ML20046D031

Monthly Operating Repts for July 1993 for Dresden Nuclear Power Station

ML20046D031
Person / Time
Site:	Dresden
Issue date:	07/31/1993
From:	COMMONWEALTH EDISON CO.
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ML20046D028	List: ML20046D027 ML20046D031
References
NUDOCS 9308160055
Download: ML20046D031 (125)

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{{#Wiki_filter:_- f i MONTHLY NRC

SUMMARY

OF OPERATING EXPERIENCE, CHANGES, TESTS, AND EXPERIMENTS PER REGULATORY GUIDE 1.16 AND le CHL 50.59 FOR DRESDEN NUCLEAR POWER STATION COMMONWEALTII EDISON COMPANY IDR July,1993 UNIT DOCKET LICENSE { 1 050-010 DPR-2 k 2 050-237 DPR-19 3 050-249 DPR.25 .t i b T 'l t 4 l t 9308160055 93081i q5 DR ADOCK 05000010 23 PDR p

TABLE OF CONTENTS July,1993 NRC REPORT 1.0 Introduction 2.0 Summary of Operating Experience 2.1 Unit 2 Monthly Operating Experience Summary. 2.2 Unit 3 Monthly Operating Experience Summary. 3.0 Opwating Data Statistics 3.1 Monthly Operating Data Report - Unit 2 3.2 Monthly Operating Data Report - Unit 3 3.3 Average Daily Power Level Data - Unit 2 3.4 Average Daily Power Level Data - Unit 3 3.5 Unit Shutdown and Power Reduction Data - Unit 2 3.6 Unit Shutdown and Power Reduction Data - Unit 3 3.7 Station Maximum Daily Load Data 4.0 Unique Reporting Requimnents 4.1 Main Steam Relief and/or Safety Valve Operations - Unit 2 and Unit 3 4.2 OIT. Site Dose Calculation Manual Changes 4.3 Major Changes to the Radioactive Waste Treatment 4.4 Failed Fuel Element Indications 4.4.1 Unit 2 4.4.2 Unit 3 5.0 Plant or Pmmdure Changes, Tests, Experimmis, and Safety-Rdated Maintenance 5.1 Amendments to Facility License or Technical Specifications 5.1.1 Unit 2 5.1.2 Unit 3 5.2 Changes to Procedures which are Described in the Final Safety Analysis Report (FSAR) (Units 2 and 3) 5.3 Significant Tests and Experiments Not Descriled in the FSAR (Units 2 and 3) 5.4 Safety Related Maintenance (Units 2 and 3) 5.5 Completed Safety-Related Modifications 5.6 Temporary System Alterations Installed 5.7 Other Required 10 CFR 50.59 Evaluations (Units 2 and 3)

1.0 Introduction Dresden Nuclear Power Station is a three reactor generating facility owned and operated by the Commonwealth Edison Company of Chicago, Illinois. Dresden Station is located at the confluence of the Kankakee and Des Plaines Rivers, in Grundy County, near Morris, Illinois. Dresden Unit 1 is a General Electric Boiling Water Reactor with a design net electrical output rating of 200 megawatts electrical (MWe). The unit is retired in place with all nuclear fuel removed from the reactor vessel. Therefore, no Unit 1 operating data is provided in this report. Dresden Units 2 and 3 are General Electric Boiling Water Reactors with design net electrical output ratings of 794 MWe each. Waste heat is rejected to a man-made cooling lake using the Kankakee River for make-up and the Illinois River for blowdown. The Architect-Engineer for Dresden Units 2 and 3 was Sargent and Lundy of Chicago, Illinois. This report for July,1993, was compiled by Kevin W. Sykes of the Dresden Regulatory Assurance Staff, telephone number (815) 942-2920, extension 2704. n i l l

2.0

SUMMARY

OF OPERATING EXPERIENCE R)R July,1993 4 2.1 UNIT 2 MONTHLY OPERATING EXPERIENCE

SUMMARY

l t i 07/01/93 to 07/31/93 Unit 2 entered the month critical and on line and continued through the end of the month. Reactor power was limited to approximately 730 MWe throughout the month, however,. due to reactor level oscillations. This oscillation problem is being investigated under Problan j Investigation Report (PIR) 12-2-93498 (NTS #237-200-9349800). i At 0950 hours on July 25, a lightning strike caused a fire on MCC51, making it necessary to ~ i place the plant in the lake bypass mode of operation. Unit load was reduced from 725 MWe i to 530 MWe in an attempt to maintain the discharge cooling take blowdown temperature below the NPDES limit. The limit was exceeded, however, for 7 minutes. This event is heing investigated under PIR 12-2-93499 (NTS #237-200-9349900). l 1 i .l a i

2.0

SUMMARY

OF OPERATING EXPERIENCE FOR July,1993 2.2 UNIT 3 MONTIILY OPERATING EXPERIENCE $UMMARY 07/01/93 to 07/31/93 Unit 3 scrammed at 0358 hours on July 10,1993 due to condenser low vacuum. This event was reported to the NRC under Licensee Event Report (LER) 050249/93-014. Unit startup occurred on July 15,1993 (reactor critical at 1207 hours on July 15; generator closed to grid at 0714 hours on July 16). At 0950 hours on July 25, a lightning strike caused a fire on MCC51, making it necasary to place the plant in the lake bypass mode of operation. Unit load was reduced from 779 MWe to 533 MWe in an attanpt to maintain the discharge cooling lake blowdown temperature below the NPDES limit. The limit was exceeded, however, for 7 minutes. This event is being investigated under Problan Investigation Report (PIR) 12-2-93-099 (NTS #237-200 - 09900).

I l 3.0 OPERATING DATA REPORT 3.1 OPERATING DATA REPORT - DRESDEN UNIT TWO DOCKET No. 050-237 DATE August 1,1993 COMPLETED BY K. W. Sykes TELEPIIONE (815) 942-2920 OPERATING STATUS

1. REPORTING PERIOD: July,1993

2. CURRENTLY AUTIIORIZED POWER LEVEL (MWth): 2,527 MAXIMUM DEPENDABLE CAPACITY (MWe NET): 772 DESIGN ELECTRICAL RATING (MWe Net): 794

3. POWER LEVEL TO WIIICII RESTRICTED (IF ANY) (MWe Net): N/A

4. REASONS FOR RESTRICTIONS (IF ANY): N/A j

REPORTING PERIOD DATA CUMULATIVE l FARAMETER Tills MONTII YFAR TO DATE 5. IlOURS IN PERIOD 744 5087 202,775 6. TIME REACTOR CRITICAL Gloun) 744 2016.7 150,844.4 j ] 7. TIME REACTOR RESERVE S11UTDOWN Oloun) 0 0 0 8. TIME GENERATOR ON.LINE Gloun) 744 1941 144,430 9. TIME GFNERATOR RESERVE SIIUTDOWN Gloun) 0 0 0 10. TIIERMAL ENERGY GENERATED (MWII Grou) 1,652.937 3.816,279 297,851,099 11. ELECTRICAL ENERGY GENERATED (MMlle Grou) 519,162 1,201.414 95,066,892 13. ELECTRICAL FNERGY GENERATED OfWile NN) 491,854 1.130,944 90,878,786 13. REACTOR SERVICE FACTOR (%) 100 39.6 74.4 14. REACTOR AVAILABILITY FACTOR (%) 100 39.6 74.4 15. GENERATOR SERVICE FACTOR (%) 100 38.2 71.2 16. GENERATOR AVAILABILITY FACTOR (%) 100 38.2 71.2 17. CAPACITY FACTOR (ESING MDC Nd) (%) RS.6 28.8 58.1 18. CAPACITY FACTOR (USING DER NN) (%) 83.3 28.0 56.4 19. IORCED OUTAGE FACTOR (%) 0 3.5 12.1 20. SilUTDOWNS SCIIEDULED OVER TIIE NEXT 6 MONTIIS (Type, Date and Duration of Each) NONE. 21. IF SillJrDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP N/A I l

3.0 OPERATING DATA REPORT 3.2 OPERATING DATA REPORT - DRESDEN UNIT TilREE DOCKET No. 050-249 DATE August 1,1993 COh!PLETED BY K. W. Sykes TELEPIIONE (815) 942-2920 OPERATING STATUS

1. REPORTING PERIOD: July,1993

2. CURRESTLY AUTIIORIZED POWER LEVEL (51Wth): 2,527 51AX151Uh1 DEPENDABLE CAPACITY (h1We Net): 773 DESIGN ELECTRICAL RATING (h!We Net): 794

3. POWER LEVEL TO WIllCII RESTRICTED (IF ANT) (h!We Net): N/A

4. REASONS FOR RESTRICTIONS (IF ANY): N/A i

REPORTING PERIOD DATA 5. IIOURS IN PERIOD 744 5087 193.104 6. TIME REACTOR CRITICAL (floun) 615 3.443.8 139.033.1 7 TofE REACTOR RESERVE SIIUTDOWN Gloun) 0 0 0 8. TIME GENERATOR ON-LINE Gloun) 597 3.368.8 133.642.2 C. TIME GI'NERATOR RESERVE SI!UTDOWN Oloun) 0 0 0 10. TIIERMAL ENERGY GENERATED (MWilt Gross) 1.339,442 7.716,703 275.246.158 11. ElICTRICAL ENERGY GENERATED (MWIIe Grms) 424,413 2.467.169 88.444,376 13. ELFETRICAL ENERGY GENERATED ofWIIe NN) 401.487 2,372,225 83.982.405 13. REACTOR SERVICE FACTOR (%) 82.7 67.7 72.0 14. REACTOR AVAILABILITY FACTOR (4) 82.7 67.7 72.0 5. GENERATOR SERVICE FACTOR (%) 80.2 66.2 69.2 16. GENERATOR AVAILABILITY FACTOR (4) 80.2 66.2 69J 17 CAPACITY FACTOR (USING MDC NW)(%) 69.8 60.3 56.3 18. CAPACITY FACTOR (USING DER Net) (%) 68.0 58.7 54.8 19. IT)RCED OUTAGE FACTOR (%) 19.8 33.8 12.0 20. SHUTDOWNS SCHEDULED OVER THE NEIT 6 MONTHS (Type, Date and Duration of Each) NONE. 21. IF SHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP N/A

i J 3.3 AVERAGE DAILY UNIT 2 POWER LEVEL DOCKET No. 050-237 UNIT Dresden 2 DATE August 1,1993 COMPLETED BY K. W. Sykes TELEPlIONE (815) 942-2920 MONTII: July,1993 f DAY AVERAGE DAILY NET DAY AVERAGE DAILY NET POWER LEVEL (MWe) POWER LEVEL (MWe) 1 686 18 648 2 680 19 656 3 690 20 660 1 4 657 21 677 5 678 22 691 6 683 23 692 7 681 24 693-8 662 25 601 9 675 26 559 10 674 27 506 11 647 28 556 12 671 29 620 13 670 30 686 14 681 31 686 15 685 16 684 17 666

3.4 AVERAGE DAILY UNIT 3 POWER LEVEL DOCKET No. 050-249 UNIT Dresden 3 DATE August 1,1993 COMPLETED BY K. W. Sykes TELEPIIONE (815) 942-2920 MONTII: July,1993 I DAY AVERAGE DAILY NET DAY AVERAGE DAILY NET POWER LEVEL (MWe) POWER LEVEL (MWe) 1 740 18 538 2 745 19 647 3 741 20 732 4 736 21 747 5 737 22 750 6 727 23 753 7 725 24 733 8 732 25 599 9 721 26 529 10 651 27 510 11 0 28 567 12 0 29 756 13 0 30 745 14 0 31 723 15 0 16 217 17 502 0

f ) l P 3.5 UNrr 2 SHUTDOWNS AND POWER REDUCTIONS REPORT MONTH JULY.1993 NO. DATE TYPE (1) DURATION REASON (2) METHOD LICENSEE SYSTEM COMPO-CAUSE & (HOURS) OF EVENT CODE (4) NENT CORREC-SHUTTING REPORTF CODE (5) TIVE DOWN ACTION REACTOR 'It) j (3) PREVENT i I RECUR-RENCE 3 930725. F 0 A 4 NA NA NA SEE NOTE 1 bel.DW i .I I l i NorrE 1 - Due to a fire on MCC51 from a lightrdng strike, Unit 2/3 was placed in lake bypass opocration. A load reduction was necessary to keep the lake temperature below the NPDES limit. This event is being investigated under PIR 12 2-93499 (NTS #237-20093-09900). j I TABLE KEY: (I) (3) F: Forced Method: S: Scheduled

1. Manual

2. Manual Scram (2)

3. Automatic Scram Reason:

4. Other (Explain)

A Equipment Failure (Explain)

5. Load Reduction B Maintenance or Test

-l C Refueling (4). D Regulatory Restriction Exhibit G Instructions for Preparation of Data Entry Sheets for Licensee Event Reports (LER) File (NUREG4161) E Operator Training & Licensing Exam F Administrative (5) G Operational Error Exhibit I Sarne Source as above; I H Other (Explain) i 1

3.6 UNTT 3 SHUTDOWNS AND POWER REDUCTIONS REPORT MONTH JULY,1993 NO. DATE TYPE (1) DURATION REASON (2) METHOD LICENSEE SYSTEM COMPO-CAUSE & (HOURS) OF EVENT CODE (4) NENT. CORREC-SHUTTING REPORT # CODE (5) TIVE DOWN ACTION REACTOR TO (3) PREVENT RECUR-RENCE 1 930710 F 147 A 3 050249/ NA NA SEE NOTE 93014 1 BELOW 2 930725 F 0 A 4 NA NA NA SEE NOTE 2 BELOW NOTE 1 - The most probable cause for this Low Condenser Vacuum Scram is imtiating a Condenser Circulating Water flow reversal with a small margin to Condenser Vacuum alarms and trips. It is emphasized that the first flow reversal was initiated within procedurally acceptable backpressure margins. De occurrence of the Turbine Low Vacuum and Condenser low Vacuum alarma prompted the NSO to reverse flow a second time, the condenser vacuums had not recovered from the first reversal, thus compounding the transient and probably led to exceeding the Condenser low - Vacuum scram setpoints in the *B" Condenser Hood. The procedure limits for condenserback-pressure and unit load to conduct the flow reversal - were adhered to. The procedure did not give adequate guidance to ensure sufficient margin was available to conduct the flow reversal, therefore procedure inadequacy was the root cause. Prior to Reactor Startup a temporary procedure change was in place on DOP 4400-8, Circulating Water Flow Reversal, delineating a minimum Condenser vacuum on the Low Hood of 26 in hg and 27 in, hg on the High Hood. A caution will also be added to warn the operator that reversing Condenser flow a second time during a vacuum transient when no equipment problems are present should not be attempted. (NTS # 249- !80-93 4 1401) NOTE 2-Due to a fire on McC51 from a lightning strike. Unit 2/3 was placed itilake bypass operation. A load reduction was necessary to keep the lake temperature below the NPDES limit. his event is being investigated under PIR 12-2-93499(NTS #237-200-9349900). i l 1 (L:\\WK Proc \\pintmgr\\gfs93\\0045.93)

TABLE KE(: (1) (3) F: Forced Method: S: Scheduled

1. Manual

2. Manual scram (2)'

3, Automatic Scram y Reason:

4. Other (Explain)

A Equipment Imilure (Explain)

5. Load Reduction B Maintenance or TestC Refueling C Refueling (4)

D Regulatory Restriction Exhibit O Instructions for Preparation of Data Entry Sheets for Licensee Event Reports (LER) File, (NUREG4161) E Operator Traimng & Ucensing Exam F Administrative (5) O Operational Error Exhibit I Same Source as above, 11 Other (Explain) r t i t h i r - (L:\\WK_ Proc \\pintmgr\\gfs93\\0045.93)

-l 3.7 COMMONWEALTII EDISON COMPANY DRESDEN NUCLEAR POWER STATION MAXIMUM DAILY ELECTRICAL LOAD FOR TIIE MONTil OF July,1993 Dresden 2 Day _ Ifour Endine IGVe 1 1100 736,400 2 1000 740,700 3 0100 734,800 4 0100 730,400 5 1000 737,200 6 1100 730,100 7 1000 725,000 8 0300 724,300 9 0100 724,700 10 0100 722,400 11 1200 736,600 12 1000 734,600 13 1000 737,200 14 1300 735,700 15 0800 735,600 16 0800 735,400 l 17 0900 735,800 18 1200 735,500 19 0300 734, BOO 20 2400 737,400 21 0200 737,800 -l 22 2400-735,400 23 2300 736,900 24 1100 737,400 25 0100 735,000 26 1400 633,100 27 2400 552,300 28 1800 648,000 29 1600 733,500 30 2100 733,900 31 0700 735,800 j

I t 'I i 6[ + 3.7 COMMONWEALTE EDISON COMPANY DRESDEN NUCLEAR POWER STATION MAXIMUM DAILY ELECTRICAL IAAD FOR THE MONTE OF July,.-1993 Dresden 3- -l t +' Day _ Eour Endino KWe t 1 1400 806,500 2 1100 .808,200 3 1300 805,100 l 4 0100 801,300 5 0400 782,000 1 6 1300 795,000 't 7 0400-781,500 i 8 '1300-. 794,300' i 9 0300 758,600-10 0100 736,800 11 0 12 0 13 0 14 0 15 0 16 2400 428,700 ~! 17 2300 608,100-18 2400 609,200 19 2400 804,900 20 1100 807,800 21 0100 '804,100: 22 -1800 797,600 23-1300 798,700-24 1200 789,100 25 0100 782,400 26 1600 583,900 27 2400 .542,200 28 2400 803,300 29 1800 807,600-30 1000 809,900 31 1600 806,800 i-. -l i .j

i '4 i 1 l (; s i L . 4.0 uniges muromrIns mageInsemurrs t i I 4.lMAIN SIEAM RELIEF VALVE OPERATIONS b ^ NOEnt L-4.2OFF-SITE DOSE CALCULATION MANUAL (ODCM) CHANGES Norie .l .j l I i l l-1 I Y 1 I l i l 1 1 i l i i'! 1-I l l l i

4.3 MAJOR CHANGES TO TIIE RADIOACTIVE WASTE TREATMENT SYSTEMS DURING July,1993 None. j 4.4 FAILED MJEL ELEMENT INDICATIONS 4.4.1 Unit 2 l t Unit 2 fuel performance during July,1993, continued to show no indications of leaking fuel. This is based on the sum ~ -l of the activities of the six (6) Noble Cases as measured at the Recombiner. Therefore, Unit 2 had excellent fuel performance. 4.4.2 Unit 3 l l Unit 3 fuel performance during July,1993, continued to show no indications of leaking fuel. This is based on the sum - of the activities of the six (6) Noble Gases as measured at the Rectwnhiner. Therefore, Unit 3 had excellent fuel _j perfonnai%. r h J k l .I d c. e 1-T e .g -i-' w w.

5.0 PLANT OR PROCEDURE CIIANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE 5.1 Amendments to Facility License or Technical Specifications during July,1993.

N0ne, i

u I l 1 i

i 5.2 Additional Changes to Procedures Which are Described in the FSAR (Units 2 and 3) during the months prior to July 1993. Only those procedures that required a new or an additional 10 CFR 50.59 review of changes are included. Changes implemented after June 30,1993 are expected to be reported along with the FSAR updates in accordance with 10 CFR 50.71(e). PROCEDURE TYPE PROCEDURE NUh1BER PROCEDURE TITLE / SUAth!ARY OF DESCRIITION CIIANGES i Dresden Emergency 0100 Reactor Control See attached Operating Procedure (DEOP) DEOP 0200-01 Primary Containment See attached Control DEOP 0300 41 Secondary Containment See attached Control DEOP 0400-01 Reactor Piessure Vessel See attached l Dooding DEOP 0400-02 Emergency Reactor See attached Pressure Vessel Depressurization DEOP 0400-03 Steam Cooling See attached DEOP 0400-04 Primary Containment See attached j Flooding j DEOP 0400-05 Failure to Scram See attached DEOP 0500 4 4 Containment Venting See attached Dresden Radiation 2000-07 Liquid Discharge hionitor See attached Surveillance (DRS) Calibration l I

DEOP 0100 Reactor Control (Revision 02 -implemented May 21,1993) The changes are described below. Changes 1, 3, 4, 5, 6, 8, 10, 12, 13, 16, and 22 were grammatical changes only and are, therefore, not discussed. CIIANGE #2 This change is for Unit 2 only and is located in the second part of the override in Section RPV/P. The change consists of deleting the caution statement "Do not use the Medium Range RPV water level instrument during rapid depressurization below 500 psig." This caution statement is no longer applicable to Unit 2 because the Reactor Vessel Water Level Instrumentation System (RVWLIS) modification moved the instruments' reference leg to outside the drywell. The flashing phenomena which was associated with this instrument can no longer occur; therefore, the caution is not needed. CIIANGE #7 "Ihis change involved extending the !!nes on the second cune in both NPSII figures down to the x-axis so as not to cause confusion for the operator when trying to determine if the NPSII is adequate. CIIANGE #9 This change was to correct for improper referencing. In the first step after the first override in Section RPV/L in the list of systans under Core Spray, the infonnation statement attached reads, "... Figure 100-F & Mgure 100-G)". This statement was changed to read "... Figure 100-F)" htsause Figure 100-G does not apply to Core Spray. CIIANGE #11 This change was to correct the wording for the LPCI Subsystems specified in the table in the third step following the second override in S(ction - RPV/L. The designations of "LPCI Pump A or B" and "LPCI Pump C or D" were changed to "LPCI Loop A" and "LPCI Loop B", respectively. The ~ present wording was not acceptable because it may mislead optrators into thinking that one pump per loop is sufficient when both pumps should be running. CIIANGE #14 The major change to this procedure is a direct result of the Reactor Vessel j Water Level Instrument Systems (RVWLIS) modification which resulted in the moving of the Yarway reference leg from within the Drywell to outside the Drywell and the removal of the Narrow Range pressure compensation (Unit 2 only). These changes resulted in the Afinimum Useable Indicating Levels for Detail 100-A being changed to reflect the most limiting. temperature combination which is based on the hottest Reactor Building temperature. The new MUIL numbers were calculated in accordance with j Emergency Procedure Guidelines Appendix C, WS-14. Although only Unit

l 2 was affected by the RVWLIS mod several Unit 3 MUIL rumbers were I also changed after a review of the original Appendix C calculations found than to be incorrect. CIIANGE #15 In Detail 100-A, the three locations that state Drywell temperature points "9 and 10" were changed to "9 or 10" since each point represents the Drywell temperature in the vicinity of the two instrument systems. This change is in agreement with the original intent of Detail 100-A that any one temperature point is enough to inop ferel instrumentation. 4 i I CIIANGE #17 In Detail 100-A, the warning on not using the medium range indications when depressurizing below 500 psig was deleted (Unit 2 only) because the reference leg is now h>cated in the Reactor Building. This change was made i because the warning is no longer applicable to Unit 2 level instrumentation. CilANGE #18 In Detail 100-A, on Hg 100-B, "RPV Saturation Pressure" was changed to "RPV Saturation Tanperature" to confonn with the EPGs and to place the emphasis on the effects of Drywell and Reactor Building temperature in respect to the curve. CilANGE #19 In Detail 100-A, on Hg 100-B, to the Y-Axis was added "OR Rx Bldg Temp (*F)*" to correct the deficiency, and to emphasize the fact that this curre is also based on Rx Bldg temperature in addition to Drywell temperature. Added "* Rx Bldg temperature required to be obtained locally". This change is also applicable to both Units since each Unit's level instruments are susceptible to boiling in their instrument legs. CIIANGE #20 Due to the RVWLIS mod, Reactor Building lanperature is now the driving tanperature behind the accuracy of the level instruments (Unit 2 only). Added to Table 100-C "* Rx Bldg temperature required to be obtained kically". CIIANGE #21 In Detail 100-A, Hg 100-C (now Table 100-C) was adjusted to prevent overlap of the temperature ranges (i.e. 200-300,300-400). Ranges now read 201-300,301-400, etc. This change rectifies a previous human factor concern. CIIANGE #23 Added normal instrument scales to Detail 100-A, Hgure (now Table) 100-C. This change is a procedure enhancement to aid the operator. CilANGE #24 Added two Medium Ranges (Unit 2 only) to Detail 100-A, Figure (now

t Table) 100-C. " Medium Range /All Rx Bldg Temps" is more limiting since it assumes a worst case Reactor Building Temperature of 350*F. " Medium Range /Rx Bldg Tanps 200*F or less *" requires operator to obtain a local instrument run temperature. This chsnge was made to provide a method that will enable the operator to recover some of the useable indicating level range that was lost due to the RVWLIS modification. CIIANGE #25 Added the following % Detail 100.A: " Reactor Building 1 't =rature near the instrumnat runs is at or above the RI'V Saturation Tunperature (Fig 100-B)" This change anphasizes the fact that the RPV Saturation Temperature curve is based on Reactor Building temperature as well as Dr>well tanperature, it lists Reactor Building temperature above saturation tanperature as a condition for declaring RPV water level instruments inoperable. Safety Evaluation 1. The probability of an occurrence or the consequer.ce of an accident, or malfunction of equipment important to safety as previously evaluated in the ISAR has not increased. There are no FSAR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. CitANC E #2 This change renoved a caution statanent concerning the Medium Range RPV water level instruments that is no longer applicable to Unit 2. The level instmments now function as originally intended, and the operator is able to use the instruments for all operating modes. A failure mechanism has been climinated from these instruments; therefore, the reliability of the instrument is improved and the FSAR accident analyses is improved because the medium range level instrument is available. CIIANGE #7 This change resulted in the extension of the second curve with a line down to the x-axis in the NPSII figures. This change ensures time operator safely operates the Core Spray and LPCI pumps by knowing when adequate NPSil no longer exists for the specified flow. The changes result in safe operation of the Core Spray and LPCI systems; therefore, the requirements of the 13AR are unchanged. (L:\\WK Proc \\pintmgr\\gfs93\\0045.93) I

m CIIANGE #9 This change was to correct for the improper referencing of Figure 100-G to the Core Spray system. This figure did not apply to Core Spray so it was removed from the statement. Since the change was to correct an er or and not to change the intent of the procedure, the FSAR remains unclumged. CIIANGE #11 This change resulted in the renaming of the two LPCI subsystems for clarificatio s. Since the changes were in the nomenclature used and not of procedural intent, the FSAR ) remains unaffected. CIIANGE #14 This change incorporates the new Minimum Useable Indicating Levels for Detail 100-A that were calculated in accordance with Emergency Procedure Guidelines Appendix C, WS-14. By using these new values the Reactor Vessel Water Level Instrument - System will be operated as designed. Th. level instruments will 'e declared inoperable as appmpriate bas 4 on these new MUIL values. The overall FSAR accident analyses is improved because operator response will be based on accurate level indication. CIIANGE #15 This change removes any confusion concerning the use of Drywell temperature points and ensures correct operator response. The level instruments will be declared inoperable as appropriate based on these temperature readings. The overall l FSAR accident analyses is improved because operator response l. will be based on accurate level indication. l l l CIIANGE #17 l This change deletes the warning on not using the medimn range j-level instruments (Unit 2 only) because it is no longer applicable. - I The change ensures that the medium range level instrument is available for use during all plant operating modes and eliminates a failure mnhr.nism from occurring on the instrument; therefore the rel: ability of the instrument is improved and the FSAR accident analyses is improved because the medium range level instrument is available. CIIANGE #18 This change conected a figure title and had no efhd on the use of the figure. This change has no impact on the FSAR. CIIANGE #19 i This change corrects a previous omission by including Reactor

Building Temperature with Drywell Tanperature on Figure 100-B. Elevated temperature in the Drywell or Reactor Building can afTect the Reactor Vessel Water Level Indication System. Separation and redundancy of the instrument lines and racks minimizes the possibility of a complete loss of indication. This change does not affect the design or normal use of RVWLIS. This change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systems design limits and reliability is no longer assured. Correct operator response based on accurate level information will improve the overall FSAR accident analyses. CIIANGE #20 This change corrects a previous omission by including Reactor Building Temperature with Drywdl Temperature on Figure 100-B. Elevated temperature in the Drywell or Reactor Building can affect the Reactor Vessel Water Levd Indication System. Separation and redundancy of the instrument lines and racks minimizes the possibility of a complete loss of indication. This change does not affwt the design or normal use of RV%LIS. This change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systans design limits and rdiability is no longer assured. Corrwt operator response based on accurate level information will improve the overall FSAR accident analyses. Cl!ANGE #21 This change rectifies a previous human factor concern. It clears up a deficiency in the table and has no affect on the FSAR. CIIANGE #23 This change is a procedure enhancement to aid the operator.11 provides useful information that can be used when operating the Reactor Vessd Water Level Indication System. It has no impact on the FSAR. CIIANGE #24 This change was made to provide a method that will enable the operator to recover some of the useable indicating level range that was lost due to the RV%LIS modification.11 provides tle operator with a reasonable option that still allows proper use of the levd instruments in accordance with the EPGs. By providing the operator with accurate level information to base resportses on the orvall FSAR accident analyses is improved. CIIANGE #25 This change anphasizes the fact that the RPV Saturation Temperature curve is based on Reactor Building temperature as wc!! as Drywell temperature. Elevated temperature in the Drywdl or Reactor Building can afTect the Reactor Vessel Water 4

L Level Indication System. Separation and redundancy of the instrument lines and racks minimius the possibility of a complete loss of indication. This change does not affect the design or normal use of RVWLIS. Tids change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systems design limits and reliability is no. longer assured. Correct operator response based on accurate level information will imprm e the overall FSAR accident analyses. 3. The margin of safety, as defined in the basis, for any Technical - Specification,was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were alTected. I DEOP 0200-01 Primary Containment Control (Revision e2 - implemented May 21,1993) The changes are described below. Changes 1,2,4,7,13,16, and 17 were grammatical changes only and are, therefore, not discussed. CIIANGE #3 This change involved swapping some logic steps in order to be consistent with other sections of the flow chart. In Section DWTF the position of the "Before" step was swapped with the position of the override located below it. This was done in order to be consistent with Section PC/P and the placing of its first "Before" and override. CIIANGE #5 The major change to this procedure is a direct result of the Reactor Vessel Water Level Instrument Systems (RVWLIS) modification which resulted in the moving of the Yarway reference leg from within the Drywell to outside - the Drywell and the runoval of the Narrow Range pressure compensation i (Unit 2 only). These changes resulted in the Minimum Useable Indicating levels for Detail 200-1-A being changed to reflect the most limiting temperature combination which is based on the hottest Reactor Building temperature. The new MUIL numbers were calculated in accordance with Emergency Procedure Guidelines Appendix C, WS-14. Although only Unit 2 was alTected by the RVWLIS mod several Unit 3 MUIL numbers were also changed after a review of the original Appendix C calculations found them to be incorrwt. j CIIANGE #6 In Detail 200-1-A, the three locations that state Drywell temperature points "9 and 10" were changed to "9 or 10" since each point represents the Drywell temperature in the vicinity of the two instrument systems. CIIANGE #8

i J l In Detail 200-1-A, the warning on not using the medium range indications when depressurizing below 500 psig was deleted (Unit 2 only) because the reference leg is now hicated in the Reactor Building. CIIANGE #9 In Detail 200-1-A, on Fig 200-1-B, "RPV Saturation Pressure" was changed to "RPV Saturation Temperature" to confonn with the EPGs and to place the trnphasis on the effects of Drywell and Reactor Building temperature in. respect to the curve. CIIANGE #10 In Detail 200-1-A, on Fig 200-1-B, to the Y-Axis was added "OR Rx Bldg 'lemp ('F)*" to correct the deficiency, and to emphasize the fact that this curve is also based on Rx Bldg temperature in addition to Drywell tanperature. Added "* Rx Bldg temperature required to be obtained locally". This change is also applicable to both Units since each Unit's level instruments are susceptible to boiling in their instrument legs. CIIANGE #11 Due to the RVWLIS mod, Reactor Building tem,,erature is now the driving tanperature behind the accuracy of the level instruments (Unit 2 only). Added to Table 200-1-C ** Rx Bldg temperature required to be obtained k>cally". CIIANGE #12 In Detail 200-1-A, Fig 200-1-C (now Table 200-1-C) was atUusted to prevent overlap of the temperature ranges (i.e. 200-300,300-400). Ranges now read 201-300,301-400, etc. CIIANGE #14 Added normal instrument scales to Detail 200-1-A, Figure (now Table) 200-1 C. C11ANGE #15 Added two Medium Ranges (Unit 2 only) to Detail 200-1-A, Figure (now Table) 200-1-C. Medium Range /All Rx Bldg Temps is more limiting since it assumes a worse case Reactor Building Temperature of 350*F. Medium Range /Rx Bldg Temps 200'F or less

• requires operator to obtain a h> cal instrument run temperature.

CIIANGE #18 Added the following to Detail 200-1-A: " Reactor Building temperature near the instrument runs is g1 or above the RPV Saturation Ternperature (Fig 200-1-B)" This change unphasizes the fact that the RPV Saturation Temperature curve is based on Reactor Building temperature as well as Drywell temperature. It lists Psdor Building temperature above saturation temperature as a condition for declaring RPV water level instruments inoperable. j -l

i Safety E-aluation 1. The probability of an occurrence or the consequence of en accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. There are no FSAR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Mnal Safety Analysis Report has not been created. CIIANGE #3 This change involved swapping some logic steps in order to be consistent with other sections of the flow chart. It did not change operator actions and does not impact the FSAR because the procedure still functions as intended. CIIANGE #5 This change incorporates the new Minimum Useable Indicating Levels for Detail 200-1-A that were calculated in accordance with Emergency Procedure Guidelines Appendix C, WS-14. By using these new values the Reactor Vasel Water level Instrument Systan will be operated as designed. The level instruments will be declared inoperable as appropriate based on these new MUIL values. The overall FSAR accident analyses is improved because operator response will be based on accurate level indication. CilANGE #6 This change removes any confusion concerning the use of Drywc!! temperature points and ensures correct operator response. The level instruments will be declared inoperable as appropriate based on these tanperature readings. The overall FSAR accident analyses is improved because operator response will be based on accurate level indication. CIIANGE #8 This change deletes the warning on not using the medium range level instruments (Unit 2 only) because it is no longer applicable. ] The change ensures that the medium range level instrument is i available for use during all plant operating modes and eliminates a failure mechanism from occurring on the instrument; therefore, ) the reliability of the instrument is improved and the FSAR accident analyses is improied because the medium range level i

instrument is available. CIIANGE #9 This chunge corrected a figure title and had no effect on the use of the figure. This change has no impact on the FSAR. CIIANGE #10 This change corrects a previous omission by including Reactor Building Temperature with Drywell Temperature on Figure 200-1B. Elevated temperature in the Drywell or Reactor Building can affect the Reactor Vessel Water Level Indication System. Separation and redundancy of the instrument lines and racks minimira th: possibility of a complete loss of indication. This change does not afTect the design or normal use of RVWLIS. This change does prompt the operator to declare level + instrumentation inoperable when conditions exceed the systems design limits and reliability is no longer assured. Correct operator response based on accurate level information will improve the overall FSAR accident analyses. CIIANGE #11 This change corrects a previous omission by including Reactor Building Tanperature with Drywell Temperature on Figure 200-1B. Elevated temperature in the Drywell or Reactor Building can affect the Reactor Vessel Water Level Indication System. Separation and redundancy of the instrument lines and racks' minimizes the possibility of a complete loss of indication. This change does not a!Tect the design or nonnal use of RVMLIS. This change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systems design limits and reliability is no longer assured. Correct operator response based on accurate level information will improve the overall FSAR accident analyses. CIIANGE #12 This change rectifies a previous human factor concern. It clears up a deficiency in the table and has no alTect on the FSAR. CIIANGE #14 This change is a procedure enhancement to aid the operator. It provides useful information that can be used when operating the Reactor Vessel Water Level Indication System. It has no impact on the FSAR. i

CilANGE #15 This change was made to provide a method that will enable the operator to recover some of the useable indicating level range that was lost due to the RVWLIS modification. It provides the operator with a reasonable option that still allows proper use of the level instruments in accordance with the EPGs. By providing the operator with accurate level information to base responses on the overall FSAR accident analyses is improved. CIIANGE #18 This change emphasizes the fact that the RPV Saturation Tanperature curve is based on Reactor Building temperature as well as Drywell temperature. Elevated temperature in the Drywell or Reactor Building can affect the Reactor Vessel Water Level Indication System. Separation and redundancy of the instrument lines and racks minimizes the possibility of a complete loss of indication. This change does not affect the design or nonnal use of RVWLIS. This change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systems design limits and reliability is no longer assured. Correct operator response based on accurate level information will improve the overall FSAR accident analyses. 3. The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were affected. DEOP 0300-01 Secondary Containment Control (Revision 02 - Implemented May 21,1993) The changes are described below. Changes 1 and 2 were grammatical changes only and are, therefore, not discussed. CliANGE #3 Corrected the reference to a monitored parameter to make it consistent with Table 300-1-B and the other tables in DEOP 300-L In the Reactor Building Radiation section at the first IF/TIIEN changed " maximum normal radiation" to " maximum normal level". CIIANGE #4 The major change to this procedure is the addition of Detail 300-1-D RPV WATER LEVEL INSTRUMENTS. The EPGs list this as an applicable caution to be used when responding to secondary containment tanperature problems. Since increasing Reactor Building tanperatures can affect the RPV Water Level Instrument runs this Detail provides necessary information for determir.ing if the level instruments are still operable. (L:\\WK Proc \\pintmgr\\gfs93\\0045.93)

CIIANGE #5 In the Reactor Building temperature section at the Afonitor and Control step, added " Refer to Detail 300-1-D to determine which water level instruments may be used". This is a EPG required reference and conforms to similar usage in the other DEOPs. CIIANGE #6 The Reactor Vessel Water Level Instrument Systems (RVWLIS) modification moved the Yarway reference leg from within the Drywell to outside the Drywell and removed the Narrow Range pressure compensation (Unit 2 only). These changes resulted ir, the Afinimum Useable Indicating Levels for Table 300-1-F being changed ;o reflect the most l' uniting temperature combination which is based on the hottest Reactor Building temperature. The new NUIL numbers were calculated in accordance with Emergency Procedure Guidelines Appendix C, WS-14. Although only Unit 2 was affected by the RVWLIS mod several Unit 3 MUIL numbers were also changed after a review of the original Appendix C calculations found thern to be incorrect. Table 300-1-F is now unit specific. CilANGE #7 In Detail 300-1-D, the warning on not using the medium range indications when depressurizing below 500 psig is required for Unit 3 only because the reference leg for Unit 2 is now located in the Reactor Building. CIIANGE #8 T:.ble 300-1-F for Unit 2 has two Afedium Ranges that is used to compensate for a loss of useable indicating level due to the RVWLIS modification. Afedium Range /All Rx Bldg Temps is more limiting since it assumes a worse case Reactor Building Temperature of 350'F. hiedium Range /Rx Bldg Temps 200 F or less

• requires operator to obtain a localinstrument run temperature.

Safety Evaluation 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. There are no FSAR affated accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Hnal Safety Analysis Report has not been created. CilANGE #3 This change was made for consistency and has no efTect upon the FSAR. CilANGE #4 This change provides necessary information that is used to determine if RPV water level instrumentation is operable. By (L:\\WK_ Proc \\pintmgr\\gfs93\\0045.93)

-{ ensuring that the operator is has accurate level infor. nation to base responses on the overall FSAR accident analyses is improved. CIIANGE#5 This change ensures that Detail 3001-D is referenced when the Reactor Building temperature section is entered. This will ensure that all operator resp (mses will be based on accurate level infonnation. This wi!I improve the overall FSAR accident analyses. CHANGE #6 This change incorporates the new Minimum Useable Indicating Levels for Detail 300-1 D that were calculated in accordance with - Emergency Procedure Guidelines Appendix C, WS-14. By using ~ these new values the Reactor Vessel Water Level Instrument System will be operated as designed. The level instruments will be declared inoperable as appropriate based on these new MUIL values. The overall FSAR accident analyses is improved because operator response will be based on accurate level indication. CIIANGE #7 The omission of a warning on not using the medium range level instruments (Unit 2 only) is appropriate because it is no longer applicable. It ensures that the medium range level instrument is available for use during all plant operating modes and eliminates a failure mechanism from occurring on the instrument; therefore, the reliability of the instrument is improved and the FSAR accident analyses is improved because the medium range level instrument is available. For Unit 3 the warning ensures that the medium range level instrument is declared inoperable when conditions warrant and operator responses are based on accurate level indication. This will improve the overall FSAR accident analyses. CIIANGE #8 The use of two Medium Ranges for Unit 2 provides a method that will enable the operator to recover some of the useable indicating level range that was lost due to the RVWLIS modification. It provides the operator with a reasonable option that still allows proper use of the level instruments in accordance with the EPGs. By providing the operator with accurate level infonnation to base responses on, the overall FSAR accident analyses is improved. 3. The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were affected. ] 1

DEOP 0400-01 Reactor Pressure Vessel Flooding (Revision 01 - Implemented May 21,1993) The changes are discussed below. Changes I,3,4, and 6 were grammatical changes only and are, therefore, not discussed. The second change is for Unit 2 only and is located in four (4) places in the - procedure. It occurs in both steps requiring the operator to open ADSVs and both steps requiring the operator to depressurize the RPV. The change consists of deleting the caution stating, "Do not use the Medium Range RPV water level instrument during rapid RPV depressurization below 500 psig." This caution is no longer applicable to Unit 2 because the Reactor Vessel Water Level Instrumentation Systmi (RVWLIS) Modification moved the instruments' reference leg from inside the drywell to outside the drywell. The repositioning of the instruments' reference leg to outside the drywell eliminated the reference leg flashing phenomena that the removed caution - was warning the operator about. The fifth change involves the rewriting of the last step before the last diamond in the right side flowpath. The wording presently reads, "... RPV water level until RPV water level indication is restored." It was subsequently changed to "... RPV water level." The remaining part of this step is contained in the diamond below, i.e., the operator terminates the. injection and then proceeds to the next step where he waits to see if level - will be restored within the Core Uncovery Time Limit. As the step now reads and the flowcharting convention used, it implies that the operator-should wait untillevel is restored in this step. If this was the case, the operator would never proceed in the flowchart due to the flowcharting convention used. Therefore, to eliminate this problem, the step was returned to its original wording prior to the last revision where this portion was inadvertently left off the master flowchart. Obviously the previous author intended that this change not be made as the above explanation bears out. The seventh change is rewording two steps to properly clarify their intent. The steps are located in both the ATWS and non-ATWS columns of the flowchart in the emergency depressurization portions where less than 5 ADSVs can be opened. The steps state if "... more than 50 psi above..." instead these steps were changed to read ".. 50 psi or more above..." bnause the actions directed by these steps to emergency depressurize. In emergency depressurization the 50 psi is the low pressure point at which the . reactor is considered depressurized. This value as specified in the EPGs and PS'IGs is inclusive not exclusive. The steps presently exclude the 50 psi; therefore, the problem was rectified by making the step read 50 psi inclusive. The eighth change is to the last wait octagon in the non-ATWS column of the procedure where the operator is waiting for the RPV water level (L:\\WK_ Proc \\pintmgr\\gfs93\\0045.93)

instruments to be determined to be available. The change involved adding tne reactor building temperature to the infonnation for checking on availability of the RPV level instruments. The EPGs are concerned with the temperatures near the cold reference legs being less than 212 "F, not with the areas to be monitored. The additional information for the opentor ensures all areas where the instrument runs tre located are checked. Safety Evaluation 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. There are no FSAR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. The second change ccmcerning the removal of the caution statement within the steps directing the operators to open all ADSVs and the steps directing the operators to augment depressurization with other systems was removed as a direct result of the RVWLIS modification. The caution warned the operators not use the medium range level instruments on rapid RPV depressurizations because the reference legs would flash to steam and level indication would he lost. Since the RVWL.IS modification moved the reference legs to'outside the'dr3well, the reference leg flashing phenomena can no longer occur. The possibility of an accident or malfunction of a type different from those evaluated in the FSAR being created by this change does not exist because the change ensures that the medium range reactor water level instrument is used during all plant operating mods. The operator now has two more level indicators that he can use to monitor reactor level, and therefore, the probability of an accident is reduced. i The fifth change involting the rewording of the third to last step of the procedure from "... RPV water lesel until RPV water level indication is restored." to "... RPV water level." This change does not impact upon the FSAR analyses because it allows the step to function properly in the flowchart as it was originally intended. This step, the previous step and the following step work in combination to attempt to reestablish accurate level indication as the EPGs and PSTGs require. The step is not presently properly flowcharted because it implies that the operator should wait for level indication to be restored. This is incorrect because the following step has the operator do this action. Operators interpret this step as the way the new version reads; in addition, the new venion conforms to the proper flowcharting conventions,

i the EPGs and the PSTGs. The FSAR analyses nmains unchanged i com the presions revision because the step still performs th ! same action, it just was not properly formatted. The object e t this procedure is to restore accurate RPV water level indica ion when RPV water level cannot be determined. A loss of all RPV water level indication implies multiple failures which the FSAR does not cover. Therefore, the actions taken by the operators as a result of a loss of RPV water level indication would not impact upon the FSAR analyses. The seventh change involves rewording two steps such that the value of 50 psi is induded within the action to be taken. Presently, the step does not include 50 psi as the lower bounding pressure. The EPGs and PSTGs both direct that additional systems be utilized for emergency depressurization at a value of 50 psi, inclusive. The change assures that the procedure properly agrees with these documents. The actions that this step directs the operator to do places the reactor in a more consenative state. Since it is predicated on a failure of equipment, the actions assure that the FSAR analyses is not impacted by assuring the reactor is properly depressurized. The eighth change involves the addition of infonnation to the wait octagon. The concern over the cold reference leg tanperature was only partially addressed in this step. It addressed the drywell temperature but not the reactor building tanperature. The change includes the reactor building tanperature in the step so the operator can monitor all areas that affect the RPV level instruments. Therefore, when the availability of the RPV level instruments is being determined, all areas that potentially impact upon the instruments' cold reference leg temperature is properly monitored. Thus, the FSAR analyses remains unalTecIed because proper equipment operability is 3 assured prior to taking the reactor to a condition where accurate level detennination can be made. Lastly, since multiple failures have occurred to place the plant in a ccmdition requiring the RPV to be flooded, the FSAR analyses does not apply. 3. The margin of safety, as defined in the basis, for any Tecimical Specification,was not reduced.' There are no Technical j Specifications where the requiranent, associated action items, associated survei!!ances, or basis were affected. DEOP 0400-02 Emergency Reactor Pressure Vessel Depressurization (Revision 01 - Implemented May 21,1993) - The changes are discussed below. Change I was a grammatical change only and is, therefore, not discussed. t

i i l The second change is for Unit 2 only and is located in four (4) places in the - procedure. It occurs in both steps requiring the operator to open ADSVs. and both steps requiring the operator to depressurize the RPV. The change consists of deleting the caution stating, "Do not use the Medium Range RPV.- water level instrument during rapid RPV depressurization below 500 psig." - r This caution is no longer applicable to Unit 2 because the Reactor Vessel Water Level Instrumentation System (RVWLIS) Modification moved the instruments' reference leg to outside the dr3vell. l The third change involves the rewriting of the decision diamonds in both ATWS and Non-ATWS trains concerning the comparison of RPV pressure and drywell pressure. The decision diamonds presently read, "Is RPV ' pressure more than 50 psi above drywell pressure?" It is being changed to "Is RPV pressure 50 psi or more above drywell pressure?" The reason for this cl.ange is that the Emergency Procedure Guidelines (EPGs)' l (NEDO-31331) Step C2-1.4 and Dresden's Plant Specific Technical Guidelines (PSTGs) includes 50 psi as the lower limit; presently, the procedure does not accurately reflect what the approved documents specified this step to read. The change assures that the operaton will continue to emergency depressurize the reactor if RPV pressure is greater l than or equal to 50 psi over drywell pressure. The 50 psi is based on the lowest RPV pressure at which an SRV will remain fully open with its control switch placed in the open position. Safety Evalulation 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment.mportant to safety as previously evaluated in the FSAR has not increased. There are no FSAR affected accidents. 2. The possibility for an acciden'. or malfunction of a different type than any previously evaluated in the Mnal Safety Analysis Report j has not been created, j ) The second change concerning the removal of the caution j statement within the steps directing the operators to open all ADSVs and the steps directing the operators to augment J depressurization with other systems was removed as a direct - result of the RVWLIS modification. The caution warned the : operators not use the medium range level instruments on rapid RPV depressurizations because the reference legs would flash to - steam and twel indication would be lost. Since the RVWLIS modification moved the reference legs to outside the dr3vell, the reference leg flashing phenomena can no longer occur. The. I possibility of an accident or malfunction of a type different from those evaluated in the FSAR being created by this change does I not exist because the change ensures that the medium range ~ i reactor water level instrument is used during all plant operating modes. The operator now has two more level indicators that he i

can use to monitor reactor level, and therefore, the probability of an accident is reduced. The third change concerns the restructuring of two decision diamonds that ask if the RPV has been fully depressurized. The EPGs and PSTGs base a depressurized reactor on the lowest RPV pressure at which an SRV will remain fully open with its control switch placed in the open position. When RPV pressure is below this value, depressurization is considered complete and RPV pressure reduction need not be augmented by use of additional systems. The value that is utilized to detennine whether the RPV - is pressurized as specified in the EPGs and PSTGs is a reactor pressure of 50 psi or more higher than the drywell pressure. The DEOPs were written stating that the depressurization be augmented with additional s3 stems at a reactor pressure of more than 50 psi higher than the drywell pressure. These two statements are not the same. Because in the fint statement,50 psi is inclusive in the pressure value specified. Therefore, to correct this deviation, the DEOPs were written to c(mfonn to the - EPGs and PSTGs. There is no adverse affect on plant systems or functions because the step remains essentia!!y the same. Since the object of this step is to assure complete depressurization, the inclusion of 50 psi accomplishes this requirmient. Also, no mw accidents or malfunctions are created because if an event were to occur such that all five (5) ADSVs did not open, the plant can still be safely depressurized. In addition, the multiple failures that are implied by these procedural steps indicate an accident outside of the design basis and therefore, do not impact upon the FSAR. 3. The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were afferted. DEOP 0400-03 Steam Cooling (Revision 01 - Implemented May 21,1993) The changes are discussed below. Change I was a grammatical change only 1 and is, therefore, not discussed. The second change is for Unit 2 only and is located in two(2) places in the procedure. It occun in the step requiring the operator to open ADSVs and the step requiring the operator to depressurize the RPV. The change consists of deleting the caution stating, "Do not use the Medium Range RPV water level instrument during rapid RPV depressurization below 500 psig." This caution is no longer applicable to Unit 2 because the Reactor Vessel Water Level Instrumentation System (RVWLIS) Modification moved the instruments' reference leg to outside the dr3well. L.

f The third change involves the rewriting of the last decision diamond in the procedure concerning the comparison of RPV pressure and drywell pressure when emergency depressurizing. The daision diamond presently reads, "Is RPV pressure more than 50 psi above drywell pressure?" It is being changed to read, "Is RPV pressure 50 psi or more above drywell pressure?" The reason for this change is that the Emergency Procedure Guidelines (EPGs) (NEDO-31331) and the Dresden's Plant Specific Technical Guidelines (PSTGs) includes 50 psi as the lower limit. Prmently, the procedure does not accurately reflect what the approved documents specified this step to read. The change assures that the operators will continue to emergency depressurize the reactor if RPV pressure is greater than or equal to 50 psi over drywell pressure. The 50 psi is based on the lowest RPV pressure at which an SRV will remain fully open with its control switch placed in the open position. Safety Evaluation 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. There are no FSAR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. The second change concerning the removal of the caution statement within the steps directing the operators to open all ADSVs and the steps directing the operator to use other systems to depressurize was removed as a direct result of the RVWLIS modification. The caution warned the operators not use the medium range level instruments on rapid RPV depressurizations because the reference legs would flash to steam and level indication would be lost. Since the RVWLIS modification moved the reference legs to outside the drywell, the reference leg flashing phenomena can no longer occur. The possibility of an accident or malfunction of a type different fmm those evaluated in the FSAR being created by this change does not exist because the change ensures that the medium range reactor water level instrument is used during all plant operating modes. The operator now has two more level indicators that he can use to monitor reactor level, and therefore, the probability of an accident is reduced. The third change concerns the restructuring of the last decision - diamond that asks if the RPV is fully depressurized. The EPGs and PSTGs base a depressurized reactor on the lowest RPV pressure at which an SRV will remain fully open with its control switch placed in the open position. When RPV pressure is below this value, depressurization is considered complete and RPV i pressure reduction need not be augmented by use of additional systems. The value that is utilized in determine whether the RPV

is pressurized is spwilled in the EPGs and PSTGs is a reactor - pressure of 50 psi or more higher than the drywell pressure. The DEOPs were written stating that the depressurization be augmented with additional systems at a reactor pressure of more than 50 psi higher than the drywell pressure. These two statements are not the same. Because in the first statement, 50 psi is inclusive in the pressure value specified. Therefore, to correct this deviation, the DEOPs were written to confonn to the EPGs and PSTGs. There is no adverse affect on plant systems or functions because the step remains essentially the same. Since the object of this step is to assure complete depressurization, the inclusion of 50 psi accomplishes this requirement. Also, no new accidents or malfunctions are created because if an event were to i occur such that all live (5) ADSVs did not open, the plant can still - be safely depressurized. In addition, the multiple failures that are implied by this procedural step indicates an accident outside of the design basis and therefore, do not impact upon the f3AR. - 3 3. The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were affected. DEOP 0400-04 Primary Containment Hooding (Revision 01-Implanented May 21,1993) The changes were grammatical changes only and are, therefore, not discussed. The safety evaluation follows below. Safety Evaluation 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. There are no f3AR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. All of the changes are grammatical in nature, setting up proper flowchart structure, to establish consistency between procedures or for the purposes of step clarification. All of these changes do not alter the intent of the procedure; therefore, these changes do not impact systems or functions so as to estate the possibility of an accident or malfunction of a type different from those evaluated in the FSAR.

3. The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. TI ere are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were affected. DEOP 0400-05 Failure to Scram (Revision 02 -Implemented Afay 21,1993) The changes are discussed below. Changes I,2,3,5,6,13,14,17, and 23 were grammatical changes only and are, therefore, not discussed. CIIANGE #4 The second curve in the NPSII curves war extended down to the X-Axis so as to remove any confusion from the curves. CIIANGE #7 In the Reactor Power section at the step referencing Table 400-5-K, deleted " Verify ARI is reset and concurrently...". This step also refennees DEOP ' 500-5 which addresses the failure of ARI to AUTO reset. By removing ARI reset as a prerequisite to entering DEOP 500-5, DEOP 500-5 can be used to mitigate the ARI malfunction. CIIANGE #8 In the Reactor Power section at the step referencing Table 400-5-K, added a downward arrow with the words "...while continuing here". This change provides specific direction to continue in the flow chart while trying to - insert control rods. This'is in agreement with the EPGs. CIIANGE #9 In the RPV water level section after the second override and first diamond changed "I. Bypass the low RPV water level Drywell Pneumatics and h1SIV isolations (DEOP 500-2)" to "I. Bypass the low RPV water lesel 31SIV isolation (DEOP 500-2)". Drywell Pneumatic's have a backup Nitrogen system that automatically aligns itself on low pneumatic pressure. The intent of this step was to restore pneumatic supply before the AISIVs drifted shut on low pressure. Since the backup pneumatic supply system maintains the required pressure to the h1SIVs there is no need to bypass the isolation signal to the Drywell Pneumatic isolation valves. CIIANGE #10 In the RPV water level section after the second override and first diamond deleted "2. Restore Drywe!I Pneumatics to the containment". Drywell Pneumatic's have a backup Nitrogen system that automatically aligns itself on low pneumatic pressure. The intent of this step was to restore pneumatic

supply before the MSIVs drifted shut on low pressure. Since the backup pneumatic supply system maintains the requimd pressure to the MSIVs there is no need to " restore Drywell Pneumatics to the containment" since technically they were never lost. CIIANGE #11 In the RPV water level section after the swond override and first diamond added a downward arrow and the words "...while continuing here". This change provides specific direction to continue in the flow chart while trying to bypass the low RPV water level MSIV isolation. This is in agreement with the EPGs. CIIANGE #12 In the RPV water level section after the second override following the RPV rapid injection Caution changed " Maintain total ECCS flow (IIPCI and LPCI) below..." to " Maintain total ECCS flow (IIPCI, LPCI and Core Spray) below...". After Emergency Depressurization with RPV water level being restored the last IF/TIIEN in the Return from DEOP 400-2 section returns level control to the RPV level section. With Core Spray already injecting in the RPV it is appropriate to add the Core Spray pump to the total ECCS flow for detennining the ECCS Vortex Limit. CIIANGE #15 The major change to this procedure is a direct result of the Reactor Vessel Water Level Instrument Systems (RVWLIS) modification which resulted in the moving of the Yarway reference leg from within the Drywell to outside the Drywell and the removal of the Narrow Range pressure compensation (Unit 2 only). These changes resulted in the Minimum Useable Indicating Levels for Detail 400-5-A being changed to reflect the most limiting temperature combination which is based on the hottest Reactor Building temperature. The new MUIL numbers were calculated in accordance with Emergency Procedure Guidelines Appendix C, WS-14. Although only Unit 2 was affected by the j RVWLIS mod several Unit 3 MUIL numbers were also changed after a i review of the original Appendix C calculations found them to be incorrect. CIIANGE #16 In Detail 400-5-A, the three h> cations that state Drywell temperature points "9 and 10" were changed to "9 or 10" since each point represents the Drywell temperature in the vicinity of the two instrument systems. l CIIANGE #18 In Detail 400-5-A, the warning on not using the medium range indications

p i 'i l when depressurizing below 500 psig was deleted (Unit 2 only) because the reference leg is now located in the Reactor Building. CIIANGE #19 In Detail 400-5-A, on Fig 400-5-B, "RPV Saturation Pressure" was changed to "RPV Saturation Temperature" to conform with the EPGs and to place the emphasis on the effects of Drywell and Reactor Building temperature in respect to the curve. CIIANGE #20 In Detail 400-5-A, on Fig 400-5-B, to the Y-Axis was added "OR Rx Bldg Temp (*F)*" to correct the deficiency, and to emphasize the fact that this curve is also based on Rx Bldg temperature in addition to Drywell temperature. Added "* Rx Bldg temperature required to be obtaincel locally". This change is also applicable to both Units since each Unit's lesel instruments are susceptible to boiling in their instrument legs. CIIANGE #21 Due to the RVWLIS mod, Reactor Building temperature is now the driving temperature behind the accuracy of the level instruments (Unit 2 only). Added to Table 400-5-C "* Rx Bldg ' emperature required to be obtained j t locally". CIIANGE #22 In Detail 400-5-A, Fig 400-5-C (now Table 400-5-C) was adj"sted to prevent - overlap of the temperature ranges (i.e. 200-300, 300-400). Ranges now read 201-300,301-400, etc. CIIANGE #24 Added normal instrument scales to Detail 400-5-A, Figure (now Table) 400-5-C. CIIANGE #25 Added two Medium Ranges (Unit 2 only) to Detail 400-5-A, Figure (now Table) 400-5-C. Medium Range /All Rx Bldg Temps is more limiting since it i assumes a worse case Reactor Building Temperature of 350*F. Medium Range /Rx Bldg Temps 200'F or less

• requires operator to obtain a local instrument run temperature.

CIIANGE #26 Added the following to Detail 400-5-A: " Reactor Building temperature near the instrument runs is at or above the .RPV Saturation Tanperature (Fig 400-5-B)" This change emphasizes the fact that the RPV Saturation Temperature (LdWK Proc \\p!ntmgr\\gfs93\\0045.93)

i curre is based on Reactor Building temperature as well as Drywell tmiperature. It lists Reactor Building tanperature above saturation temperature as a condition for declaring RPV water level instruments inoperable. Safety Evaluation I. The prohahility of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as - previously evaluated in the FSAR has not increased. There are no FSAR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. CIIANGE #4 TMs chnge rtsulted in the extension of the second curve with a lir.e down to the x-axis in the NPSil figures. This change ensures the operator safely operates the Core Spray and LPCI pumps by knowing when adeounte NPSII no longer exists for the specified flow. The changes result in safe operation of the Cwre Spray and LPCI systems; therefore, the requirements of the FSAR are unchanged. CIIANGE #7 This change ensures that a failure of ARI to AUTO reset does not hinder the use of Alternate Rod Insertion methods. Became proper operator response to a failure to scram transient has been strengthened, the overall ISAR accident analyses is improved. CIIANGE #8 This change provides specific direction to continue in the flow chart while trying to insert control rods. Because proper operator response to a failure to scram transient has been strengthened, the overall FSAR accident analyses is improved. CIIANGE #9 This change removes an unnecessary step to allow the Drywell pneumatics systems to operate as they were designed. There is no impact on the FSAR. CIIANGE #10 .1

This change removes a step that is no longer required. Since the Drywell pneumatic's systans are operating as they were designed there is no impact on the FSAR. CHANGE #11 This change reinforces the intent of the EPGs to lower RPV level to control reactor power while bypassing the MSIV low level isolation. Because proper operator response to' a failure to scrams transient has been strengthened the overall FSAR accident. analyses is impreved. ' CHANGE #12 This change adds Core Spray to the total ECCS flow list used in determining the ECCS Vortex Limit. By ensuring that the ECCS pumps operate within the ECCS Vortex Limit potential damage to the pumps is minimized. Because pump availability has been. increased the FSAR accident analyses has improved. CIIANGE #15 This change incorporates the new Minimum Useable Indicating Levels for Detail 400-5-A that were calculated in accordance with Emergency Procedure Guidelines Appendix C, WS-14. By using these new values the Reactor Vessel Water Level Instrument. Systan will be operated as designed. The level instruments will be declared inoperable as appropriate based on these new MUIL values. The overall FSAR accident analyses is improved because operator response will be based on accurate level indication. CIIANGE #16 - This change removes any confusion concerning the use of Drywell temperature points and ensures correct operator response. The - level instruments will be declared inoperable as appropriate based on these temperature readings. The overall FSAR accident l analyses is improved because operator response will be based on accurate level indication. ] CIIANGE #18 This change deletes the warning on not using the medium range level instruments _(Unit 2 only) because it is no longer applicable. The change ensures that the medium range level instrument is available for use during all plant operating modes and eliminates a failure mechanism from occurring on the instrument; therefore,. the reliability of the instrument is improved and the ISAR j accident analyses'is improved because the medium range level i I i

instrument is arallable. CIIANGE #19 This change corrected a figure title and had no effect on the use of the figure. This change has no impact on the FSAR. CIIANGE #20 This change corrects a previous omission by including Reactor Building Temperature with Drywell Temperature on Figure 400-5B. Elevated temperature in the Drywe!I or Reactor Building can alTect the Reactor Vessel Water Level Indication System. Separation and redundancy of the instrument lines and racks minimizes the possibility of a complete loss of indication. This i change does not affect the design or normal use of RVWLIS. This change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systems design limits and reliability is no longer assured. Correct operator response based on accurate level information will improve the overall FSAR accident analyses. CIIANGE #21 This change corrects a previous omission by including Reactor Building Temperature with Drywell Temperature on Figure 400-5B. Elevated temperature in the Drywell or Reactor ' Building can alTect the Reactor Vessel Water Level Indication System. Separation and n:dundancy of the instrument lines and racks minimizes the possibility of a complete loss of indication. This change does not affect the design or normal use of RVWLIS. This change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systems design limits and reliability is no longer assured. Correct operator response based on accurate level infonnation will improve the overall FSAR accident analyses. CIIANGE # 22 This change rectifies a previous human factor concern. It clears up a deficiency in the table and has no effect on the FSAR. CIIANGE #24 This change is a procedure enhancesnent to aid the operator. It provides useful information that can be used when operating the Reactor Vessel Water Level Indication System. It has no impact on the FSAR. ClIANGE #25 L

i This change was made to provide a method that will enable the-opemtor to recover some of the useable indicating level range that was lost due to the RVWLIS modification. It provides the operator with a reasonable option that still allows proper use of the level instruments in accordance with the EPGs. By providing the operator with accurate level information to base responses on the overall FSAR accident analyses is improved. CIIANGE #26 This change emphasizes the fact that the RPV Saturation Temperature curve is based on Reactor Building tanperature as well as Drywell temperature. Elevated tanperature in the Drywell or Reactor Building can affect the Reactor Vessel Water Level Indication System. Separation and redundancy of the instnnnent - lines and racks minimizes the possibility of a complete loss of indication. This change does not affect the design or normal use of RVWLIS. This change does prompt the operator to declare level instrumentation inoperable when conditions exceed the systems design limits and reliability is no longer assured. Correct operator response based on accurate level information will improve the overall FSAR accident analyses. i 3. The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surteillances, or basis were affected. DEOP 050044 Containment Venting (Revision 06 - Implemented May 21,1993) The changes are described below. CIIANGE #1 PURPOSE was reworded to stipulate the two methods used to reduce Ilydrogen Concentration in Primary Containment (direct venting and purging with Nitrogen / Air). The DR was removed since this procedure is - used to control both Ilydrogen concentration and Primary Containment pressure. CIIANGE #2 USER REFERENCES added Electrical Print 12E-2512 (12E-3512A), P.C.I./APCV - System Atmospheric Control Inboard / Outboard Valves. The j electrical print was tad to resolve the pmper sequence for placing the APCV system in service. It was also used to detennine the steps required to' l bypass Group 2 interkw:ks when using the APCV system as a vent path for i Nitrogen purging. ) i-

~ CIIANGE #3 USER REFERENCES added DAP 07-04, Control of Temporary System Alterations. DAP 07-04 is referenced in the LIMITATIONS AND ACTIONS Section and is required to be listed here. CIIANGE #4 USER REFERENCES added DOP 1600-21, Draining Augmented Primary - Containment Vent Systan. DOP 1600-21 is a new procedure that removes from DEOP 0500-04 the steps used for draining the APCV system after use. The DEOP directs the operator to refer to the DOP when time and radiation dose rates permit. Since draining APCV is not an immediate concern of the operator it has been removed from this procedure. CIIANGE #5 PREREQUISITES had D.1 rewritten to require direction from the Shift Supervisor /DEOPs to enter this procedure. Since this support procedure is NOT to be used for normal venting but only for emergencies, DEOP 0200-01 and DEOP 020042 were added as the procedures that provide instruction for entering DEOP 0500-04. CIIANGE #6 PREREQUISITES had D.4 added to stipulate requirements for venting when the DEOP procedures do NOT state " Irrespective of the offsite radioactivity release rate...". Included in D.4 is a NOTE that specifies how - long and under what conditions an at=cspheric sample is considered valid si Drisden. Conditions that require the Station Director's authorization to vent were also listed. CIIANGE #7 LIMITATIONS AND ACTIONS added F.3 which provides a brief explanation on why venting, irrespwtive of offsite radioactive release rates,'_ is performed. It also points out that the operator should be trying to limit the total amount of the release by controlling pressure or Ilydrogen concentration below applicable limits. Continuous venting is NQI authorized unless the applicable limits are continuously being exceeded. Tids change has also been incorporated throughout the procedure as cautions and steps with directions on how to control helow the applicable limits. This clarifies and expands the direction given by DEOP 0200-01 and DEOP 0200-02. CIIANGE #8 LIMITATIONS AND ACTIONS added F.4 which specifies limits for SBGT operation and the specific actions required if they are exceeded. Tids information is relevant to this procedure since SBGT is used for both i venting and purging. Operated under emergency cemditions with probable i I i

fuel damage, deposits of fission pruducts on the SBGT filters and charcoal absorbers is expected. CIIANGE #9 LIhflTATIONS AND ACTIONS added F.5 which states that venting Primary Containment at high pressure can present a radiological hazard to personnel in the Reactor and Turbine Buildings. Radiation Protection is to be notified of any venting of Pnmary Containment at high pressure. This is. a reasonable safety precaution and has been incorporated as a step to announce evacuation of all unnecessary personnel due to changing radiological conditions. Since Chemistry and not Radiation Protection is involved in sampling and providing recommendations for the release it is appropriate to contact Radiation Protection to advise them of potential hazards. CIIANGE #10 LIh11TATIONS AND ACTIONS added F.6 which restricts simultaneous tenting of Unit 2 and 3 through the Augmented Primary Containment Vent (APCV) System. Since the APCV has a common line to the chimney from each unit this restriction will retain separation between the units and maintain operation of the system within its c.igina' design limits. This limitation has been incorporated into the procedure as a caution whenerer. the APCV system is in use. CIIANGE #11 LIh1ITATIONS AND ACTIONS added F.7 which states that repressurization of Primary Containment with Nitrogen is not allowed at Dresden. This limitation has been built into the procedure's IF/TIIEN Logic that requires Primary Containment to either be depressurized or to have pressure decreasing. A NOTE is placed before any step that introduces Nitrogen into Priman Containment stating that repressurization with Nitrogen at Dresden is NOT allowed. This limitation also applies to Air purging but since ACAD is no longer used and Air purging requires Primary Containment depressurized there is no reason to include it here as a limit. This is in agreement with the BWR Owner's Group Emergency 1 Procedure Guidelines. CilANGE #12 LIhtITATION AND ACTIONS added F.8 which restricts air purging of Primary Containment to only when Primary Containment is depressurized. Since the use of ACAD is no longer allowed at Drtsden the only air available for purging Primary Containment is from the Reactor Building to Drywell vent valve. To keep from contaminating the Reactor Building, Primary Containment must be depressurised before this sent valte is opened. This requirement has been incorporated into the procedure as cautions and IF/THEN statanents prior to air purging. 1 i i-

CIIANGE #13 LIMITATIONS AND ACTIONS added F.9 which restricts venting through SBGT and the 2-inch vent valves to less than 26 psig in Primary Containment. An Engineering study has detennined that the pressure drop across the 2-inch vent valves with 26 psig in Primary Containment is sufficient to keep from exceeding the SBGT maximum allowable internal pressure of 1.5 psig. This restriction has been incorporated into the procedure as IF/FIIEN statements that require securing venting / purging through SBGT when Primary Containment pressure exceeds 26 psig. CIIANGE #14 Added section to incorporate Unit 2 APCV modification for venting Primary Containment. Removed section that allowed the use of the Reactor Building Ventilation system to control the Primary Containment Pressure 1 imit. Unit 2 was using this method for depressurizing because they did not have APCV available. In addition the Reactor Building Ventilation system was not designed to handle the high pressure and was expected to overpressurize and rupture the Drywell Purge Fan Ducts. The new section was merged with the existing Unit 3 APCV section. Letter from Pacific Nuclear. (XCE-065-187) specifically states that the Vent and Purge System will no longer be needed for pressure contnil. CIIANGE #15 Revised option of venting to Reactor Building Ventilation. Venting to Reactor Building Ventilation will only be performed in conjunction with air purging Primary Containment for Hydrogen control. Previously Reactor Building Ventilation was being used for Primary Containment pressure and Ilydrogen control. Because ACAD is no longer used at Dresden, air purging now requires that Primary Containment be depressurized. The caution describing overpressurization and rupture of the Dr3vell Purge Fan Ducts has been deleted since Reactor Building Ventilation will no longer be used for venting when Primary Containment is pressurized. A letter from Pacific Nuclear (XCE-065-187) specifically states that "upon initiation of the APCV, the primary containment pressure will not increase". The need for retaining Reactor Building Ventilation as an alternate to APCV pressure control no longer exists. CIIANGE #16 - Revised steps used to bypass the Reactor Building Ventilation isolation. Reactor Building Ventilation was to be used to purge Primary Containment " irrespective of the offsite radioactivity release rate". The old procedure faihd to bypass all of the isolations that would prevent accomplishing this purpose. The isolations not previously bypassed included: Reactor Building Ventilation Exhaust Ifigh Radiation, Refueling Floor High Radiation, and signals from the una!Tected unit. The NOTE stating that the Reactor j Building Vent Exhaust RE Refuel Floor IIigh Radiation isolations are NOT _) bypassed has been delett4. i

= L CIIANGE #17 Established SBGT as the primary vent path for Ilydrogen control. Besides allowing a more rapid lineup through the 2-inch vent lines without having to lift leads or install jumpers, SBGT will also limit the total amount of the release. This in conjunction with the standard practice of venting the Torus first is the preferred method for venting Primary Containment. Reactor Building Ventilation will be used for air purging when Primary Containment is depressurized. CIIANGE #18 Established APCV as an optional vent / purge path for Ilydrogen control. APCV allows venting Primary Containment at pressures greater than the SBGT pressure limit of 26 psig. Using the 18-inch vent valves instead of the 2-inch valves allows Nitrogen purging without repressurization of Primary Containment. Using APCV for Ilydrogen control removes the need to retain Reactor Building Ventilation as a possible path for Nitregen purge. Since Nitrogen purging may be done at pressure the potential for rupturing the Drywell Purge Fan Ducts by using Reactor Building Ventilation has been i averted. Because the APCV Systmi has no fans associated with it, air purging is not available. Only Nitrogen may be used to purge Primary Containment with APCV. CIIANGE #19 Changed flow path for Nitrogen purging of Primary Containment. Old procedure had path from Nitrogen makeup - Torus Makeup Viv - Torus - Torus /Drywell Vacuum Breakers - Drywell-Drywell Vent Viv Reactor Building Ventilation QR SBGT. The vacuum breakers require a 1.5 psid d/p from Torus to Drywell to open. In order to open the Torus would have to be pressurized at least 1.5 psig above Drywell pressure. Since repressurizing with Nitrogen is no longer allowed this method of purging is not acceptable. This method would also require repeated " puffing" of the vacuum breakers and they do not appear to have been analyzed for this purpose. The new flow path Nitrogen purges the Drywell and Torus separately and does not require the use of the vacuum breakers. Primary Containment is NOT pressurized using this lineup. CIIANGE #20 Changed flow path for Air purging of Primary Containment. Old procedure had path from Reactor Building -- Vent Viv -+ Torus Purge Viv -+ Torus -- Torus /Drywell Vacuum Breakers - Drywell - Drywell Vent Viv - Reactor Building Ventilation QR SBGT. The vacuum breakers require a 1.5 psid d/p from Torus to Drywell to open. Because ACAD is no longer used the 1.5 psid has to come from the suction of either SBGT or Reactor Building Ventilation. Since both Reactor Building Ventilation and SBGT draw on the rest of the Reactor Building and cannot be isolated by closing the Rx Bldg /SBGT dampers, the efficiency of this flow path is questionable. This method would also require repeated " puffing" of the vacuum breakers and (L:\\WK_ Proc \\pintmgr\\gfs93\\0045.93) m

i they do not appear to have been analyzed for this purpose. The new flow path Air purges the Drywell and Torus separately and does not require the use of the vacuum breakers. CIIANGE #21 Restricted use of SBGT te less than 26 psig in Primary Containment through the 2-inch vent valves. Letten from Sargent & Lundy dated February 16,1993 and hfarch 23,1993 specifically state that the " containment can be vented through the SBGT system at a pressure of 26 psi" with the stipulation that the " venting has to be done through the 2-inch - bypass line". This restriction has been incorporated as a LIh!!TATION AND ACTION as IF/TIIEN statements that require securing - senting/purty:g through SBGT when Primary Containment pressure exceeds 26 psig. The steps utilizing 5BGT for Primary Containment pressure control when greater than 2 psig have been deleted. DEOP 0200-01 only allows use of SBGT for pressure control when Primary Contairunent pressure is less than 2 psig. Between 2 psig and the Primary Containment Pressure Limit no release:: are authorized; pressure is controlled by spraying the Torus /Drywell and emergency depressurizing. Venting is not allowed ' until Primary Containment integrity is threatened. At this pressure SBGT ~ cannot be used since Primary Containment pressure is well.above 26 psig. r CIIANGE #22. Added various steps and cautions to linut the total amount of the radioactive release by requiring operators to monitor and control the. venting / purging below applicaMe Primary Containment limits. Previously this procedure did not provide specific direction concerning this. Continuous senting is NOT authorized unless the applicable limits are continuously c.eing exceeded. CIIANGE #23 In general, added specific steps to complete required actions. Provided recovery steps to make it easier to mcre from one section to another. Listed specific criteria to be met before secuing Imeup. r CHANGE #24 Added PRIhfARY CONTAINhfENT PRESSURE LIhflT DEOP IP.gure 200-1-G to APCV pressure control section. Hgure will be used to limit the - l total amount of the radioactive release by requiring operators to monitor and control Torus bottom pressure below the Primary Containment Pressure Limit. CHANGE #25 ~ Removed APCV draining section and cre ted separate procedure. Referenced DOP 1600-21, Draining Augmented Priman Containment Vent System. DOP 1600-21 is a new procedure that deletes from DEOP 0500-04 (L:\\WK_ Proc \\pintmgr\\gfs93\\0045.93)

the steps used for draining the APCV system after use. The DEOP directs the operator to refer to the DOP when time and radiation dose rates permit. Since draining APCV is not an immediate concern of the operator it has been removed from this procedure. CIIANGE #26 Added requirement to observe Main Chimney SPING Monitor when venting or purging Primary Containment. This information will be used to determine if the Tech Spec LCO release rate limit is being exceeded. If it is being exceeded and the DEOPs do not state " Irrespective of the offsite radioactivity release rate..." then the venting / purging operation will be stopped. This expands the direction given in DEOP 0200-02 anti places it in a more uscable format for the operators. CilANGE #27 Revised section describing Oxygen /llydrogen requirements for Nitrogen and Air purging. If Drywell DE Torus Oxygen concentrations are not less than 5% then Air purging is required. If Drywell AND Torus Oxygen concentrations are both below 5% then Nitrogen purging is required. This is in accordance with the latest EPG revisions. CIIANGE #28 Placed steps in the Nitrogen purging sections that will secure the Nitrogen purge before starting the Air purge. This conserves the Nitrogen supply for future use and also keeps it available for use by the other Unit. Safety Evaluation 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. The probability of damage to fuel during a refueling accident is nqt increased by this procedure revision. All purge and vent equipment is being operated within its design limits. DEOP 500-04 does nel inenase the probability of a malfunction of fuel handling equipment. It is expected that DEOP 500-04 will not be used until several hours after the DEOPs are initially entered. During that time any fuel handling on the unaffected unit will have been secured. Since the equipment is not being used, the consequences of it malfunctioning has aqi increased. The probability of a LOCA occurring is apet increased by this procedure revision. All purge and vent equipment is being operated within its design limits. DEOP 500-04 does not increase E(L:BVK Proc \\pintmgr\\gfs93\\0045.93)

the probability of a malfunction of equiopment imp artant to safety. Although Group 2 and Reactor Building Ventilation isolations are overridden by this procedure, the operator is continuously monitoring the release rate within established guidelines. DEOP 500-04 is written to control the flydrogen and pressure generated from a LOCA. Any oft-site dose increase will be according to established Primary Containment Ilydrogen/ Pressure limits. The consequences of a malfunction of equipment important to safety has not increased. 2. The possibility for an accident or malfunction of a different type than any presiously evaluated in the Final Safety Analysis Report has not been created. This change ensures proper operation of the vent / purge valves, SEGT, APCV, and Reactor Building Ventilation for Primary Containment pressure and 51ydrogen control in accordance with established operating practices. In addition, the proper operation of the APCV system removes the threat of contaminating the Turbine Building and damaging the Drywell Purge Fans. This procedure change prohibits repressurization with Nitrogen or Air and prutects Primary Containment from overpressurization due to a Ilydrogen burn. SBGT is protected from damage due to excessive pressure. The total amount of the radioactive release is limited to what is necessary to protect containment. Proper operation of these systems does not create the possibility of an accident or malfunction of a type different from those evaluated - in the FSAR. 3. The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. There are no Technical Specifications where the requiranent, associated action itans, associated surveillances, or basis were affected. DRS 2000-07 Liquid Discharge Monitor Calibration. This procedure provides instructions for the Technical Specification required calibration of the liquid - discharge monitors. (Revision 02 - Implanented June 7,1993) Changes: the procedure change corrects an error of understanding of the Eberline electronics, which had multiplied the value read from the Data Acquisition Module by a factor of 2. The calibration constant - determination has been corrected per Eberline. Also, the method of - determining the alarm setpoints on the Service Water Monitor was clarified to eliminate problans if the monitor was reading background or slightly negative. Safety Evaluation L The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as (L:\\%RProc\\pintmgr\\gfs93\\0045.93)

1 1 l I i accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. There are no FSAR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created. The indication function of the liquid radwaste monitors did not change by adjusting the calibration constant of the Service Water Monitors or by the methodology of calculating the alarm setpoints. The Radwaste Dischari:e Monitor is not affected by the stated changes since the alarms and the method of calculating the alarms is contained in DOP 2000-110. 3. The margin of safety, as defined in the basis, for any Technical Specification, was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were alTected.

5.3 Significant tests and experiments not described in the FSAR (Units 2 & 3) Significard special procedure involving tests not descrilxxl in the EAR which were approved during the month of July,1993, are listed bdow: None. f r p T 4 i i i l 4

5.4 Eafety Related Maintenance (Unit 2 and 3) Safety related maintenance activities for July,1993, are sumr.sarized in the attached tables, i I ? P i e i ' I 4 1 i i

.ORESDI N UN11 2 p 9 .,c,, SAFETY RELAILD MAINIENANCE NATURE OF IER OR UUTAGE MALFUNCTION ' EQUIPMENT MAINTENANCE NUMBER CAUSE RESULT CORRECTIVE ACTION 3616-8/0-502 CORRECTIVE N/A HALUE MAUAL_CLDEF 7Al FW HTR 00242't IEFFORMED HELDING PER HPS 1-1-D, WIFS, }UBESIDEDRN: WELD _MAELAND.DAP_3-2._EERFORMEli ALL PNVS PER PMI SHEETS. 2:1622n CORRECIIVE N/A REMOVED. 0LD, XMITTER. FAIN 1ED DRACKLT. XMITTER. TORUS TO REACTOR BLOG D02572 DIFF INSTALLED NEW XMI1TER. REIUDED. 100K A9 LEFT VALUES. PhV PASSED. 2-7800-20-1 PR5V5Nilyl N/A ---= REE0VSU 5ftIEhKEl B0fi;0VER.~~NIRE 67496 MOTOR CONTROL CENTER 2r '005438 t; WAS NISMARKED/AS 67494. REV PlH IN PKG li 10 VERIFY CORRECT HIRES ON PRINT T E R M IMLI.0tLP_ DIN TE._____ _.__ _ _ 2-1347A PREVENTIVE N/A SWITCH ISOLATION CONDENSER HIGH D06255 ISOL ATED 2-134 7A,D. FUNCT CHEEKED HIlli LEVEL DEMINERALI7ER WATER SWITCilES SAT. ALARM IN CONTROL RM SATISFACTORY. .I ' 2-1347B' ' PREVENTIVE N/A LEVEL _ ISOLATED 2-134 7A,0. FUNCI CHECKED WITH ' ShlITCH"IEDLATION CQ@ERSER HIDH ' 006M5 DEMINE'RALIlER WATER SWITCllES SAT. ALARM Td CDN1ROL'RM SATIBFAC10RY. ~ 2-1600 PREVENTIVE N/A ' ORVWELT CDVER~%ND 5RIEED BLOCK 006021 INSTALLED DAYWELL COVER B SHII:LD DLOCKS REMOVE-INSTALL FER DHP~l'600-05~ALL PERTINENT UTEPS. ADDRESSED ALL GC HOLD PTS AND l'MV'S. h305IUlT46:35 PREVENTIVE ^N/A ~ - - - - - - --~~~---- VLV" PARTS CLHD/ INSPECTED'OR'RFFl. ACED, VALVE MANUAL GATE DRIVE INSERT D06886 RISER ISO INCLUDED PKND, USKT WEDOE, SIEN a BNN1. ULUE CHECK ACCEPTADLE. VI.V 10ROUED 10 PROPER VLV,LLFT_IN AS FOUND C0110I1104. 2-305-101-50-51 PREVENTIVE N/A VLV MANUAL GATE DRIVE INSERT 006890 ESTABL FEEZE SEAL TO INSPECT / REPAIR VLV RISER ISO INTERNALS. PERFORMED ALL PNV's PER PHI ~ SHEET AND RETURNED ALL TOOLS AND EQUIP in7HCIR~~ PROPER ~LOCAT11)Ni - ~~~ 2-305-102-42-07 PREVENTIVE N/A PERFORMED HORK FER MMP/TRAVELIIR, VLU MANUAL GATE DRIVE-WITHORAW 006902 ~~ DNP 306-05, DMS 305-01. PERFORMED l'MVS RISER ISO PER ' PRT7HEET ~~~ ~ ~ 2-305-102-50-11 PREVENTIVE N/ VLV" MANUAL" GATE DRIVE; WITHDRAW D06905 ~~ ~~ ~~~ ~ ~ ~A PERFORMED HORK PER DNP 300-05 AND RISER 150 ~ MNP/ TRAVELER. FERFORMED Al l PleiVS OF riiT SHEET. / e f 23 ~+

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NATURE OF L ER Uli oui AGE MotfuNCilON EQUIPRENT MAINTENANCE HUMBER CAUSE RESULI 10RRECTIVE ACTION ~2-2301-64 PREVENTIVE M/A lVLV GLOBE A0 HPCI STP VLV ABOVE D07828 HURK COMPLETEf1 f ER illIS NMR AHH IC It st.ai unn ASut NP 1D0/328 01. Gr 2-2301-2 PREVENTIVE N/A TRAPHVLA unAIN FUT~B' DU7903 PERFORMED TRAP INSPEC TION /SURVElt.L ANCF iu AFPROVETTTRUCEDURES^ AND INGTRUCTIniG 2*150P22A PREVENTIVE N/A - ~ - - - - - -~~- -- OPERATOR MOV 1501-22A D03149 REPLCD f10 51501'22A~MTR PINION GR GEI, L REPLCD SPRINGPACK COVER. WORKl:0 IN CONil V UNCTION H/D72212. ALL GAOKETS LISTED IN 072212. 2-1308-17 PREVENTIVE N/A OPERATOR AIR ISOL CDSR STEAM 008494 INSTAfLED 7 FACKING RINGS, 1 DNAIDI D UENT TO MN STEAM GRAF01L, 1 ERAIDED, 1 ORALDILe 1 EUAJI! Elf s CHGD OUT DIAPilRAGM. STRKO VLV R VERIFIEn t-i"~STt7t TRAVEt-E 37'PSIG LIFT ' PRFS3URI. 2-1301-36 PREVENTIVE. N/A . VALVE' CHECK 190 CONDENSER HELL, 008889 REMOVED CONNET FROM VALVF AND INTERNALS. RhiER INLET REINSTALLED INTERNALS, TOROUED BONNET 10 M FT7tUS. 2-2301-35 PREVENTIVE N/A vntvt. Un i t. r10 HFCI~FUNP SUvi U03709 NORK CUMPLETED PER THIS HHR AUU FCII FR WPT UDU70W-01 F ~~~' -[OM-TORUSISOL bO7UU4 U-77U cuRRECTIVt. N/n AMPLIFIER CARD LOCAL POWER 009003 UMKD ITRM70209C UV UIt' 700-11 DEiCTR RANGE MONITOR AC CONNECTOR DAD, NELDS Cl!KD OL9tHG Uttit OUTAGE. REPLAI:FD PWR SUPPLY. t;t;MPlillEU DIP 700-4/05 24 BAD / DIS 700-9 FOR US 24 2-2301-20 PREVENTIVE N/A VALVE CHECK HPCI COND STORAGE D09475 HORK COMPLETED IN ACCORDANCE H)Til 1H1D TANK SUCTION NWR AND FCII NORK PACKAGE 009475-01 2-1501-65C PREVENTIVE N/A VALVE CHECK C LPCI PUMP MIN 009477 REMOVEli DONNET FOR INSikCiION AND RE I NS I AL LEU. _....___,..._ -_. '2-1501-63C PREVENTIVE N/A VREVE~ CHECK U~'LFCI~FUMP 009478 REMOVED /INGFE CTED VALVE. CLEAN!-D VALVI 01SCHARGE FL ANGE!TT REINGTALLED~ CHECK VAL VE. 10RQUED DOLTS 10 450 FT/LDS. ItAP 1J-1rs INITI1 AILD FOR NEW STUDS AND Nti1L m,, ,,=.a. _e..:,se a Q ef 2 D s

HA luke Ol' t rk flu olll AGE MAlfitNCilOH EQUIPMENT NAINIENANCE UUNDLR CAUSE RESUI r I:ORRECTIVE ACiluu 2-1501-638 PREVENTIVE N/A Ri:NOVI.D CHELK UhLUE. INSIALLEN ilI# Vnt. ' Cfu d_.LPC L_E' UMP __ fl09_4ZY..- _ _ _.... _. _ __ _ REMOVED /OIGAGSEMBLED CllECK VALVE. r IOROUED PER DAP 11-20.

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~ ~ ' - ~ ~ LEG LT-646B FULLDHING HTElS G5 TilRU G12. 60 DREG 5Efl DC HOLD POINIS AND PNV'S. REINSTAlt.ED FLOW CHECK VALVE. 2 D253;179 FREVENTIVE N7A ~ =-- -PERFORMEU DHP'0263-01. VALUE FLOH LIMITING CHECK RPV 009494 -INSTRUMENTS 2-0263-15B PREVENTIVE N/A VALUE FLOW LIMITING CHECK VAR D09495 VALVE AND POPPET CLEANLD, TESifD UKAY. 1ELEOR_2202 -_. 2-0263-138 PREVENTIVE H/A

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NAIURE OF LER OR DuioGE MAliUNCi[ON

1. EQUIPCENT MAINTENANCE NUMBER CAUSE RESUL1 CORRECTIVE ACituN l2-1601-55 PREVENTIVE N/A vfiDLENDID DW N2 PUR0E_VLV OB D10631 RFMOVED OLD UOLENDID. RECEIVED NEN

- - - - - - - - ~ ~ - - - - - - +9ei ~ 90LENulD, UENCH TESTED, INSI Al I Ett tUOGED Q+ TQ BOETErfTAFFO AND SPL1CED.'MEGUFRFD~ CIRCUIT. FOR 1 MINUTE CLEADED 00S. 2-201 PREVENTIVE N/A REACTUR1ANUAl~CONTR0E'HI5c D10440 - ~ ' CET TIMER CONTACTS. FOUNE W1RE ON IERHS-l ' NUT Dif PRINTTh1RES~TO SE f UF LAP 9 IN YO2-2P PANtL. REPLACED LUGS ON TERHS 08, MA AND 7A. 3-2301-1 FREVENTIVE N/A RAP HPCItDRAIN-POT 4A' 010673 DUERH5UEEU~5TEAH TRAP' DHP-~2300:5.~ f %- l 2-2303-MSC PREVENTIVE N/A MOTOR HPCI SPEED CHANGER D10715 REMOVED MSC FROM FRONT Sill. DISASSEMClED / CLEANED & INSPECTED. REASSEMDIED M/ UNIO gg, p PARTS /NEW OREASE. MSC INSlALLtO IN ORlG LUCATIUN7EINKAGE CONNECTEDT~~~~ ~-~ 5-7328-6B PREVENTIVE N/A SWOR BRK DW COOLER DLOWER 'D10763 INSPECTED.AND TESTED NEW OREAKER. i2-0/.44-6 DDTAINED ARCH-CHUTES ON 5/12/V3 ANO INSTAL EEU UREAKER'IffTO' CUDICLE 7 !2-7320-6D PREVENTIVE N/A SWUH UHK UM t00LER'BLOWEN UIO764 Cl.EANED CUUE, EllECKED DKRS PHYG COND r(OO

2-5734-D n:.bT~ VERIFIED' SETTINGS AND TRIP. USING

[i NtJLTI-AMP SYSTFM PERFORME0 1ESING. PASSED. - IOl~6Ul~'UV PREVENTIM. N7ff - - ~ - - - - - -SOLENDID DW AIR NAKE UP VLV 0B D10785 CENCil TEUTED~ DIC RENOVED OLDS REPLACED Hf TH NEW SOLENGID. RFPLACCD H tRES NIlit RAYCHEM FIf LD WIRES 10 Sul ENDID SPI ICI-AND Jt)NClIDN PER OC. 2-2301-31 PREVENTIVE N/A SOLENDIO HPCI DRAIN POT A. TRAP D10793 PER WORK INSTRUCTION 3, DIO N0f HAVE SYPASS PROBLEMS WITH JOB. 2-7329-6B PREVENTIVE N/A SWOR BRK DW COOLER BLOWER D10941 ClEANIB CUME, BREAKER TEHIED AND 2:5734-U INSTAlLEO. SENT OUT FOR RMS 9 UNIT ANU UVERHAUL. - -~ ~ ~ ~ ' 1 2-0595-146 PREVENTIVE N/A REL F CUNU VACUUH~ PUMP CONTROL U11031~ ~ REPLACED COIL / TRAVELER SfCPS 1 ll!RU 20. EEP1 ACED ~REthV~2-595;I46'HITH LIKE FOR LIKE. REPERFORMED STEPG 9-21. VERIt'IEU HATISFACTORY RESULTS. L.. - _.. __~ // o f .a.3 <a

1 NAIURE OF LEH UR 00fAGE MALFilNCIIUN EBUIPMENT MAINTENANCE NUME:ER CAUGE RESULT CORktL11VF ACTLON 2-0595-112B CURRECTIVE N/A hEFL Af:ED Cli1L R RELAY 10 VER11Y fANTAl'! RELAV NAIN STEAM VALVES 011035 ' m i t.m.uu s HAS LESS THAN, OHM HHEN MtU1 ANlt N UREATER THAN 20~0HM~WHEN UPEN; 2-995-134 PREVENTIVE N/A RFPLACI:D RFLAY COIL. CLEANED AND CllECKEh - - - - - ~ ~ - - - --------- uttav VKUTLUi10E SEUT AUTU-D11U36 ACTUATION RFE AT T'ORTACT S. z-ove-ub FREVENTIVE - NTA REMOVED'OLD COIL ~AND REFLACED HI1H HEF RELAY DRYWELL ISOLATION LOGIC 011037 TRIP GROUP II PER WORK INSTRUCTIONS. U 2-595-136 PREVENTIVE H/A 1:EPLACED RILAY COIL ANU VERIFtED rh0K R RELAV PROTECTIVE SBGT AUTO-D11038 ACTUATION CONIACT MAKE UP. ALL CONIACIS LESS 1HhN .5 OHMS CLO5Lu GRE A I E._R__T_H.A.N. __1_K I_LO. OU p 2-0595-143J PREVENTIVE N/A RELAY' TRIP RESET AND SEAL /VLV -D11042-CHECKED C01l AND CON 1 ACTS FOR VOLTAGE. 16ui-o PCUNTRUL REMOVED OLO COIL. REMOUNilD REIAY, LONDED ~ Tire TJIREUT' DOCUMENTED I TF1ED/ LANDED LE AD LOG. i2-0595-144 PREVENTIVE N/A RELAY TRIFRE6b AND SEnt7Vth v11U43 RLMVD/ REPLACED ULD CHIL. HEASUREO REGtGt 1601-57 CONTROL nNCE UF NEN CGIL".~ LIFTED 00TTOM MIf?ES DF COIL. ENERGIZED COIL AND MEASURED RESIST ANCE OF COH1 ACTS. RECONNECTED NIRES.

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^1JD TOH90El_lNUTS 10 45 FT/LDS, 2-1402-3A PREVENTIVE N/A UNPACKCD/ REPACKED W/5 RING SEI CARUCN ' VALVE MD A CORE SPRAY PUMP D11234 TORUS SUCTION SPACER, LIVE LDAD SPRINGS AND EYEBOLTU. 2-1501-5C PREVENilVE N/h VALVE MO GATE C LPCI PUMP D11242 NFh0 Veli PACKING, INSIALLEO NEN 5 RfNo

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VDEVE Mu unit u ITCI7 UMP vitz43 REMOVED / REPLACED PACEING (LIVE LOAN)

~ SUCTION TUSTarrEU TTEWEYEC0tTS t27, Hr7 UUTS AMP DUSHING. ,f h- /2 ,f 23 g -,,-=a -.-g m .+ a - - - - p .i,,. e+ t +- .mu

NAIURE Ur t.LR Hit OUTAGE MALF UNf' t 10N EQUIPMENT MAINTENANCE NUNDLR CAUSE RESULT CORRECTIVE AC1104 '2-3706 PREVENTIVE N/A X BUILD C.00L WTR FROM D11258 RFPACKED VALVF WITil LIVE 1 UAD i Ek UMP 0040-40. ga,, p ,.W 2-1201-1._- valve MO REACTOR CLEANUP INLET. PBEVENTIVE _.f / A D11264 kLPLoCED_EACKING_HITil. LIVE LDAU PER ISOLAT20N UMP 0940-40. 2-1201-2 PREVENTIVE N/A VALVE MO GATE RWCU AUX PUMP D11265 UNPACKED /RI. PACKED VALVE PCR DNP 0400-40 BYPOS!/, L1 2-0202-48 PREVENTIVE N/A > OPERATOR MOV D11279 REMOVFD MOV COVER /PERFURMED EO SURt.*. ND PROBLEM 100NU. PERFORMED MOV UIGNAIUHC UNDER 2-4 I ERER _t100_ HR.J2139. r 2-1501-18B . PREVENTIVE N/A OPERATOR RGV. 011301 1 ROTOR MOD IN PROCESS WR DV2139. '2_-1501 k @OV-E8EVENIlVE N/A 2 -OPERAID D11303 COMPLis]_E! LED _INSFEC f(ON PCR Dir 40 1. E71501-30D PEE'050iIVE N/A OPERATOR MOV D11305 PLRf ORMLU LO UllRVLIl.1 ANCE > NO PRURI EMS i OUND. FERI ORMED MOV SIGNATURI: PER N00 HR D9213Y. 2-2301-49 PREVENTIVE N/A .0PERATOR NOV HPCI CLG WTR TO D11310. DID MOD PER (1-2 MOV 2301-49. REMOVED

CDMILSIDIL

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NAIURE OF t LH U1: 001 AGE MALFUNCiION EQUIPMENT MAINTENANCE NUMDER CAUSE RESULI CORRfCTIVE ACTIllN =2-3703 CURRECTIVE N/A -- --~ UVERHntill;D VALVE AND SIGNATUR! ll. , PERATOR MOV D13395 p R LUE~HANUAE GATE URIVE-' INSERT 305-101-19-23 CORRECTIVE VA U13691 - ~~ N/A PERFORMED FREEZE ON 101 VLV FI h GTATICU RIBER ISO IRAVEELR T DMt! 0300 01 & DMF 040b-05. 2:1301~-T3B PREVENTIVE H7A OPERATOR MOV D13760 FRT DIFRHF FROM' WRNG ~ IN ri D. id 11 RMEO r. l' RF TERMED TO AS CLT. REPLCli GR1 ASE I N l.h l i is SNTH. REPLCD 1 SEALTITE 10 LM1. REPLCD GRD_OlLLMG_FRE.fE8F_Jt0LUIGN_LER P-U. 3-1501-13A PREVENTIVE N/A OPERATOR MOV D13761 COMPLETED OVERHAUL. lI 2-1500-D00R PREVENTIVE. N/A 000R:CCSW VAULT ROOM D13775 PERF liURV.B ALL REFAIRS, NEED 10 ASSIST IN TEliTINO OF CCSW VAULT ROOM 000R. ALL HORK DUNE IN 1CCORDANCF OF MMP & TRAVtR. 3 DMP 1500-01. ADDR alt. PHV ANr. OC HOIDb 2-1601-61 PREVENTIVE N/A - OPERATUR A07AEVE~DRVWEEE VENI v13777 RFhVD GPIR FROM VLV. CLEANED R INSPEC1ED BYPASS: nLL'PARTSTTORDUED' TIE RODS TO'60 FT/t M SIROKE Cilkfi & CHKD FOR LEAKS. REINSTALLEli OPERATOR ON VALVE STEM. - 2U19y rREVENTIVE N70 . OPERATOR A0 VALVE DRYWELL VENT D13'773 REMOVEIT-~'AUTREPLACED' WITil HITG CYLINDI:R 159 10RQUED ANU AIR TES frD CYLINDLI:. REINSI Al. LED. MADE NEW DRACKET FOR All< SWITCHES. 2-1601-58 PREVENTIVE N/A OPERATOR A0 VALVE DRYHELL VENT D13779 REMOVED A0 FROM VALVE. REH0 Veli 11E R00.

958.

HONED CYLINDER. REDUILT/10RQUI D 10 624. REINSTALLED. 1601-55 PREVENTIVE N/A

0PERATOR A0 VALVE DRYWELL VENT D13780 DISASSEMPLIO VALVE /01'ERATOR AND REl

UllI

2 1:EINS f 41. LEU. (LCH STAFr DID Lt.NI.

4 U LEAKTI. ~ ~ ~ ' ' ^ - 2-220-2-28 PREVENTIVE N/A PIFINU~RECIRC~DECUTCUNNECTION U13757 CONNECTED IHE DECON PROCESS HOSE TO "O" REFUEL OUTA0E RECIRC 'SCUTION LOOP ~DECorFLANGE. ~ .I - - -. - -. -... ~ / 5 of -t.3

m iG WE 5 AH ~ l i. k h l ll 1 lb VV L EPG D ~ I F YN Io RM 1.RL t' N O N T LD Ol EP E9ED F E D NEI A T RI CP S M TA 4 N ORR _. D R E) L U0 ObtI OD A CU R O PI DC U0 I AS l t UN UH SiD EC _ t O l. TGRN t 0 Ir-JA D ,St I D_ iUPO A TD Et3G a,- CA O NoiG E GET A ~ 4 1E RK N NRS I L E DL C LS E Tl s N 0C A IP U V EPD PA 5T Ol5A 0 E t a K,U M n V NG EW _ 1S PA1i f 1H L m C U M 0MAU R E L o T EC C AGN C _. E HE N Y DT ,f m RO C V PN ~ HV m C B_ EMSS_ .A N 6' A L ;E IL A i G WL t A. l m AU M ,RL n EE E VL E S U a_ D u FSI. N A_ 9 5hRI NR 3L NES m V O T PLB E L 0IRQ / 1I SGDE I K _ S MA C 1 NR T ND _ C S1NI T CR i_ AEG A IR0L L 0N

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NATURE OF IEH OR OUIAGE MALFilNCTION EQUIPMENT MAINTENANCE NUMOER CAUSE RESULT C0liREC11VE AC110N -r M-11570-277 PREVEtti1VE N/A R{RRAS2A-202 PUMP D15176 HURk I:HMitt TED IN ACCORDANCE Hil!! IHP NMR ANUFC III WORK PhG DIU176 4)1 4 02 b 2-0700-40-490 PREVENTIVE N/A ECEMENT LFRM EUCAC POWER *RANGt U15260 ~~ ~~~ HIGGLf:lt CARD 20-40-49. PE RIODJ t: UPUCAll -~~- ~ ~~~ ~

MONITOR

~~~~ SPIKING ~0CCURFD. REMOVED CARD AND CLEANED EDUE CONNEC10R. REIN 50 HIED CnRD AND VfRIFIID RE~ADINO OK. PASSID. 2:07DD APRH4 CORRECTIVE IU6

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HAHEC Hi i Ek OU 001 AGE MALFUNCTI04 EQUIPMENT MAINTENANCE NUMBER CAUSE RESULi 1:0RRE.C11VE AC TION 2-1601-33D CORRECTIVE N/A El NOVED COVER. CLEANI:D AND INSrECil:D UAl UE UACUUM DRFAMFR DRWrt i nmv7n 'TO TORUS 933D 0; RING _11uRg ACLS_AND. VALVE.INTLI NALS. ~ INSTALLED NEN 0-RINGS AND-TUR9UED 001.I9 .6 10 200 FT/LB. 2.-1402-37A -.p_._ CORRECIIVE m.__._H/.fi FER_ ATL IRAVLLER. AND PMP, VAL'!L FA3 VALUE GLUBE A CORE SPRAY DISCH D16097 HDR FILL REMOVED AND Ri f t ACED PLR NPS 1 -1 B. 2-1402-36A CORRECTIVE H/A REMOVED PACKING, CLEANED UTUFI.1NU DOX tt VALVE STOP. CHECK A CORE SPRAY D16098 STEM, FEPLACED VALVE PACKINii AND {, TIGHTENED. 2-305-120-38-03 CORRECTIVE N/A 10 WIRES LANDl:D ON TERM CIOCKi COLOR SOLENOID VLV DRIVE WITHDRAW D16308 W/SFEED 29-03 K-1 CODED., WORK PRG NEEDS l.IF1ED/I ANDEli LUG j SPRIN1S.[0, VERIFY. HIRING.AND l'ARIS. LIB 1 3/4 SEAL 111E FITTINGS. REPLAClO SEAllll! 2-1101-98 CORRECTIVE N/A VALVLGLOBE. R EBLC DISCMARGE D166D2 REMOVED NUT AND THEN REMOVED UROKEN INBD DRAIN liANDHMEEL. _ CLEANED.TliREAD11. IhnTAll.ED NLW HANDWHLEL, TIGHTENED NUT IULL llUf ENGAGEMENT. '1RUBBfCCORE~5FRAV UNE -1405-30 PREVENTIVE N/A 016702' HORK COMPLITED IN ACCORDANCE H)1H FC11 2-1403-10" HURkyFACKAut.f 6782-01 AND NWR. lI 27132F40' CURRECTIVE N7A FUT CRFAREli~f ROM 29 10'29-5^AHii 4 JH I:Os SWGR BRK TURB CLOG D16904 MCC 27-1 2-2827-1A1 2710 MCC 27-1. PUT OREAKFR FROM Dlts ?/ 10 27-t IN Bll3 29 TO 2Y-5 AND 6. _e 2-7329-60 CORRECTIVE N/A SWGR BRK RX BLOG MCC 29-5+6 016904 PUT CREAKER FROM 2V TO 29-5 AND 61H UUS 27 TO MCC 27-1. PUT OREAKER FROH DllS 27 2-7_822-5Ali6AL _ 10_21-1 IN BULL 2LT0 JCC_29-5 AND _6. 2-1501-28 CORRECTIVE N/A RtPACKED VALVE PER DNP 40-07, HE!DINH VeLVE. MANUAL _.GME ILCCSR_P_ UMP D1105L _._______ flf.R WPS.1 + D, WIR'S, HELit MAPS, DAP t DISCHARGE lt PERI?hRMED ALL APPLJCABLI: PMV'S PI-R REPACKING ON PHT SHEET. M-11540-74 CORRECTIVE N/A-- SUFFORt MKUPUEHIr4315 2" D17090~ ~~ ^ ~ WORK COMPLETED PER TilIS HHR ANU F Ct1 HORK~l ACK ADE lil70Y0-01 ~ CLASS 2 ehnee in-md-. i uim hw esau4-et-e h6w ,simw-. em.=w-b - > usi d% ed ems-em,. mui wh se es'

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NATURE OF LEH UR OUIAGE MALFUNCTION V. ,g . EQUIPMENT MAINTENANCE NUMOER CAUSE RESULT 00khECTIVE ACTION

• 2-1501-2C CORRECTIVE N/A UEPACKED VALVE TER DMP 40- 07 AND NLLfit P

' VALVE MANUAL GATE C CCSW PUMP 070157 ' Pm m R06 1 ERWPS t B, DMP 40-30, NELD MAPS 3 Ht.L" It ' INSP TiECORDi DAP 3 7.~ PERFORMFli All. p APPLICABLE PMV'S PER DNP 40-0/. 2'0220-51 PREVENTIVE N/A REMOVFD HLD DIAPHRAGM AND INSIAllED N;W OFERATUR~RX READ'UIRING EEARDF F U91514 ~-~~~~ DIAPHRAGH.~' ~ ~~~ ' ' ' ' ~ ~ ~ ADV 220-51 E-vzzo-oz FREVENTIVt. IUn HEMOVEIT"ULD~ DIAPHRAGM AND INS 1 ALLED NT W OPERATOR RX HEAD 0-RING LEAKOFF 091015 DIAPHRAGM. ADV L _, 220-52 2-0205-25 PREVENTIVE N/A VALVE MANUAL RX HEAD COOLING 091818 REMOVED /REf' LACED PACKING. VALVI CYCLLf! INB00RD TEST LIMES AND PACKING NUIS T1HHIEN10 EACll CYCLE. VALVE LEFT WIIH CLOSED POSIIION. p I-2-1705-28 PREVENTIVE N/A RAUI'MT10N 094657 PERFORMED AS FOUND CALIBRATION. REMOVED MONITOR MAIN STEAM LINE EFRDMS AND INGTALLED NEW AS Fl:R WORK TNSTR11CT10NSTPERFORMED~ A9 LEFT CAL'S VIA DIS 1700-01 REV 16. 2-1705-2C PREVENTIVE N/A PIRFORMED AS FOUND CllECKS FOR :'C MbL KUNIlUR MAIN bitAM~Elht UY4658 RADIATION LEMT REPEACED EPRDMS. PERFORMED AS LEFT CilECKS FOR 2C MSL LRM. 2 17U5:2I1 LORRECTIVE TT7A FTRFORMED AS TUUND~ CAL'S:~REPl ACFD rrRnn MONITOR MAIN STEAM LINE D94659 RADIATION CET PER DE FD1 (NDGOs REV 1. M R! ORMEli 6B LErl cal'G. 2-1501-448 PREVENTIVE N/A PUMP CENTRIFUGAL CLG SERV WTR D94903 ASSEMDLED l' UMP a ALIONED HDTOR FER [.- DMP 1501-04 REV 5. TLCH GTAFF OIU VIBRAIION REAl INGS. PASSED REAllING. 2-0719-16-33 CORRECTIVE N/A DETECTOR NEUTRON 4 ELEMENT 095319 COMPLETED DIP 700-6 FOR NiW LPUM SIRIUn STRINU T.PRM 7 6;33~~' COMPLElED DIP 700-16 FOR I.PRMS IN NEM LT'RM TITRINUS '~'~ ~ ~ ~ ~ ~ 2-0719-32-49 CURRECTIVE N/A UETECTUR' NEUTRON TELEMEN I D95319 - COMPLLIED DIP 700~6 FOR NEW LPRM S1PIUW STRING LPRM 32-49 CGMPLETED DIP700'16 FOR IPRHS IU NEU I l'RM SIRINOS. em w-.- e.,- -. - 1/ of)3 k'

r; NATURE OF [ER OR OlllAGE HALFUNCTION 4 4 EQUIPMENT MAINTENANCE NUMBER CAUSE RESULI CORRECTIVE ACTION 2-0717-40-25 CURRCCTIVE N/A

---

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COMPLETEli IIIP 700-6 3 DN NL W L:SM 51FINO i gM_g ELE.f1EN T 095319 ( 11 L t 100-16_FOR_LFRNG_IN_NEH_

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2r 7-3Q-i1 CDMELEJELLDIP_70026 00R NEN LI?RM. STRINGU GE ' TUR _NEUIRON 4 ELEMENT. CORBERTIVE N/A UV531V STRING LPRM 40-41 COMPLLIEU UIP 700-16 FOR LPRMS IN NEN s LPRM STRINGS. 2-0719-40-49 CORRECTIVE N/A COMPLETEli OIP 700-6 FOR NEW LPRM SIRINGU DETECTOR NEUTRON 4 ELEMENT 095319 STRING LPRM 40-49 COMPLETED DIP 700-16 FOR LPRti9 IN NEW P* LPRM STRINGS. -0700-24-57D PREVENTIVE N/A 1;0MPLETED DIP '/00-6 FOR NEW Li'RM SIRING:, JLEMENT LPRM LOCAL POWER RANGE D95319 MONITOR COMPLETED DIP 700-16 FOR LPRMG IN NEH R v: Li RM SIfML_ CHANGED _0DL5_NCH_ LORH'S. 3 $a203-1A PREVENTIVE N/A g g g DARD MAIN STEAM. - D96_138 FCII HORK COMPLETED IN ACCORDANCE HIlli THIS _ NHR AND_ECII._ WORK PACK AG_E_ DY6138- 01 i 2-0220-51 CORRECTIVE N/A

---

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UnLvE. AU 10rHEAU U RING ~EEARUFF.096756 CHANGED 0U1' THE CONTROL SHI TCit PER WOI:K 1 - tr7%MC, INSTRUCTIDHSTLEFT DLD~NAME~ PLATE ON NEH { pgt~r $7+ SWITCH /NEW SAID (OPEN/CLOGE), SHOULD SAY - (MONITOR / DRAIN) i M BUT 4WD l CURRECTIW. (Un AMMUEu a AI IGNEITTHFFER DMP~ 7'J01:04 REV PUMP CENTRIFUGAL D CONTAINMENT 097364 5LO SERV HTR G. INSTLD HEW IMPELLER ON SHAl1. Ri OL.1 l ~ PFR WRK INSTRCTNS. REPl ACEO S,N 3/8 "T' RUN OF 3/8 TuilNG__&_I_URREL AT f.ONNEC110U j 9-0263-22A PREVENTIVE N/A INSTALL.ED 4 NEH 710DU MitP S IN FANELG. TRIP NASTER UNIT ATHS REACTOR D98478 PBLSSWE_ PERFORMED DIS 260-3 AND DIS 250-4 iOR S PNV'S. .-.m . _~._ $-0263-228 PREVENTIVE N/A INS 1ALLED 4 NEW 71001) HTU'S IN PANI LL irRIP HASTER UNIT ATHS REACTOR D98478 PERFORME D DIO__2_6_0__3 A_ND 01S _260 4 f OR 263-22C PREVENTIVE N/A INSI All_ED 4 NEN 710D11 MTIP S IN PANELU. IF MAUTER7 NIT ATNS~REACTUR-v7347U >RESSURE PERr0RMED DIS 260=3~AND DIS'27,0-4~FOR PMV. L F H -u.m_- _._ue___u.m uma__hm_-__.-ms.m .u___ ___.s__C__._2__-a._2_.__ m i ___..n._-_ _,2._ a ..._._.2.

w 1' y. NATURE (IF LER OR Util AGE M ALF lINC T it!N sii.:, JEQUIPMENT MAINTENANCE NUMLER CAUSE RESULI CORREC11VE ACTION 2-0263-22D PREVENTIVE N/A INSTA! LED 4 NEti 710DU MTil's It! PotJELS. TRIP NASTER UNIT ATHS REACTOR D9847G ITRFORMED 015 260-3 AND DIS 240-4 FOR f[gttbuNr. FMV. 8. -q, b m.. e m---.. se.%- ( 1'h_s n' Y E - ~ ~.-- h-.-- =w .-w. -a, b mem. 1.. + ..-e% ew,a--m E s L_. 3 i. ~ m-han-r-+- -rism.... h ~~~ - - ~ ~ ...-w. A3 of 23

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1 J 5.5 C-N Safety Rdated Modification (Units 2 and 3) i Modifications reported to Regulatory Assurance which were authorized for operation during the months prior to July 1993 are listed. Changes implemented after June 30,1993 will be reported with the FSAR updates in accordance with 10 CFR 50.71(c). For case of reference, the changes have been identified by their design change contrul modification number. Previously, only inodifications that had been completely closed out were reported. l Modification No. Description M12-2-92-021 This modification involved repairing the cracked Access Ilote Covers. The repair eliminated two types of cracks: (1) circumferential cracking, which developed along the weld affected area and (2), radial cracking, which could pmpagate to the vessel or shroud wall. The original plates and the weld. affected area were Electric Discharge Machined out. The new plates were designed in accordance with ASME Section III, Subsection NG. (Authorized for Operation - June 8,1993) Safety Evaluation 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. The probability of an accident will not be increased as there are no new failure modes associated with the AIIC of EDMing out the weld affected area and potential radial cracks. The new cover plate will be J-Bolted in place rmd tack welded to prevent movement. This repair will not alTect any components or systems that mitigate or prevent accident conditions. This repair is also designed using ASME Section III design requirements as guidance, material requirements, and construction standards. The design encompasses stress limits which will ensure adequate margins for all design basis ctmditions. The replacement cover has bem designed for a 40 year life. t The probability of a malfunction of equipment important to safety will not increase. The purpose of the AHC is to close the access holes in the support shelf. Closure is required to maintain two-thirds core reflooding height following a LOCA. During normal operation, the AHC joins with the shroud support shelf to maintain a structural barrier to direct core flow. The replacement coven src safety related, seismic class 1. The covers were designed to remain in position under all normal and accident conditions. The replacement coven were designed for differential pressure loads that exceed 100% Core Flow Conditions. The reactor vessel thermal cycle drawing was used as the temperature input for the design. Stress calculations were performed in accordance with

adequate maq; ins of safety in the design. The replacement AIIC design wi!! not impair the function or structural integrity of the shroud support shelf. The consequences of a malfunction of equipment important to safety will not increase. This repair will not effect any fission product barrier, increase any radiation source term, or prevent any component from performing its safety related function. This repair will prevent the original welded cover from detaching from the shroud support shelf and damaging fuel, being sucked down a recirculation pump suction line, or preventing the core from being able to maintain two-thirds core coverage in the event of a LOCA. This repair will also cut out the circumferential cracks in the weld affected area which is postulated to be initiating the radial cracks which can propagate to the vessel wall. 2. The possibility for an accident or malfunction of a different type than any previously eraluated in the Final Safety Analysis Report has not been created. This repair will not create an accident or malfunction different than evaluated in the SAR. ASME Section III Subsection NG was used to assure reliability and adequate margins of safety it. de design. The materials of construction are compatible with the vessel internals for a 40 year life. There has ~ been no new malfunctions that have been associated with this repair, nor does this introduce a new method of impacting other RPV internals. 3. The margin of safety, as defined in the basis, for any Technical Specification, was not reduced. This repair will not significantly effect any accident of transient safety analysis which forms the basis for the Technical Specifications. The only limit potentially impacted is the MCPR safety limit. The MCPR safety limit considers the effect of core flow uncertainty. The amount of ' bypass flow is negligible during steady state conditions. The total cost flow uncertainty therefore runains below the value used to generate the safety limits. See the General Electric Safety Evaluation for Dresden Unit Two AllC repair. Therefore, the current MCPR safety limit is valid and the basis for the Technical Specifications will not be effected provided no more than one double tap per recirculation loop and two single tap jet pump flow instrumentation are out of service.

i j 5.6 Tanporary System Altwations Installed (Unit 2 and Unit 3) A "Tesnporary System Alteration" refers to electrical jumpers, lifted leads, remosed fuses, fuses turned to non-conducting position, fuses moved from nonnal to reserve holder, tanporary power supplies, test switches in alternate positions, tanporary blank flanges, and spool pieces. Alterations ccmtrolled and documented as part of a routine out.of-service or other procedure, alterations which are a normal feature of system design, and hoses installed as part of a venting or draining process are not included. M410 wine information rdlects temrmrary alterations which were installed durine the months prior to July 1993. Chances implanented aftw June 30.1993 are croected to be rtworted storte with the FSAR updates in accordance with 10 CFR 50.71(c). Log Number Description 11-13-93 (NWR #19642) This temporary alteration attached a ruhher hose to the vent tap line and ran the hose to the EllC system floor drain. This was done to provide the Service Water Receiier Tank with a vent path, as the existing one does not work. (Date installed - June 29,1993) Safety Evaluation 1. The probability of an occt.rrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. There are no FSAR affected accidents. 2. The possibility for an accident or malfunction of a different type than any previously evaluated io the IPmal Safety Analysis Report has not been created. Adding a sent path will not adversely impact the systan or function. Currently, h service watrer radiation monitor vent line C+es not work, and it does not have a way to sent. Adding an additional vent line will allow the tank to properly vent. Because it replaces and performs the same function as the existing vent path, no additional failures will be introduced. Also, the EllC system floor drain will not he affected because the drainage from the vent line will not exceed the floor drain capability of 5 epm. 3. The margin of safety, as defined in the basis, for i any Technical Specification,was not reduced. There are no Technical Specifications where the i requirement, associated action items, associated surveillances, or basis were affected.

5.7 other Units 2 and 3 Required 10 CFR 50.59 Evaluations Other required 10 CPR 50.59 evaluations include Set Point Changes (SPC), Rigging Evaluations and changes to equipment not reported in Sections 5.2 thror.gh Section 5.6. Changes implemented after June 30, 1993 will be reported with the FSAR update in accordance with 10 CFR 50.71(e). Item Description None}}