replacement of the snubbers for supports/restraints with rigid restraints, which eliminates the possibility of snubber malfunction, has been evaluated to ensure that:
- l. Piping stresses for the modified condition have remained within the stress limits allowed in the ASME, Section III Code.
- 2. The existing and modified support/restraints have been demonstrated to be adequate for revised stress analysis loads in accordance with the applicable codes.
(ii) As a result of this modification, there is no possibility for an accident or malfunction of a different tvpe than any evaluated previously in a safety analysis report because the modification has no adverse impact on the piping stresses and the associated piping supporting systems, which remain within accepted limits.
(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 nor does it change the design flow or functional requirements of This modification does not affect the pressure.
the systems.
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boundary integrity of the piping.
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'n the Technical Specifications and prior NRC approval for the implementation of this PCM is not required.
t PCM page 367-288 1 of 2 4I MFIV MAINTENANCE LIFT BEAMS ABSTRACT This engineering package is being issued in response to RFD 292-288. This . package will provide the engineering documentation required to install lifti'ng beams above the A and B train main feedwater isolation valves (MFIV's) in the Unit 2 Steam Trestle.
The lifting beams will be used periodically during plant outages to facilitate maintenance activities on the valve operators.
The lifting beams do not perform any safety Related related function.
since the However, this PC/M is classified Quality lifting beams will be located within Therefore, Quality Related the steam trestle in the vicinity of safety related items.
design requirements have been applied to this modification.
The safety evaluation has been performed and it has been determined that this PC/M does not constitute an unreviewed safety question as defined in 10CFR50.59. Furthermore, the implementation of this PC/M does not require a change to plant Technical Specifications and does not affect plant operations or safety. Based on the above, implementation of this PC/M does not require prior NRC approval.
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dl i'l DQ./ -~G Safet Analvsx ae2of P 8 2 This package addresses the addition of lifti'ng beams above the A and B train MFIU's in the Unit 2 Steam Trestle. The lifting beams will be used during Plant outages to facilitate maintenance activities on the valve operators.
As described in FSAR 3.8.4.1.9, the steam trestle provides support and missile protection for the main steam and feedwater piping and the auxiliary feedwater pumps. In addition, the steam trestle provides sufficient main steam mass and energy blowdown area to accommodate a main steam line break outside conta'inment.
With respect to title 10 of the Code shall of Federal Regulations, be deemed to involve an Part 50.59, a proposed change unreviewed safety question:
of (i) if accident the or probability malfunction of of occurrence or the consequences an equipment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; basis for any or (iii)
Technical if the margin Specification of safety is reduced.
as defined in the The modifications included in this engineering package do not involve any unreviewed safety questions because:
(i) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated ,in the FSAR are not increased by this modification because it does not affect the availability, redundancy, capacity, or function of any equipment which is important to safety. The integrity of the steam trestle missile protection and the main steam mass and energy blowdown area available to accommodate a main steam line break outside containment will not be affected.
(ii) There is no possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR because the lifting beams perform no safety function, will only be used when the applicable main feedwater train is out of service for maintenance, and no changes have been made to any operational design. Failure of the lifting beams cannot occur because the modification has been designed for the design basis conditions (iii) This modification I
does not affect the margin of safety as defined in the bases for any Technical Specification sin'ce the lifting beams do, not affect the bases for any Technical Specification.
The lifting beams do not perform any safety related function.
However, this PC/M is classified as Quality Related since failure of the lifting beams could potentially affect safety Trestle. related items since the lifting beams are located in the Steam
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The implementation of this PC/M does not require a change to plant Technical Specifications.
J 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.
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PCM 389-288 Page 1 of 6 FUEL ROD DESIGN-REGION G ABSTRACT This engineering package addresses a design modification to the St Lucie 2 Region G fuel that implements a debris resistant fuel and burnable absorber rod assembly design change. The design change consists of the lengthening of the lower end ,caps of the fuel and burnable absorber rods from 0.641 to 3.370 inches and the shortening of the lower end fittings by 0 7 inches. The basic concept of this feature is to displace the active portion of the fuel and burnable absorber axially upward. Any debris that would be caught by the lowermost Inconel grid would now interact with the inert lower and cap instead of the fuel or burnable absorber rod cladding.
The engineering package has been classified as safety related. This classification was selected since the fuel and burnable absorber rod cladding is the first barrier to preventing or mitigating the consequences of accidents that could cause undue risk to the health and safety of the public.
The safety evaluation concluded that the design changes implemented by this engineering package do not affect plant safety or operation.
Since the design changes do not pose an unreviewed safety question, require a change to the Technical Specifications, or reduce existing margins of safety, prior commission approval is not required to implement them.
Issued to cover administrative changes only. This supplement does not address any changes to plant design, plant safety or operation, and does not require a change to the Technical Specifications.
The purpose of the supplement was to change the section 11.0 drawing list to reflect JPN's numbering system and format. Also, PC/M and page numbers were added to Attachment IV.
SAFETY EVALUATIO page 2'f 6 This engineering package addresses the debris resistant fuel rod assembly design change to be used with the Region G fuel for St.
Lucfe 2. A nuclear safety related classiffcatfon was selected for thfs engfneerfng package. The safety phflosophy upon which the design and safety analysis is based fs one of multiple fission product barriers between the public and the fuel. The fuel cladding, therefore, fs the 'first barrier to providing adequate protection to the publfc.
The'esign modification consists of the lengthening of the fuel and burnable absorber rod assemblies lower end caps. The basic concept of this feature fs that the fuel column fs moved axially upward 2.029 inches relative to the Regfon F fuel column. As a result any debris caught by the Inconel grfd would interact wfth the fnert lengthened end cap instead of the actfve fuel regfon of the fuel rod cladding.
Title 10 of the'Code of Federal Regulations Section 50.59 states that the licensee may make changes to the facility as described fn the FSAR without prior Nuclear Regulatory Coamfssfon approval unless the proposed change involves a change to the technical specfffcatfons or an unrevfewed safety questfon. A proposed change involves an unrevfewed safety question ff:
a) the probability of occurrence or the consequences of an accfdent or malfunction of equipment important to safety
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previously evaluated fn the safety analysfs report fs increased. or b) the possfbilfty for an accident or malfunction of a different type than any evaluated previously fn the safety analysis report may be created, or c) 'he margins of safety as defined fn the basis for any technfcal specification fs reduced. ~ ~ C4 The modfffcatfon proposed fn this engfneerfng package neither involves a change to the plant technical specifications nor an unrevfewed safety question. Plant operation with the Region G fuel design is not a 'sa'fety concern. The bases for this conclusfon are
'addressed below.
Fuel S stem ue~<
PCM of 389-288'age 3 6 The debrfs resfstan feature fncorporated into the R on G fuel design fs a lengthened end cap resulting in Che movement of the fuel and burnable absorber columns axfally upward. The changes. resulted fn modfffcatfons to the fuel rod assembly, burnable absorber rod assembly, fuel bundle assembly, lower end fitting, gufde tube connectfons, and guide Cube assembly. These modifications are described fn the Progect Scope section. Figures 3-1 through 3-4 show, respectively, the detafls of the lower end ffttfng change, the long end cap design, the fuel rod assembly design, and the fuel bundle assembly design (more detailed information for the Region G design fs provided, respectively, in Orawing Numbers E-F16-E142-C01, O-FUEL-116-012, E-FUEL-116-222, and E-15072-711-401 listed fn Section 11.0).
The modifications described above require an 'evaluation of the mechanical and thermal designs of the fuel. The mechanical modfffcatfons made to the lower end fitting and the other fuel components were evaluated with the conclusion that none of the stress criteria are exceeded and that the design changes are acceptable. Similarly, the thermal performance of the new fuel was The results of the burnup-dependent fuel performance 'valuated..
analysis are used fn the LOCA and non-LOCA safety analysis. The thermal design evaluatfon concluded that the Cycle 5 assemblies thermal performance fs bounded by the results from the preceding analysis (see Reference 6.6).
Nuclear Oesf n The fuel column in the fuel rod assembly fs displaced upward with respect to the previous design by 2.029 inches. This change was explfcftly modeled fn the core neutronfcs analysis. These analyses results for the Region G fuel design indicate the following:
- 1. The effect on the axial power distribution fs negligfble (The increase fn axial peak at begfnning of cycle (SOC) fs less than 1.4 percent).
- 2. The effect on the fntegrated radial power dfstribution fs unobservable. That fs', there fs no discernable difference between Che maximum radfal peaking factor calculated with and wfthout the long end cap. The maxfmum integrated radial peakfng factors, therefore, are unchanged.
Pl "i'1 369-255 page 4 'of 6 C-E fs currently evaluating the effect of the juxtaposftfoned Region G fuel on the excore detectors'esponse. The result of this evaluatfon will determfne whether an additional bias needs to be applied. The additional bias, if needed, wfll mafntafn consistency
between the peripheral axfal shape index calculated from CECOR and that measured by the excore detectors.
" The effect of the upward displacement of the fuel column on the,:..",.",-
fncore nuclear fnstrumentatfon fs currently being evaluated. The result of this evaluation will determine whether any modifications are required.
Thermal-H draulfc Oesk n The thermal-hydraulic parameters for the debris resfstant fue1 design are essentially the same as those for the reference cycle as shown fn Reference 6.6. The thermal-hydraulic input data used to model the reference cycly were shown to bound the 'debris resfstant.
fuel design specfffc data.
Safet Anal sfs The fuel performance, nuclear, and thermal-hydraul fc evaluations demonstrated a neglfgfble fmpact on the fnput data to the safety analyses.
'he ECCS response to a large break LOCA was evaluated to include-the effects of the debris resfstant design. The system thermal hydraulic evaluation included the effect of the increased elevation of the bottom of the active core relatfve to the bottom of the pressure vessel. These evaluations concluded 'that the increase fn peak clad temperature due solely to the debris resfstant desfgn fs less than 10'F for the lfmftfng break size. The overall hot rod heatup evaluation, however, demonstrated that the peak clad temperature does not exceed th8 reference cycle 3 value of 2107'F for the limiting break size.
PCii 389-288 page 5 of 6 The KCCS response to a small break LOCA was also evaluated and resulted fn a peak clad temperature of 1814'F. The'evaluation concluded that the fncrease fn peak clad temperature for the worst.
small break LOCA fs sufficfently small (less than 53'F higher'han reference cycle 3 value of 1761'F) to assure that it remains
'he non-lfmftfng relative to the large breaks. The small break LOCA analysis was performed using a very c'onservatfve bounding approach.
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'"That fs, the system thermal-hydraulic calculation was not performed.
the hot rod heatup calculation was performed for this analysis.. 'nly The hot rod heatup calculation assumed that the two-phase level relative to the bottom of the pressure. vessel was the same for the base case and the debrfs resistant design. This results fn more core uncovery for the new fuel desfgn than actually would be calculated ff the effect of the increased elevatfon of the active core relative to the bottom of the pressure vessel was explfcftly modeled in the system-thermal hydraulic .analysfs. That fs, the small break LOCA calculation maximized the core boil off which
. directly impacts the clad temperature transient response. The large
'reak LOCA results remain lfmftfng as the peak clad temperature for the worst case fs about 300'F higher than for the small break LOCA worst case.
The non-LOCA safety analysis demonstrated that the analysis consequences would be unaffected. The plant setpofnts addressed in the technical speciffcatfons are unaffected by this design change.
The standards put forth fn'10 CFR 50.59 can be negatively answered.
- 1) Mill. the probability or consequences of an accfdent or malfunctfon of equipment important to safety previously reported fn the FSAR be fncreased2 The debris resfstant fuel design does not involve any increase in the probabflfty of an accfdent or malfunction previously evaluated. The impact on input data to the safety analysfs is negligfble. As a result the change in consequences of
PCN 389-288 page 6 of 6 accidents and mal functions evaluated is negligible. The setpofnts fn the plant remain unchanged. Therefore,'hanges to the Region G fuel desfgn has no impact fn the probability or consequences of an accident or mal function previously evaluated.
- 2) Will the possibility of an accident or malfunction of a type different than that previously evaluated in.the FSAR be created?
The debris resistant fuel design has been demonstrated to be compatfble to the current design from mechanical, thermal, nuclear, and thermal-hydraulic considerations. The fuel fs manufactured fn accordance with equality Class 1 requirements as fs the current fuel. Therefore, the Region G fuel does not .
create'he possibility of a new or different kind of accident.
or malfunction important to safety as.neither plant operation fs affected nor any other system fs impacted fn the plant.
3} N11 the margin of safety as defined fn the basfs of a technfcal specfffcatfon be reduced?
The technical specifications riqufrements are met by the Region G fuel because there fs no sfgnfffcant fncrease fn the probability of exceeding a safety lfmft. further, no changes to the current bases for. the technfcal specifications 'are required. The safety analysfs conclusfons from the reference cycle have not changed. As a result the plant setpoints remain unchanged. Therefore, there fs no reduction on the margin of safety as defined fn the bases of the technical specfffcations.
Prfor NRC approval fs not required to implement this engineering package as the design changes do not pose an unreviewed safety question or require a change to the technical specifications. '
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PCM 404-288 page 1 of 2 PIPE WHIP RESTRAINT REMOVE PCM 091-288, "RCS Hot Leg Nozzle Replacement" > provides for the replacement of five (5) RTD instrument nozzles in the Reactor Coolant System hot legs. In order to implement that modification and reduce personnel exposure, automated tooling and welding equipment wi11 be utilized. An interference exists between this equipment and pipe whip restraints RE-RC-30 and RE-RC-31 on the pressurizer surge line.
This Engineering Package (EP) provides the details for modifications to restraints RE-RC"30 and RE-RC-31 to permit the removal of portions of the restraints and their subsequent reinstallation in order to accommodate the operation of the automated welding equipment The modifications will be implemented in two phases. Phase I will include the .
dismantling and removal of the two pipe'hip restraints. Phase II, to be implemented after the implementation of PCM 091-288, will consist of the reinstallation of the restraints and the restoration of their design function.
Both Phase I and Phase II will be implemented when the plant is shut down for refueling (i.e. Mode 5 or 6).
This modification does not involve an unreviewed safety question, has no effect on plant safety or operation, and does not require a change to any plant Technical Specification. This Engineering Package is classified as Nuclear Safety Related since the pipe restraints are Seismic Category I structures required to protect ad)scent safety related equipment against the effects of a postulated pipe rupture.
Su lement 1 s supplement provides the details for the restoration of pipe whip restraints RE-RC-30 and RE-RC-31. The modification does not involve an unreviewed safety question or a change in any plant Technical Specification.
It has no effect'. on plant safety or operation and does not affect, amend, or change the original safety evaluation. Since the modification involves Seismic Category I structures required for the protection of safety-related equipment, this supplement retains the original Nuclear Safety Related clhssification
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PCM 404-288 SAFETY EVALUATION page 2 of 2 arith 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: (1) if the probability of occurrence or conse-quences of an accident or malfunction of equ1pment important to safety previously evaluated in the Safety Analysis Report may be increased; (ii) if a possibility fox an accident or malfunction of a diffexent type than any evaluated in the Safety Analysis Report may be created; or (iii) if the margin of safety as defined in the bases for sny Technical Specification is reduced.
This Engineering Package (EP) provides the requirements for modifi-cations to pipe whip restraints RE"RC-30 and RE-RC-31 to permit the removal and reinstallation of port1ons of the restraints in order to accommodate the implementation of PCM 091-288. The is essential to the protection of ad)scent safety related restraints'ntegxity equipment against the effects of a postulated rupture in the pressurizer surge line. Accordingly, this EP is classified as Nuclear Safety Related.
This modification does not give rise to an unreviewed safety question. The foU.owing are the bases for this conclusion:
(1) The probabil1ty of occurrence or the consequences of a previously evaluated accident or malfunction of equ1pment important to safety is not increased by this modification.
The affected pipe whip restraints are not a factor in the probability of any such accident or malfunction. The function of the restraints is to mitigate the effects of a pipe rupture accident; the implementation of this modification will ensure their continued capabQ.ity to pexform this function. There-fore, this modification cannot increase the consequences of any design basis event.
(ii) The possib1lity of an acc1dent or malfunction of equipment of a different type than any evaluated previously 1s not created.
The effect of the modifications will be to maintain the design function of the affected pxessurizer surge line whip restraints and as such will have no impact on the function of the pressurizer'uxge line or any other safety related system.
Thexefore, a failure of any safety related component which could cause, contribute to, or become a factor in a new type of acc1dent cannot result from this modificat1on.
(111) The margin of safety as defined in the bases for any Techzdcal Specificat1on is not reduced. This modification will be implemented when the plant is shut down for refueling, and at 1ts completion, the original design function of the xestraints will be restored. Therefore, no Technical Specification wiU.
be impacted by this modification.
The 1mplementation of this PCM does not require a change to plant Technical Specifications.
The foregoing constitutes, per 10CFR50.59(b), the written safety evaluation which provides the bases that this change does not involve an unreviewed safety quest1on or a change in the Technical Spec1fications and prior NRC approval for the implementation of this PCM is not required.
t PCN page 035-289 1 of 2 0 ICW PIPING Sc RESTRAINTS OF FIS 21-9B ABSTRACZ This Engineering Package (EP) provides for the elimination of two 4 inch diameter by 12 inch long fiberglass spool pieces located on the Intake Cooling Water (ICW) piping which act as tap connections for the sensing lines for FIS-21-9B, and any necessary rework of associated piping, up to the instrument root valves. A future supplement wiH. xework the piping from the root valve to the instruments Until that time, the instrument piping should be considered nonseismicslly qualified. An evaluation has shown that complete severance of. the instrument line does not affect the ability of the ICW system to perform its safety function and ICW flow can be confirmed by other means.
These changes are required as dispositions to NCR's 2-120 & 2-121, as one fiberglass spool piece is cracked and the other is needed elsewhere in the plant.
Insulating gaskets and bolt sleeves will be used to avoid galvanic interaction between the carbon steel ICW piping and aluminum-bronze instrument sense lines.
The modifications considered in the EP are in the ICW System. The ICW System performs a safety related function, therefore, this modification is classified as safety related The safety evaluation has shown that this EP does not constitute an unreviewed safety question and prior NRC approval is not required fox implementation.
The implementation of this EP does not require a change to the plant Technical Specifications and does not reduce the margin of safety for any Technical Specification.
The implementation of this EP will have no impact on plant safety or operation.
SAFETY EVALUATION With respect to Title 10 of the Code of Federal Regulation, Part 50.59, a proposed change shaU. 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 ox malfunction of a different type than any evaluated previously in the safety analysis repoxt 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 EP will eliminate two 4 inch by 12 inch long fiberglass spool pieces branching off of 'iameter the'Intake Cooling Water (ICW) piping which act as a tap connections for the sensing lines for FIS-21-9B, and any necessary rework of associated piping, up to the instrument root valves. Insulating gaskets and bolt sleeves will be used to avoid galvanic interaction between the carbon steel ICW piping and aluminum-brome instrument sense line.
0198L/
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PCM 035-289 =
page 2 of 2 In the interim period between performance of this modification and completion of seismic qualification in the supplement to this PCM, the instrument root valves shall be normally closed. In the unlikely event that the pressure boundary between the ICW header and the instrument root valves fails, the hest removal capabilities of the ICW/CCW Systems will not be adversely affected because:
- 1. The break locaton is downstream of the CCW heat exchanger therefore, total flow through the heat exchanger is not affected.
- 2. The ICW System is an open system and therefore system fluid inventory is not a concern.
- 3. Flooding of the CCW Building is not a concern because all equipment is located above elevation 23.66 feet on pedestals (Ref FSAR Section 3.4.1). The maximum possible flood elevation inside the CCW Building is 23.5 feet (Ref. Drawing 2998&&77, Sh 20 of 3) which is the elevation that the water would spill out of the outside air intakes.
The ICW System performs a Safety Related function, namely, transfer of decay heat from the Component Cooling Water system to the ultimate heat sink (Atlantic Ocean), therefore the modifications included in this Engineering Package (EP) are considered to be safety related.
This EP does not involve an unreviewed safety question; the following are the bases for this Justification:
i)'he probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report ts not increased since the modification is only to eliminate fiberglass spool pieces and replace with functionally equivalent insulating kits.
ii) As a result of this modification, there is no possibility for an accident or malfunction of a different type than any previously evaluated because no new equipment is added and the function and location of existing equipment does not change. This is further defined in Attachment 7.3 iii) This modification does not reduce the margin of safety as defined in the basis for any Technical Specification of the because it neither it change the changes the design parameters ICW nor does ICW design flow or functional requirements.
The implementation of this PCM does not require a change to or impact the plant Technical Specifications.
The foregoing constitutes, per 10CFR50.59(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, therefore, prior NRC approval for the implementation of this PCM is not required.
CM 044-289 page 1 of 1 STM GEN TUBE PLUQWE DESIGN ABSTRACT This PCM documents Engineering review and concurrence for the use of Combustion Engineering expanded type plugs in the St. Lucie Unft 2 steam generators. This PC/M also provides the information necessary to as-build affected documents.
Since the steam generator tubes are nuclear safety related, the tube plugs described herein are also nuclear safety related.
Based upon a failure mode evaluation and 10 CFR 50.59 review, thfs modification does not involve an unreviewed safety question nor require changes to the technical specifications. Therefore, prior NRC approval fs not required for implementation of this modification. The modification has no adverse affect on plant safety or operability.
SAFETY EVALUATION This modi'ffcatfon involves documenting the maintenance practice of plugging steam generator tubes. Steam generator tubes are nuclear safety related, therefore this engineering package fs classified as nuclear safety related.
The PC/M provides engineering concurrence for. the use of the Combustion Engineering expanded tube plug design (previously utilized on the Unit 2 steam generators), the required 50.59 .review of the modification, and the information required for update of affected documents.
10 CFR 50.59 allows a change to a nuclear facility without prior NRC approval if an unr eviewed 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 ts 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).
ii) The consequences of a previously postulated design basis accfdent or malfunction of equipment important to safety previously evaluated in the FSAR are not made more severe for the same reasons given in (i) and since no existing accident mitigation equipment or. systems are altered by this modfficatfon.
iii) 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.
iv) 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 fs less than assumed in the Cycle Five 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
'equired, the addition of the Combustion Engineering tube plugs to the Unit 2 steam generators does not require prior NRC approval.
PCM 171-985 age 1 of 1 HYPOCHLORITE SYS Z&C ENHANCEMENTS ABSTRACT This engineering package covers the modifications and details required to.support the enhancements to the Hypochlorite System instrumentation.
The hypochlorite system is used to inject a hypochlorite solution into the seawater upstream of t4e intake structure to control slime formation.
The modifications and details consist of improvement of quality of instrument air supply, replacement of 4 level switches per unit, improvement to the bubbler system and addition of level gauge glasses to the rectifiers oil tanks. All the modifications will oe performed within the Hypochlorite Skid boundaries.
Based on the design of the Hypochlorite Injection System, this Engineering Package has been classified non-safety related.
This modification will improve the overall performance of the instrumentation, providing accuracy and repeatability in, all the automatic and alarm functions and it will reduce maintenance of the hypochlorite system.
SAFETY EVALUATION I
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 imporatnt to safety previously evaluated in the safety analysis report may be increased; or (ii) if the 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 basis for any technical specification is reduced.
The Hypochlorite injection system is not a safety related system. The instrumentation enhancements to be performed have no impact on any safety related plant systems and/or operations. The modifications improve the equipment operation without changing the original design intents Chlorine in the form of Sodium Hypochlorite is used to control biological fouling in the Circulating, Water System by use of a hypochlorite generating system serving both St Lucie Units 1 and 2 ~
No other chemicals will be added to the circu1ating water flow.
This modification to this package will not increase the probability or consequences of an accident. This system is not used in any accident mitigation scenario and therefore the systems failure will have no impact on plant safe shutdown. This modificaton is not described in the Technical Specifications and therefoxe, the implementation of this PCM does not require a change to the, plant Technical Specifications.
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 Nuclear Regulatory Commission approval for the implementation of this PCM is not required.
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CM 111-986 Page 1 of 2 SIMULATOR BLDG PIPING TIE-INS ABSTRACT This engineering package is being issued to cover the addition of. the St Lucie Simulator Training Facility fire protection, service water and sanitary piping tie-ins. No aspect of this project will add to, modify or otherwise affect any plant safety related system. Fire system modifications associated with this project that tie into the plant, fire loop up to the first isolation valve shall be classified as "Quality Related" QA/QC required. The remainder of the fire protection, all service water and all sanitary piping shall be Non-Nuclear Safety Related.
The addition of the Simulator Training Facility fire protection, service water and sanitary tie-in lines do not pose any unreviewed safety questions nor involve any changes to plant Technical Specifications.
Su ement 1 This revision incorporated engineering and construction comments by adding references, drawings and s pecifications to the ap pro priate sections and making editorial changes. These changes do not affect the conclusions of the original safety evaluation.
PCM 111-986 page 2 of 2 SAPEIT BVALUhTION The St. Lucie plant fire protection loop is defined as a Quality Related system. Those portions of this modification that tie into the fire loop up to and including the first isolation valve have been designated as Quality Related and conform w'ith the requirements oi the original fire loop. The remainder of the fire protection piping added by this modification has been designated Non-Nuclear Safety Related Quality Group D. Those portions of the modification providing service water and sanitary piping and tie-ins are classified as.Non-Nuclear. Safety Related. These components tie into the existing plant service water and sanitary systems which are also classified as Non-Nuclear Safety Related Quality Group D.
A faQure mode analysis was performed on the Non-Nuclear Safety Related portions of the modification. Based on this analysis, failure of the service water and sanitary piping or components and those portions of the fire main downstream of the first isolation valve will not inhibit the operation of any safety related equipment or components. These materials are allocated remote from any safety related equipment or components and as such cannot fall on or hit such components. &allure of the service water line will cause loss of service water to the simulator building. Failure of the sanitary line will inhibit the use.of the Simulator Building sanitary system. Failure of the downstream fire main piping will not inhhit the functional capabilities of the fire loop since the post indicator valve, located upstream of these portions of the system provides adequate isolation capabilities to ensure functional integrity of the fire loop.
Those portions of the modification providing fire protection piping tie-ins to the first isolation valve can affect the functional capabilities of the fire loop and therefore can affect fire protection capabilities for Safety Related equipment and components. As addressed in the Design Analysis, these portions of the modification have been desgned and construction requirements have been specified to comply with the necessary Quality kelated requirements. Since the equipment affected-by this modification is not considered by the FSAk in determining the probability of accidents or possible types of accidents or in the evaluation of the consequences of accidents, it can be concluded that the probability of occurrence of accidents previously aadressea in the FSak is unchanged and the possibility 'f new accidents not considered in the FSAk has not been created.
Therefore, the potential fa>lure mode of this system and degree of protection provided to nuclear safety related equipment remains unchanged.
Based on this information, it can be demonstrated that an unreviewed safety question as defined by 10CFB50.59 does'not exist since the consequences of all analyzed accidents remains unchanged. Additionally, with respect to Nuclear Safety, no new accidents or malfunctions are introduced as a result of this modification. Finally, the margin of safety as defined in the Technical Specifications has not been reduced nor have changes to the Technical Specifications been required.
ln conclusion, this modification is acceptable from the standpoint of nuclear safety since it does.not involve an unreviewed safety question nor require changes to the'Technical Specifications. Thus implementation of this modification does not require prior NnC approval
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