ML030080594

Report of Changes, Tests, & Experiments - 10 CFR 50.59

ML030080594
Person / Time
Site:	Calvert Cliffs
Issue date:	01/03/2003
From:	William Holston Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML030080594 (18)

Text

Calvert Cliffs Nuclear Power Plant Constellation Generation Group, LLC 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 Constellation Energy Group January 3, 2003 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

SUBJECT:

Document Control Desk Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Report of Changes, Tests, and Experiments - 10 CFR 50.59 In accordance with 10 CFR 50.59(b)(2), Calvert Cliffs Nuclear Power Plant hereby submits a report containing brief descriptions of changes, tests, and experiments approved under the provisions of 10 CFR 50.59.

Attachment (1) of this report includes 50.59 evaluations recorded and approved between January 1, 2001 and July 31, 2002. Items in the report are sorted by 50.59 identification number. We will be submitting another report covering August 1, 2002 through December 31, 2002 at a later time.

Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, William C. Holston Manager, CCNPP Engineering Services WCHIEMT/bjd

Attachment:

(1)

Calvert Cliffs Nuclear Power Plant Report of Changes, Tests, and Experiments

[10 CFR 50.59(b)(2)]

cc:

J. Petro, Esquire J. E. Silberg, Esquire Director, Project Directorate I-1, NRC D. M. Skay, NRC H. J. Miller, NRC Resident Inspector, NRC R. I. McLean, DNR 7

ATTACHMENT (1)

CALVERT CLIFFS NUCLEAR POWER PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS

[10 CFR 50.59(b)(2)]

Calvert Cliffs Nuclear Power Plant, Inc.

January 3, 2003

50.59 's Approved by POSRC 1/1/2 001 - 7/31/2 002 DOCUMENTID SUBJECT REV

SUMMARY

SE00154 0002 ESI 99502468-003 INSTALLED A CONTAINMENT OUTAGE DOOR (COD) ON THE EXTERIOR OF EACH CONTAINMENT STRUCTURE AT THE EQUIPMENT HATCH (EH) OPENING AS A SUBSTITUTE FOR THE EQUIPMENT HATCH DOOR (EHD) WHEN SETTING CONTAINMENT CONTAINMENT CLOSURE DURING OUTAGE CONDITIONS. IN ADDITION, THIS MODIFICATION CONDUCTED A ONE-TIME PNEUMATIC PROOF TEST OF OUTAGE DOOR THE COD USING THE EXISTING EQUIPMENT HATCH DOOR AS A TEST BOUNDARY. ADDED A DESCRIPTION OF THE COD TO UFSAR PROJECT CHAPTER 5.1, REVISED THE DESCRIPTION OF CLOSURE CONDITIONS IN UFSAR SECTION 14.18 (FUEL HANDLING INCIDENT), AND REVISED APPLICABLE UFSAR FIGURES TO SHOW THE COD. SAFETY EVALUATION SE00154 CONCLUDED THAT NO UNREVIEWED SAFETY QUESTION EXISTED BECAUSE THE MODES 5 AND 6 EVENT CONDITIONS ARE MUCH LESS SEVERE THAN THE MODE 1 EVENTS AND THE EQUIPMENT INTENDED FOR MODES 5 AND 6 USE (THE COD) WAS DESIGNED USING THE CODE ALLOWABLES FOR DESIGN OF THE CORRESPONDING MODE 1 EQUIPMENT(THE EHD), BUT AT THE APPROPRIATE CONDITIONS EXPERIENCED DURING MODES 5 AND 6 CASUALTIES. THIS MODIFICATION IS CONSISTENT WITH THE "PROGRAMMED ENHANCEMENTS" DISCUSSED IN GL 88-17.

SE00260 0000

SUMMARY

THIS ACTIVITY INVOLVES 6 CHANGES TO SECTION 9.7 OF THE UFSAR TO ADDRESS DISCREPANCIES OBSERVED DURING THE UFSAR REVIEW PROJECT. THEY ARE:

A. REMOVE DESCRIPTION OF A TEMPERATURE RANGE FOR NEW FUEL. THERE IS NO UPDATE UFSAR REQUIREMENT.

B. REMOVE REFERENCE TO A COVER FOR THE SPENT FUEL POOL (SFP). THERE IS NO COVER.

C. STATE SECTION 9.7 THAT THE DEMINERALIZER BEDS ARE NOT REGENERATED RATHER THAN NON-REGENERABLE. REMOVE STATEMENT THAT THE SYSTEM IS DESIGNED TO ELIMINATE THE NEED OF A TRANSFER CASK.

D. ADD A CLARIFICATION THAT 5.0 WT% ENRICHMENT FUEL MAY NOT BE STORED IN THE SFP RACKS.

E. ADD A BRIEF DESCRIPTION OF THE MANUAL MODE OF OPERATING THE SPENT FUEL HANDLING MACHINE.

F. CLARIFY THAT THE TWO HALVES OF THE SFP ARE NOT IDENTICAL. THE CHANGES ARE ALL MINORIEDITORIAL AND HAVE NO EFFECT ON ANY PLANT FUNCTION OR SAFETY ANALYSIS OR MARGIN OF SAFETY. THUS, THE CHANGE IS NOT CONSIDERED A USQ.

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SUMMARY

SE00345 0003

SUMMARY

THIS SAFETY EVALUATION CONSIDERED THE OPERATION OF UNIT 2 CYCLE 13. MODIFICATIONS TO THE FUEL ASSEMBLY AND THE RELOAD CORE DESIGN WERE CONSIDERED. THE USE OF A THIRD FULL BATCH OF ERBIUM FOR UNIT 2 AS A BURNABLE UNIT 2 CYCLE 13 ABSORBER WAS CONSIDERED. THE PRE-TRIP STEAM LINE BREAK EVENT, AND SEIZED ROTOR EVENT WERE EVALUATED USING NRC RELOAD SAFETY APPROVED DNB CONVOLUTION METHODOLOGY TO PREDICT THE PERCENTAGE OF FUEL FAILURES. THE EVALUATION ASSUMED FrT EVALUATION FOR and FxyT LIMITS EQUAL TO 1 65 AND RESULTED IN FUEL FAILURES LESS LIMITING THAN THAT PREVIOUSLY REPORTED. THE EOC REFUELING CHANGES ASSOCIATED WITH FrT AND FxyT LIMITS EQUAL TO 1 65 ARE IMPLEMENTED IN THE UNIT 2 CYCLE 13 COLR AND ARE BORON VERIFIED TO BE APPLICABLE TO UNIT 2 CYCLE 13. THE LOSS OF LOAD EVENT WAS REANALYZED FOR A DECREASE IN THE RANGE OF CONCENTRATION TURBINE STOP VALVE CLOSURE TIMES. THE ANALYSIS CONCLUDED THAT THE PEAK RCS AND STEAM GENERATOR PRESSURES AND THE LINEAR HEAT RATE DO NOT EXCEED THE NRC ACCEPTANCE LIMITS. THE EXCESS HEAT REMOVAL EVENT WAS EVALUATED FOR AN INCREASE IN FEEDWATER FLOW AND A DECREASE IN FEEDWATER ENTHALPY. THIS EVALUATION CONCLUDED THAT THE PREVIOUSLY REPORTED RESULTS WERE MORE LIMITING. THE POST-TRIP STEAM LINE BREAK EVENT WAS RE-EVALUATED WITH REGARD TO COOLABILITY FOR AN INCREASE IN THE SAFETY INJECTION SWEEPOUT VOLUME. IT WAS DETERMINED THAT FUEL FAILURE DUE TO VIOLATION OF THE DNB SAFDL OR DUE TO EXCEEDING THE CENTER LINE MELT (CLM) LIMIT DID NOT OCCUR FOLLOWING THE RETURN TO POWER FOR THE FULL POWER CASES, THUS AVERTING THE NEED TO INVOKE THE CORE COOLABILITY LIMIT. THE RESULTS OF THIS EVALUATION WERE BOUNDED BY THE PREVIOUSLY REPORTED RESULTS FOR THE POST-TRIP STEAM LINE BREAK EVENT. A TOTAL OF 48 CEAs WERE CREDITED FOR REFUELING BORON CONCENTRATION. THE UNIT 2 CYCLE 13 SAFETY ANALYSES ACCOUNTED FOR ALL THE RELOAD CORE DIFFERENCES. REVISION 0001 OF THIS SAFETY EVALUATION CONSIDERED THE IMPACT OF NOT INSTALLING THREE INCORE INSTRUMENTS (ICls). IT IS CONCLUDED THAT THE CORE MONITORING SYSTEM WAS NOT ADVERSELY EFFECTED BY NOT INSTALLING THESE ICIs. REVISION 0002 OF THIS SAFETY EVALUATION CONSIDERED THE IMPACT OF SPALLATION AND GRID TO ROD FRETTING ON UNIT 2 CYCLE 13. REVISION 0003 JUSTIFIES THE NEW UNIT 2 CYCLE 13 REFUELING BORON CONCENTRATION REQUIREMENT, AS OF JANUARY 2001. THE UPDATED REFUELING BORON CONCENTRATION CREDITS THE NEGATIVE REACTIVITY OF 0 CEAS. IT IS CONCLUDED THAT OPERATION OF UNIT 2 CYCLE 13 DOES NOT INVOLVE AN UNREVIEWED SAFETY QUESTION.

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SE00385 0000

SUMMARY

IN SUPPORT OF THE CCNPP UNIT 1 STEAM GENERATOR REPLACEMENT (SGR), ESP ES199601526-110 PROVIDES FOR THE REMOVAL AND MODIFICATION OF EXISTING SECTIONS OF THE SURFACE AND BOTTOM BLOWDOWN SYSTEM PIPING/SUPPORTS IN SAFETY THE VICINITY OF THE UNIT 1 STEAM GENERATORS. THIS ESP SUPPLEMENT ALSO ADDRESSES THE REINSTALLATION OF THESE EVALUATION FOR SECTIONS OF PIPING/SUPPORTS AND ANY BLOWDOWN PIPING OR SUPPORT MODIFICATIONS, INCLUSIVE OF NEW PIPING/SUPPORTS ESP ES199601526-FOR A SECOND RSG BLOWDOWN NOZZLE, NECESSARY TO ACHIEVE FIT-UP WITH THE REPLACEMENT STEAM GENERATORS (RSGs).

110, UNIT 1 SGR EVALUATION OF THE DESIGN AND FUNCTION OF THE BLOWDOWN SYSTEM FOR THE REPLACEMENT STEAM GENERATOR (RSG)

BLOWDOWN COMPONENTS IS CONTAINED IN ESP ES199601526-000, "REPLACEMENT OF CCNPP STEAM GENERATORS". OUTAGE IMPLEMENTATION OF THIS ESP WILL BE LIMITED TO UNIT I COLD SHUTDOWN (MODE 5), REFUELING SHUTDOWN (MODE 6), OR WITH THE REACTOR DEFUELED AND ALL FUEL REMOVED FROM CONTAINMENT (DEFUELED), WITH THE FURTHER MODE 5 RESTRICTION THAT ACTIVITIES AFFECTING THE SECONDARY SIDE BOUNDARY MUST AWAIT CLEARANCE BY PLANT OPERATIONS TO ASSURE THAT THE STEAM GENERATORS ARE NO LONGER REQUIRED TO BE MAINTAINED AVAILABLE FOR RCS DECAY HEAT REMOVAL EQUALLY, SUCH SECONDARY SIDE BOUNDARY-AFFECTING ACTIVITIES MUST BE COMPLETED AND THE RSGs BE RESTORED TO OPERABILITY AT THE CONCLUSION OF THE STEAM GENERATOR REPLACEMENT OUTAGE PRIOR TO PLANT OPERATIONS RELYING ON THE STEAM GENERATORS TO FUNCTION AS RCS DECAY HEAT REMOVAL DEVICES. THE LICENSING AND DESIGN BASES FOR THE BLOWDOWN SYSTEM PIPING/SUPPORTS AFFECTED BY IMPLEMENTATION OF ESP ES199601526-110 ARE TO MAINTAIN THE INTEGRITY OF THE SECONDARY SIDE PRESSURE BOUNDARY/CONTAINMENT BARRIER, AND TO DIRECT BLOWDOWN FROM THE STEAM GENERATORS TO TO OTHER BLOWDOWN SYSTEM COMPONENTS (E.G., WATER RECOVERY, SG SAMPLING). ESP ES199601526-110 CONTAINS AN ASME CODE RECONCILIATION WHICH DEMONSTRATES THAT IMPLEMENTATION ACTIVITIES (I E., PIPE CUTTING, PREPARATION AND WELDING, PIPING AND PIPE SUPPORT MODIFICATION DESIGN, AND WELD TESTING/INSPECTION) WILL BE PERFORMED IN ACCORDANCE WITH ASME CODE REQUIREMENTS, AS RECONCILED TO THE ORIGINAL PLANT CONSTRUCTION CODE. THIS CODE RECONCILIATION INCLUDES EVALUATIONS FOR PIPING AND SUPPORT DESIGN, FABRICATION, INSTALLATION AND INSPECTION/TESTING TO VERIFY THAT INSTALLATION EQUIVALENCY TO THE ORIGINAL CONSTRUCTION CODE HAS BEEN MAINTAINED. ESP ES199601526-110 IMPLEMENTATION WILL ENSURE THAT THE POST-MODIFICATION BLOWDOWN SYSTEM PIPING IS ENGINEERED-EQUIVALENT AND RESTORED TO THE ORIGINAL CONSTRUCTION CODE, RESULTING IN NO CHANGES TO THE CCNPP UFSAR DESIGN BASES ACCIDENT OR EQUIPMENT MALFUNCTION ASSUMPTIONS. AS A RESULT OF THESE MEASURES, PIPING RELIABILITY IS REESTABLISHED EQUIVALENT TO PRE-ESP IMPLEMENTATION RELIABILITY SO THAT THE PROBABILITY OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAR IS NOT INCREASED. THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAR WILL NOT BE INCREASED. EQUALLY, NO ACCIDENT INITIATORS OF ACCIDENTS OR EQUIPMENT MALFUNCTION MECHANISMS OF A DIFFERENT TYPE THAN ANY PREVIOUSLY EVALUATED IN THE SAR ARE CREATED. IN THE REVIEW CONDUCTED TO PERFORM THIS 10 CFR 50 59 SAFETY EVALUATION, THE FOLLOWING TECHNICAL SPECIFICATION BASES WERE IDENTIFIED AS APPLICABLE FOR REVIEW FOR THE IMPLEMENTATION OF ESP ES199601526-110: B3.4.7, B3 4 8, B3 6.1., B3 6.3, B3.7.14, AND B3.9.3. THE MARGIN OF SAFETY AS DEFINED IN THESE TECHNICAL SPECIFICATION BASES WAS DEMONSTRATED NOT TO BE REDUCED. BASED ON THE 10 CFR 50.59 SAFETY EVALUATION PERFORMED FOR ESP ES199601526-1 10, IT IS CONCLUDED THAT AN UNREVIEWED SAFETY QUESTION DOES NOT EXIST, AND THAT THIS ESP MAY BE PERFORMED WITHOUT PRIOR NRC APPROVAL.

Tuesday, December 31, 2002



Page 3 of 16o

DOCUMENT ID SUBJECT REV

SUMMARY

SE00386 SAFETY EVALUATION FOR ESP ES199601526 107, UNIT 1 SGR STEAM GENERATOR SUPPORTS Tuesday, December 31, 2002

SUMMARY

IN SUPPORT OF THE CCNPP UNIT I STEAM GENERATOR REPLACEMENT (SGR), ESP199601526-107 PROVIDES THE DESIGN AND IMPLEMENTATION FOR TEMPORARY AND PERMANENT MODIFICATIONS to the STEAM GENERATOR 11 AND STEAM GENERATOR 12 SUPPORTS (SG UPPER LATERAL SUPPORTS, SG SNUBBERS, AND SG SLIDING BASE) DURING THE UNIT 1 STEAM GENERATOR REPLACEMENT OUTAGE. THE DESIGN INPUTS REQUIREMENTS (DIR) TO THIS ESP SUPPLEMENT CONTAINS ENGINEERING REVIEWS WHICH DEMONSTRATE THE DESIGN ACCEPTABILITY OF BOTH TEMPORARY AND PERMANENT MODIFICATIONS TO THE SG SUPPORTS, AND FURTHER ENSURE THAT DURING IMPLEMENTATION WINDOWS FOR ESP ES199601526-107 ALL DESIGN AND LICENSING BASIS FUNCTIONS for the SG SUPPORTS ARE MAINTAINED. FOLLOWING PLACEMENT OF THE RSGs WITHIN THEIR SG CUBICLES, THE SG SUPPORTS WILL BE REATTACHED AND INSPECTED TO ENSURE PROPER FUNCTION. AS PART OF THESE INSPECTIONS, HOT GAP MEASUREMENTS WILL BE PERFORMED DURING MODE 3. AS FURTHER DISCUSSED In the DIR, THE GOVERNING PRINCIPLE FOR DETERMINING UNIT 1 MODE FOR PERFORMING THESE ACTIVITIES WILL BE WHETHER STEAM GENERATOR SUPPORT IS REQUIRED TO MAINTAIN THE SEISMIC FUNCTION OF THE SG AND THE RCS. AS A RESULT, MOST OF THE IMPLEMENTATION FOR THIS ESP SUPPLEMENT WILL BE PERFORMED DURING THE UNIT 1 DEFUELED CONDITION HOWEVER, THE DIR ALSO PROVIDES THE BASIS FOR TEMPORARY REMOVAL/RESTORATION OF CERTAIN SG SNUBBERS IN MODES 5 AND 6, WHEN PIPE RUPTURE LOADS WILL NOT BE PRESENT. THE NRC-APPROVED APPLICATION OF LEAK-BEFORE-BREAK OF PRIMARY COOLANT PIPING FOR LARGE BREAK LOCAs IS UTILIZED TO ALLOW THE BENT SHIM PLATE FROM EACH SG SLIDING BASE SUPPORT TO BE PERMANENTLY DELETED WITH THE PURPOSE OF FACILITATING PROPER FITUP AND FUNCTION OF THE RSGs TO THEIR RESPECTIVE SG SLIDING BASE SUPPORT. THE MODIFIED SG SUPPORTS MEET ALL DESIGN AND LICENSING BASIS REQUIREMENTS AS THE PRE-MODIFICATION SG SUPPORTS. THE DESIGN AND LICENSING BASIS EQUIVALENCE ASSURES THAT THERE ARE NO CHANGES TO THE CCNPP UFSAR DESIGN BASES ACCIDENT OR EQUIPMENT MALFUNCTION ASSUMPTIONS. THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATION BASES IS NOT REDUCED. BASED ON THE 10 CFR 50.59 IMPLEMENTATION SAFETY EVALUATION PERFORMED FOR ESP ES199601526 107, IT IS CONCLUDED THAT AN UNREVIEWED SAFETY QUESTION DOES NOT EXIST, AND THAT THIS ESP MAY BE PERFORMED WITHOUT PRIOR NRC APPROVAL.

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DOCUMENT ID SUBJECT REV

SUMMARY

SE00387 0000

SUMMARY

IN SUPPORT OF THE CCNPP UNIT 1 AND UNIT 2 STEAM GENERATOR REPLACEMENT (SGR), ESP ES199601526-004 PROVIDES FOR BUILDING OF AN ORIGINAL STEAM GENERATOR STORAGE FACILITY (OSGSF) ON THE CCNPP OWNER CONTROLLED SAFETY AREA, JUST NORTH OF THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI). BECAUSE CHANGES TO THE UFSAR WOULD EVALUATION FOR BE REQUIRED TO REFLECT THE PRESENCE OF THE OSG STORAGE FACILITY ON UFSAR FIGURE 2.3-1 AND UFSAR DESCRIPTIONS OF ESP EO199601526-ONSITE RAD WASTE STORAGE (SECTION 11.1.2), A SAFETY EVALUATION IS PERFORMED. THE OSGSF WILL BE A 2-BAY REINFORCED 004, ORIGINAL CONCRETE STRUCTURE CONSTRUCTED WITH WALLS AND ROOF OF SUFFICIENT THICKNESS TO PROVIDE RADIATION SHIELDING OF STEAM GENERATOR THE OSG LOWER ASSEMBLIES DURING STORAGE AND MAINTAIN THE DOSE LIMITS SPECIFIED IN 10 CFR 20, "STANDARDS FOR STORAGE FACILITY PROTECTION AGAINST RADIATION". THE OSGSF IS DESIGNED FOR SECURE STORAGE OF THE OSG LOWER ASSEMBLIES FOR THE TIME PERIOD FOLLOWING INITIAL USE THROUGH TO END-OF-PLANT LIFE (YEAR 2038) AND COMMENCEMENT OF PLANT DECOMMISSIONING THE ESP ES199601526-004 DESIGN INPUT REQUIREMENTS (DIR) EVALUATION CONTAINS THE NECESSARY ENGINEERING REVIEWS DEMONSTRATING THAT BUILDING AND USE OF THE OSGSF MEETS ALL REGULATORY REQUIREMENTS FOR ONSITE STORAGE OF THIS CATEGORY OF RADIOACTIVE WASTE. BECAUSE OF THE MUTUAL EXCLUSIVITY OF THE OSGSF AND IMPORTANT TO SAFETY EQUIPMENT DESCRIBED IN THE SAR, THERE ARE NO CHANGES TO THE CCNPP UFSAR DESIGN BASES ACCIDENT OR EQUIPMENT MALFUNCTION ASSUMPTIONS, AND THE PROBABILITY OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAR IS NOT INCREASED. FOR THE SAME REASON, THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF EQUIPMENT IMPORTANT TO SAFETY PREVIOUSLY EVALUATED IN THE SAR WILL NOT BE INCREASED. EQUALLY, NO ACCIDENT INITIATORS OF ACCIDENTS OR EQUIPMENT MALFUNCTION MECHANISMS OF A DIFFERENT TYPE THAN ANY PREVIOUSLY EVALUATED IN THE SAR ARE CREATED. THE MARGIN OF SAFETY AS DEFINED IN THE TECHNICAL SPECIFICATIONS BASES IS NOT REDUCED. THE ENVIRONMENTAL EVALUATION OF THE CALVERT CLIFFS NUCLEAR POWER PLANT STEAM GENERATOR REPLACEMENT PROJECT PERFORMED IN ADHERENCE TO TECHNICAL SPECIFICATIONS APPENDIX "B",

ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL) TECHNICAL SPECIFICATIONS CONCLUDED THAT NO UNREVIEWED ENVIRONMENTAL SAFETY QUESTION EXISTS AND THAT IMPLEMENTATION OF ESP ES199601526-004 MAY BE PERFORMED WITHOUT PRIOR NRC APPROVAL. BASED ON THE 10 CFR 50.59 SAFETY EVALUATION PERFORMED FOR ESP ES199601526-004, IT IS CONCLUDED THAT AN UNREVIEWED SAFETY QUESTION DOES NOT EXIST, AND THAT THIS ESP MAY BE PERFORMED WITHOUT PRIOR NRC APPROVAL.

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DOCUMENT ID SUBJECT REV SUMMA R Y SE00388 0001 IN SUPPORT OF THE CCNPP UNIT I STEAM GENERATOR REPLACEMENT (SGR), ESP ES199601526-109 PROVIDES FOR THE REMOVAL SAFETY AND MODIFICATION OF EXISTING SECTIONS OF THE FEEDWATER SYSTEM PIPING (PIPING CLASS DB-1) ATTACHED TO STEAM GENERATORS SGl1 AND SG12 THIS ESP SUPPLEMENT ADDRESSES THE INSTALLATION AND REMOVAL OF TEMPORARY SUPPORTS EVALUATION FOR AND THE SUBSEQUENT REINSTALLATION OF THE PIPING SECTIONS TO THE REPLACEMENT STEAM GENERATORS (RSGs), INCLUDING ESP ES199601526-ANY NECESSARY PIPING MODIFICATIONS TO ACHIEVE FIT-UP. IN ADDITION, THIS ESP SUPPLEMENT ALLOWS THE INSTALLATION OF 109, UNIT 1 SGR 1" INSPECTION PLUGS IN LINES 16" DB-1-1018 AND 16" DB-1-1019 FOR THE PURPOSE OF RADIOGRAPHIC INSPECTION OF FIELD BUTT FEEDWATER WELDS. OUTAGE IMPLEMENTATION OF THIS ESP WILL BE LIMITED TO UNIT 1 COLD SHUTDOWN (MODE 5), REFUELING SHUTDOWN (MODE 6), OR WITH THE REACTOR DEFUELED AND ALL FUEL REMOVED FROM CONTAINMENT (DEFUELED), WITH THE FURTHER MODE 5 RESTRICTION THAT ACTIVITIES AFFECTING THE SECONDARY SIDE BOUNDARY MUST AWAIT CLEARANCE BY PLANT OPERATIONS TO ASSURE THAT THE STEAM GENERATORS ARE NO LONGER REQUIRED TO BE MAINTAINED AVAILABLE FOR RCS DECAY HEAT REMOVAL. ATTACHMENT AND REMOVAL OF TEMPORARY SUPPORTS TO THE FEEDWATER SYSTEM PIPING WITHOUT APPLICATION OF RESTRAINING FORCES IS CONSIDERED TO BE AN ACTIVITY NOT AFFECTING THE PRESSURE BOUNDARY OF THE FEEDWATER SYSTEM. ALL FEEDWATER SYSTEM SECONDARY SIDE PRESSURE BOUNDARY-AFFECTING ACTIVITIES MUST BE COMPLETED AND THE RSGs BE RESTORED TO OPERABILITY AT THE CONCLUSION OF THE STEAM GENERATOR REPLACEMENT OUTAGE PRIOR TO PLANT OPERATIONS RELYING ON THE STEAM GENERATORS TO FUNCTION AS RCS DECAY HEAT REMOVAL DEVICES. THE LICENSING AND DESIGN BASES FOR THE FEEDWATER SYSTEM PIPING ARE TO MAINTAIN THE INTEGRITY OF THE SECONDARY SIDE PRESSURE BOUNDARY/CONTAINMENT BARRIER, TO FUNCTION TO REMOVE RCS DECAY HEAT USING THE STEAM GENERATORS AS A HEAT REMOVAL MECHANISM, AND TO MINIMIZE THE LIKELIHOOD OF A HIGH ENERGY LINE BREAK (I.E., MAIN FEEDWATER LINE BREAK).

ESP ES199601526-109 CONTAINS AN ASME CODE RECONCILIATION WHICH DEMONSTRATES THAT IMPLEMENTATION ACTIVITIES (I E.,

PIPE CUTTING, PREPARATION AND WELDING, PIPING MODIFICATION DESIGN, AND WELD TESTING/INSPECTION) WILL BE PERFORMED IN ACCORDANCE WITH ASME CODE REQUIREMENTS, AS RECONCILED TO THE ORIGINAL PLANT CONSTRUCTION CODE. THIS CODE RECONCILIATION INCLUDES RSG MODIFICATION EVALUATIONS FOR PIPING DESIGN, FABRICATION, INSTALLATION AND INSPECTION/TESTING TO VERIFY THAT INSTALLATION EQUIVALENCY TO THE ORIGINAL CONSTRUCTION CODE HAS BEEN MAINTAINED. ESP ES199601526-109 IMPLEMENTATION WILL ENSURE THAT ANY REPLACEMENT FEEDWATER SYSTEM PIPING IS ENGINEERED-EQUIVALENT AND RESTORED TO THE ORIGINAL CONSTRUCTION CODE, RESULTING IN NO CHANGES TO THE CCNPP UFSAR DESIGN BASES ACCIDENT OR SSC MALFUNCTION ASSUMPTIONS. AS A RESULT OF THESE MEASURES, FEEDWATER PIPING RELIABILITY IS REESTABLISHED EQUIVALENT TO PRE-ESP IMPLEMENTATION RELIABILITY SO THAT THE LIKELIHOOD OF OCCURRENCE OF UFSAR-EVALUATED ACCIDENTS/MALFUNCTIONS OF SSCs IMPORTANT TO SAFETY IS NOT INCREASED. THE CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION OF SSCs PREVIOUSLY EVALUATED IN THE UFSAR WILL NOT BE INCREASED.

EQUALLY, NO ACCIDENT INITIATORS OF ACCIDENTS OF A DIFFERENT TYPE OR SSC MALFUNCTION MECHANISMS WITH A DIFFERENT RESULT THAN ANY PREVIOUSLY EVALUATED IN THE UFSAR ARE CREATED. NO FISSION PRODUCT DESIGN BASIS IS ALTERED FROM IMPLEMENTATION OF THIS ESP SUPPLEMENT. THE CHANGE IN DESIGN PRESSURE FOR SECTIONS OF FEEDWATER SYSTEM PIPING FROM THE FEEDWATER PENETRATIONS TO THE STEAM GENERATORS FROM 1500 PSIG TO 1400 PSIG DOES NOT COMPROMISE THE SAFETY OR DESIGN BASES OF THESE FEEDWATER LINES, AS WAS CONCLUDED IN SAFETY EVALUATION NO 90-B-045-164-RO (DECEMBER 19, 1990) AS ADDRESSED IN THE DIR FOR ESP ES199601526-109, THERE IS NO EFFECT OF ANY REPLACEMENT STEAM GENERATOR STEAM PRESSURE INCREASE ON THE CCNPP CALCULATION M-90-143 EVALUATED WORST CASE SCENARIO PRESSURE POSTULATED TO BE EXPERIENCED BY THIS FEEDWATER PIPING TO ENSURE THAT A 1400 PSIG DESIGN PRESSURE REMAINS BOUNDING FOR THIS FEEDWATER PIPING. THE ANALYSIS METHOD USED TO MAKE THIS DESIGN PRESSURE DETERMINATION WAS BASED ON ACCEPTED ENGINEERING PRACTICES, AND WAS IN KEEPING WITH THE DESIGN ANALYSIS AND ENGINEERING ASSUMPTIONS USED TO ESTABLISH THE ORIGINAL UFSAR-DESCRIBED DESIGN PRESSURE VALUE, INCLUDING EVALUATING THE IMPACT OF USING THE RSGs. THE 10 CFR 50 59 SAFETY EVALUATION PERFORMED FOR ESP ES1 996015265-109 CONCLUDES THAT NO Tuesday, December 31, 2002 Argeuu 6 yu1

DOCUMENTID SUBJECT REV

SUMMARY

10 CFR 50 59(c)(2) CRITERIA ARE MET, AND THAT THIS ESP SUPPLEMENT MAY BE PERFORMED WITHOUT PRIOR NRC APPROVAL.

SE00394 0001 AS A RESULT OF STEAM GENERATOR TUBE DEGRADATION OVER THE LIVES OF THESE COMPONENTS, CALVERT CLIFFS NUCLEAR SAFETY POWER PLANT INCORPORATED (CCNPPI) ELECTED TO REPLACE THE CCNPP UNIT 1 AND UNIT 2 ORIGINAL COMBUSTION LAFTIN F ENGINEERING STEAM GENERATORS (OSGs) WITH REPLACEMENT STEAM GENERATORS (RSGs) MANUFACTURED BY BABCOCK AND EVALUATION FOR WILCOX CANADA (BWC). THE RESULTANT BENEFITS OF THE RSGs ARE IMPROVED CORE HEAT REMOVAL AND INCREASED REACTOR ESP ESi99601526-COOLANT SYSTEM (RCS) FLOW, IMPROVED MARGIN AGAINST STEAM GENERATOR TUBE RUPTURE, REDUCED PERSONNEL 000 RSG EXPOSURE DUE TO FEWER AND LESS FREQUENT STEAM GENERATOR TUBE INSPECTIONS AND REPAIRS, AND INCREASED COMPONENT 10 CFR AVAILABILITY OF UNITS 1 AND 2 DUE TO IMPROVED STEAM GENERATOR TUBE INTEGRITY. DESIGN AND LICENSING BASIS 5059 EVALUATION EVALUATION FOR THE RSG COMPONENTS IN THEIR INSTALLED CONFIGURATIONS IS CONTAINED IN THIS 10 CFR 50 59 EVALUATION AND THE DESIGN INPUT REQUIREMENTS (DIR) OF ENGINEERING SERVICE PACKAGE (ESP) ES199601526-000, REPLACEMENT OF CCNPP STEAM GENERATORS. SEPARATE 10 CFR 50 59 SCREENINGS ARE PERFORMED FOR ESPs ES199601526-116 (-216), UNIT 1 (UNIT 2) RSG MODIFICATION, WHICH ADDRESS THE IMPLEMENTATION ACTIVITIES ASSOCIATED WITH PHYSICAL REMOVAL OF THE OSGs AND INSTALLATION OF THE RSGs. THE PRIMARY BASIS DOCUMENT FOR EVALUATING THE DESIGN AND MANUFACTURE OF THE RSGs IS FTI DOCUMENT NO. 77-5005336, REPLACEMENT STEAM GENERATOR REPORT FOR BALTIMORE GAS AND ELECTRIC UNITS 1 AND 2. THE RSG REPORT DESCRIBES RSG DESIGN AND FABRICATION, AND DESCRIBES THE USE OF THE RSGs AT CCNPP. THE CONTENTS OF THE REPORT SUPPORT THE CONCLUSION THAT THE RSGs WILL SUPPORT NORMAL AND TRANSIENT PLANT OPERATION WITH NO ADVERSE IMPACT AND THAT THE EXISTING LICENSING BASIS IS MAINTAINED WITH THE RSGs. UTILIZING THE RSG REPORT AND OTHER SUPPORTING INFORMATION, THE 10 CFR 50 59 EVALUATION PERFORMED FOR ESP ES199601526-000 CONCLUDED THAT NO 10 CFR 50.59(c)(2) CRITERIA EXISTS THAT WOULD REQUIRE A LICENSE AMENDMENT REQUEST. THEREFORE, THIS ESP SUPPLEMENT MAY BE PERFORMED WITHOUT PRIOR NRC APPROVAL Tuesday, December31, 20(12 rage i oj 10 Tuesda, December 31, 2002 rage 7 oj lo

DOCUMENT ID SUBJECT REV SUMMA R Y SE00413 0002

SUMMARY

SOME ALLOY 600 RCS INSTRUMENT NOZZLES ARE BEING MODIFIED BY ES199800338. THESE NOZZLES ARE CURRENTLY WELDED ON THE INTERIOR OF THE RCS USING A PARTIAL PENETRATION WELD. THIS MODIFICATION WILL INSTALL MECHANICAL INSTALLATION OF NOZZLE SEAL ASSEMBLIES (MNSA) ON EACH OF THE HIGH-RISK INSTRUMENT NOZZLES. THE MNSA PROVIDES BOTH PRESSURE MECHANICAL INTEGRITY AND STRUCTURAL INTEGRITY FOR THE NOZZLE. CONSEQUENTLY, IT REPLACES THE FUNCTION OF THE INTERIOR WELD, NOZZLE SEAL SUCH THAT THE WELD IS NOW CLASSIFIED AS CLADDING. UFSAR FIGURE 4-8 IS AN OUTLINE OF THE PRESSURIZER. THIS FIGURE IS ASSEMBLES (MNSA)

BEING ANNOTATED TO INDICATE THE PRESENCE OF THE MNSA. THERE ARE NO OTHER IMPACTS ON THE UFSAR. THE MNSA IS A ON RCS NOZZLES FULLY QUALIFIED ASME SECTION III DESIGNATED APPURTENANCE. IT IS CONCLUDED THAT THE MNSA HAS NO IMPACT ON THE PROBABILITY OF OCCURRENCE OF ANY ACCIDENT OR MALFUNCTION PREVIOUSLY DESCRIBED IN THE UFSAR. FURTHERMORE, THERE IS NO IMPACT ON THE RCS RESPONSE TO ANY ACCIDENT OR MALFUNCTION. THEREFORE, IT HAS BEEN CONCLUDED THAT THE CONSEQUENCES OF ACCIDENTS OR MALFUNCTIONS PREVIOUSLY DESCRIBED IN THE UFSAR ARE NOT CHANGED A REVIEW OF THE FAILURE MODES AND EFFECTS OF THE MNSA CONCLUDED THAT THEY ARE NO DIFFERENT THAN THOSE OF THE EXISTING WELDED NOZZLE DESIGN. THE ONE EXCEPTION IS THAT A MNSA LEAK DOES NOT RESULT IN THE POTENTIAL FOR NOZZLE EJECTION EVEN IF LEFT UNNOTICED AND UNCORRECTED FOR A LONG PERIOD (I E., MULTIPLE FUEL CYCLES, WHICH IS NOT CREDIBLE). THERE IS NO IMPACT ON THE MARGIN OF SAFETY OF ANY TECHNICAL SPECIFICATION. THE MNSA IS DESIGNED AND FABRICATED IN ACCORDANCE WITH THE 1989 EDITION OF ASME SECTION I1I. THESE ITEMS ARE CLASSIFIED AS AN APPURTENANCE UNDER NCA 9000. THIS MEANS THAT THE MNSA IS AN NPT STAMPED ITEM THAT REQUIRES THE DEVELOPMENT OF A DESIGN REPORT (NCA-3200)

AND A HYDROSTATIC TEST (NB-6000) THESE ITEMS HAVE BEEN CLASSIFIED AS CLASS 1 (I E., REACTOR COOLANT PRESSURE BOUNDARY) IN ACCORDANCE WITH THE CRITERIA IN 50 55A AND 50 2. CONSEQUENTLY, THE PROVISIONS OF ASME SECTION III SUBSECTIONS NCA AND NB ARE APPLICABLE. ABB-CE SUBMITTED AN INQUIRY TO ASME SECTION III REGARDING THE ABILITY TO DESIGN A MECHANICAL CONNECTION FOR NOZZLE ATTACHMENTS. THIS INQUIRY IS NI-97-015 THE ASME CONCURRED WITH THIS INQUIRY.

SE00421 0001 APPROVE CHANGES TO CHARGING AND LETDOWN TEMPERATURES IN CHAPTERS 9 AND 10 OF THE UFSAR.

SUMMARY

. IT WAS IDENTIFIED THAT THE NORMAL CHARGING SYSTEM TEMPERATURES DOWNSTREAM OF THE REGENERATIVE HEAT EXCHANGER ARE HIGHER THAN THE VALUES LISTED IN UFSAR TABLE 9-3. MORE DETAILED REVIEW HAS DETERMINED THAT THE REGENERATIVE HEAT EXCHANGER TRANSFERS HEAT FROM LETDOWN TO THE CHARGING SYSTEM MORE EFFECTIVELY THAN ASSUMED BY THE UFSAR. THEREFORE, THE LETDOWN AND CHARGING TEMPERATURES LIST IN TABLE 9-3 ALL REQUIRE REVISION.

ADDITIONALLY, THE LETDOWN TEMPERATURE INTO THE LETDOWN HEAT EXCHANGER IS LESS THAN THAT SHOWN IN TABLE 9-5.

CONSEQUENTLY, THE COMPONENT COOLING WATER TEMPERATURES SHOWN IN THIS TABLE ARE ALSO REDUCED. THE EFFECTS OF THE REVISED TEMPERATURES HAVE BEEN CONSIDERED WITH REGARD TO COMPONENT QUALIFICATION (E.G., PIPE STRESS),

SYSTEM OPERATION AND CONTROLS, POTENTIAL HIGH ENERGY LINE BREAK EFFECTS, AND SUPPORT SYSTEM IMPACTS. ALL CHANGES WERE FOUND TO HAVE NO ADVERSE IMPACT ON ANY OF THESE AREAS. THEREFORE, IT WAS CONCLUDED THAT THERE IS NO CHANGE IN THE PROBABILITY OR CONSEQUENCES OF ANY MALFUNCTION OR ACCIDENT. FURTHERMORE, NO ITEM IS BEING OPERATED IN A DIFFERENT MANNER OR IN A MANNER THAT IS OUTSIDE ITS NORMAL DESIGN ENVELOPE. THEREFORE, THE POSSIBILITY OF A NEW TYPE OF MALFUNCTION OR ACCIDENT IS NOT CREATED. THE TECHNICAL SPECIFICATIONS AND ASSOCIATED BASIS ARE UNAFFECTED BY THESE CHANGES. THE MARGIN OF SAFETY AS AFFORDED BY THESE DOCUMENTS IS NOT AFFECTED BY THE CHANGES HEREIN. BGE HAS CONCLUDED THAT THIS ACTIVITY DOES NOT REPRESENT AN UNREVIEWED SAFETY QUESTION AS DEFINED BY 10CFR50.59.

Tuesday, December31, 2002 rage 50110 Page 8 oj 16o Tuesday, December 31, 2002

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SUMMARY

SE00442 0001

SUMMARY

THIS ACTIVITY MAKES A CORRECTION TO THE UFSAR, WHICH CURRENTLY DESCRIBES THAT THE UNIT 1 AND UNIT 2 UFSAR CONTAINMENT TENDON GALLERIES ARE EQUIPPED WITH SMOKE DETECTION SYSTEMS. ALTHOUGH THE DETECTION IS NOT ISCPAN CURRENTLY INSTALLED, AND HAS NEVER BEEN INSTALLED, THIS EVALUATION ADDRESSES THE REMOVAL OF SMOKE DETECTION DISCREPANCIES OF SYSTEMS FROM THE UFSAR FOR THE CONTAINMENT TENDON GALLERIES. THE TENDON GALLERIES CONTAIN NO SAFE SHUTDOWN TRM AND TABLE 9-20 OR APPENDIX R COMPONENTS, AND ESSENTIALLY NO COMBUSTIBLE MATERIALS THIS ACTIVITY DOES NOT INCREASE THE PROBABILITY OR CONSEQUENCES OF MALFUNCTION OF ANY SAFE SHUTDOWN EQUIPMENT PREVIOUSLY EVALUATED IN THE UFSAR, NOR DOES IT INCREASE THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED IN THE UFSAR. FINALLY, THIS ACTIVITY DOES NOT CREATE THE POSSIBILITY FOR A MALFUNCTION OR ACCIDENT OF A TYPE NOT PREVIOUSLY EVALUATED BY THE UFSAR, AND DOES NOT INCREASE THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY TECHNICAL SPECIFICATION REQUIREMENT. THERE ARE NO APPLICABLE TECHNICAL SPECIFICATION REQUIREMENTS ASSOCIATED WITH THIS ACTIVITY, SINCE FIRE DETECTION SYSTEMS ARE NOT ADDRESSED IN THE TECHNICAL SPECIFICATIONS. THIS ACTIVITY DOES NOT REPRESENT AN UNREVIEWED SAFETY QUESTION (USQ).

SE00450 0001

SUMMARY

THIS EVALUATION ADDRESSES THE PLANT SAFETY IMPACTS OF A PROPOSED CHANGE in the CONDITIONS AT WHICH PORVs ARE STROKE TESTED. THE CURRENT STP STROKE TESTIS PERFORMED IN MODE 5 AT A PRESSURIZER PRESSURE OF 210 TO STROKE TESTING 224 PSIA WITH DECAY HEAT REMOVAL PROVIDED BY THE SHUTDOWN COOLING SYSTEM. THE CHANGE INVOLVES PERFORMING THE OF PRESSURIZER STROKE TEST AT A HIGHER PRESSURE OF 500 TO 600 PSIA WHILE IN MODES 4 OR 3, WITH 2 (SAME LOOP) OR 4 RCPs IN OPERATION.

PORVS IN MODES 3 THIS PROPOSAL ORIGINATED AS A CORRECTIVE ACTION ASSOCIATED WITH ROOT CAUSE ANALYSIS PD200000006 INVOLVING PORV AND 4 WITH SEAT LEAKAGE AT THE CLOSE OF REFUELING OUTAGES. MANIPULATION OF INSTRUMENT CHANNELS P-1 03/P-1 03-1 DURING THE PRESSURIZER TEST WILL DISABLE: 1) AN OPEN PERMISSIVE INTERLOCK (OPI) ASSOCIATED WITH THE SHUTDOWN COOLING SYSTEM (SDCS)

PRESSURE AT 500 ISOLATION VALVES SI-651 AN SI-652 RESPECTIVELY AND 2) AN OPEN SIGNAL SENT TO SAFETY INJECTION TANK ISOLATION VALVE TO 600 PSIA PAIRS SI-614/624 AND SI-634/644. UNDER THE NEW TEST CONDITIONS, DECAY HEAT REMOVAL WILL BE PROVIDED BY THE STEAM GENERATORS WITH THE REACTOR COOLANT PUMP(S) IN SERVICE. ALL TECHNICAL SPECIFICATIONS APPLICABLE TO THE PLANT MODE AT THE TIME OF THE TEST WILL BE COMPLIED WITH THE PORVs WILL BE TAKEN OUT OF AUTOMATIC CONTROL ONE AT A TIME IN ORDER TO PERFORM THE MANUAL STROKE TESTS UTILIZING ALLOWED OUT OF SERVICE DURATIONS IN THE T.S. LCOs. THIS EVALUATION CONCLUDES THIS ACTIVITY DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION. STROKE TESTING THE PORVs IN MODES 4 OR 3 ALIGNS CCNPP'S APPROACH TO DEMONSTRATING PORV RELIABILITY MORE CLOSELY TO PAST NRC EXPECTATIONS ASSOCIATED WITH GENERIC LETTER 90-06.

SE00451 0000

SUMMARY

This evaluation supports the changes to several plant installed setpoints and monitoring limits for changes due to the Unit 2 Cycle 14 UEC14) Reload, improvements in the safety analysis, needed margin in the LOCA analysis and RCS cold leg temperature differences. The changes UNIT 2 CYCLE 14 are as follows: - Reduction In the PLHR limit from 14.3 kw/ft to 13.9 kwft to obtain positive results for the LOCA analysis - Expansion of the peaking SETPOINTS AND factor tents (Fxy & Fr) between 80-20% power to accommodate potential operating conditions. -Expansion of the DNB LCO tent due to increase LOCA RELOAD analysis margin from the ABB-NV correlation. -An added penalty to the TM/LP trip to account for RCS cold leg temperature differences. These SAFETY EVALUATION changes will be implemented on both Unit 1 and Unit 2. COLR Figures and TS Bases for 3 2.2 & 3.2.3 were revised to include the requirement to operate by the LTIL when the peaking factors exceed 1.65. This requirement was Incorrectly eliminated from the TS during the transition to ITS.

These changes will retum the requirement. All changes have been analyzed In accordance with approved methodology. All safety analysis will continue to perform within their allowable limits.

Tu1aDeebri,

((2rae=01 Tuesda, December 3'1, 2002 rage 9 oj. 6

DOCUMENTID SUBJECT REV

SUMMARY

SE00452 0000

SUMMARY

THIS SAFETY EVALUATION EVALUATED THE OPERATION OF UNIT 2 CYCLE 14 IN MODES 1 THROUGH 6, EXCEPT FOR SETPOINT AND LOCA CHANGES. SAFETY EVALUATION SE00451 WILL EVALUATE THE UNIT 2 CYCLE 14 SETPOINT AND LOCA CHANGES UNIT 2 CYCLE 14 IN MODES I THROUGH 6 SAFETY EVALUATION SE00441 APPROVED THE REDUCTION OF INCORE INSTRUMENTS (ICI) FROM 45 TO 35 RELOAD SAFETY STRINGS, REDUCTION OF CORE EXIT THERMOCOUPLES (CET) FROM 45 TO 35 AND INSERTION OF TWO PLATINUM TEST ICIs. SAFETY EVALUATION (NOT EVALUATION SE00443 APPROVED CREDITING THE NEGATIVE REACTIVITY OF THE CEAs DURING AN INCORE SHUFFLE. UNIT 2 CYCLE INCLUDING 14 (U2C14) WILL BE THE FIRST UNIT 2 CORE TO IMPLEMENT A FULL BATCH OF VALUE ADDED PELLETS FUEL ASSEMBLIES (92 FRESH SETPOINTS OR VAP ASSEMBLIES). UNIT 1 CYCLE 15 WAS THE FIRST FULL BATCH IMPLEMENTATION OF VALUE ADDED PELLET FUEL AT CALVERT LOCA)

CLIFFS. U2C14 IS THE FIRST CYCLE TO USE THE APPROVED ABBNV CRITICAL HEAT FLUX CORRELATION FOR DNB ANALYSES. U2C14 IS THE SECOND RELOAD TO USE THE IMPROVED ENDF/B-VI CROSS SECTION LIBRARY IN LIEU OF THE TRADITIONAL ENDF/B-IV CROSS SECTION LIBRARY. LEAD FUEL ASSEMBLY (LFA) IRT4 WILL BE REINSERTED IN U2CA14 FOR A THIRD CYCLE OF IRRADIATION. THIS LFA CONTAINS ADVANCE FUEL DESIGN FEATURES INCLUDING VALUE ADDED PELLETS, MIXING GRIDS, I-SPRINGS AND ADVANCED CLADDING MATERIAL. MODIFICATIONS TO THE FUEL ASSEMBLY DESIGN AND THE RELOAD CORE DESIGN WERE CONSIDERED IN THE SAFETY ANALYSIS. ERBIUM CONTINUES TO BE USED AS AN INTEGRAL BURNABLE ABSORBER FOR THIS 24 MONTH FUEL CYCLE.

ACTIONS HAVE BEEN TAKEN TO MINIMIZE GRID TO ROD FRETTING WEAR AND SPALLATION. A COLR FOR U2C14 HAS BEEN DEVELOPED PER THE REQUIREMENTS OF TECH SPEC 5.6.5. THE UFSAR AND TRM WILL BE UPDATED WITH THE APPROPRIATE UNIT 1 AND 2 RELOAD DESIGN CHANGES. IT WAS CONCLUDED THAT OPERATION OF UNIT 2 CYCLE 14 IN MODES I THROUGH 6 DOES NOT REQUIRE PRIOR NRC REVIEW.

SE00453 REVISION TO CHAPTER 11 OF UFSAR

SUMMARY

UFSAR SECTIONi 1.1 TEXT IS BEING REVISED TO CLEARLY STATE THE PRESENT METHOD OF OPERATION USED BY PLANT PERSONNEL TO OPERATE THE WASTE SYSTEMS. THIS CHANGE IS COVERED UNDER UCR 209. OTHER UFSAR CHANGES HAVE BEEN MADE AND ARE GROUPED UNDER THE FOLLOWING UCRs THIS OTHER INFORMATION IS BEING PRESENTED FOR COMPLETENESS SINCE MANY OF THE CHANGES OVERLAP SECTIONS OF THE SAR.

A) UCR00l 15: UN-RETIRES EQUIPMENT RETIRED BY THE 1988 BECHTEL SAFETY EVALUATION. NOTE: ES199701700-000 EVALUATES THIS ITEM FOR ACCEPTABILITY.

B) UCR00206: RESOLVES STEAM GENERATOR BLOWDOWN ISSUES IDENTIFIED BY IRl-064-777 (IR199702040) AND IR3-035-029 (IR199901049).

C) UCR00208:

MAKES EDITORIAL CHANGES TO CHAPTER 11 UNDER NEI 98-03 GUIDANCE TO REMOVE EXCESSIVE DETAIL.

D) UCR00209" OTHER CHANGES WHICH NEED TO BE JUSTIFIED UNDER SE00453. THE SOLID, LIQUID AND GASEOUS RADWASTE PROCESSING SYSTEMS AT CALVERT CLIFFS ARE OPERATED IN A MANNER IN WHICH THE RADIOLOGICAL RELEASES FROM THE PLANT ARE MAINTAINED ALARA.

ALL OF THE SYSTEMS ARE DESIGNED WITH VARIOUS PUMPS, TANKS, FILTERS, ION EXCHANGERS AND OTHER COMPONENTS THAT ENABLE THE RADIOACTIVE CONTAMINANTS IN THE WASTE STREAMS TO BE REDUCED PRIOR TO DISCHARGE TO THE ENVIRONMENT.

PERSONNEL AT CALVERT CLIFFS OPERATE THIS EQUIPMENT IN VARIOUS CONFIGURATIONS, AS ALLOWED BY THE UPDATED FINAL SAFETY ANALYSIS REPORT, USING APPROVED PLANT PROCEDURES. BY UTILIZING ALL AVAILABLE PROCESSING OPTIONS, THE SITE IS ABLE TO ROUTINELY MAINTAIN RADIOACTIVE RELEASES TO A SMALL FRACTION OF THE REGULATORY LIMITS.

Tuesday, December31, 2002 Page 1(10110 Tuesday, December 31, 2002 Page 10/o.f 16

DOCUMENTID SUBJECT REV

SUMMARY

SE00454 0000

SUMMARY

ES200100054-000 EVALUATES MOVING THE CHEMICALS APPROVED FOR CONDENSATE TREATMENT FROM THE UFSAR SECTION 10 2.2 AND LISTING THE APPROVED CHEMICALS TO THE APPROPRIATE CHEMISTRY SECTION PROCEDURE, VIA UCR00214.

REVISE SAR CP-0217 IS A CHEMISTRY SECTION CONTROLLED PROCEDURE AND IS UNDER THE UMBRELLA OF THE APPENDIX B PROGRAM.

SECTION 1022 TO ADDITIONALLY, CARBOHYDRAZIDE IS ADDED TO THE LIST OF APPROVED CHEMICALS FOR CONDENSATE TREATMENT. THE ANALYSIS REMOVE DETERMINED THERE IS NO ADVERSE EFFECT TO ANY COMPONENTS ASSOCIATED WITH THE CONDENSATE FEEDWATER SYSTEM BY CHEMICALS USED IN MOVING THE LIST TO CHEMISTRY CONTROLLED PROCEDURE, CP-0217 AND ADDING CARBOHYDRAZIDE to the LIST.

FEEDWATER CARBOHYDRAZIDE IS AN IMPROVEMENT OVER USING HYDRAZINE (ON THE APPROVED LIST), AS IT PASSIVATES IRON AND TREATMENT AND SCAVENGES OXYGEN STARTING AT A TEMPERATURE AS LOW AS 90 DEGREES F AND CARBOHYDRAZIDE IS NOT TOXIC. AT REFERENCE ELEVATED TEMPERATURE, CARBOHYDRAZIDE BREAKS DOWN TO HYDRAZINE AND CARBON DIOXIDE. WITHIN THE LIMITS OF CP-0217, CHEMISTRY TECH NEITHER COMPOUND IS DETRIMENTAL TO ANY MATERIAL USED In the CONDENSATE/FEEDWATER SYSTEM INCLUDING INCONEL 600 PROCEDURES.

AND INCONEL 690. THERE IS NO CHANGE TO ANY EVALUATION TO THE ACCIDENTS EVALUATED IN THE SAR. THE CONSEQUENCES OF AN ACCIDENT WITH RESPECT TO THE PUBLIC CONCERNING RADIOLOGICAL DOSE OR CHEMICAL DISCHARGE DO NOT INCREASE AS A RESULT OF THIS CHANGE.

SE00455 0002 Summary: THIS ESP CHANGES THE SAFETY CLASSIFICATION OF 0-RE-5350 TO SR-PB, AND REASSIGNS CONTACTS FROM RELAYS 0-RY 5350A AND 0-RY-5350B TO SR-1 E RELAYS 1 R35OAXl AND 1 R350B/X1. THROUGH THIS DESIGN CHANGE, CONTROL ROOM ISOLATION CONTROL ROOM AND FILTRATION IN RESPONSE TO MANY UFSAR ACCIDENT SCENARIOS WILL BE ADDRESSED THROUGH SIAS AND CRRMS EMERGENCY INITIATION. FOR THE SINGLE UFSAR ACCIDENT SCENARIO THAT IS NOT MET THROUGH SIAS INITIATION (A FUEL HANDLING VENTILATION ACCIDENT, 'FHA'), THE CONTROL ROOM RADIATION MONITOR WILL CONTINUE TO PROVIDE FILTRATION AND ISOLATION INITIATION.

SYSTEM MODS -

IN THIS CASE, SINGLE-FAILURE-PROOF OPERATION IS NOT REQUIRED SO THE USE OF NSR EQUIPMENT IS SATISFACTORY. ALL SIAS INITIATION SAFETY EVALUATION FORM QUESTIONS HAVE BEEN ANSWERED 'NO', THEREFORE, THIS ACTIVITY DOES NOT DECREASE THE MARGIN OF SAFETY AND IT DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

SE00456 MODIFICATION A SPARE ELECTRICAL PENETRATION FACILITATE VENTING AFTE 0000 Summary: This package modifies a spare 12" electrical penetration, 2ZED09, such that it can be utilized when the plant is at non-power operations to vent containment air at the end of the 10-year Integrated Leakage Rate Testing (ILRT). The spare penetration will not be used during power O0OF operations. During power operations it will retain its safety related status such that the failure effect requirements of containment are maintained This design change utilizes sealed closed barriers in the form of blind flanges that are designed to ASME III Class 2 design criteria and tested to 10 CFR 50 Appendix J leak testing requirements. This modification does not increase the probability of occurrence or Increase the consequences of an TO accident or a malfunction of equipment important to safety previously evaluated in the SAR. This modification does not create a new accident or R ILRT malfunction and does not decrease the margin of safety of any Technical Specification.

SE00462

SUMMARY

DEGRADED THERMAL PERFORMANCE IN THE UNIT 1 STEAM GENERATORS (SGs) HAS MADE IT DIFFICULT TO ACHIEVE RATED FULL POWER OUTPUT. SIMILAR PERFORMANCE ISSUES IN UNIT 2 PROMPTED A CHANGE TO O1-12A TO ALLOW PARTIAL FEEDWATER BYPASS OF 26 NB FEEDWATER HEATERS. THE PROPOSED ACTIVITY WILL PROVIDE UNIT I OPERATIONS WITH THE SAME FLEXIBILITY TEMPERATURE IN ORDER TO MAXIMIZE POWER OUTPUT. THE PROPOSED ACTIVITY HAS BEEN ANALYZED IN ACCORDANCE WITH APPROVED REDUCTION UNIT 1 METHODOLOGY. ALL SAFETY ANALYSES WILL CONTINUE TO PERFORM WITHIN THEIR ALLOWABLE LIMITS. NO NEW FAILURES HAVE BEEN CREATED AS A RESULT OF THE PROPOSED ACTIVITY. THEREFORE, THIS ACTION DOES NOT REQUIRE PRIOR NRC APPROVAL.

Tuesday, December 31, 2002 Page 11 of 16

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SUMMARY

SE00463 0000

SUMMARY

DURING THE PERIOD OF TIME WHILE THE PLANT IS COOLING DOWN AND DEPRESSURIZING, PRIOR TO BEING ABLE TO GO ON SHUTDOWN COOLING, VERY LITTLE CLEAN-UP OF REACTOR COOLANT IS POSSIBLE DUE TO THE HIGH PRESSURE DROP REMOVE INTERNALS ACROSS THE LETDOWN CONTROL VALVES. AFTER THE RCS HAS BEEN BROUGHT DOWN TO A PRESSURE OF 1000 PSIA BOTH FROM 1-CV-110P OR VALVES ARE PLACED IN FULL OPEN. HOWEVER, THE DISK PACKS IN THE VALVES THAT ARE USED TO REDUCE PRESSURE ACROSS 0 DURING THE VALVES PREVENT ALL BUT MINIMAL FLOW. THIS ACTIVITY WILL ALLOW ONE OF THE TWO LETDOWN VALVES TO BE ISOLATED SHUTDOWN TO AND THE DISK PACK TO BE REMOVED SO THAT FLOW CAN BE INCREASED. THE PIPING DOWNSTREAM OF THE LETDOWN VALVES IS IMPROVED CLEAN-DESIGNED FOR 600 PSIG, AND IS OVER PRESSURE PROTECTED FOR 600 PSIG BY 1-RV-345. THE LETDOWN VALVE WITH ITS UP INTERNAL REMOVED, MAY BE PLACED IN SERVICE WHEN THE RCS IS BELOW 600 PSIG. THE OBJECTIVE IS TO MATCH THE LETDOWN FLOW TO THE CHARGING FLOW. THE BACK PRESSURE REGULATING VALVE DOWNSTREAM OF THE LETDOWN VALVES WILL BE USED TO MATCH THE LETDOWN FLOW TO THE FLOW FROM ONE, TWO, OR THREE CHARGING PUMPS. THIS WILL IMPROVE THE CLEAN-UP RATE OF THE REACTOR COOLANT SYSTEM. THIS DOES NOT CREATE A CONDITION ADVERSE TO NUCLEAR SAFETY AND DOES NOT REQUIRE PRIOR NRC APPROVAL.

SE00464 0000

SUMMARY

ESFAS DOOR INTERLOCKS HAVE FAILED FREQUENTLY DUE TO MAINTENANCE PERSONNEL NOT BEING ABLE TO TROUBLESHOOT THE CIRCUITRY EFFECTIVELY. TO PROPERLY TROUBLESHOOT THE INTERMITTENT SPURIOUS ALARMS, THE ESFAS RETIRES IN-PLACE CABINETS WOULD HAVE TO BE TAKEN OUT OF SERVICE FOR EXTENDED PERIODS OF TIME, WITH NO GUARANTEES OF SUCCESSFUL THE ESFAS REPAIRS. THE DOOR INTERLOCKS PERFORM NO SAFETY-RELATED FUNCTIONS FOR THE PLANT AND DO NOT AFFECT THE SAFETY ACTUATION RELATED FUNCTIONS OF ESFAS, ADMINISTRATIVE PROCEDURE AND SKILL OF THE CRAFT PREVENT OPENING MORE THAN ONE CABINET DOOR CABINET DOOR SIMULTANEOUSLY, AND NO OTHER SAFETY-RELATED CABINETS IN THE PLANT POSSESS SIMILAR DOOR INTERLOCK INTERLOCK CIRCUIT CAPABILITIES. FOR THESE REASONS, RETIRING THE ESFAS DOOR INTERLOCKS IS JUSTIFIED. RETIRING THE ESFAS DOOR INTERLOCKS IN-PLACE PROVIDE THE MOST NON-OBTRUSIVE METHOD OF IMPLEMENTING THE MODIFICATION. IN THIS MANNER, NO ADVERSE AFFECTS TO THE SAFETY RELATED FUNCTIONS OF ESFAS WILL RESULT FROM THE IMPLEMENTATION OF THIS ACTIVITY. FOR THESE REASONS, THIS MODIFICATION WILL RETIRE IN PLACE THE ESFAS ACTUATION LOGIC AND ACTUATION RELAY CABINET DOOR INTERLOCKS. THESE INTERLOCKS PREVENT THE SIMULTANEOUS OPENING OF MORE THAN ONE ACTUATION CABINET DOOR. THE INTERLOCK SOLENOIDS WILL BE RETIRED IN PLACE, AND THE MANUAL KEY-LOCKING CAPABILITIES MAINTAINED.

SE00465 0001 THE SUBJECT DESIGN CHANGE (ES200100849), AUTHORIZES THE INSTALLATION OF AN AUTOMATIC SIGNAL TO SHUT THE UNIT 1 AND UNIT 2 STEAM GENERATOR SURFACE AND BOTTOM BLOWDOWN CONTAINMENT ISOLATION VALVES (1 (2)CV4010, 1(2)CV4011, MODIFY CONTROL 1(2)CV4012, AND 1(2)CV4013), UPON RECEIPT OF AN AUXILIARY FEEDWATER ACTUATION SIGNAL (AFAS). A PERMISSIVE FROM THE CIRCUIT TO SG VALVES' HANDSWITCHES, CLOSED POSITION, IS REQUIRED TO RESET AFAS IN THE CONTROL ROOM. THE ABILITY TO RESET AFAS IN BLOWDOWN THE CABLE SPREADING ROOM IS NOT CHANGED. THIS MODIFICATION DOES NOT CONSTITUTE AN UNREVIEWED SAFETY QUESTION.

CONTROL VALVES THIS MODIFICATION DOES NOT REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS BASES. THIS MODIFICATION DOES TO CLOSE ON A REQUIRE A CHANGE TO THE UFSAR.

AFAS INITIATION.

Tuesday, December31, 2002 1'age 120110 lae 12o 0 Tuesday, December 31, 2002

DOCUMENT ID SUBJECT REV

SUMMARY

SE00467 0000

SUMMARY

AFTER EACH REFUELING, STARTUP PHYSICS TESTS ARE CONDUCTED TO VERIFY THAT THE OPERATING CHARACTERISTICS OF THE CORE ARE CONSISTENT WITH THE DESIGN. THESE TESTS ARE DESCRIBED IN CHAPTER 13 4 OF THE CHANGE REVIEW UFSAR. FOR THE POWER DISTRIBUTION MEASUREMENTS, POWER IS DETERMINED IN EACH OF THE FUEL ASSEMBLIES, AND CRITERIA FOR COMPARED TO PREDICTED VALUES. THE DEVIATIONS BETWEEN THE MEASURED AND PREDICTED VALUES ARE COMPARED TO STARTUP TESTS REVIEW CRITERIA. THE POWER DISTRIBUTION MEASUREMENT IS PERFORMED AT 30% RTP, 60% RTP, 85% RTP, AND FULL POWER. THE POWER DISTRIBUTION REVIEW CRITERIA IS CHANGED FROM +1-15% FOR INTERIOR FUEL ASSEMBLIES, +/-20% FOR PERIPHERAL FUEL ASSEMBLIES TO +/-15% / +/-0.15 RPD FOR ALL FUEL ASSEMBLIES FOR THE 30% RTP MEASUREMENT; AND CHANGED FROM +/-10% FOR INTERIOR FUEL ASSEMBLIES, +1-15% FOR PERIPHERAL FUEL ASSEMBLIES TO +1-10% / +/-0.10 RPD FOR ALL FUEL ASSEMBLIES FOR THE 60% RTP, 85% RTP, AND FULL POWER MEASUREMENTS.

SE00468 0000

SUMMARY

THIS SAFETY EVALUATION CONSIDERS THE REDUCTION OF THE NUMBER OF INSTALLED INCORE INSTRUMENTS (BOTH ICIs AND CETs) IN CALVERT CLIFFS UNIT 1. AN EVALUATION WAS PERFORMED TO SHOW THAT THE ELIMINATION OF 10 INCORE REDUCTION OF ICr's INSTRUMENTATION ASSEMBLIES FROM THIS UNIT WOULD NOT IMPACT THE ONLINE CORE MONITORING SYSTEM OR INADEQUATE AND CETS FROM 45 CORE COOLING SYSTEM FROM PERFORMING THEIR FUNCTIONS. THIS EVALUATION ADDRESSES ALL OF THE NRC REQUIREMENTS TO 35 FOR UNIT 1 FOR CHANGES TO INCORE DETECTION SYSTEM - DETECTION OF A FUEL MISLOADING, VALIDITY OF TILT ESTIMATES, ADEQUATE (2002 RFO)

CORE COVERAGE, ENSURING UNCERTAINTIES WILL MEET TS LIMITS, AND RESTORATION OF THE SYSTEM UPON REFUELING. THIS EVALUATION HAS SHOWN THAT REDUCING FROM 45 TO 35 INCORE INSTRUMENT ASSEMBLIES CONFIGURATION WILL NOT RESULT IN MORE THAN A MINIMAL INCREASE IN THE PROBABILITY OR CONSEQUENCES OF MALFUNCTION, IN MORE THAN A MINIMAL INCREASE IN THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED IN THE SAR, IN A NEW MALFUNCTION OR ACCIDENT, OR CONSTITUTE A CHANGE IN METHODOLOGY. THIS IS ALSO TRUE OF THE IMPLEMENTATION OF THE TWO PLATINUM TEST DETECTOR STRINGS. AS A RESULT, THE PROPOSED MODIFICATIONS DO NOT CONSTITUTE AN ACTIVITY THAT REQUIRES NRC PRIOR REVIEW.

SE00469 0000

SUMMARY

THIS ACTIVITY IS TO ALLOW THE IMPELLER DIAMETER OF THE CONTAINMENT SPRAY PUMPS TO BE CHANGED FROM 10 1/16 INCHES TO 10-1/4 INCHES. THE NEED TO CHANGE THE IMPELLER DIAMETER WAS IDENTIFIED UNDER ISSUE REPORT 1R3-005-690 CHANGE DIAMETER WHICH IDENTIFIED THAT THE UNIT 2 CONTAINMENT SPRAY PUMPS HAD ALREADY BEEN INSTALLED WITH 10-1/4 INCH IMPELLERS OF CONTAINMENT DURING START-UP TESTING, BUT THAT DESIGN DOCUMENTATION WAS NOT APPROPRIATELY UPDATED AT THAT TIME. ALSO, ISSUE SPRAY PUMP REPORT IR3-028-403 IDENTIFIED THE NEED FOR MORE PERFORMANCE MARGIN FROM THE UNIT 1 PUMPS WHICH HAD THE 10-1/16 IMPELLER FROM 10 INCH DIAMETER IMPELLER. THE MAIN EFFECT OF THE LARGER DIAMETER IMPELLER IS TO INCREASE THE FLOW FROM THE 1/6 INCHES TO 10 1/4 CONTAINMENT SPRAY PUMP. THIS PROVIDES A POSITIVE BENEFIT FOR THE LOCA, MSLB EVENTS, AS WELL AS FOR THE CORE INCHES. REVISE FLUSH DUTY OF THIS PUMP DURING POST-LOCA OPERATION. TOO MUCH CONTAINMENT SPRAY FLOW CAN ADVERSELY AFFECT THE FIGURE 6-5 TABLE CONDITIONS ON MINIMUM ECCS BACKPRESSURE DURING A LOCA; HOWEVER, IT WAS SEEN THAT THE INCREASED FLOW RESULTING 6-6, AND SECTION FROM THE LARGER IMPELLER DIAMETER IS STILL WELL BELOW THE CONSERVATIVE MAXIMUM VALUE ASSUMED IN THE ACCIDENT 6.4.5 OF THE UFSAR.

ANALYSIS. THE IMPACT OF THE HIGHER HEAD OF THE LARGER DIAMETER IMPELLER ON THE SYSTEM PRESSURE WAS REVIEWED, AND FOUND NOT TO EXCEED THE DESIGN RATING OF THE PIPING SYSTEM. THE EFFECT OF THE HIGHER BHP REQUIREMENTS OF THE PUMP WITH THE LARGER DIAMETER IMPELLER WERE REVIEWED, AND FOUND TO BE BOUNDED BY EXISTING INPUTS TO THE DIESEL LOADING CALCULATIONS AND SYSTEM RELAY DESIGN REQUIREMENTS. FINALLY, THE EFFECT OF THE HIGHER FLOW FROM THE PUMP HAS BEEN INCORPORATED INTO THE RWT VORTEXING ANALYSES AND MINIMUM TIME TO RAS ANALYSES. THEREFORE, THIS CHANGE IS ACCEPTABLE, AND DOES NOT RESULT IN AN UNREVIEWED SAFETY QUESTION.

TusaDcmbr1 02raeii01 Page 13 oj I c Tuesday, December 31, 2002

DOCUMENT ID SUBJECT REV

SUMMARY

SE00470 0000 SE00470 EVALUATES THE PHYSICAL, MECHANICAL, AND METHODOLOGY CHANGES BEING IMPLEMENTED WITH THE NEW FUEL PURCHASED FOR THE 2002 REFUELING OUTAGE (BATCH IV/ UIC16). THE SIGNIFICANT CHANGES INCLUDE FULL BATCH UNIT 1 CYCLE 16 IMPLEMENTATION OF THE TURBO GRIDS (MIXING VANES AND I-SPRINGS), THE FIRST USE OF THE WESTINGHOUSE STANDARD ZIRLO FUEL ASSEMBLY FUEL CLADDING, AND NUMEROUS PROCESS AND DESIGN CHANGES TO THE FUEL PINS THAT ARE NECESSITATED BY THE MECHANICAL TRANSITIONING OF FUEL MANUFACTURING OPERATIONS TO THE WESTINGHOUSE COLUMBIA FACILITY. THE SEISMIC DESIGN CHANGES METHODOLOGY FOR ALL CATEGORY I EQUIPMENT LOCATED IN CONTAINMENT (INCLUDING THE NSSS AND FUEL ASSEMBLIES) HAS BEEN UPGRADED TO A 3-D METHOD. PRIOR NRC APPROVAL IS NOT REQUIRED BECAUSE THE DIFFERENT METHOD IS ONE THAT HAS ALREADY BEEN REVIEWED AND APPROVED BY THE NRC FOR THE INTENDED APPLICATION. TO SUPPORT ZIRLO CLAD IMPLEMENTATION, CALVERT CLIFFS IS REQUIRED TO PROVIDE THE NRC WITH A PROPOSED SITE SPECIFIC ZIRLO FUEL DUTY LIMIT.

ADDITIONALLY, THE NRC MUST ISSUE A TECH SPEC CHANGE TO ADD A REFERENCE TO THE ZIRLO TOPICAL REPORT TO OUR TECH SPEC "LIST OF APPROVED METHODOLOGIES". PRIOR NRC REVIEW AND APPROVAL IS NOT REQUIRED FOR ANY OTHER ACTIVITIES BEYOND THE TECH SPEC CHANGE ADDRESSED IN THE PREVIOUS PARAGRAPH.

SE00471 0000 SE00471 EVALUATED THE OPERATION OF UNIT 1 CYCLE 16 IN ALL PLANT MODES. UNTIL ISSUES ASSOCIATED WITH THE POST-TRIP SLB ARE RESOLVED, THE CYCLE LENGTH IS LIMITED TO 5,000 MWD/MTU. A FUTURE REVISION TO THIS 50.59 WILL JUSTIFY THE UNIT 1 CYCLE 16 REMAINDER OF THE 24 MONTH FUEL CYCLE. THE ANALYSES ACCOUNTED FOR THE UICI6 FUEL MANAGEMENT AND ALL OF THE RELOAD PHYSICS PHYSICAL CHANGES IMPLEMENTED IN THE BATCH 1V FRESH FUEL ASSEMBLIES (TURBO GRIDS, ZIRLO CLAD, AND FUEL PIN DESIGN AND TRANSIENTS CHANGES). THE EFFECTS OF REPLACEMENT STEAM GENERATORS HAVE BEEN INCLUDED IN ALL ANALYSES. NO RPS SETPOINT SAFETY EVALUATION CHANGES WERE REQUIRED A COLR FOR UIC16 HAS BEEN DEVELOPED PER THE REQUIREMENTS OF TECH SPEC 5 6.5. THE EVALUATION CONCLUDED THAT NRC APPROVAL WAS REQUIRED FOR THE FOLLOWING ACTIVITIES. NRC ISSUE TECH SPEC CHANGE TO ADD ZIRLO TOPICAL REPORT TO THE TECH SPEC "LIST OF APPROVED METHODOLOGIES" (TS 5 6 5). (NRC APPROVED ON 4/8/02) NRC ISSUE TECH SPEC CHANGE TO ADD LOCA TOPICAL REPORTS TO THE TECH SPEC "LIST OF APPROVED METHODLOOGIES" (TS 5 6 5). (NRC APPROVED ON 4/8/02) NRC APPROVE THE "LOSS OF FEEDWATER EVENT'. (NRC APPROVED ON 2/26/02) NRC ISSUE TECH SPEC CHANGE FOR MINIMUM RCS FLOWRATE FROM 340,000 GPM TO 370,000 GPM. (NRC APPROVED ON 3/1/02) NRC ISSUE TECH SPEC CHANGE TO STEAM GENERATOR LEVEL-LOW TRIP SETPOINT TO BECOME >/=50" BELOW NORMAL WATER LEVEL. (NRC APPROVED ON 3/1/02)

SE00472 AN ERROR WAS DISCOVERED IN THE SAFETY ANALYSIS WHICH IMPACTS ALL CHAPTER 14 EVENTS IN WHICH AN INITIAL AMOUNT OF THERMAL MARGIN MUST BE PRESERVED IN ORDER TO PROTECT EITHER NOT EXCEEDING THE DNB SAFDL OR LIMIT THE COLR AND BASSS CONSEQUENCES OF THOSE EVENTS WHICH DO EXCEED THE SAFDL THIS MARGIN IS REFERRED TO AS REQUIRED OVER POWER REVISION TO MARGIN (ROPM). THE ERROR RESULTED IN THE NEED FOR ADDITIONAL ROPM THAN ORIGINALLY CALCULATED. IN ORDER TO IMPLEMENT MORE PRESERVE THE ADDITIONAL NEEDED ROPM THE PROPOSED ACTIVITY WILL REVISE COLR FIGURE 3 2.3 AND BASSS SETPOINTS TO RESTRICTIVE FR REDUCE THE ALLOWED FrT. NO ADDITIONAL MARGIN CREDITS ARE NECESSARY. ALL CHANGES HAVE BEEN ANALYZED IN LIMITS ACCORDANCE WITH APPROVED METHODOLOGY. ALL SAFETY ANALYSES WILL CONTINUE TO PERFORM WITHIN THEIR ALLOWABLE LIMITS.

Tuesday, December31, 2002 ragei oj io

-age I4 o j 10 uesday, December 31, 2002

DOCUMENT ID SUBJECT REV

SUMMARY

SE00473 0000 THE REASONS FOR THIS ACTIVITY ARE CHANGES IN THE CONTAINMENT COATING SYSTEM AND IN THE STEEL HEAT SINK. THIS ACTIVITY REQUIRES THE REANALYSIS OF THE CONTAINMENT RESPONSE TO THE LOSS OF COOLANT AND MAIN STEAM LINE BREAK CONTAINMENT ACCIDENTS (LOCA AND MSLB), AND THE ASSOCIATED UPDATE TO UFSAR CHAPTER 14.20 AND OTHER AFFECTED PORTIONS OF THE RESPONSE TO DBA UFSAR AND TECHNICAL SPECIFICATION BASES. BOTH THE REPLACEMENT STEAM GENERATOR AND THE ORIGINAL STEAM FOR REPLACEMENT GENERATOR DESIGNS WERE CONSIDERED LOCA AND MSLB ARE THE DESIGN BASIS ACCIDENTS (DBAs) THAT ESTABLISH THE STEAM ACCEPTABILITY OF THE CONTAINMENT DESIGN AND DETERMINE THE ENVIRONMENTAL CONDITIONS FOR THE DESIGN OF SAFETY GENERATORS AND RELATED COMPONENTS WITHIN THE CONTAINMENT. CALVERT CLIFFS NUCLEAR POWER PLANT, INC. (CCNPPI) PERFORMED THE REVISED COATING ANALYSIS WITH THE STATE-OF-THE ART CONTAINMENT CODE GOTHIC, USING CONSERVATIVE INPUTS AND ASSUMPTIONS. THE SYSTEM REANALYSIS DEMONSTRATES THAT THE CONTAINMENT DESIGN PRESSURE OF 50 PSIG IS NOT EXCEEDED. THE REANALYSIS ALSO DEMONSTATES THAT THE TEMPERATURE OF THE INNER SURFACE OF THE CONTAINNMENT WALL, OF STRUCTURAL MEMBERS IN CONTAINMENT, AND OF SAFETY RELATED COMPONENTS IN CONTAINMENT REMAIN WELL BELOW THE RESPECTIVE DESIGN TEMPERATURES FOR THESE COMPONENTS. TO ENSURE THE QUALIFICATION OF THE STATE-OF-THE-ART CONTAINMENT CODE GOTHIC FOR SAFETY AND LICENSING ANALYSIS OF THE CONTAINMENT RELATED ISSUES, CCNPPI PERFORMED AN EXTENSIVE BENCHMARK AGAINST THE NRC APPROVED COPATTA CODE FOR ALL DBAs SHOWING EXCELLENT AGREEMENT. CCNPPI DOCUMENTED THE BENCHMARK ACTIVITIES IN A QUALITY ASSURED DESIGN CALCULATION. APPLICATION OF THE GOTHIC CODE TO RESOLVE LICENSING ISSUES RELATED TO THE CONTAINMENT RESPONSE IS IN FULL COMPLIANCE WiTH THE SUPPLEMENT 1 TO GENERIC LETTER 83-11. CCNPPI ALSO REVIEWED SERs ISSUED TO McGUIRE AND FARLEY NUCLEAR PLANTS REGARDING NRC APPROVAL OF GOTHIC FOR CONTAINMENT RESPONSE ANALYSIS. THE NRC WAS INFORMED OF CCNPPI USE OF GOTHIC FOR CONTAINMENT ANALYSIS IN A LETTER DATED FEBRUARY 15, 2002.

SE00474 SUPPORTTA 2-01 0039 THIS ACTIVITY EVALUATES DISABLING A PRIMARY SYSTEM RESISTANCE TEMPERATURE DETECTOR (RTD) INPUT TO THE REACTOR PROTECTION SYSTEM (RPS). THE SYSTEM AVERAGES THE TEMPERATURE OF THE 11 AND 12 HOT LEG RTDs FOR FOUR DIFFERENT CHANNELS. THIS ACTIVITY REMOVES ONE RTD FROM ONE OF THE CHANNELS SO THAT THREE OF THE CHANNELS WILL STILL HAVE A Thot AVERAGES (TWO INPUTS) AND ONE CHANNEL WILL HAVE A SINGLE Thot INPUT. TO ACCOMPLISH THIS, THE OUTPUT OF THE TEMPERATURE TRANSMITTER INTO RPS WILL BE DISABLED. THIS ACTIVITY WILL MAINTAIN FOUR OPERABLE RPS CHANNELS AND TWO OPERABLE SUB-COOLED MARGIN MONITORS (SCMM). THE FUNCTION OF THE RPS WILL BE UNAFFECTED BECAUSE THE TWO HOT LEGS ARE AT APPROXIMATELY THE SAME TEMPERATURE, THEREFORE SYMMETRICAL EVENTS WILL BE DETECTED BY THE SINGLE RTD. THERE ARE NO ASYMMETRICAL EVENTS THAT RELY ON Thot INPUTS TO TRIP THE REACTOR. BASED ON THE ABOVE EVALUATION, THIS ACTIVITY DOES NOT INCREASE THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT OR MALFUNCTION PREVIOUSLY EVALUATED IN THE SAR, NOR DOES IT CREATE A NEW TYPE OF ACCIDENT OR MALFUNCTION NOT PREVIOUSLY EVALUATED IN THE SAR. THIS ACTIVITY DOES NOT RESULT IN A REDUCTION OF THE MARGIN OF SAFETY IN THE TECHNICAL SPECIFICATIONS, THEREFORE THIS ACTIVITY IS NOT AN UNREVIEWED SAFETY QUESTION.

Tuesday, December ii, UY2 rage i., oj io Tuesday, December 31, 2002 I-age Ii oji o

DOCUMENT ID SUBJECT REV S UMMAR Y SE00475 0000 THIS ACTIVITY INCREASES THE RCS MAXIMUM ALLOWED LITHIUM CONCENTRATION FROM 3 50 PPM TO 5 25 PPM. THIS CHANGE IS APPLICABLE TO BOTH UNIT 1 AND UNIT 2. RAISING THE MAXIMUM ALLOWED [Li] TO 5 0 PPM WILL ENABLE BOTH UNITS TO START INCREASE RCS EACH CYCLE WITH A HIGHER pH AND REACH ANY MID-CYCLE pH TARGETS SOONER. THIS WILL REDUCE RCS CORROSION RATES LITHIUM LIMIT TO 50 WHEN THEY ARE MOST CRITICAL; BOG WHEN PROTECTIVE OXIDE LAYERS ARE BEING FORMED. IT WILL ALSO MINIMIZE MID-CYCLE PPM pH CHANGES, WHICH CAN DISTURB EXISTING PROTECTIVE OXIDE LAYERS AND CAUSE EXCESSIVE CRUD RELEASES. THE REDUCTION IN RCS CORROSION RATES WILL ALLOW CCNPP TO MAXIMIZE RCS CHEMISTRY CONTRIBUTION TO RCS INTEGRITY, WHILE MINIMIZING OUT-OF-CORE RADIATION LEVELS AND THE POTENTIAL FOR AXIAL OFFSET ANOMALY WITHOUT ADVERSELY IMPACTING CLADDING CORROSION.

SE00476 0000 INCREASING THE STACK HEIGHT OF THE AG-IN-CD NEUTRON ABSORBER CURRENTLY BEING USED IN THE FULL LENGTH FULL STRENGTH CEA TIPS FROM EIGHT INCHES TO TWELVE INCHES WILL RAISE THE LOWEST B4C PELLET FOUR INCHES. RAISING THE REPLACEMENT LOWEST B4C PELLET FOUR INCHES WILL EXTEND THE TIME IT TAKES IT TO ACCUMULATE SUFFICIENT FLUENCE AND SWELL SUCH CEAS WITH 12" AG-THAT IT INDUCES LIMITING CEA CLADDING STRAIN. THIS WILL EXTEND THE SERVICE LIFE OF THE CEAs ALLOWING CCNPPI TO ONLY IN-CD SOLID SLUG REPLACE THEM ONCE MORE BEFORE END OF PLANT OPERATION. REPLACING FOUR INCHES OF B4C WITH AG-IN-CD NEUTRON LENGTH ABSORBER RESULTS IN APPROXIMATELY A 0.3% DECREASE IN THE NEUTRON ABSORPTION CAPACITY OF THE CEA. NO APPRECIABLE IMPACT WAS DETERMINED FOR NOMINAL OPERATING CONDITIONS. FOR SEVERE ASI CONDITIONS, A SMALL DECREASE IN SCRAM WORTH WAS DETERMINED. HOWEVER, THE GENERIC SCRAM WORTHS DETERMINED WHEN CCNPPI ADOPTED THE BOUNDING ANALYSIS PROCESS, REMAIN BOUNDING. A POST-TRIP STEAM LINE BREAK EVALUATION SHOWED THAT A Fq PENALTY WOULD BE REQUIRED, DEEPENDING ON WHICH COMPUTER CODE WAS USED FOR ANALYSIS. CHANGES TO THE BOUNDING ANALYSIS HAVE BEEN INITIATED TO ENSURE THESE CRITERIA ARE CAPTURED FOR FUTRUE CORE RELOADS. ANALYSIS FOR THE CURRENT UNIT 1 CYCLE 16 RELOAD HAVE SHOWN THE CEA CHANGE TO HAVE NO APPRECIABLE IMAPCT.

SE00477 0000 THIS ACTIVITY REVISED THE UNIT 1 CYCLE 15 (UlC15) CORE OPERATING LIMITS REPORT (COLR) REFUELING BORON CONCENTRATION (RBC) TO REMOVE THE REQUIREMENT TO CREDIT THE NEGATIVE REACTIVITY OF CEAs BEYOND A CORE BURNUP ELIMINATION OF OF 16731 MWD/MTU. THE OPERATORS WILL CONTINUE TO HAVE AT LEAST 33 MINUTES FROM THE TIME OF INITIATION OF THE MODE CEA CREDITING 6 BORON DILUTION EVENT TO TERMINATE THE DILUTION AND PRESERVE THE MINIMUM SPECIFIED SHUTDOWN MARGIN.

FROM REFUELING BORON CONCENTRATION FOR EOC U1C15 TO SUPPORT CORE OFFLOAD

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=a I tIJi Tuesday, December 31, 2002

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