ML20059G922

Submits Per 10CFR50.59(b)(2),10CFR50.59 Rept of Facility Changes,Tests & Experiments for Dbnps,Unit 1

ML20059G922
Person / Time
Site:	Davis Besse
Issue date:	11/01/1993
From:	Storz L CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2186, NUDOCS 9311090187
Download: ML20059G922 (88)

Text

{{#Wiki_filter:- -i l l 1 CENTERIOR ENERGY 300 Mod. son Avenue Louis F. Sforz Tc;eco. OH 43652-0001 Wee President-Nuclect. 419-249-2300 Ocvis-Besse Docket Number 50-346 i License Number NPF-3 Serial Number 2186 y November 1, 1993 l l United States Nuclear Regulatory Commission Document Control Desk Vashington, D. C. 20555 i Subj ect: 10 CFR 50.59 Report of Facility Changes, Tests and Experiments 5 Gentlemen: l The Toledo Edison Company hereby submits, pursuant to 10 CFR l 50.59(b)(2), the 10 CFR 50.59 Report of facility changes, tests and experiments for Davis-Besse Nuclear Fover Station, Unit 1. i Those changes, tests and experiments identified via the safety review r process during the reporting period of January 23, 1992, through May 1, i 1993, are attached. This report includes facility changes that occurred during the Eighth Refueling Outage which concluded on May 1, 1993. . provides an executive summary of those changes, tests and l experiments contained in the attachment. The attached safety evaluation summaries do not involve an unreviewed safety question. If you have any further questions concerning this matter, please contact Mr. V. T. O'Connor, Manager-Regulatory Affairs, at (419) 249-2366. Very truly yours, gy . ( JMM/amb Attachment i cc: J. B. Martin, Regional Administrator, IRC Region III J. B. Hopkins, NRC Senior Project Manager S. Stasek, DB-1, NRC Senior Resident Inspector Utility Radiological Safety Board I / 08n021 4 L O, OpemSng Componics -r 9311090187 931101 g j C:ewand De:.tnc liiummat.rg Tomoo rdson PDR ADOCK 05000346 c1 P PDR b

Docket Number 50-346 License Number NPF-3 Serial Number 2186 g La e- . Attachment Page 1 ATTACHMENT I i 10CFR50.59

SUMMARY

SHEET l Number Title i DCR 88-0252 As-Built Revisions to Auxiliary Shutdown Panel Drawings l DCR 90-0008 Revision to Emergency Diesel Generator Figure l DCR 92-0011 TSC Datalink i DCR 92-0019 Revision of the EDG Loading Tables DCR 92-0029 Revision of USAR Figure 10.4-4 DCR 90-0030 Revision of USAR Figure 9.2-5 l DCR 92-0031 Vaste Evaporator Package 4 DCR 92-0037 De-rate Cyberex Inverters YV2 and YV3 5 DCR 92-0050 Revision of DC HCC Single-Line Diagram l DCR 92-0055 Deletion of Circuit Details From FHAR Appendix C-3, J " Circuit Coordination Summary" j q DCR 92-0061 Miscellaneous Vaste Drain Tank Nitrogen Supply Pressure Control Valve Abandoned in Place DCR 92-0067 Revision of the EDG Loading Tables i DCR 92-0069 Miscellaneous Drawing Changes j ] FCR 81-0167-05 Condensate Demineralizer Vaste Transfer Pump FCR 86-0404-01 Replacement of Vater Treatment Sump Pumps with Portable l Pumps 3 FCR 86-0425-09 Remove Installed Instrumentation Associated with the Motor Operators to Valves FV6397 and FU6398 4 3 FPR 80-0189-004 Revision of USAR Figure 9.1-5 FPR 90-1881-901 Replacement of COM-5 Relay in AC109 l FFR 91-1535-901 Change to Reactor Coolant Pump Motor Flywheel d FPR 92-0495-903 Replacement of Vatt and Voltage Transducers in Diesel Generator Electrical System MOD 87-1273 Replacement of Steam Admission Valves to the Auxiliary Feed Pump Turbine

Docket Number 50-346 2 l License Number NPF-3 s'

• Serial Number 2186 Attachment Page 2 l

NUMBER TITLE MOD 88-0027 Upgrade Megavatt Electric Input to the Integrated Control System HOD 88-0034 Installation of 3-Vay Ball Valves at Inlet of Each Instrument Air Dryer MOD 88-0055 Installation of Vafer Check Valves in the Component Cooling Vater and Clean Vaste Monitor Tank Room Floor l' Drains MOD 88-0113 Modification of output Logic of Division One Generator. Protection System 1 l MOD 88-0207 Increase of the Auxiliary Boiler Feed Pump Recirculation Flow MOD 88-0227 Installation of Synchronism Check Relays for AC101 and AD101 l MOD 89-0003 Replacement of Relief Valves DH1508 MOD 89-0094 Modification to Emergency Diesel Generator Room Sprinklers MOD 89-0109 Alternate AC Power Source for Station Blackout MOD 90-0030 Replacement of Synchro Che:k Relays 25Cl, 25D1 in the 4160 Electrical System MOD 90-0045 Removal of Internal Components for Check Valves SV85, SV101 and Removal of SV93 MOD 91-0013 Reviring of DH7A and DH7B MOD 91-0015 Control Rod Drive Group Power Supply Programmer Replacement MOD 91-0020 OSTG Tube Sleeving and Plugging MOD 91-0049 Revision to the Auxiliary Building Blowout Panel Release Pressure MOD 91-0052 Service Vater System Modification MOD 92-0004 Repair of Steam Generator 1-2 Continuous Vent Line Nozzle MOD 92-0027 Modification of Main Steam Piping Supports MOD 92-0032 Removal of DV61, DV161, and DW162 j MOD 92-0038 & Replacement of Letdown Cooler 1 and 2 FPR 92-0492-901 1 l \\

Docket Number 50-346 License Number NPF-3 .f Serial Number 2186 8 Attachment Page 3 NUMBER TITLE MOD 92-0044 Removal of Vent Path from Abandoned Drumming Station MOD 92-0047 Eliminate Local Control / Indication for MS106, MS106A, t MS107 and MS107A PCAO 92-0372 Use-As-Is Disposition for Station Batteries l 4 SE 91-0081 Modification of the Hydrogen Supply Piping to the Makeup Tank by the Addition of a Manual Isolation Valves i SE 93-0025 Cycle 9 Reload and Core Operating Limits Report l UCN 92-032 & Revision to USAR to Add Loads and Revise Indicating-UCN 92-033 Lights to the 125V DC Essential Panel Description i l UCN 92-052 Revision of Battery Load Tables i UCN 92-056 Delete Control and DC Pover Circuits From AC-Power i Circuits Paragraph in the USAR UCN 92-067 Alternate Amines and Steam Generator Feedvater Quality 1 UCN 92-079 Change to USAR Section 8.3.1.1.12 a UCN 92-084 Purification Demineralizer Filter Bypassed l UCN 92-089 Removal of Certain Electrical Drawings and Figures from j the USAR j UCN 92-096 Remove Vendor Specific Information from the Electrical l Conductor Descriptions in the USAR l UCN 92-108 Use of Certified Mill Test Report Yield Strength for 1 Structural Steel i 1 Elimination of Quality Assurance Director In-Line Review i UCN 93-001 of Measuring and Test Equipment Evaluations l l 1 UCN 93-013 Containment Hydrogen Generation and Control I UCN 93-028 Nuclear Operations Reorganization UCN 93-032 Nuclear Operations Reorganization UCN 93-033 operation with the Heater Drain Pumps Out-of-Service UCN 93-034 Boron Dilution Following a Cold Leg Pump Discharge LOCA i UCN 93-035 Correction of Quench Tank Cooler Performance Data A l UCN 93-041 Mitigation of Potential Critical Crack of an Auxiliary J Feedvater Line in the Annulus

Docket Number 50-346 License Number NPF-3 i _ Serial Number 2186 l i a Attachment [ Page 4 { NUMBER TITLE UCN 93-052 Revision to USAR Section 12 1 UCN 93-066 Changes to High Energy Line Break Analysis-for PCAQ 92-0195 l DB-PF-10119 Disabling Relaying During Full Load Testing of the l Station Blackout Diesel Generator l DB-SP-03372 & Deletion of Requirement for Replacing Orifices During DB-SP-03377 Makeup System Surveillance Procedures i f ) I j i 3 l e ,-e .~

'l t f SAFETY EVALUATION

SUMMARY

FOR DCR 88-0252 (SE 92-0002). TITLE: l I "As-Built" Revisions to Auxiliary Shutdown Panel Drawings. t CHANGE: Revise Auxiliary Shutdown Panel design drawings to depict internal wiring and external wiring on the same set of drawings. These changes are depicted on 3 USAR Figure 7.4-9. REASON FOR CHANGE: l Drawing Change Request (DCR) 88-0252 was originated in response to PCAQR 88-0281, to revise Auxiliary Shutdown Panel 4 Ign_ drawings to depict internal 8 (Vendor) wiring and external (field) wiring c. the same set of drawings. At the time that DCR 88-0252 was originated the internal (Vendor) wiring of the panel was depicted on two different Vendors drawings. l SAFETY EVALUATION

SUMMARY

l J} A walkdown w e performed to verify all of the wiring within the panel. No l modifications to the Auxiliary shutdown Panel are being performed for this activity, it is a paper change only to reflect the "As-Built" condition of.the panel in a different format-on design drawings. l 6 Several s'tfferences were noted between the drawings and the walkdown l information. These differences were reviewed to determine if any of the j "As-Built" conditions affected the function of the panel. The results of this review determined that none of the functions of the panel are adversely l l affected. l The drawing changes do not affect-the safety function of the Auxiliary Shutdown Panel. .j 1 1

n n SAFETY EVALUATION SU111ARY FOR DCR 90-0008 (SE 92-0006) TITLE: Revision to Emergency Diesel Generator Figure CHANGE: Revised USAR Figure 9.5-8A to depict the Emergency Diesel Generator starting air valves DA1147A, DA1147B, DA1148A, and DA1148B as normally closed. REASON FOR CHANGE: Drawing discrepancies associated with drawings M-017B, OS-041B, and ISID2-0017B were discovered. These drawing deficiencies were evaluated on a component / system basis to determine: a) uhat drawing changes, if any, would be required to ensure consistency with the design basis documents and the as-built drawings, and b) If the discrepancies would compromise plant safety or operation. SAFETY EVALUATION

SUMMARY

DA1147A. DA1147B, DA1148A, and DA1148B are Diesel Generator Air Receiver Discharge line Solenoid valves. These valves are energized to open when the Diesel Generator is being started. The valves will remain open until the engine reaches 200 rpm. Once the engine reaches 200 rpm, the valves will be deenergized to close. These solenoid valves are normally closed, and the change performed by this DCR is only to reflect as-built. conditions.

l i l 6 SAFETY EVALUATION

SUMMARY

FOR DCR 92-0011 (SE'92-0060) .j TITLE: i l TSC Datalink l l CHANGE: .l DCR 92-0011 revised Appendix C-3 of the FHAR to reflect current design and plant configuration. Appendix C-3 previously showed circuit ACYAU14A supplying cabinet C5457K and circuits ACYAU14E and ACYAU14C supplying the TSC Datalink. The actual plant configuration has C5754K powered by circuit ACYAUl3A and l circuits ACYAU14B and ACYAU14C supplying junction boxes JT5708 and JT5707, respectively. REASON FOR CHANGE: il' l Original reference to circuit ACYAU14A in Appendix C-3 is in error since~the l circuit does not exist. The TSC Datalink is powered by outlets in the Control l Room which are circuits not identified in Appendix C-3. { SAFETY EVALUATION

SUMMARY

-l l The circuit revisions contained in DCR 92-0011 will not affect the-safety function of the 120 V Instrumentation AC System because the affected circuits l are still protected by fuses which are properly sized to protect the cable, supt'ly the necessary current to the load, and are coordinated with the upstream l fuse. There is no effect on implementation or an increase in hazards because. DCR 92-0011 is a " drawing change only' document. These changes to Appendix C-3 -i of the FHAR will reflect as-built conditions'and will have no adverse effect on j safety. i l l i i l l l I P j

u s i .5 e SAFETY EVALUATION

SUMMARY

I FOR DCR 92-0019 (SE 92-0020) P TITLE: l Revision of the EDG Loading Tables f r CHANGE: 1 l Document Change Request (DCR) 92-0019 changed electrical drawings E-1042 l sheets 1 and 2 and E-1043 sheets 1 and 2. These drawings constitute the load i tabulation drawings for the Davis-Besse Emergency Diesel Generators (EDG). These drawings also constitute Table 8.3-1 in the Updated Safety Analyses Report (USAR). t l REASON FOR CHANGE: } f Changes to drawings E-1042 and E-1043 are to: f 1) Revise the continuous brake horsepower (bhp) ratings for motors in the HPI -j system, DHR system, MDFP, Containment Spray system. Containment Recirculation System and the Makeup System-2) Document on the EDG Loading Tables that Containment Lighting is not included.in the total lead because breakers BE1168, BE1167 BF1114 and BF1115 are required to be open during performance of DB-OP-06901 .j

• Containment Closeout Checklist" and following a containment entry per DB-HP-01001, " Containment Entry";

-j 3) Change the kW and kVA for the EDG Soak Back Pump motors from 1.1 and 1.5 I to 1.0 and 1.7, respectively; I 4) Editorially change notes 7 and 10 to clarify their intent; ) i' 5) Change note 11 by replacing " Decay Heat Pumps" with ' Containment Spray Pumps" since for a very small LOCA the Containment Spray Pumps will not be -f used if Safety Actuation Level 4 is not reached; i 6) Increase the Step 1 Cumulative Lead to include the manual loads which could be on before a Loss of Station Power (LOSP). These manual loads are ] identified on the EDG Loading Tables by having the Step 1 EW and KVA ratings shown ' boxed-in" l l-

+ - ~.. l l ~ l 3 4 SAFETY EVALUATION

SUMMARY

i There will be no adverse effects on safety as a result of the proposed EDG l Loading table revisions. No physical changes to the plant are performed by l I this Document Change Request. The proposed changes provide accurate EDG loading based on more accurate horsepower requirements and a more realistic accounting of loads which should appear in the Tables' Cumulative Load. The j revised load tabulations do not result in any cumulative loadings which exceed l the EDGs' continuous rating. i The High Pressure Injection, Decay Heat Removal and Containment Spray Pump motors all have 1.15 service factors. Although the bhp's recorded on the EDG l Loading Tables exceed the motors' rated horsepower, the bhp remains within the 8 permissible power loading for these motors. The bhp increase for the Containment Recirculation Fan and the Motor Driven Feedwater Pump motors remains below the rated horsepower. The likelihood of LOCA and/or LOSP occurring during a containment entry at l power is considered remote enough to warrant not including Containment. Lighting ,) in the cumulative Load for each EDG. Inclusion of these manually loaded i lighting transformers in the cumulative load would show that the EDG 1-2 l remains within the machine's continuous rating and EDG 1-1 does not exceed the j 2000 hour rating. f t t l ? l l I r e t I r i r 1 i t i i

4 i i SAFETY EVALUATION

SUMMARY

FsR l DCR 92-0029 (SE 92-0038) 1 TITLE: Revision of USAR Figure 10.4-4 l CHANGE: i Revised valve numbers for the cooling tower central flume isolation valves. t REASON FOR CHANGE: i DCR 92-0029 revised drawings M-12D, 05-16 and E-48B to correct the indicated line numbers for CT1589 and CT1590 the cooling tower central flume isolation valves. } i l SAFETY EVALUATION SUMtiARY: This change will correct a drawing error by depicting CT1589 and CT1590 in the correct circulating water lines in the cooling tower. The revised drawings will eliminate inconsistency between the control room controls and the design drawings and will improve reliability by providing accurate information on l j design drawings. This DCR will have no effect on safety. 4 J 4 4 .h 4 1 n i

i i

I

.i SAFETY EVAhUATION

SUMMARY

-j FOR

l DCR 92-0030 and UCN 92-062 (SE 92-0035) l i

i TITLE: Revision of USAR Figure 9.2-5 l l CHANGE: i Drawing Change Request (DCR) 92-0030 revised various drawings to delete equipment not installed in the plant. USAR Change Notice (UCN) 92-062 updated. l USAR Figure 9.2-5. i REASON FOR CHANGE: t Facility Change Request (FCR) 79-0125 requested installation of an Emergency Shower and Eyewash in the turbine building with a supply line from the Domestic Water System. Although this FCR was never implemented and the line was not installed, several drawings were incorrectly updated to show the equipment. SAFETY EVALUATION

SUMMARY

The domestic water system does not perform any functions important to safety. I i 4 This drawing change will not create or impact any existing hazards because no l hardware changes or design changes are involved. The reliability of the i domestic water system is not affected. Therefore, this DCR will have no effect on safety. I i i .i I I J i ii h i .t

SAFETY EVALUATION

SUMMARY

FOR DCR 92-0031 AND UCN 92-068 (SE 92-0047) L TITLE: Waste Evaporator Package CHANGE: Document Change Request (DCR) 92-0031 revises plant documentation to show the Waste Evaporator Package as abandoned in place. USAR Change Notice (UCN) 92-0068 revises Figures 7.3-9, 9.2-2, 9.2-4A, and 10.1-2A to show the correct valve positions to depict the Waste Evaporator Package as isolated. Additionally, UCN 92-0068 revises the description of the miscellaneous liquid radwaste system and its operating procedures in USAR Section 11 to reflect use of the Duratek skid in place of the Waste Evaporator. REASON FOR CHANGE: The Waste Evaporator Package has been abandoned in place, and the Duratek Skid-is currently used for processing miscellaneous liquid radwaste. SAFETY EVALUATION

SUMMARY

Isolation of the Waste Evaporator Package does not affect the function of the Miscellaneous Waste System because the Duratek system is used for radioactive waste separation. The use of the Duratek System has improved the reliability of the Miscellaneous Waste System. Revising the position of valves in the Nitrogen, Component Cooling Water, Demineralized Water, Primary Water and Auxiliary Steam Systems to normally closed, do not affect the operation or reliability of their respective system. The Miscellaneous Liquid Radwaste System design and operating changes have been reviewed and determined not to affect radiological consequences.of unexpected and uncontrolled releases. These changes do not create or impact any existing hazards because no hardware-changes or design changes are involved. The reliability of the Miscellaneous Liquid Radwaste System or any other system is not affected.

f SAFETY EVALUATION

SUMMARY

FOR DCR 92-0037 (SE 92-0043) i TITLE: l l De-rate Cyberex Inverters YV2 and YV3 CHANGE: DCR 92-0037 revised Cyberex inverter rating shown on plant configuration j documents from 10 kVA to 8 kVA. These changes are depicted on USAR j Figure 8.3-25. REASON FOR CHANGE-I l t ~ The purpose of DCR 92-0037 was to revise plant configuration documents based on the condition described by PCAQR 92-0189. The condition described by j J PCAQR 92-0189 is that factory acceptance tests for the Cyberex inverters did not meet the requirements of specification 7749-E-20Q. Inverter output voltage l was not maintained within limit s for the entire range of input DC with the Inverter operating under full load. SAFETY EVALUATION SUMFIARY: The present load on Cyberex inverters YV2 and YV3 is within the de-rated value of 8 kVA or 66 Amps @ 120 Vac. During normal operation YV2 is observed to carry approximately 55 Amps and YV3 less than 15 Amps. These loads do not increase during accident conditions. The effect of de-rating the Cyberex inverters is a reduction in load growth margin for YV2 and YV3. Based on the present loading of YV2 and YV3, factory l tests indicate that the inverters will perform in accordance with specifications. Reliability is considered to be unaffected by de-rating the inverters. All other inverter output ratings are not changed by the proposed action. De-rating the Cyberex inverters will ensure that additional load will not be added to the point where supplied equipment is affected. Therefore all equipment supplied by the Cyberex inverters is not affected by the proposed action. The proposed action will not effect the 120 Vac Instrument AC System or any loads supplied by the system and is considered safe.

l f ~I SAFETY EVALUATION

SUMMARY

l l FOR DCR 92-0050 (SE 92-0063) i ' TITLE: Revision of DC MCC Single-Line Diagrams l l i CHANGE: } I Document Change Request 92-0050 revised drawing E-6 sheets 3 and 4, which are ) depicted in the USAR as Figures 8.3-46 and 8.3-47. E-6 sheets 3 and 4 are the single line diagrams for essential DC Motor Control Centers 1 and 2. Item 9 on i these drawings describe the four current meter relays operating between the I station batteries and the DC bus. In addition, this DCR corrects the quality class of the non-essential. portion of DC MCC1, as required by PCAQ 92-0210. ? REASON FOR CHANGE: The directional current setting of item 9 on drawing E-6 is inconsistent with the actual field setting and with the Relay Setting Manual. None of the loads connected to the affected portion of DC MCC1 are. safety related, therefore this portion was overclassified as being essential. SAFETY EVALUATION SUlfiARY: There will be no effect on safety as a result of the proposed changes. No physical changes to the plant are performed by this DCR. One change corrects drawing E-6 sheets 3 and 4 to show the actual setting for four directional current relays. During normal operation the required DC current comes from either the battery chargers or the regulated rectifiers, not the batteries. At a preset level of current flow from the batteries, there is an alarm in the control room. This DCR changes the depicted preset level or setpoint from 100A' to 25A, but does not affect the actual setpoint, which presently is 25A. The setpoint of 25A provides earlier indication of a potential problem with the DC ] power supplies. 1 I The other proposed change corrects the quality class of circuits downstream of the 500A fuses. D118 is a downstream breaker which is not coordinated with the upstream 500A fuses. D118 feeds the turbine generator emergency bearing oil 1 pump, which is not a safety related load. IEEE Standard 384 permits connection of non-safety related loads to class 1E power sources, but requires that " shorts in the non-1E side shall not degrade the IE side." a i 4 ~

l i

~! Review of the other circuits connected ta the 500A fuses (Dill through D117) shows that none of these circuits are nr. clear safety related Therefore, there .( is no requirement to design and maintiin this portion of DC MCC1 to class IE 1 standards. Accordingly, this section is being reclassified. l Y s i r i s l a 1 ~ ~ ~ ,, ~ .j

.. = ) 1 SAFETY EVALUATION

SUMMARY

FOR DCR 92-0055 AND UCN 92-078 (SE 92-0049) l TITLE:

1 1

Deletion of Circuit Details from FHAR Appendix C-3, " Circuit Coordination i Summary" i CHANGE: This UCN deletes circuit details from FHAR Appendix C-3, " Circuit Coordination' l Summary", updates the associated title and notes, and corrects typographical j errors. ] l REASON FOR CHANGE: j FHAR Appendix C-3 contains the following information: Power Supply - Circuit i Number - Component Number and Description - Feed and Load Rating, Type, and i Reference Drawing - Coordination Status - Notes, j Since coordination calculations exist for all of these circuits, coordination can be documented by summarizing the results, rather than by repeating portions I of the coordir2 tion analysis. Therefore, it is decided to delete the rating, i type, and drawing for the feed and load portions of Appendix C-3. I SAFETY EVALUATION

SUMMARY

i There will be no effect on safety, as there will be no physical changes as a. result of this change, and the documentation which is'being deleted merely t repeats information which is retained in electrical calculations and design' drawing E-2014. Basic design practices include consideration of acceptable voltage drops, ampacity, and coordination for all circuits. When circuits important to safety are involved, this " consideration" is documented in the form of calculations. Several coordination calculations exist, which cumulatively address all of the circuits listed in Appendix C-3, Drawing _E-2014 and the existing related electrical calculations combined with t the design practices discussed above ensure that proper coordination analyses-will always be available, even if any of these circuits are modified in the future. -j l 1 1 I e l l

t i -i SAFETY EVALUATION S"MMARY FOR DCR 92-0061 (SE 93-0011) s TITLE: i .j Miscellaneous Vaste Drain Tank Nitrogen Supply Pressure Control Valve Abandone In Place i CHANGE: ? Removed the Miscellaneous Waste Drain Rank (HWDT) nitrogen supply pressure 'j control valve (PCV 1569) from service (abandon in place). i t REASON FOR CHANGE: Ongoing problems have been experienced with this valve including sporadic contrcl and " leaking-by" O SAFETY EVALUATION

SUMMARY

i With PCV1569 isolated via isolation valves NN38 and NN189, the only potential- { effect that this change would have on the nitrogen system would be the failure of these otherwise open isolation valves which would have no effect on the j ability of the Containment Isolation valve to perform its function. The only radwaste related USAR analyzed accident is the vaste gas decay tank rupture. This change will not in anyway affect.the vaste gas decay tank. There are no USAR evaluated malfunctions of equipment important to safety-l pertaining to the equipment affected by this change. Therefore this change can not be considered as a possible initiator of.a USAR evaluated accident or to increase the probability of occurrence of a USAR evaluated malfunction of equipment important to safety. j -r l l i I ~f i r l l l \\~ l I l 1 .1; l

4 l l SAFETY EVALUATION

SUMMARY

j FOR [ DCR 92-0067 (SE 92-0075) TITLE: f Revision of the EDG Loading Tables r CHANGE: Revision of the EDG Load Tables for the Containment Air Coolers. l 1 REASON FOR CHANGE: f The brake horsepower for the Containment Air Cooler (CAC) fan motor high speed j was changed from 126 bhp to 131 bhp. This change is to correct a discrepancy. [ between a calculation and the drawings. l l SAFETY EVALUATION SUMt4ARY: j There will be no adverse effects on safety as a result of the propoeed EDG l Load Table revisions. No physical changes to the plant are performed by this Document Change Request. The proposed changes provide accurate EDG loading based on current readings. L As a result of this revision, the operators have a more accurate tool to use when pertorming EDG load shedding and manual load' additions The revised load i 4 tabulations do not result in cumulative loadings which exceed the EDG's l continuous rating. [ The actual c"Trent readings for the CAC motors are within the nameplate rating-of the motors. i i l i . ~. -v --- i v.

i. 1 j SAFETY EVALUATION

SUMMARY

FOR j DCR 92-0069 (SE 93-0013) TITLE-I Miscellaneous Drawing Changes CHANGE: This DCR consolidated document discrepancy corrections and editoria1' changes to drawings incorporated in the USAR. j REASON FOR CHANGE: 1. Changed BE1130 and BE1198 (Spares) to show future units. These breakers j were not being used, and were removed. This change affects USAR Figure 8.3-23. l 2. Corrected drawings which depicted the old Emergency Communications Center or " red barn" as "ECC". When the new ECC was built, existing plant-drawings were not revised to delete "ECC" from drawings showing the red barn. This change affects USAR FJgure 8.3-17. 3. At E4 Sheet 4, increased the indicated trip rating of BDFS and BCES from 1200A to 1300A to coincide with the field, Electrical Load Management System (ELMS)'and Relay Setting Manual (RSM). This change has no effect on any existing design inputs. This change affects USAR Figure 8.3-17. SAFETY EVALUATION

SUMMARY

l Removal of unused (spare) breakers BE1198 and BE1130 has no effect on the-electrical design or operation of the plant. Existing seismic analyses of MCCs assume maximum utilization of available MCC space. Review shows that maximum utilization is the most limiting for MCC seismic analyses. Therefore, removal of BE1198 and Bell 30 is bounded by the existing seismic analysis, and has no-effect ou safety. 4 The miscellaneous drawing changes associated with this DCR have no effect on the design or operation of the plant, and none of the affected components have any function which is important to safety. In addition, the RSM, ELMS, and related calculations have been reviewed to ensure other design documents are correct. Therefore, the miscellaneous drawing changes associated with the DCR will have no effect on safety. l Therefore, there will be no effect on safety. l I I

~. ~ t SAFETY EVALUATION

SUMMARY

FOR FCR 81-0167-05 and UCN 92-066 (SE 92-0037) f TITLE: Condensate Demineralizer Vaste Transfer Pump j 't CHANGE: I UCN 92-066 updates the text and USAR Figure 3.6-20 of the USAR to identify that i the installed Condensate Demineralizer Waste Transfer Pump is not a positive j displacement pump. i i REASON FOR CHANGE: Condensate Polishing Demineralizer (CPD) Waste Transfer Pump P135-1 was removed f under temporary modification TM 81-1222 to allow use of a portable positive i displacement pump for transferring resin slurry. FCR 81-0167 was generated to 1 install a permanent positive displacement pump and filter. This FCR has been re-evaluated and determined to be unnecessary. Field work accomplished under this FCR was installation of flanged connections to permit use of a portable positive displacement pump. Therefore. FCR 81-0167 Supplement 05 updates ,j documents to:

1) reflect field work accomplished under this FCR, and

2) reflect permanent removal of P135-1.

i SAFETY EVALUATION SGIMARY: Permanent removal of P135-1 will not adversely affect the function of'the CFD Waste Transfer Pumps because the redundant pump P135-2 is available to decant water. Installation of flanged connections and use of a portable positive i i. displacement is an enhancement because the original certrifugal transfer pumps are not suitable for transfer of resin slurry. Use of the portable positive l displacement transfer pump is controlled through procedure. Since decant and resin transfer operations are only performed periodically, redundant pumps are j not necessary for reliable system operation. The affected pump suction and j discharge piping is low energy piping and., therefore. does not create a pipe break hazard. The system volume is low and does not create a flooding concern. l Reg. Guide 1.143 excludes the condensate cleanup system from the special design j requirements for radioactive waste systems. I 4 l ~i 3 I b 4 9 i I

-l L SAFETY EVALUATION

SUMMARY

FOR FCR 86-0404 (SE 92-0028) l t TITLE: Replacement of Water Treatment Building Sump Pumps with Fortable Pumps i CllANGE: USAR Figure 9.3-4 was revised to delete reference to installation of a sump pump in the Water Treatment Building. REASON FOR CHANGE: The purpose of this supplement is to allow FCR 86-0404 to be closed 'as-is'. l The Water Treatment Building Sump Pumps, MP171A and B, were removed from j service and replaced with portable sump pumps. FCR 86-0404 Supplement 00 specified that the pumps would be replaced via a future supplement. It has been determined, that the replacen.ent of the sump pumps is unnecersary and that the portable sump pumps are adequate. SAFETY EVALUATION

SUMMARY

j This modification will have no effect on safety. Since the Vater Treatment. j Building contains no nuclear safety related equipment, flood protection is not j necessary, however, a portable pump will provide a means to remove liquid'from i the sump. l l l I i -l 1

.. ~ SAFETY EVALUATION

SUMMARY

FOR j FCR 86-0425-09 (SE 90-0147) TITLE: Remove Installed Instrumentation Associated with the Motor Operators to Valves l FW6397 and FW6398 l 1 CifANGE : Supplement 09 of FCR 86-0425 deleted the addition of motor operators to valves l FW6397 and FW6398 Motor Driven Feedwater Pump (MDFP) Isolation Valves. I REASON FOR CHANGEt The addition of motor operators was identified in supplement 0 of FCR 86-0425 l as being added at a later time. To support the. original intention to add motor ~l operators to FW6397 and FW6398, instrument rack R2408 was modified to include l the appropriate switches (disconnect, open/close etc.) in the Motor Driven Feed Pump (MDFP) mimic. FCR 86-0425, supplement 09 will modify the mimic to .l remove the instrumentation associated with the motor operators and spare the l power cables for the motor operators in place. l SAFETY EVALUATION

SUMMARY

Removing the requirement of motor operators on Isolation Valves FW6397 and l FW6398 does not impact safety or the function of the'MDFP System. The valves will remain operable by manual action rather than remote action. The Target, Rock. solenoid operated valves (FW6459 and FW6460) provide the system with remote control capabilities intended by the addition of motor operators on FW6397 and FW6398. FW6397 and FW6398 are also locked open valves. This ensures the valves can not be mispositioned tc inhibit the MDFP f rom performing it's intended function. [ 'I Removing the instrument controls from rack R2408 associated with motor operators will eliminate possible confusion when operating the MDFP from instrument rack R2408. i ) ? J k 6 1 .,,..I

SAFETY EVALUATION

SUMMARY

FOR .i FPR 80-0189-004 AND UCN 92-0070 (SE 92-0040) f TITLE: 1 Revision to USAR Figure 9.1-5 j CHANGE: UCN 92-0070 will revise USAR Figure 9.1-5 to show the correct pipe elevation. REASON FOR CHANGE: l FCR 80-0189 recently lowered the top of the over-flow pipe in the Spent Fuel- '{ Pool (SFFI from 602'-O' to 601'-9". This change was initiated to prevent:high l water level from flooding of the transfer system winch motor. i SAFETY EVALUATION

SUMMARY

i The normal level of the SFP water is 60l*-6". The setpoints of the high and low level alarms are 601'-7' and 601'-2" respectively. Per USAR l a Section 9.1,3.9.1 the low level alarm assures a minimum of 23' of water is maintained above the fuel assemblies, and the high level alarm is provided to j prevent overfill. Therefore lowering the overflow pipe level to 601'-9' will J l maintain the overflow level above the high level alarm setpoint and will not j cause level to be less than the minimum of 23' required by'the Technical j Specifications. USAR Section 15.4.7 discusses the effect of a fuel handling accident. This analysis assumes that 99% of the iodine' released from the fuel assembly is j essumed to remain in the water. According to the Bases for. Technical specifications 3/4.9.10 and 3/4.9.11, the minimum water level is required to ensure sufficient water depth to remove 99% of the assumed iodine gap activity ] released from the rupture of an irradiated fuel' assembly. Since the i radiological consequences evaluation associated with the fuel handling accident assumed water depth was at the minimum required, the lowered level of the j overflow pipe will not affect the radiological consequences for this accident. a i n i ~, ,..m.. _..-,.,.,.I

l l l SAFETY EVALUATION

SUMMARY

l FOR FPR 90-1881-901 (SE 92-0065, R01) j TITLE: Replacement of COM-5 Relay in AC109 ) CHANGE: FPR 90-1881-901 replaced a Westinghouse CDM-5 relay (P9-11480) in Unit 9 of the ~! ~4.16 kV essential switchgear C1. The associated switchgear, AC109, provides l power to the Service Water Pump 1-3. l l REASON FOR CHANGE I I I i l Class IE Westinghouse COM-5 relays are no longer available, therefore an Asea j Brown Boveri (ABB) COM-5 relay is being used as a replacement. The ABB relay is functionally equivalent, but has a solid state high dropout. instantaneous trip (SSC-T) unit vice the electromechanical (ITH) unit in the existing l Westinghouse relay. 'l 'I SAFETY EVALUATION

SUMMARY

l This replacement has no effect on safety. The Material Engineering Evaluation f provided with this FPR shows that the ABB COM-5 relay is a functionally ) equivalent and seismically acceptable replacement for the Westinghouse COM-5 relay. Only the SSC-T unit of the ABB COM-5 current sensing elements is different from the corresponding units in the installed Westinghouse COM-5 relay. Inadvertent actuation of that unit can not trip AC109 prematurely, and j failure to actuate will not preclude a trip, since the device contains an alternate instantaneous trip device. The SSC-T unit interacts with the rest of the device through a current transformer and a set of contacts,.in essentially i the same way the Westinghouse equivalent (ITH) unit interacts. Also, the 1 l alternate instantaneous trip device curve does not cross (and therefore j coordinates with) the curve for the nearest upstream overcurrent protection l i device. -j The purpose of the COM-5 relay is to limit a locked rotor and/or fault damage j to downstream devices, and to establish a protective zone which prevents a j downstream locked rotor and/or fault from disabling otherwise unaffected l equipment. Analysis based on the COM-5 time characteristic curve,-shows that these goals are accomplished without taking credit for the SSC-T unit. l The SSC-T device is an operational convenience. It has no important influence on any safety function. Therefore, it is immaterial whether its design is solid state or electromechanical. The remaining portions of the ABB COM-5 relay are the same as is currently installed. Accordingly, this replacement has no adverse affect on safety.

i s L l o SAFETY EVALUATION

SUMMARY

i FOR FPR 91-1535-901 (SE 93-0008) l TITLE: l Change to Reactor Coolant Pump Motor Flywheel l CHANGE: l Change in Reactor Coolant Pump Motor Flywheel fabrication, material and testing. l l J REASON FOR CHANGE: i Reactor Coolant Pump Motor (RCPM) 2-1 was being replaced as part of the RCP preventive maintenance program. L SAFETY EVALUATION

SUMMARY

The new RCPM is completely interchangeable with the existing RCPM, therefore

i there is no impact on the Reactor Coolant Pump (RCP) or on RCS flow.

.l I The changes to the flywheel do not change the ability of the motor to start l l under design conditions or the running speed of.the RCPM. Additionally the changes do not affect the motor rotating element mass moment of inertia, and therefore the motor coastdown for loss of offsite power-is unaffected. Therefore the attributes of the RCPM provided by the flywheel assumed in the i bases of the Technical Specifications are unchanged. l None of the functions important to safety are affected by the changes in l flywheel fabrication, material and testing. Engineering evaluation j demonstrates the changes do not increase the probability of a missile j associated with flywheel failure, therefore no new hazards are introduced. The existing hazards in containment will not have an increased effect an the ) p new flywheel material because the chemical-and physical properties of the new l flywheel material are essentially the same as the existing flywheel material properties. The reliability of the RCPM is unaffected by the' changes in flywheel material, l fabrication and testing. The new flywheel material has greater isotropy than the rolled plate material. The increase in isotropy is verified by the l increased material testing. This increase in uniformity of properties is desirable for rotating elements. 1' l l l 1 1 1 I j 1 l l l l

l l t SAFETY EVALUATION

SUMMARY

j FOR { FPR 92-0495-903 (SE 92-0054, R01) t t TITLE: j Replacement of Watt and Voltage Transducers in DJesel Generator Electrical System CHANGE: r Replacement of Watt and Voltage Transducers, ET 6262, ET 6221, JT 6221, l l ET 6263 ET 6231, JT 6231 in the Diesel Generator Electrical System. l REASON FOR CHANGE: These transducers are being replaced since the current transducers are obsolete. j SAFETY EVALUATION

SUMMARY

l The transducers are proposed to be replaced by new, seismically installed j safety related transducers which will perform the identical function. The j l transducers have input to ground, and input to output isolation strength of .j 2500 VAC. i The replacement of the obsolete transducers with new transducers will not have any adverse effect on the safety of the system. ) I 1 i l I i ) ) l l l l l I ~

SAFETY EVALUATION

SUMMARY

FOR MOD 87-1273 (SE 88-0077, R02) TITLE: Replacement of Steam Admission Valves to the Auxiliary Feed Pump Turbine CHANGE: This modification replaced the existing Masoneilan Steam Admission Valves, MS5889A and MS5889B, to the Auxiliary Feed Pump Turbine (AFFTs) with Valtek-valves. REASON FOR CHANGE: The existing Masoneilan valves have exhibited excessive leakage. SAFETY EVALUATION SU.9Jt RY: This change replaces the Masonellan valves with comparable Valtek valves and - does not affect the safety function of the Auxiliary Feedwater or Main. Steam Systems. The Masonellan and Valtek valves were purchased to the same requirements and both valves were specified for this specific application. The-Masonellan valves are control valves with a balanced trim design. The two seated design of the old Masonellam valves is not as leak tight as a' single seat design of the new Valtek valves. The Valtek valves are globe valves with a single seat design for on/off service. The Valtek valves are 'Q' and are seismically qualified. Valve operation will not change. The electrical modifications to the conduits supplying the steam admission valves in the AFW pump rooms will not impact plant safety because the conduits will still be supported seismically and the wiring scheme will remain in its present configuration. Pursuant to the above, it is concluded that replacement of the subject valves is safe.

l SAFETY EVALUATION

SUMMARY

FOR MOD 88-0027 (SE 90-0145) I TITLE: i Upgrade Megawatt Electric Input to Integrated Control System j r CHANGE: i The purpose of this MOD was to revise the circuitry which provides the Megawatt Electric (MWE) signal to the Integrated Control System (ICS). This revision l utilized redundant sensors and incorporate automatic selection of valid input signals to ICS. .l t REASON FOR CHANGE: The reason for this change is to preclude the initiation of a plant transient as a result of a failure of the Megawatt Electric input signal-to ICS. This change was also recommended by Babcock and Wilcox Owners Group. { l SAFETY EVALUATION

SUMMARY

The circuitry changes within the scope of this Modification are outside of. " Safety System" functional boundaries. However, a failure of the MWE signal [ may still challenge plant safety systems as a result of ICS perturbations. 5 Evaluation of the value of this Modification requires the existing system configuration to be compared with the proposed tonfiguration. Results of the comparison show that this Modification will result in a net j increase in reliability of the input signal to ICS. Failure of the selection i I and validation circuitry will result in a default to the "X" string which is equivalent to the existing conf 3guration. Failures in the system, from the l input of the redundant sensors to the output of the respective mV/V conversion modules in UNI, are protected by the selection.and validation circuitry. l i -) l 1 l l l l

.m I i i SAFETY EVALUATION

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l FOR MOD 88-0034 (SE 91-0007) ' TITLE: Installation of 3-Way Ball Valves at Inlet of Each Instrument Air Dryer [ 'h i CHANGE: i This modification installed 3-way ball valves at the inlet of each Instrument. Air (IA) dryer Each receives an air signal directly from the IA header. As pressure drops below a given setpoint the 3-way valve will bypass the dryers j diverting full flow to the system. A check valve and a new header tie-in will j prevent backflow into the failed dryer from the IA header. Once the valves have diverted to bypass the dryers they will remain in the bypass mode until manually reset. t REASON FOR CHANGE: Stuck open valves in the IA dryers can provide significant air leakage to depressurize the system to a low setpoint and trip the plant on loss of IA. , l A loss of IA header pressure generated a plant trip on December.7, 1987. The root cause for the air loss was a failed open purge valve which diverted a j large amount of dryer inlet air out of the system. A identical event on l May 23, 1985 caused a secondary plant upset however, because of its duration i did not generate a plant trip. iq Besides enhancing air dryer reliability and reducing pressure drops in the l system this modification will also eliminate other concerns related to air dryers and associated auxiliary components by; increasing bypass capacity of valve IA2041, installing a check valve upstream of T122 air receiver, and reconnecting the bypass valve IA2041 upstream of IA14 but downstream of after filters. i SAFETY EVALUATION

SUMMARY

i The proposed changes will not adversely affect the safety function of any structure, system, or component for the following reasons: 1. None of the engineered safety features depend on the supply'of instrument air for its operation. i 2. The changes proposed do not affect the station or instrument air containment isolation valves. 'l 3. The changes proposed will increase the reliability of the station and instrument air system by ensuring adequate bypass flow is available to maintain system pressure above 75 psig during a failure of the operating air dryer. It will not create any new failure modes. Spurious opening of the new bypass valve is equivalent to a failure of the dryers to remove l the moisture from the air. Pressure indication of centrol air to the valve operator will be available for monitoring. l l I i y

j ? j SAFETY EVALUATION

SUMMARY

FOR MOD 88-0055 (SE 92-0011) l TITLE: j Installation of Wafer Check Velves in the Component Cooling Vater and Clean Waste Monitor Tank Rooms Floor Drains f CHANGE: Modification 88-0055 installed wafer check valves in two floor drains in the i Component Cooling Water (CCW) Room and one floor drain in each of the Clean Waste Monitor Tank. rooms. Modification 88-0055 also capped off one equipment l drain in the CCW Room. This cap will be drilled to allow existing equipment l drain lines to continue discharging into the drain. l [ 3 REASON FOR CHANGE: The purpose of this modification is to provide _a means of preventing backflow of contaminated fluid from the floor drains and equipment drains being modified. -i SAFETY EVALUATION

SUMMARY

The proposed action will have no effect on the ability or reliability of any i system, structure or component to perform its safety function because the wafer check valves being installed are essentially identical to those in use in the l plant Negative Pressure Area. These check valves are passive devices that open l upon accumulation of approximately one-third of an inch of water. Blockage of a floor drain in the unlikely event of catastrophic failure of a wafer check j valve would be no more severe than blockage-of the floor drain due to accumulation of debris. Also, the capping of the equipment drain will have no-I effect because the equipment drain lines which use this drain will pass through i the cap and the equipment will drain in the same manner as before capping. J i I i l 1 i I

l [ j SAFETY EVALUATION

SUMMARY

'l FOR l MOD 88-0113 (SE 88-0628) ) TITLE: R Modification of Output Logic of Division One Generator Protection System 'f CHANGE: ) i MOD 88-0113 reduces the number of unnecessary challenges to the Station's l Class 1E standby power supplies. These challenges are reduced by modifying the 1 output logic of the division one Generator Protection System. Modifying the division one output logic will result in one Station tripping scheme when the Generator Protection System (GPS) operates, j REASON FOR CHANGE: 1 In the event of a significant system disturbance or the malfunction of a

j generator relay the full load rejection capability of the unit would be

-l exercised. Failure of this capability would result in a challenge to the l Emergency Diesel Generator. -l l SAFETY EVALUATION

SUMMARY

j i The changes required to implement MOD 88-0113 will not adversely affect the j safety function of any Station equipment. Providing Station electrical i auxiliaries the opportunity to always transfer to the power system grid when j the GPS operates, reduces the number of challenges to the Station's Emergency j Diesel Generators. Should the power system grid be incapable of supporting i Station electrical auxiliaries, relays exist which will detect that grid I condition and cause appropriate changes in the Station electrical line up. The critical electrical parameters which determine the grids capability of supplying Station electrical auxiliaries are frequency and voltage. Under-f regnency relays are provided by the 13.8KV Bus A and Bus B. Undervoltage relays are provided at the 13.8KV buses and at the essential 4.16KV buses C1 and D1. Thus, if the power system grid is not capable of supplying the Station-Electrical Auxiliaries the existing under-frequency and undervoltage relays i will initiate signals that result in start signals to the Emergency Diesel Generators. 4 1 l

SAFETY EVALUATION

SUMMARY

FOR MOD 88-0207 (SE 92-0031, R01) TITLE: Increase of the Auxiliary Boiler Feed Pump Recirculation Flow CHANGE: This modification increased the Auxiliary-Boiler Feed Pump recirculation flow from 35 gpm to approximately 200 gpm and required the line size be changed from one to three inches in diameter. The valves in the affected portion were also changed and the orifice size was increased to 15/16 inches from 3/8 inches. Only one orifice was installed instead of the.two previously used. The orifice was placed in the common line to the Deareator instead of each pump having its own orifice. REASON FOR CHANGE: The problem to be corrected by this modification was the excessive failures occurring to the pump bearings and seals. According to the Vendor (Ingersol Rand letter) the recirculation flow is insufficient at lower boiler power levels, and it causes vibration and mechanical damage to the pump seals and bearing-. SAFETY EVALUATION

SUMMARY

The change performed by MOD 88-0207 to increase the recirculation flow of the Auxiliary Boiler Feed Pump has no effect on any safety function. This Safety Evaluation is required due to a change in the USAR Figure 10.1-2 to show one l orifice instead of two in the recirculation line. Although the t otal recirculation flow is increased along with Auxiliary Boiler Feed Pump flow the capacity of the pump will not be exceeded and no effect on the Auxiliary Steam System capacity will be produced. I

l 4 SAFETY EVALUATION

SUMMARY

j FOR MOD 88-0227 (SE 91-0008) TITLE: Installation of Synchronism Check Relays for Breakers AC101 and AD101 I-CHANGE: Plant Modification 88-0227 adds synchronism (sync) check relays to prevent manual closing of the emergency diesel generator output breakers (AC101 and { AD101) if the voltages on either side of these breakers are not in a predetermined frequency, phase, and magnitude. relationship, { ~! REASON FOR CHANGE: l i This modification implements Babcock and Wilcox Owners Group recommendation ] TR-144-PES. This recommendation directs members to evaluate hardware which j will reduce the likelihood of extensive diesel generator damage as a result of 1 paralleling out of sync. SAFETY EVALUATION

SUMMARY

Adding the sync-check relays will not adversely affect the safety function of j l the EDGs or the output breakers. The synch-civck relay permissive contacts are not being connected in series with contacts which automatically close.the EDG output breakers on loss of essential bus voltage. Thus, the state of the synch-check relay permissive contacts cannot affect auto closing of these output breakers. The new sync-check relay is provided with features which, if selected, will defeat the normal sync check function. Selection of the HLDB (Hot Line--Dead-Bus) feature permits loading the EDG with a bus at less than or equal to 50 percent voltage. Thus, manual loading of a " Dead Bus", as backup to auto closing, is permitted when HLDB is selected. Installation of these sync-check relays in.the electrical control and relay boards will not degrade the Seismic I category classification of these. cabinets. Relay " contact bounce" will not result in undesirable circuit operation as sync-check relay contacts are permissives, normally isolated from the breaker's closing coil. l i -l i i =, c..

t { SAFETY EVALUATION

SUMMARY

j FOR ~ MOD 89-0003 (SE 91-0055. Rol) { TITLE: l Replacement of Relief Valve DH1508 { CHANGE: Relief Valve DH1508 was replaced. The previous relief valve had a socket welded end connection. The valve was connected to the inlet and discharge pipes by unions which were used for removing the valve for maintenance. The l replacement valve has a flanged connection, therefore, the piping was modified for the new flanged end connections. The flanges allow for the removal of the i I new valve for maintenance similar to unions. i REASON FOR CHANGE: f i Relief Valve DH1508 is a Longergan valve which is no longer manufactured. MOD 89-0003 replaced the existing valve with a Anderson Greenwood relief valve. SAFETY EVALUATION

SUMMARY

i l I The replacement of relief valve DH1508 will not adversely affect the safety functions of the DHR System. The replacement relief' valve is ASEE Section III Class 2 and has the same design rating and setpoint as the existing valve. The modified piping and new relief valve have been seismically analyzed and are i acceptable. l l The replacement relief valve has a higher capacity than the existing relief I valve. The higher capacity valve will reduce the possibility of overpressurization from valve leak-by because more leakage into the low l pressure piping can be relieved by the new valve. The' increased capacity of the relief valve has no impact on the Reactor Coolant Drain Tank and Containment Vent Header System because the increased flow rate is within the capability of the collection system, t 1 l l i 1 >,,.r y y 7 yy.. g. -m__ 9 ,,i_ ytr-r y a w

I l L ) i SAFETY EVALUATIOW SUl@iARY FOR L MOD 89-0094 (SE 90-0148, ROI) i i TITLE: I i Modification to Emergency Diesel Generator Room Sprinklers CHANGE: The design for MOD 89-0094 provided for the following changes to the air supply. l to the Emergency Diesel Generator rooms sprinkler _ system pre-action valve .j l numbers FP114A and FP115At i E i l-1. Replaced pressure indicator numbers PI 8823C and P18824C with gauges l having a range.of 0 to 36 ounces. The previous pressure indicators l have a range of 0 to 250 psi, j l l 2. Relocated pressure indicator numbers P1 8823C and PI 8824C.and their 1 respective isolation valves, FP114J and FP115J. These pressure j indicators were located downstream of check valves FP114U and FP115U, but should have been located upstream of these check valves in order j for the system to operate properly. I 3. Added flow restrictions (R08823 and R08824) upstream of pressure l switch numbers PSL 8823 and PSL 8824. l 4. Added " drip pots" to remove moisture build-up from the air supply l lines. l I t REASON FOR CHANGE: l i These changes were made to bring the Number 1 and 2 Emergency Diesel Generator rooms pre-action sprinkler system supervisory air supplies into compliance with. l the pre-action valve manufacturer's recommended installation configuration. 'l l I SAFETY EVALUATION SUt2iARY: Replacing and relocating the pressure indicators and adding a restriction i orifice to the EDG rooms pre-action valve supervisory tir supply lines will not adversely affect plant safety. This change does not affect any safety related j system or component. These changes will increase the reliability of these suppression systems by allowing the supervision system to work correctly. -l These changes will also ensure that the systems can be properly tested. i The " drip pots" are being added to remove moisture build-up from the air supply .i l lines. Moisture build-up was identified as a possible cause of the system { L pressure regulators failing to properly control the supervisory air system l pressure. Failure to properly control supervisory air pressure.could lead to 'l an increased time period required to detect damage to the sprinkler system i piping and/or sprinkler head (s). i i l i 6 3 e-m.. . - -..., - - ~ r m,-

l SAFETY EVALUATION'

SUMMARY

-l FOR MOD 89 0109, Supp 01 and 03 (SE 91-0066) { TITLE: Alternate AC Power Source for Station Blackout i CHANGE: l l Supplement 01 installed raceways for routing power and control cables to the. High Voltage Switchgear Room B and the Control Room, respectively The remote l Station Blackout Diesel Generator (SBODG), control panel was also installed by. I Supplement 01. j i Supplement 03 installed the SBODG, it's support equipment (auxiliaries) and makes connections between the SBODG and the station's nonessential 4.16kV- / distribution system. i REASON FOR CHANGE: 3 Plant Modification 89-0109 added an alternate AC source, SBODG, to comply with the requirements of 10CFR50.63. The SBODG is to be utilized in the event of a complete loss of alternating current (AC) electric power. SAFETY EVALUATION

SUMMARY

Installation of the SBODG, SBODG auxiliaries, Unit AD213, SBODG building services, the remote SBODG control panel in the Control Room and the raceway for SBODG power and control cables will not adversely affect the safety function (s) of any SSCs. The remote SBODG control panel and sections of raceway routed through the Auxiliary Building will be seismically supported to ensure that a Seismic II over Seismic I condition is not created. All raceway penetrations through Auxiliary Building walls and floors will be evaluated to ensure these structures are not unacceptably weakened. Barriers which must be opened to allow passage of raceway will be provided with an appropriate penetration seal to ensure that the barriers safety function is not adversely affected. A core drill from the exterior of the Auxiliary Building into the Control Room will also be provided with a missile shield so as not to invalidate the stations missile protection discussed in USAR Section 3.5. l The new 4.16kV switchgear cubical, AD213, is being seismically mounted in j Room 323 so as not to create a hazard for existing essential distribution ] equipment. Control power for AD213 is provided by the station's train 2 DC. I distribution system via the DC bus within the D2 switchgear. Breaker AD213 will be maintained in the closed position to ensure that a' path from the alternate AC source to the station's distribution system is available. a

_--_=y 1 i The safety function of the station's Drain and Discharge System vill not be adversely affected by the drain and discharge system provided for the SBODG 4 building since the safety related discharge path and the SBODG building discharge path do not connect. l The SBODG building is provided with fire detection and fire suppression. 1 d. i The SBODG is capable of supplying either of the station's essential 4 16kV i buses through nonessential Bus D2 and is available within ten minutes of the onset of station blackout. During an emergency, the SBODG can be manually l l started and loaded onto Bus D2 from the Control Room - auto starting and l loading is not provided. Operation of circuit breakers to make an SBODG line up to either essential bus can also be accomplished from the Control Room. The l SBODG is capable of unattended operation at rated full-load for at least four j hours. 'l l 3 i i F t e l I i r i l i i 6 .l l 2 9 e I a i d

SAFETY EVALUATION SUtDiARY FOR MOD 90-0030 (SE 92-0039) TITLE: Replacement of Synchro Check Relays 25C1, 25D1, in the 4160V Electrical System CHANGE: a) Replacement of Synchronizing Check Relays 25C1 and 25D1 in the 4160 volts electrical system. b) Downgrading the classification of the relays from Nuclear Safety Related - (Q) to Non-Safety Related (non-Q). c) Installation of isolation fuse for the BD transformer voltage signal from potential transformer PT-2, located in AC103, to relay 25/Cl; and installation of isolation fuse for the AC transformer voltage signal from potential transformer PT-2, located in AD103, to Relay 25/D1. REASON FOR CHANGE: The originally installed Westinghouse relays type CVE did not perform satisfactorily and were a maintenance burden. These relays had a history of contact velding, because in their application, the relays were continuously energized. SAFETY EVALUATION SUt24ARY: Synchro check relays 25/C1 and 25/D1 are provided to cross connect the busses C1 and D1 with transformers BD and AC respectively. Relays 25/Cl and 25/D1 do not perform a safety function, nor do they indicate performance of a safety system. The new relays replacing 25/C1 and 25/D1 will perform identical function, and therefore there is no change in the existing functional design of the circuit. The new relay is considered to be more reliable than the existing relay. In the event of a fault in the relay circuit, the relay will be isolated from the rest of the PT2 circuit. Therefore as a result of this equivalent componant replacement, there will not be any adverse effect on plant safety. l l

1 2 j SAFETY EVALUATION SU194ARY ) FOR MOD 90-0045 (SE 92-0005, R01) TITLE: Removal of Internal Components for Check Valves SW85, SW101 and Removal of SW93 -i 5 CHANGE: Removed check valve SW93 and removed the internals of check valves SW85 and SW101. REASON FOR CHANGE: FCR 84-115 locked closed gate valves SW86, SW94, and SW102 which are down stream from check valves SW85, SW93, and SW101 respectively. Thus, the-- function of check valves SW85, SW93, and SW101 of preventing backflow between l the two Service Water trains was eliminated. Instead, locked closed gate { valves SW86 SW94, and SW102 provide the isolation between the two Service Water trains now. SW85 SW93, and SW101 check valves are left subjected to no flow conditions during normal plant operations which may lead to the corrosion of the i internals and subsequently valve failure. Therefore, the internal components of the bolted bonnet check valves SW85. SW101 and the complete welded bonnet check valve SW93 was removed. SAFETY EVALUATION

SUMMARY

j The check valves are installed in seismic piping. Therefore the reduction of weight by removing the internals of check valves SV85, SW101,'and the removal of valve SW93 was reviewed against the current seismic analyses. The review concluded that the removal of the valve internals or the valve would not [ adversely affect any seismic qualification of the piping. The removal of internals of check valves SW85, SW101 and the removal of check l valves SW93 would not adversely affect plant safety because the valves now do j not perform the function of preventing backflow. The modification will enhance l the safety of the plant because it will eliminate the possibility of the check valves f ailing, while preserving the option to supply Service Water co train 2 ECCS room coolers from Service Water train 1 supply header during abnormal-line up. l I J i 4 ~

~ l j SAFETY EVALUATION

SUMMARY

FOR MOD 91-0013 (SE 93-0012) -TITLE: 1 Rewiring of DH7A and DH7B l CHANGE l Moved the connection of the control and indication circuits upstream of the breakers on BE1157 (DH7B) and BF1148 (DH7A), such that position indication is 1 available with the breakers open or closed. ) i l REASON FOR CHANGE: License Amendment 174 allows breakers BE1157 and BF1148 to be'left open during j normal operations to eliminate an operator action for a serious' control room j fire. SAFETY EVALUATION

SUMMARY

1 Moving the connection point of the control / indication circuit to the line side from the load side of breakers BE1157 and BF1148 will provide power to the valve position indication lights in the control room, regardless of breaker position. Properly sized fuses will be seismically installed in the primary leads to the control power transformer to ensure that any faults in the control / indication circuit will be isolated from the 480 vac system. Eecause the control circuits for breakers BE1157 and BF1148 are not de-energized, caution labels will be placed on the breakers stating control circuit'is energized to prevent misunderstanding and possible personnel injury.. There will be no effect on the operation of DH7A and DH7B'or their associated MCCs with this change. Moving the connection point of the control / indication circuit to.the line side of 480 vac circuit breakers BE1157 and BF1148 will not. invalidate the Emergency Core Cooling system valve position information readouts available to the j operator as described in USAR Table 7.5-1. l i 1 .- { l j )

... ~.. i SAFETY EVALUATION

SUMMARY

FOR l MOD 91-0015 (SE 92-0046, R01) TITLE: Control Rod Drive Group Power Supply Programmer Replacement CHANGE: i This modification replaced the existing electromechanical programmer controller. circuitry and programmer motor hardware in the Control Rod Drive Control System (CRDCS) with a functionally equivalent microprocessor based system. REASON FOR CHANGE: The purpose of this replacement was to improve CRDCS reliability and reduce the maintenance required for the current configuration. -) SAFETY EVALUATION

SUMMARY

There is no adverse effect on safety. The proposed modification does not affect the primary trip (safety) portion of the ORDCS. The proposed modification installs equipment with the same functions as the existing equipment with two exceptions The first exception is the deletion of. Progranmer Lamp fault indication. This is due to the new design not using incandescent lamps and therefore not needing to detect a Programmer Lamp fault. As the programmer lamp fault circuitry has no safety function and programmer lamps themselves are being replaced by solid state circuits there is no effect on safety by deleting Programmer Lamp fault circuitry. The second exception is deletion of Direction Error circuitry. As it now functions the existing Direction Error circuitry detects failures only in the command logic portion of the Programmer Controller (it does not sense the direction of phase sequencing or actual rod motion). The microprocessor hardware and sof tware functions replace the conmand logic circuitry. Since it would take several coincidental failures within the microprocessor for the same failure to. occur and since the Direction Error circuitry has no safety function, the deletion of this function has no effect on safety. -The Electronic (Secondary)-Trip function is maintained by the new circuitry. It differs from the existing circuit in that the Programmer Lamp and Photodiodes have been removed. Instead of de-energiring the Programmer Lamp the Trip "C" and "Da contacts external to the Programmer Controller will de-gate the gate drives directly. The Programmer Lamp and Photodiodes are high maintenance items. Their removal enhances the reliability of the system. 1

~., t i l L The proposed modification installs a microprocessor based Programmer Controller. Rod speed is controlled by reference to the processor's quartz j crystal controlled clock. The existing Programmer Controller maintains rod. l speed by means of a synchronous motor. Rod speed can vary in some.small proportion to deviations in AC line frequency in the existing-Programmer l Controller. The new Programmer provides the same Run and Jog rod speeds as before, and these are not adjustable. j A comprehensive review of all accident analyses in USAR Chapter 15 was performed to determine what ef f ect a faster or slower Control Rod movement speed would have on acceptance criteria associated with these accidents. This I evaluation concluded that the variations in Control Rod speed which could result from a Programmer Failure are physically limited by the existing Control i Rod Drive Mechanisms to a maximum value for which the reactivity insertion rate was bounded by the existing plant safety analyses. Therefore no analyzed accidents are affected. I l l 4

~.. _ ~ l ) -r SAFETY EVALUATION

SUMMARY

j FOR l MOD 91-0020 (SE 93-0004, R02) TITLE: OTSG Tube Sleeving and Plugging CHANGE: Eighty inch mechanical sleeves are designed to span defects in the upper tubesheet, upper tube span, or at the 15th tube support plate. Alloy 690 backup plugs will be installed behind existing Ally 600 explosive' plugs in three tubes in the-lower head. Mechanical plugs are to be used for plugging steam generator tubes as required j by eddy current examination including tubes that have been sleeved. ] 1 REASON FOR CHANGE: 1 Sleeves are to be installed in both of Davis-Besse's OTSGs as a preventive measure to prevent tube leaks or to repair tubes with defects. Tube sleeving will be used as an alternative to plugging which will allow the tube to remain in service with structural integrity of the tube maintained.and only a small reduction in flow and heat transfer capabilities. i SAFETY EVALUATION

SUMMARY

Mechanical Sleeves The proper installation of a 80 inch mechanical sleeve actually improves the pressure boundary safety function because a double barrier is established 2 between the primary and secondary fluids which reduces the risk of a tube rupture. Eighty inch mechanical sleeve is seismically qualified. The structural adequacy of the OTSG sleeve was evaluated for internal and external pressure in accordance with the ASME code. The specified minimum wall thickness is conservative when compared to requirements and the sleeve exceeds the strength of the original tube for external pressure loading. The required minimum sleeve wall thicknesses were calculated for both roll expanded and ] non-roll expanded sections of the sleeve. Mechanical tests were performed on a series of assemblies to demonstrate the structural adequacy of the sleeve and sleeve-to-tube joints., Testing and analysis was also performed to determine the response of the sleeved tubes to flow-induced vibrations. It was concluded that flow induced vibration will not be detrimental in an OTSG tube sleeved in the upper span, even if the tube is i ~? completely severed anywhere between the upper tubesheet and the fif teenth tube support plate. i i

i l 1 l I As the second barrier, the 80 inch alloy 690 mechanical sleeve provides enhanced general corrosion resistance as compared to the original alloy 600 l tubing. Interaction between Alloy 600 and 690 is not a concern due to the j similarity between the composition of these alloys. Also, cobalt release rates are two to three times lower for alloy 690 as compared to alloy 600. j A 80 inch mechanical sleeve affects heat transfer capability of the tube to l remove the reactor coolant heat produced during normal power operations and affects the capability to provide. primary flow paths and heat transfer capability for both normal and emergency cooldown. These potential effects were evaluated and their magnitudes are not adverse to safety, j Expected flow rate reduction from this modification will not cause a significant reduction in the margin between the actual flow rate and the limit l l as specified in the Technical Specifications. 1 An evaluation of the effects of plugging up to 600 tubes and sleeving 400 tubes I in each steam generator was conducted. This analysis considered the impact on. normal operating conditions and on accident analysis. The effect during normal i' condition was determined to be a decrease in RCS flow and a reduction in main steam superheat temperature. The expected reduction in flow rate would be less j than 1 percent and would leave adequate margin between the Technical Specifica+1on limit and the actual RCS tlow. Also the maximum difference in flow between the loops would not result in a asymmetric flow pattern that invalidates current analysis. The effect on superheat temperature is a reduction of about 30F. Therefore, an adequate margin to the design value of 350F for superheat will be maintained. Also the distribution of plugged tubes will not result in steam with a quality less than one exiting the steam generators. Eighty inch mechanical sleeving does not adversely affect the safety of affected Systems, Structures, and Components (SSC). Eighty inch mechanical sleeving actually improves the pressure boundary function and is preferable to plugging when considering hydraulic ef fects and heat transfer surface area. Welded Tube Plugs The design of the welded plugs meet all applicable requirements. The welded tube plugs will be used in tubes which have already been plugged with explosive plugs, therefore there will be no effect on reactor coolant flow j or on the heat transfer capability of the steam generater. I The tubesheet cladding has a nominal thickness of 5/16 inches and a minimum thickness of 1/4 inches. Since the depth of the spotface to be machined on the cladding is.055" to.025' the cladding will not be violated. The maximum diameter spot f ace will not violate the minimum weld leg width (ni any adjacent tube. Therefore, there are no adverse effects on adjacent tubes from installing the spotface. The effect of enlarging the hole in the tubesheet has been evaluated as having no adverse effects on the tubesheet.

f i i g The tube end will no ' longer be c'onnected to the tubesheet following the 'l machining process. However.the tubes that will have welded plugs. installed ij already have two explosive plugs located approximately 12 inches and 6 inches -l into the tubesheet. The explosive plugs have been verified through testing to i cause the tube to contact the tubesheet, which prevents the tube from vibrating in the tubesheet. Therefore, there are no adverse effects on vibration or stability of the tubes that will have welded plugs installed. { Mechanical Plugs l The design of mechanical rolled plugs meet all applicable requirements. Results of qualification tests performed on the-mechanical rolled plugs are [ ~ documented. The tests included of hydrostatic tests. leak tests, thermal-cycling, transverse fatigue tests, pressure cycling and axial pull tests. The plugs did not move during the axial pull test, nor did they eject during the l hydrostatic test. l The mechanical roll plug was designed so that the leak rate will not exceed j 1 gallon per minute (GPM) with a total of 202 of all " *** E'" 3 plugged. This 1 GPM limit is equivalent to.0144 in /hr per plug for a steam generator with 16,016 tubes. The final average leak rate _ measured for plugs i after qualification testing was.0037 in #/hr, or 402 of the allowable rate. For slee3e plugs the average leak rate measured after qualification testing was '{ .0032 in /hr which is 222 of the allowable rate. Therefore the contribution to leak rate from the number of tubes to be installed by this modification will 1 ,t be significant. Even if 20% of all tubes are plugged the leakage would be i's than the allowable per Technical Specification. i i I 1 3 e 'i I l i 4 i l 1 I

~.. SAFETY EVALUATION

SUMMARY

FOR MOD 91-0049 AND UCN 92-059 (SE 92-0029, R01) l } TITLE: .j .L Revision to the Auxiliary Building Blowout Panel Release Pressure CHANGE: Modification 91-049 implemented the plant changes required to revise the blowout panel release pressure in Auxiliary Building Rooms 303 and 314, and i revised the applicable design documents. j i UCN 92-059 will revise the USAR to identify the changes in annulus pressurization, revise the maximum allowable external pressure for the .j Containment Veseel, and indicate the change in the Rooms 303 and 314 blowout j panel release point. ) i REASON FOR CHANGE: 1.: The purpose of this safety evaluation is to evaluate the effects of revising .{ the blowout panel release pressure from 1.0 psi to 0.65 psi. This release pressure change is being performed to lower the maximum pressurization in the i Annulus, following a pipe break, to a value equal to the Containment Pressure l Vessels ASME code allowable. l l f SAFETY EVALUATION

SUMMARY

i The existing setpoint was calculated based on a constant EVS System flow rate-_ l of 8000 CFM and no outleakage through the negative pressure boundary, although the calculated pressures are significantly positive during initial periods i following a LOCA. The EVS fen performance curve indicates that the flow l through EVS could be significantly higher than ihe constant flow assumed previously.- The Standard Review Plan 6.2.3 assumptions allow credit.for fan' E performance. { The Containment Annulus Post-LOCA pressures were evaluated by taking credit for ~ additional flow through EVS based on fan performance curves. Calculations i still conservatively assume no outleakage through the negative pressure boundary. The maximum calculated pressure using system flow resistance corresponding to the minimum EVS flow (7200 CFM) allowed by DENPS Technical i specifications is less than 0.64 psig. The calculated pressure using system .l resistance for nominal EVS flow (8000 CFM) is less than 0.55 psig. Since i outleakage is not assumed and additional conservatisms exist in the 'l calculation, it is concluded that if the blowout panel setpoint is changed to 0.65 psig, the blowout panels will not prematurely lift during a postulated LOCA. Sensitivity studies have shown that very small amounts of outleakage are highly effective in limiting the initial positive pressure inside the negative j pressure boundary following a LOCA. Therefore, during the "one at a time" l replacement, there will always be adequate fastener capacity to maintain the blowout panel integrity. i i I i

r The allowable external design pressure using the equations provided in-the'1986. edition of the ASME B&pV code is 0.67 psig and the theoretical structural l capacity of the containment Vessel as stated in.USAR Table 3.6 is 1.93 psig. f L 'The pressurization of Auxiliary Building rooms will also be reduced by lowering the blowout panel release pressure. This has no detrimental effect to any structure, system, or component. t Tornado differential pressure produces a higher design load on the Shield j Building than the postulated pipe break pressurization. Therefore,'the Shield Building's structural integrity is not compromised. The change in the Auxiliary Building blowout panels release pressure does not adversely effect any safety related structures, systems, or components. I f t s I J i 1

~_ - i l' i SAFETY EVALUATION

SUMMARY

FOR HOD 91-0052 (SE 92-0055) p TITLE: 'l i i Service Water System Modification l CHANGE: \\ t This modification provided for the installation of two flow elements with l removable spool pieces, seven differential pressureLindicators, three containment fire suppression connections, sixteen removal spool pieces and ) three full body access points into the Service Water System (SW). l I l REASON FOR CHANGE: 1-t i l The installation of flow elements in the discharge piping of the CCW Heat Exchangers will enhance Service Water System flow balancing and. quarterly pump { t e s t in g. The installation of the spool pieces in the ECCS Room Cooler Return i lines will allow access to this piping for cleaning / inspection of.this piping some of which may occur during normal operating cycles with limited impact on l plant operation. Installation of the spools in the CAC supply and return lines will allow access to this piping for cleaning / inspection during outages. l A fire suppression connection will be installed in each of the three'CAC supply l lines to provide a water source for a Containment fire hose connection during. .j refueling outages. l r SAFETY EVALUATION

SUMMARY

)') Installation of flow elements at the location of existing restriction orifices RO-1497 and RO-1499.will not affect the hydraulic capability of the system because the bore diameter of the new elements is identical to the bore of the 1 existing restriction orifices. Installation of the Containment fire suppression connections will not affect the normal or emergency operation of the Service Water System because use of these connections will be limited to outage situations when the CACs are not required to be operable. Addition of the fire suppression connections does not affect the classification of the Service Water System as a " closed" system inside containment, as delineated in Standard Review Plan with respect to containment isolation because the normally closed valve and pipe cap prevent direct communication with the containment atmosphere. Note that administrative controls will be applied to ensure that system integrity is maintained. Therefore, this MOD does not affect containment integrity and these valvos need not be considered, and tested as, containment isolation valves.

4 9 Instrument tubing for FDI-11104 and PDI-11105 (located.in Service Water Valve l Em #1), FDI-11106 and PDI-11107 (located in Clean Waste Receiver Tank Rm #1), ~ PDI-11210 and FDI-11211 (located in Waste Gas Compressor Rm #2).has been designed non-seismic. Failure of the tubing following a seismic. event will not adversely impact the capability of the Service Water System nor create any flood concerns. Additionally, an EIT walkdown was performed to ensure that water spray from these installations will not adversely affect any safety I related equipment. The tubing for PDI-11108 has been designed to seismic category I requirements to prevent wetting of the Decay Heat and HPI train 2 . pump motors. Note that the pipe taps for all of the affected instruments are j designed to seismic category I requirements, and therefore, failure of this piping is not considered. i i i i i i r

-~.- l l l SAFETY EVALUATION

SUMMARY

I FOR MOD 92-0004-01 (SE 93-0023, R01) i TITLE: f Repair of Steam Generator 1-2 Continuous Vent Line Nozzle i CHANGE: I Modification of the Once Through Steam Generator (OTSG) 1-2 upper head by metal i removal to a new contour and elevation, enlargement of the stud holes and 'l Installation of corrosion resistant threaded inserts and the addition of l stainless steel cladding to outside the boit hole circle. .l _l Medification of the flanged joint of the Reactor Continuous Vent Line and i OTSG 1-2. l l Removal of the whip restraint held in position with the flange studs and nuts. f l Replacement of a segment of Reactor Continuous Vent Line. The replacement is limited to the line from the OTSG flange to the next flange upstream. .l REASON FOR CHANGE: l During the Eighth Refueling Outage significant corrosion due to borated water leakage was discovered at the Steam Generator 1-2 upper hand-hole flanged j connection to the continuous vent line. l 1 SAFETY EVALUATION SUleiARY: P RCS Pressure Boundarv f o A reduction in eam Generator primary side wall thickness Minimum wall thickness calculations, fatigue calculations and 'f reinforcement considerations for the hot leg and vent line were evaluated j against the original B&W calculation and ASME Code requirements and found to be acceptable. o Stresses induced by welding clad to the nozzle area' l 1 The modification to the OTSG upper primary side inspection opening is ~i comprised of removal of low alloy shell material and stainless steel cladding below the depth of damage that resulted from the leakage. The I area of metal removal vill be verified to be sound material by dye penetrant exam. The area of metal removal will be cladded with 309L stainless steel. The total surface area of cladding will be limited.to 100 square inches. Cladding will be applied to the Steam Generator nozzle area using a f tempered bead cladding repair technique specified in NG-4622.10 of ASME Boiler & Pressure Vessel Code 1986 Edition no addenda. This authorized weld repair is designed to maintain acceptable stress levels. i

i. p p 1 o Installation of threaded inserts in the Steam Generator nozzle for flange [ studs. r The threaded inserts made of SA479, TP316L material.were evaluated for f strength, minimum thread engagement and material capability and found j acceptable. In addition the insert holes were evaluated for their effects on minimum wall thickness and fatigue and were found acceptable. l Change in design of the flange joint from metal at the bolt circle.to two j o flat face flanges with a flexatallic gasket with backing rings inside the j bolt circle and flange material change to SA-182 type F316L. l The flange design was evaluated against ASHE Code requirements and found. [ acceptable. New stud torques were calculated based on desired gasket j loading. The missile effects from these studs are some what greater than j the studs in a standard 5" inspection opening because their weight is greater, but this is no greater concern than.the other Steam Generator l l closure studs since proper procedural controls are used to ensure the l studs are not torqued beyond the limit s of the material. I Approximate 1" addition in length of the continuous vent line l o " CCA-23 has been The addition of approximately 1" in length to line 2 evaluated against the original ASME calculations and found to have no impact. The new veldments, NDE, and hydrostatic testing will be done in accordance with Code requirements. j Whip Restraint Instrumentation on the steam generator required to maintain the steam generator heat transfer capability is not required to mitigate the consequences of a break at the steam generator, with a total break flow area greater than or equal to 0.05 sq. ft. It is noted that postulated break of the CVL at'other. 4 locations would have the same effects on the steam generator instrumentation as j the postulated terminal end break. Therefore, removal of whip restraint R-16 will not increase the probability of malfunction of equipment important to safety as previously evaluated in the USAR. The environmental consequences of j this pipe. failure are bounded by other pipe breaks. i Removal of whip restraint R-16 does not have any effect on safety functions of the reactor coolant system because the restraint does not serve any function as long as the RCS is not breached. Removal of whip restraint R-16 f rom the CVL introduces pipe whip and jet impingement effects due to the postulated tenminal end break at the OTSG connection. The Engineering Inspection Team (EIT) walkdowns have reviewed and evaluated the potential adverse consequences of this pipe failure. It was concluded that no conditions are created which would adversely affect the safe shutdown of the plant. The structures, systems, and components which are required to mitigate the consequences of this break are not adversely affected j by the removal of whip restraint R-16. l 1 r -.,I

t SAFETY EVALUATION

SUMMARY

FOR l MOD 92-0027 (SE 93-0007) l TITLE: i Modification'of Main Steam Piping Supports CHANGE: Modified supports (deletion, addition. or change) to five percent of the Main l Steam pipe supports in Rooms 124, 235, 236, 304, 314,'and 427. } 'f REASON FOR CHANGE: I Potential Condition Adverse to Quality Report (PCAQR) 92-0195 was written to identify concerns with the High Energy Line Break (HELB) evaluation associated - with the Main Steam supply line to Auxiliary Feedpump Turbines. Specifically, it was discovered that the environmental analysis utilized non-conservative assumptions and that the corrected calculations change the worst case -l environmental conditions of the Auxiliary Building. - i SAFETY EVALUATION

SUMMARY

[ The new design of the piping system (supports modified) meets all applicable ll code requirements, (ASME Section III, ANSI B31.1 and BTP HIB 3-1). Utilizing-j the criteria in Generic Letter 87-11 (Issue MEB 3-1 Rev. 2), the postulated break and crack locations were identified for the new configuration. fThe l results of this evaluation revealed that no new break / crack locations were ) identified and, in fact, all breaks / cracks in Rooms 124, 235, 236, 304, 314 i and 427 were deleted. Therefore, no new postulated adverse environments will be created and many previously analyzed harsh environments will ru) longer need to be considered. Based on the above, the new piping support configuration will not adversely affect plant safety. i j l l l l l i i l l

- - ~. SAFETY EVALUAT10H-

SUMMARY

'I FOR j MOD 92-0032 (SE 92-0059) l i TITLE: i L l Removal of DW61. DW161 and DW162 1 CHANGE: Isolation valves DW61, DW161 and DW162 were removed and the piping was capped. I i REASON FOR CHANGE: ] Demineralized Vater hose connection isolation valves DW61 DW161 and DW162 are' l located on Containment elevation 603' adjacent to the reactor head stand. l These valves that supply piping and support will interfere with the proposed shielding of the reactor head and vent line when stored.at the reactor head' j stand. l l 1 SAFETY EVALUATION

SUMMARY

l 1 Removing a portion of the Demineralized Water system that supplies two Demin ) Water hose connections for the Containment will not affect plant safety. This non-seismic line originally supplied the reactor head stand Demin Water ring j header which was removed by a SCC. This SCC also installed these hose connections for convenience. Ilowever, other existing hose connections that serve elevation 603

• are available for use and these valves are unnecessary.

l } 5 'l p l l' l

t Lj l { r SAFETY EVALUATION

SUMMARY

f f FOR MOD 92-0038 (SE 92-0074) l FPR 92-0492-901 (SE 92-0072) TITLE:

i i

Replacement of Letdown Cooler 1 and 2 l I CHANGE: j l Letdown Cooler #2 and #1 were replaced with equivalent coolers. The only significant differences between the existing and the replacement coolers are I (1) the replacement cooler uses type 316L stainless for the tubes, manifold and sleeve versus type 304 stainless, (2) the tube wall. thickness for the j replacement cooler is 0.072 inch versus 0.065 inch (3) the replacement cooler is certified to a later edition of the ASME Code and.(4) the replacement cooler is heavier. REASON FOR CHANGE: Letdown Cooler #1 was experiencing leakage. Replacement of Letdown Cooler #2 f would preclude leakage for this cooler. i -r r SAFETY EVALUATION

SUMMARY

j -i This activity will not affect any of the functions important to safety provided-by the Makeup and Purification System or the Component Cooling Water System. The vendor has evaluated the effects of the change in material and tube f thickness on the cooler performance. The change in wall thickness changes the differential pressure and the velocity, but both are within the requirements of the design specification. B&W Calculation 32-1167226-00 evaluates the adequacy-of the cooler primary side nozzles with the change in material. The design Specification Reconciliation for SMUD Specification.DS-M-0001 and for .{ specification M-51oN, Revision 1 evaluates the adequacy of use of a later edition of the ASME Code. .A seismic analysis has been performed for the. t replacement cooler. These evaluations found the replacement cooler to be acceptable. \\ i n l i i 'l l I i

II l i SAFETY EVALUATION

SUMMARY

l FOR MOD 92-0044-(SE 92-0057) TITLE: l Removal of Vent Path from Abandoned Drumming Station l i CHANGE: This modification removed a portion of piping that provided a vent path from the abandoned Drumming Station to the station vent via the Ga4eous Radwaste l j! System. REASON FOR CHANGE: l t The Drumming Station has been abandoned in place and serves no operational function. This MOD is required to eliminate physical interference with the f Service Water pipe rework. l l SAFETY EVALUATION

SUMMARY

I Removing a portion of pipe from an abandoned system will not affect plant l safety. The downstream piping will be capped to provide the pressure boundary l for the Gaseous Radwaste System piping that the vent line discharges to, while the upstream piping will be capped for cleanliness reasons. The seismic. 1 qualification of the remaining piping is not affected by this MOD. { Additionally, an Engineering Inspection Team walkdown was performed to ensure -l that the removal of this piping did not create any new adverse environment to j 4 safety related equipment. No other consequence could result from removal of a l portion of vent piping from a system that is not in use. l i J j + 1 d v

SAFETY EVALUATION

SUMMARY

{ FOR -MOD 92-0047 (SE 92-0068)' TM 92-0007 (SE 92-0025). } TITLE: Elimination of Local Control / Indication for MS106, MS106A, MS107 and MS107A CHANGE: _j i Modification 92-0047 provides for field wiring changes and drawing revisions-i for the permanent elimination of the local control station and associated j position indication for Main Steem to Auxiliary Feed Pump Turbine Isolation j Valves MS106, MS106A, MS107 and MS107A. Local Control Stations NV01060, NV0106A, NV01070 and NV0107A were disconnected and bypassed at the respective terminal boxes EV01060, EV0106A, EVO1070 and EV0107A. The pushbuttons, i l indicating lights, rocker switches and boxes were also removed from the field l by this MOD. .l l l REASON FOR CHANGE: L This modification is being implemented in order to prevent a potential failure l of this equipment during a postulated high energy line break. 5 l SAFETY EVALUATION

SUMMARY

l There are no safety functions for MS106 (MS106A, MS107, M5107A) which j necessitate the local electrical operation of these valves nor the use of the local position indication lights. I The ability of the valves to respond to a Steam Feedwater Rupture Control l system actuation signal is maintained and unaffected. Similarly, the ability j to operate the valves, from the control room hand switches, is maintained and j unaffected. .i l The FRAR has been reviewed and determined to have no adverse impact on its .i analysis. This modification does not impact the ability for remote

• manual" closure when required by station conditions, as identified in USAR 7.4.1.3.1.

Remote manual operation vill not be adversely affected. Also, this modification does not affect the ability to block and reposition these valves, as necessary, follawing an SFRCS actuation. This modification vill not adversely impact the redundancy or diversity of'the AFW System or SFRCS. i This modification does not impact the interlock from the Decay Heat Removal System which inhibits the opening of MS106 (MS107) when DH12 (DH11) is open. l In conclusion the safety functions of MS106. MS106A, MS107, and MS107A'are not adversely affected by the functional elimination of the local electrical control station and local position indication. i l i

SAFETY EVALUATION

SUMMARY

FOR PCAQR 92 0372 (SE 93-0044) TITLE: 3 Use As-Is Disposition for Station Batteries CHANGE: The results of PCAQR 92-0372 evaluation were to: o Continue to use the station batteries as de-rated by the manufacturer and evaluated by calculation C-EE-002.01-010/R05 Revise the USAR description of the batteries and loads to agree with the o C-EE-002.01-010/R05 REASON FOR CHANGE: PCAQR 92-0372 was initiated to document a 10CFR Part-21 ncfication. This defect concerns the inability of the batteries to perform t. yublished specifications, in particular the 1-minute discharge rate. At 1600 Amps, battery terminal voltage is less than the published value of 1.75 Vo'ts per cell (105 Vdc for 60 cells). The cause was attributed to a phenomenon known in the battery industry as " Coup de Fouet

• and involves the rate of reaction when converting chemical energy to electrical energy under high current conditions.

SAFETY EVALUATION

SUMMARY

USAR Section 8.3.2.1.2 states that the 8 hour and 1 minute rating of the batteries is 187.5 and 1600 Amps. These values vill be changed to 187 and 1400 Amps. The basis for 187.5 Amps is judged to be due to battery data available at the time of the FSAR preparation. The change from 187.5 Amps to 187 Amps is considered insignificant. i USAR section 8.3.2.1.2 presently states that the battery one minute load is 1130.3 Amps. This value is in slight disagreement with the total of battery IP load list presently in the USAR (1090.3 Amps). The one minute load of 1130.3 Amps, wher L itiplied by design factors of 1.11 (battery' temperature of 600F) and 1.;> (battery aged to 80% of original capacity) is corrected to 1568 Amps. This value is slightly below the original rating of 1600 Amps and i exceeds the revised rating of 1400 Amps. During revision of C-EE-002.01-010 the Reactor Coolant Pump DC Oil Lift Pumps were removed from the battery load lists. The lift pumps are turned on for RCP starting and are turned of f shortly af ter the RCPs are running. The normal-state of the RCP DC oil lift pumps is OFF. The effect of removing the RCP DC 011 Lift Pumps from the DC system load list is a reduction of 247 Amps from the first (1 minute) step and 75 Amps from the second (29 minute) step. e

I l The effect of the reducing the battery one minute rating by 200 Amps (for Coup de Fouet) is more than offset by removing the RCP DC Oil Lift Pumps from the -j load list Revision 05 of calculation C-EE-002.01-010 indicates that battery-IP is still the most heavily loaded. battery, but with a first minute step of 820 amps. This value increase to 1138 Amps when multiplied by the temperature correction factor (1.11) and aging factor (1.25). However, the worst case one minute load is well below the de-rated value of 1400 Amps. i The net effect on the battery analysis (calculation C-EE-002.01-010) of l including the battery defect and removing the RCP DC 011 Lift Pumps is an increase in margin of one minute capability above one minute load requirements. Calculation C-EE-002.01-010/R05 verifies that battery voltage is well above the design lower limit of 105 Vdc for all steps of the design duty cycle. l i l l 'NM '.iCui -..,-m- -. =. ..e----i.- v-wgruser--unverw-

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~i = SAFETY EVALUATION

SUMMARY

l FOR SE 91-0081,'R01 .j TITLE: Modification of the Hydrogen Supply Piping to the Makeup Tank by the Addition of Manual Isolation Valves j l CHANGE: To provide a positive means of isolating the Makeup Tank (HUT)Lfrom the - Hydrogen supply piping external to the Auxiliary Building, two manual valves .l upstream of G4977A (normally closed) were installed.- In addition, a j maintenance isolation valve (normally open) was installed downstream of MUS4. j REASON FOR CHANGE: Backleakage from the MUT past check valves MU385, MU187 and solenoid valves MU54 G4977A and G4978A allowed MUT water to enter the Hydrogen supply piping external to the Auxiliary Building. Previous operational history required periodic draining of condensate from the piping. On November 5, 1991 however, l with the above valves leaking the Makeup water was unintentionally drained to I the ground outside of the Auxiliary Building. SAFETY EVALUATION

SUMMARY

{ The installation of the two normally closed manual isolation valves will l provide two additional isolation boundary between the Hydrogen Supply system l and the Makeup Tank. The valves will require manual operator action to operate when establishing Hydrogen to the Makeup Tank. - The addition of the valves and appropriate supports will not create any adverse ll conditions, since the installation has been evaluated to assure compliance with the ANSI B31.1 Code and the supports for the piping have been evaluated for the imposed loading. a Disabling drain valve G222 provides additional assurance that the Makeup Tank water will not be inadvertently drained outside the Auxiliary Building. I I k i

SAFETY EVALUATION SUFMARY FOR SE 93-0025 TITLE: Cycle 9 Reload and Core Operating Limits Reports CHANGE: Cycle 9 Reload REASON FOR CHANGE: The Cycle 9 Core Operating Limits. Report (COLR), and the Reload Report'on which it is based, were developed with NRC approved methodology by the B&W Fuel Company (BWFC). The reference cycle for the analyses was cycle 8. The cycle 9 core loading, as described in the Reload Report, includes a batch 9 assembly that was repaired during 7th refueling outage (7RFO) and 64 batch 11 fuel. assemblies that are of a Mark B8B design, containing uranium enriched to 3.77 w/o U-235. This batch 11 reload also incorporates a few fuel rod design changes, such as an increased fuel density to 96% Theoretical Density, a larger outside diameter fuel pellet, a decreased active fuel length, modified lower end plugs and reduced pre-pressurization. Eight Extended Life Control Rod Assemblies (ELCRAs) will also be substituted for eight standard Control Rod Assemblies (CRAs) that will approach their design lifetime at the end of Cycle 9. Because of potential damage incurred by a CRA (CO34), one of the eight CRAs (CO38) was re-inserted.in its place, subsequent to an evaluation which concluded that the CRA could operate for one more cycle. SAFETY EVALUATION SLM1ARY: The reference fuel cycle for Cycle 9 is Cycle 8, and nuclear and thermal-hydraulics analyses were based on duration of cycle 8 to 453 (+15, -30) Effective Full Power Days (EFPDs), following an Axial Power Shaping Rods (APSRs) full withdrawal at 400 (+/-10) EFPDs and a planned coastdown. Cycle 9-was analyzed to 500 EFPDs and the applicability of the Cycle 8 RPS limits and setpoints to Cycle 9 were also verified to 500 EFPDs. There have been no facility modifications that have affected the cycle 8 RPS setpoints, so it is concluded that maintaining the same setpoints for cycle 9 would have no effect on safety. The control rod group designations for cycle 9 also remain unchanged from those of cycle 8, as reflected in the design analyses by BWFC. Eight CRAs were replaced with 8 ELCRAs. These CRAs have approached or would have slightly exceeded design limits during Cycle 9's operation, based on conservative, generic design analyses. Because of an inadvertent drop of a Control Rod Drive Mechanism's leadscrew onto CRA C034, it was deemed prudent to assume that some impact damage may have been incurred by the CRA's spider assembly. Based on an analysis, specific to four out of the eight discharged CRAs, BWFC concurred q.

with the decision to re-install one CRA (C038) for cycle 9. The specific analysis showed that the use of these four CRAs for one more cycle would have no effect on safety (since the most limiting criterion of <0.25% cladding strain would be satisfied) nor would it cause operational constraints. The design of the ELCRAs is very similar to the CRAs, and has been the state-of-the-art replacement in all of the E&W Fuel Company's Mark-B reactors. To achieve the " extended life' capability, the control rod design was modified to alleviate a cladding strain limitation. The cladding material of the ELCRA l rodlets was changed to a more corrosion resistant lnconel 625, rather than 304 l stainless steel (SS) as used in the standard CRAs' cladding. The performance of ELCRA's in other EWFC plants has been excellent, and the design changes were factored into the Reload Report Analyses. Therefore, there is no effect on safety, as was also concluded in the Safety Assessment and Significant Hazards Consideration for the Technical Specificatien 5.3.2 amendment for the CRA's design description. Furthermore, there have been no operating anomalies durg; Cycle 8 which would affect safety or fuel performance during Cycle 9. There was an indication of one, or possibly two, fuel defects in the last half of Cycle 8. Evaluations of coolant activities indicate that there is a high likelihood that the defective fuel resides in batch 8 fuel that will be discharged during 8EFO. EWFC also analyzed the effects of operation of an FA that had been repaired during 7RFD. NJ0542 was repaired by insertion of a SS rod in place of a fuel rod. The analyses, with NRC approved methodology, addressed the effect of a slight increased amount of local (in-assembly) peaking. The assembly will be in-core for its third cycle and will be on the core periphery in location F12, 1 operating at a ~70I lower radial power than during cycle 8. The assembly in that location will have over 1002 margin in Departure from Nucleate Boiling Rat io (DUER) relative to the limiting fuel assembly. Therefore, there is more than adequate DNER margin available to justify operation of the core with the reconstituted assembly, NJ0542. EWFC has also evaluated the impact of the l reuse of NJ0542 from a Loss-Of-Coolant-Accident analysis standpoint, with the cenclusion that the Cycle 9 core is insensitive to the presence of one st ainles s steel pin. The core loading for Cycle 9 is a Very-Low-Leakage (VLL) pattern. For a VLL loading scheme, high burnup assemblies are strategically placed to effectively l reduce neutron fluence to the reactor vessel, and critical welds in particular. l The reduction of neutron flux was evaluated in order to anticipate the response of ex-core Nuclear Instrumentation (NI). The implementation of the VLL reload j shuffle scheme for Cycle 9 is expected to have an insignificant impact during all aspects of reactor startup and power operation. The acc"prability of these design and manufacturing changes to the Mk-E8A FA, resulting in the Mk-E8B FA, was verified with Davis-Eesse fuel analysis methodology by use of the NRC approved ESCORE fuel rod modeling code as well as by performance of Quality Assurance fabrication surveillances and technical reviews of Reload Peport analyses. The acceptability of the mechanical design and manufacturing process changes has been <Jocumented. l l l 1 l l l

q 'l l 'j I SAFETY EVALUATION SUltiARY FOR j UCN 92-032 AND UCN 92-033-(SE 92-0033) I f t ~ TITLE: Revision to USAR to Add Loads and Revise Indicating Lights-to the 125V DC Essential Panel Description i CHANGE: i 1) USAR Change Notice (UCN) 92-032 corrects USAR Section 8.3.2.1.4 which lists the loads on 125V DC essential panels D1P, D2P, DIN, and D2N.

j i

2) USAR Change Notice (UCN) 92-033 corrects USAR Section 8.3.1.6.c which j discusses the locally mounted instruments on essential 125V DC panels. l REASON FOR CHANGE: 1) As new essential loads were added to panels D1P, D2P. DIN and D2N, the l loads identified in USAR Section 8.3.1.2.1.4.a through 8.3.1.2.1.4.d were' not updated. In addition, some of the-original construction loads were-j not listed. I l 2) Originally USAR Section 8.3.1.6.c indicated that four indicating ?ights, { two for each disconnect switch are on each 125V DC essential panel. Thie j is not correct. Each 125V DC essential panel has one indicating light for

j l

undervoltage indication.- This has been the configuration of the panel indications since insta11at2on. i l SAFETY EVALUATION

SUMMARY

1 The safety function of the 125V essential DC distribution system is not affectd by UCN 92-032 or UCN'92-033 since both UCNs only update the current DC panel load lists and panel indications listed in the USAR as they are configured in the plant. The identified loads on the essential panels have been evaluated by a DC Calculation and represent no effect on safety or the function of the 125V DC distribution system importent to safe plant operation. ~The editorial' correction by UCH 92-033 that the essentie' DC distribution panels only have one indication for v..dervoltage does not change the plant configuration, thus also does not effect the 125V DC distribution system function for safe plant operation.

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SUMMARY

l FOR l -UCN 92-052 (SE 92-0026) I ~ TITLE: 't Revision of Battery Load Tables CHANGE: UCN 92-052 is to update _the existing USAR DC system discussion and load table l to agree with calculation C-EE-002.01-010. j .i REASON FOR CHANGE: Section 8.3.2.1.2 of the USAR, titled ' Station Batteries" contains a table of loads associated with each of the station batteries. IP, IN, 2P and 2N. This section was last revised in 1989. Since then, calculation C-EE-002.01-010, revisions 00 and 01 have been issued. Both revisions of the calculation used 3 values other than those listed in the USAR table for determining minitum j battery size. SAFETY EVALUATION

SUMMARY

Revising the USAR based on calculation C-EE-002.01-010 will not adversely affect the safety function of the DC Power System nor equipment supplied by-the l' DC Power System. The results of calculation C-EE-002.01-010 verify the following as discussed under " Design Bases

• in USAR Section 8.3.2.1.2:

'The batteries are sized to supply the anticipated DC and instrument AC supply for a period of one hour after the loss of the battery charger supply...The above includes about 20 percent over capacity to compensate for the loss due to aging of the batteries over a 20-year period...The loads and operating requirements during the worst case accident with no AC power available (a loss of offsite AC power, a LOCA and no credit is taken for the battery chargers) are detailed below...' In addition, the calculation verifies that: Voltage to equipment required to operate during a LOCA followed by a loss of AC, remains above the minimum required for equipment operation; i The battery chargers are sized to recharge the batteries within 12 hours l concurrently with supply DC system loads; and 2 Available short circuit current is within the rating of the DC 2 distribution equipment. f r n . ~.

- - ~. i -i I [ Technical Specification Bases does not discuss battery or battery. charger l sizing, system voltage drop or short circuit capability in detail. It does l address maintenance testing of storage batteries in regard to demonstration of l operability. The Bases is not affected by revision of calculation C-EE-002.01-010 (or the USAR table). However, the maintenance procedure for performing Battery Service Test could be af fected by a revision to the battery I hour discharge profile or " design duty cycle". It was verified that-the test duty cycle required by Maintenance Procedure DB-ME-03002 ROO, Station Battery d Service and Performance Discharge Test, envelopes both revisions of the design 'l duty cycle. j 1 In developing the design duty cycle, calculation C-EE-002.01-010 assumes that manual load shedding takes place by plant operations personnel. The procedural i actions required by Abnormal Procedure DB-OP-02521 R02 Loss of AC Bus Power l Sources, are consistent with the load shedding assumptions in the calculation. Revision of the USAR DC system discussion-based on calculation C-EE-002.01-010 j will clarify " anticipated DC and instrument AC supply" discussed above. .j i l l l I \\ l 3 i

-i 7 SAFETY EVALUATION SUMHARY l FOR j UCN 92-056 (SE 92-0048) } i TITLE: j Delete Control and DC Power Circuits from AC Power Circuits Paragraph in.the USAR CHANGE. i Revision of USAR Section 8.3.1.2.23, AC Power Circuits, to delete reference to control and DC power circuits. REASON FOR CHANGE: j USAR Paragraph 8.3.1.2.23 (titled 'AC Power Circuits *) states that all control i and DC power circuits are protected by fuses. This statement is inconsistent I with USAR Figure 8.3-25, which shows a breaker at D118 (a DC power circuit from a fuse-protected panel), and is also inconsistent with the_ paragraph title, f which specifically excludes control and DC circuits. This UCN deletes the-reference to control and DC power circuits. This requirement is deleted. l rather than moved, because the design of protective devices _for control and DC 1 power circuits must be performed on a case by case basis, taking into account the relative importance and time-current characteristics of each load. SAFETY EVALUATION

SUMMARY

f l There is no effect on safety. since this USAR change will not cause any change -l in the design or operation of the plant Proper protection and coordination of electrical circuits is guaranteed by existing design practices, including calculations which verify acceptable voltage drops, ampacity, and coordination for all Class 1E circuits. Similar. calculations are also part of basic design l j practices for non-1E circuits, although such work may not always be formally l documented as a calculation. Methods for establishing the acceptability of electrical designs are based on applicable guidance from the ANSI /NFFA 70 i (National Electric Code), IEEE S-135/IPCEA-P-46-426-(Power Cable Ampacities), the Design Criteria Manual, the Electrical Design Guide, and other standards ') and commitments relating to electrical distribution. .) 1 l l I l l l

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SUMMARY

FOR UCN 92-067 (SE 92-0036. R01)- 1 TITLE: Alternate Amines and Steam Generator Feedwater Quality t CHANGE: Update Table 9.3-5 Section 10.4.7.2 and 10.4.7.3 in the USAR to reflect the current status for controlling steam generator (S/G) feedwater quality. REASON FOR CHANCE: Chemical control for the secondary side system at Davis-Besse Station has utilized an all-volatile treatment (AVT) since initial operation. The' original i AVT consisted of using ammonia for pH control and hydrazine for dissolved oxygen removal. In May 1989. morpholine addition was. implemented for AVT chemistry control. After changing to morpholine addition. the iron. transport l to the steam generator was reduced by approximately a factor of three. A decrease in iron transport rate due to a change in amine can be directly l l related to a proportional decrease in erosion-corrosion rate for the affected .l l piping. Iron transport to the steam generators is expected to be further i reduced using one of the following amines: ethanolamine (EA). [ l 2-amino-2-methypropanol (AMP) or 3-methoxypropylamine (MPA). The morpholine in. current use is also an amine. .j L SAFETY EVALUATION

SUMMARY

l l Changes to the chemical control for the Main Feedwater System (MFS) will; I l provide additional protection to the surfaces of affected system and therefore l will not adversely affect MFS reliability. The only safety function of the MFS which is the containment isolation is unaffected by the. change in the chemical 3 control program. Because of the wealth of information available for i morpholine. it becomes a standard against which to compare the results of other amines. i In an Electric Power Research Institute (EPRI) project 96 amines were tested to l obtain data on base strength and volatility. From this research,~10 amines had. characteristics superior to ammonia and morpholine. From these 10. three have. I been selected for field evaluation at Davis-Besse Station. 'These are EA l (Ethanolamine). AMP (2-amino-2-methylpropanol) and MPA (3-methoxypropylamine). A summary of pH and relative volatility data plotted against temperature for-the three amines as well as for morpholine shows that the effectiveness.of all three amines should be an improvement over morpholine by raising the pH in-the j moisture separator drains and condensate. l' Another EPRI project was conducted to evaluate amines in a high pressure test boiler system (HPTBS). The HPTBS provides an unusually harsh test envircnment for amine decomposition in comparison to a PWR. Extrapolation of the kinetic. data to PWR conditions indicates none of the amines tested is likely to show 1 f

more than four percent decomposition: therefore, each is sufficiently stable for use as AVT in a PWR. Babcock and Wilcox, Alliance Research Center performed heated crevice tests for. EA and AMP. Intergranular attack (IGA) was introduced in the specimens prior to testing. Creviced specimens were placed on Alloy 600 specimens (S/G tube material). The alternate amines being considered for testing did not induce IGA in Alloy 600 tubing in creviced regions. Even though MPA was not used in the heated crevice tests. Its use for field evaluation is justified based on its similar properties to EA and AMP. Constant extension rate tests (CERT) were performed by Babcock'& Wilcox, Alliance Research Center for EA and AMP, Comparing the failure surfaces and nature of the secondary cracks of the 10 percent NaOH CERT specimen and those run in EA and AMP clearly showed no evidence of stress corrosion cracking. Based on these results, EA and AMP do not appear to enhance SCC. Although MPA was not used in CERT, its use for field evaluation is justified based on its similar properties to EA and AMP, as stated above with the crevice testing. Compatibility of organic amines on PWR gasket and packing materials has been studied in an EPRI project. Testing was performed using morpholine, EA and MPA at concentrations up to 50 percent at 2000F for one month. Materials not-degraded include Dixosteam. EPDM, and Kalrez. Materials found unsuitable include Neoprene, Red Rubber. Silicone Rubber, and Viton. The main steam condensate and feedwater systems have raised faced flanges and use flexitallic gaskets which are not degraded. Although AMP was not tested in the study, it similar chemical properties to EA and MPA would make AMP suitable for use, particularly at the very low concentrations which it will be used. Resins have also been tested with aqueous morpholine and EA with no evidence of significant degradation. In amine solution concentrations up to 10 percent, which is considerably higher than those present in condensate polishing units, only small changes in ion exchange capacities, moisture retention, swelling, and bead integrity were observed. Amine compatibility with cation resins, and morpholine, EA and MPA, were assessed. The amines tested did not physically or chemically degrade the resins under the test conditions. The Ohio EPA has approved the use of EA, AMP and MPA at Davis-Besse Station. Any amine used in place of hydrazine will not be used until it has been evaluated to properly control Dissolved Oxygen (DO) in the condensate and feedwater within the station chemical control specifications. l The proposed actions will not increase the probability of occurrence of an l accident previously evaluated in the USAR because there are no physical changes i created with the use of the alternate amine which would affect the safety function of systems when these chemical are used. Therefore, adding the alternate amine is not associated with any accident initiator. Changes to USAR reflect current chemistry control which is consistent to station procedures, vendor and EPRI guidance, excluding the hydrazine specification. Hydrazine may be substituted with another amine provided its ability to scavenge dissolved oxygen is equivalent to hydrazine and not have any detrimental affect on metal i surface.

SAFETY EVALUATIO!!

SUMMARY

FOR UCN 92-079 (SE 92-0051) TITLE: Change to USAR Section 8.3.1.1.12 CHANGE: Revision of USAR Section 8.3.1.1.12 regarding purchase specifications of safety related Limitorque actuators. REASON FOR CHANGE: ~! During an Independent Safety Engineering (ISE) audit of Davis-Besse's program. for valve operator sizing and torque switch setting calculations,.ISE was~ concerned that the current USAR guideline provided an incomplete description-of the acceptance criteria used regarding valve operator terminal voltage. This j is because several valve operators require greater than.70 percent rated terminal voltage as stated in the USAR, in order to produce sufficient torque ) for the motor operated valve to perform its safety function. SAFETY EVALUATION

SUMMARY

l There will be no adverse effects on safety as a result of.the proposed USAR j Change Notice (UCN). No physical changes ~to the plant are performed by this -j UCN and no policies or guidance is created which results in'the plant or any of. 1 its equipment from operating outside of their design basis. Limitorque Actuator sizing and torque switch settings will still be determined via the' same method as in the past, and will continue to use the calculated worst case .) voltage as an input into that determination. The proposed change simply provides a clearer understanding of the methodology used in thrust calculations j with respect to reduced voltage. 1 i l l

SAFETY EVALUATION

SUMMARY

FOR UCN 92-084 (SE 92-0058) TITLE: Purification Demineralizer Filter Bypassed CHANGE: The change eliminated the use of the Purification-Demineralizer. Filter (PDF) except during heatup or cooldown. For isolation of the PDF,. valve MU94 was. changed from open to closed, and MU63 from closed to open for normalfoperation. REASON FOR CHANGE: The Purification Demineralizer Filters are being bypassed to reduce the volume of radwaste generated at Davis-Besse. SAFETY EVALUATION

SUMMARY

Both the PDF and bypass lines are sized to handle the maximum letdown flow rate The proposed change with the PDF being bypassed will reduce the volume of radwaste generated because filter replacements are reduced. Radioactive particulates in the RCS will not be increased. Although the Purification Demineralizers will be utilized for removing crud, the resin will not-become iron fouled causing increased differential pressure because the iron concentration in the RCS is less than 10 ppb: iron fouling would occur at _ concentrations on the order of 1,000 ppb. The additional amount of' activity i collected on the resin will not significantly increase the total activity on the resin. l 1 During approximately the first month of operation following a refueling outage, during the cooldown going into plant ehutdown, and outage periods, the PDF-1 should be used. These are the periods when crud is released from RCS surfaces. ] I I i i i

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SUMMARY

j FOR l UCN 92-089 (SE 92-0064) TITLE: Removal of Certain Electrical Drawings and Figures from the USAR. i CHANGE: (USAR) Change Notice UCN 92-089 proposes to remove two USAR figures and several drawings currently classified as " incorporated by reference." REASON FOR CHANGE: ) i These referenced electrical drawings and electrical figures do not contribute'- to the safety analysis and are not required per Regulatory Guide 1.70, I SAFETY EVALUATION

SUMMARY

E-10A is a listing of standard notes and symbols, The drawing symbols used at j Davis-Besse are the same as, or similar to the ones generally used in ] electrical design and in the nuclear industry, and in some cases recommended for use by the Institute of Electrical and Electronics Engineers (IEEE). I E-11A contains numbering conventions for electrical equipment. These numbering conventions are based on the Bechtel numbering conventions, and are themselves based on industry practice and IEEE convention. In addition, the Davis-Besse numbering conventions are readily apparent upon examination of Davis-Besse drawings. I E-12B is the electrical motor list. It provides no unique information needed to comply with the requirements of RG 1.70. E-13B is a list of Davis-Besse equipment numbers. This drawing is not needed to interpret any of the drawings, figures, discussions, or analyses contained in the USAR. No unique required information is provided by E-13B. E-14B is a list of solenoid valves. Similar to E-12B, this tabulation satisfies no RG 1.70 requirements. E-2, E-3. E-20. E-21 Shl, E-21 Sh2, E-22 Shl, and E-22 Sh2 are metering and relaying diagrams All required relaying and metering circuits are depicted in separate schematic diagrams; therefore this information is redundant, and not required. l 1 a

.. ~. SAFETY EVALUATION

SUMMARY

FOR UCN 92-096 (SE 92-0067) TITLE: Remove Vendor Specific Information from Electrical Conductor Descriptions in the USAR CHANGE: Remove vendor-specific details of electrical conductor' descriptions from Section 8.3.1.2.21, Electrical Conductors and Table 8.3-4 Environmental Requirements for Class IE Cable. In addition, the calculated environmental requirements are being deleted from Table 8.3-4. REASON FOR CHANGE: The USAR currently identifies particular vendors when purchasing class IE cable. Table 8.3-4 not only identifies the vendor to be selected, but lists purchase specifications which only that vendor can meet. In addition, the material composition descriptions in Section 8.3.1.2.21 also indirectly prescribe the vendor to be used. The effect of the changes described above is the flexibility to select cable based on availability and price, without compromising the required performance characteristics. The environmental requirements in Table 8.3-4 are found in other sections of the USAR. i SAFETY EVALUATION

SUMMARY

e i' This is a paperwork change only and does not involve any modification to plant systems or equipment. The affected cables include cables which will be j installed under future modifications, and existing cables which may be replaced- ) under future maintenance work orders. 9 In addition. cable performance requirements are being maintained, despite the j elimination of vendor-specific details. e l l

~ .) I i SAFETY EVALUATION

SUMMARY

l FOR UCN 92-108 (SE 92-0077) 4 TITLE: l t Use of Certified Mill Test Report Yield Strength for Structural Steel CHANGE: j The USAR will be revised to reflect the practice of using Certified Mill Test Reports. REASON FOR CHANGE: .j The USAR Section 3.8 states that the " minimum yield strength" of structural steel material, shall be used to. establish the allowable design stresses. Contrary to this, the actual yield strength documented by Certified Mill Test Reports (CMTR), have been used to establish design allowables for various. structural steel members. In many cases, the use of CMTR yield strengths provide significant increase in .j capacity when compared to minimum specified yield strengths. The use of CMTR l yield strengths has been restricted to backfit applications only. The applicable minimum material yield strength is used for new design. -j SAFETY EVALUATION

SUMMARY

The primary strength characteristic of steel used in structural design is the i yield strength. The American Society for Testing & Materiale (ASTM) specifies the methodology in determining the yield strength of the material. This methodology is used to determine if the material meets the minimum allowable value and is a true indication of the strength of the material. The value of the yield strength is dependent of the type of steel. The ASTM specifications state the minimum yield strength for ASTM A-36 structural' steel is 36 ksi. Specification 7749-C-46 required that the steel supplier submit the CMTR documenting the yield strength test for each structural member. The yield .l strengths shown on the CMTR are for a particular heat of steel, which is comprised of multiple steel members. The actual member yield strengths from a. j single heat of steel can vary by as much as 20%. This variance is due to a combination of the effects. The variances are as likely to occur in members from a heat with a CMTR yield strength value of 36 ksi as they are in heats with higher yield strengths. Therefore, the variances in yield strengths are-I inherent in the steel making process and can be considered being covered by the American Institute of Steel Construction (AISC) code factor of safety. In order to justify using a CMTR yield strength, for a structural member. a continuous documentation trail must be verified. This trail must rela +.e the beam number and size to the specific steel heat number. The appropriate CMTR package which references the heat, beam number, and size must be identified. l 1 2 ) i

The documentation of this review for Auxiliary Building and Containment Internal structure. floor steel can be found in their respective calculations. Any steel member that cannot be definitely tied to a CMTR shall use the appropriate ASTM minimum specified yield value. I The value of a CMTR yield strength is independent of factors which cause variations of yields with a heat of steel. Therefore, the use of CMTR yield strengths greater than 36 ksi, for backfit applications, is acceptable when done in conjunction with AISC factors of safety. I 1

SAFETY EVALUATION

SUMMARY

FOR UCN 93-001 (SE 93-0006) TITLE: Elimination of Quality Assurance Director In-Line Review of Measuring and Test Equipment Evaluations CHANGE: The change will remove a "non-regulatory" in-line review which has been determined to be unnecessary. REASON FOR CHANGE: Review of approximately fourteen-hundred such evaluations determined that the in-line review had not discovered any subsequent " conditions adverse to quality *, SAFETY EVALUATION

SUMMARY

The change would not alter remaining programmatic controls which include investigation and evaluation (by affected Nuclear Group department and/or Metrology Lab management) to re-establish acceptability of equipment or establish that a condition adverse to quality exists. The proposed change does not affect the sa. ty function of any SSC's. The change is solely administrative as it delete. nn additional level of review. No corrections to prior review determinations hcve been made by the QA Director's review. Therefore, elimination of this :eview is determined to be solely an administrative change without effect on safety or margin of. safety.

t SAFETY EVALUATION

SUMMARY

l FOR .j UCN 93-013 (SE 93-0024) -j I TITLc: Containment Hydrogen Generation and Control CHANGE: i .i Increase in aluminum, zine or zine-based paint in Containment and changes to j the description in the USAR of the containment hydrogen purge system. l REASON FOR CHANGE: It was requested to keep the aluminum reactor vessel core cover inside l containment. Addition of this would cause hydrogen generation rates to exceed the analyzed values. The previous perpetual need to disassemble, decontaminate, and remove this and other components from containment results in unnecessary expenditures of time and results in incrementally higher outage dose. Therefore, reanalyses of containment hydrogen generation rates have been made to allow for greater quantities of the subject materials to be stored ) J inside containment. j The containment hydrogen purge system, as described in the USAR, is being revised to make the description accurate with respect to the system function. SAFETY EVALUATION SUMHARY: f An increase in the allowable amount of aluminum, zinc, zinc-based paint inside f containment will increase the potential amount of hydrogen which is generated 1 in the post-LOCA containment environment. A Toledo Edison calculation i determined that a maximum increase in the hydrogen generation' rate of l approximately 0.05 lb-moles /hr would remain within the design capacity of both the CHD/CHP systems and the containment hydrogen recombiner. In this case, l without action, the containment hydrogen concentration limit of 3 volume percent would be reached in 17 days rather than the current 21 days. It has-l been determined that 17 days is ample time to initiate either the CHD system, i or to transport and install the hydrogen recombiner. 1n addition, in the existing USAR, the CHP system would not be needed until 110 days after the f .LOCA. With the additional proposed allowable hydrogen generation rate, the CHP j system could be needed within approximately 62 days. Although the CHP system t is fitted with particulate and charcoal filters, an earlier release would result in a higher offsite dose than a later release because less time is j available for activity levels to decrease after the LOCA. j 10CFR50.44, section (f), establishes the allowable dose associated with- ] releases through the CHP system. The existing USAR analysis does not present a j specifically calculated dose for operation of the containment purge system, but I rather provides for a range of initiation times and durations. The range of .i times extends from initiation at 21 days to initiation at 147 days, which l bounds both the existing expected initiation time of 110 days as well as the j earlier initiation at approximately 62' days (with the additional hydrogen I generation). l I 1 i I r

'.I i i Davis-Besse's original safety evaluation is more specific in that it presents a single case and also represents the basis for original NRC acceptance of the Davis-Besse combustible gas control systems. This analysis assumes initiation of purging at 24 days, a purge duration of 30 days, and an average purge flow l of 47 CFM. The initiation of purging at 24 days approximates the time at which j hydrogen concentration would have first reached 3-volume percent in the original licensing analysis. l Again making the conservative assumption that purging would commence at the f time.when hydrogen concentration first reaches 3 volume. percent (i.e. 17 days), j the method in'the original safety evaluation was repeated in calculation C-NSA-060.04-003, revision O. Assumed purge ~ flow rate was also raised in i proportion to the projected increase in hydrogen generation. rate. The results l showed that a factor of ten margin will still exist with respect to 10CFR100 j limits. Therefore, the consequences of the event-will remain within the 'l guideline values of 10CFR Part 100 and it is concluded that there is no-increase to the consequences of the event. In addition, even if a passive failure within the CHP filter were to be considered, dose rates would continue l to be bounded by the unfiltered dose rates for actuation at 21 days as' l presented by the USAR. Therefore, additional malfunctions within the system l would not lead to more adverse consequences than those already evaluated. Revisions to calculation C-NSA-060.04-003 will be used to ensure that' future additions of aluminum, zinc, and zine based paint do not cause the calculated j hydrogen generation rate to exceed the new allowable limit. USAR section 6.2.5.2.2 currently states; 'The Hydrogen Purge system functions as a backup to the two redundant full capacity Containment Hydrogen Dilution l Assemblies.' The above USAR quote was used in response to an NRC question, which addressed single failures. Supporting documentation would have made it clear that the containment hydrogen purge system did not provide additional redundancy to the j containment hydrogen ( 'on system. The above USAR section will be reworded to reflect that the centainment hydrogen purge system " backs-up" the operation-of the containment hydrogen dilution system, but is not a redundant " backup" system. I l i l

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SUMMARY

l FOR. { t UCN 93-028 (SE 93-0031)- TITLE: Nuclear Operations Reorganization I CHANGE-q The change to the USAR is the creation of a new position within the corporate i organization. l 1 ' REASON FOR CHANGE: t. l The Centerior Energy Board of Directors has created the' position of Senior Vice 3 l President - Nuclear. This new position will report to the Chairman, President and Chief Executive Officer. The Senior Vice President - Nuclear will be responsible for both the Davis-Besse Nuclear Power Station and the Perry. Nuclear Power Plant. Prior to this appointment the Nuclear Power stations j reported to the Executive Vice President - Power Generation. l l 1 1 l SAFETY EVALUATION

SUMMARY

l l l The proposed changes to the USAR consist of the creation of a new position ] within the corporate organization. i The technical qualifications necessary to operate the DBNPS continue to be i provided by the Toledo Edison nuclear organization. There continues to be - 1 l established and well-defined lines of authority, responsibility, and communication from the highest management levels through intermediate levels to and. including all onsite operational positions involved with activities affecting the safety of the plant. Therefore, ANSI N18.1-1971 remains I satisfied. l l 1 l

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SAFETY EVALUATION SUltiARY FOR UCN 93-032 (SE 93-0033, R01) UCH 93-056 (SE 93-0052) TITLE: Nuclear Operations Reorganization CHANGE: The Centerior Energy Board of Directors has approved a reorganization of the management structure at the Davis-Besse Nuclear Power Station (DBNPS). This reorganization consists of modifying the existing five director functional arrangement to four directors with the associated realignment of activities to accommodate this new structure. The Centerior Energy Board of Directors also approved a streamlining of the corporate organizatien. As a result of this streamlining, the number of vice presidents and above has been reduced from 16 to 12 with the responsibilities of the remaining corporate management personnel being Droadened. The Davis-Besse Operations Section was modified by eliminating the - Superintendent - Operations and realigning the associated activities to the Superintendent - Operations. In a previous Davis-Besse organization change, responsibilities for implementation of the site fire protection program were consolidated under the direction of the Manager - Operations. These responsibilities were shared by Systems Engineering and Operations. Responsibility for the Fire Hazards Analysis Report will be maintained in Design Engineering. REASON FOR CHANGE: As part of Centerior's efforts to improve quality and work process efficiency and to achieve organizational consolidation and realignment, a plan was made to restructure the organization. SAFETY EVALUATION SUl@iARY: The proposed changes to USAR Sections 12.3, 13.1, 13.2, 13.4, 13.5, 13.6, 13.7, and 17.2 have no effect on any st ructures, systems and components or their associated safety functions. The proposed changes are administrative in nature and do not affect the operation of any plant system. The technical qualifications necessary to operate the DENPS continue to be provided by the Toledo Edison nuclear organization. As required by Technical Specification 6.2.1.a. the new organizational structure provides well-defined lines of authority, responsibility, and communication from the highest ~ management levels through intermediate levels to and including all onsite operational positions involved with' activities affecting the safety.of the plant. The staff qualification requirements of Technical Specification 6.3.1 and the staff training requirements of Technical Specification 6.4 are not-changed by the proposed reorganization.

Cont.olidation of responsibilities for implementation of the. site fire protection program under the direction of the Manager.- Operations enhances the-overall effectiveness of the program. Appropriate staffing changes have been made to ensure.that operations personnel have the knowledge and experience to implement the fire protection program.

-. = ~ l i 6 l t l SAFETY EVALUATION

SUMMARY

l FOR I UCN 93-033-(SE 93-0036) TITLE: Operation with the Heater Drain Pumps Out-of-Service f CHANGE: i l The purpose of this safety evaluation is to demonstrate the adequacy of running f the low pressure feedwater heater drains via loop seal to the condenser, with j the heater drain pumps out-of-service and to ensure all design functions are satisfied. r REASON FOR CHANGE: 1 A review of plant efficiency was conducted when the heater drain pumps were j removed f rom service for regular preventive maintenance at the end of Cycle 8. -{ This review concluded that with all three condensate pumps on line, the plant was more ef ficient with the heater drain pumps out-of-service. ) i SAFETY EVALUATION

SUMMARY

The low pressure (LP) feedwater heater drain system / extraction steam system'is j designed to provide two equally acceptable flow paths. One is via the heater 1 l drain pumps to the condensate system and the other directly from the heater drain tanks through a loop seal to the condenser. Not running the heater drain pumps and routing the LP feedwater heater drains to the condenser is within the system design function but is not considered important to safe plant operation or safe shutdown, therefore, there is no effect on plant safety. J l

SAFETY EVALUATION

SUMMARY

f FOR UCH 93-034 (SE 93-0040) TITLE: Boron Dilution Following a Cold Leg Pump Discharge LOCA CHANGE: i The purpose of this safety evaluation is to determine if accrediting the new j flow path the reactor vessel internal gaps, as a passive boron dilution method l as a replacement for the hot leg injection method of dilution when above l 23 percent of rated thermal power (RTP) creates an unreviewed safety question. l l i REASON FOR CHANGE: j In 1991 Babcock and Wilcox found that the analysis supporting the boron l dilution methods described above had not fully modeled the reactor. vessel j internals. Consequently, when the reactor vessel internals were completely modeled, the short term passive method will provide sufficient dilution flow path thought to exist prior to opening DH-11 and DH-12, the decay heat drop .i line. An additional outcome was that the hot leg injection method of dilution was found to be potentially ineffective above certain power levels. l SAFETY EVALUATION

SUMMARY

When the RCS was originally analyzed for boron dilution requirements, the reactor vessel plenum assembly was not explicitly modeled. The analysis did not consider the plenum assembly and assumed that if the two phase mixture height above the core reached the level of the RVVVs, it would flow through the RVVVs into the downcomer. The modeling predicted that this passive dilution path would exist for a significant period of time..Therefore operator action to initiate a dilution flow path via the decay heat drop line or the auxiliary spray line could be delayed to a long-term recovery status. Upon re-examining the vessel modeling, B&W concluded that the annulus between the vessel plenum assembly and the core support shield would potentially act as a steam-liquid separator and effectively prevent liquid flow out the RVVVs. j Consequently, dilution of the core's boric acid would be limited. The reactor vessel was designed with gaps between the hot leg nozzles and the reactor vessel. The gaps were designed to be open when the reactor vessel is l I cold to permit disassembly for refueling. The gaps were designed to be. closed-by thermal expansion when the vessel is at normal operating temperature. l

~- l f Ilowever, when the reactor vessel was constructed, assembly methods and j machinery tolerances resulted in the gaps being slightly open even with the t vessel hot. Thes? gaps provide a flow path between the outlet annulus and the { inlet nozzle /downtomer region of the vessel. From this region, the liquid can j reach the break site which is postulated to be a cold leg at the pump j discharge. This provides a method of diluting the coolant which remains in the core region. i f The existence of the gaps and the flow that could occur through them were known to exist at the time of plant construction. However, because two active l methods of dilution were thought to be adequate, the capability of the gap flow I was not substantiated. Therefore, flow through the gaps was not formally 'l recognized and accepted by the NRC in any Safety Evaluation Reports (SERs) and i the gaps vete not credited as a post-LOCA boron dilution method. B&W has evaluated the as-built gaps. Using a conservative gap size and conservative assumptions and initial conditions, it was determined that there is sufficient flow through the gaps to keep the boron concentration in-the core region well below the solubility limit for initial boron concentrations of { 3500 PPM and 2800 PPM in the Core Flood Tank and the Borated Water Storage l Tank, respectively. Lower initial boren concentrations would result in lower peak concentrations. Analyses concluded that results of analyses crediting the i gap recirculation flows, show adequate core bcron concentration dilution l ) witheut any additional operator initiated systems. The existence of the i internal vessel gaps provides a passive method of ensuring that the boron concentration does not reach its solubility limit. ) In addition to this passive method of dilution, the active method of dilution through the decay heat drop line still exists. Consequently, the drop line i provides an adequate and diverse method of boric acid concentration control. If drop line flow initiation is delayed or fails to be established, the passive .{ gap flow path provides sufficient dilution by itself, as discussed above. I While the hot leg injection path is also available, it does not provide a dilution flow through the core until core boiling is suppressed when no credit I is taken for RVVV overflow or internal gap flow. This flow path would be effective for preventing precipitation only at decay heat levels following -l operation below 23 percent RTP. Therefore, this path is not as effective as the passive gap flow path or the decay heat drop line path. Consequently, the l hot leg injection path can no longer be credited as a boron dilution flow path at high power levels. Therefore, it is concluded that accrediting the reactor vessel internal gap 1 flow path as a passive method of post-LOCA horon dilution as a replacement for the hot leg injection dilution method does not have an adverse effect on safety. In addition, no longer accrediting hot leg injection as an active method of boren dilution above 23 percent of RTP does not have an adverse l effect on safety since it is being replaced by the more effective and diverse l passive method. l l l i

i i r i SAFETY EVALUATION

SUMMARY

FOR { l UCN 93-035 (SE 93 0037) l I TITLE: j ? Correction of Quench Tank Cooler Performance Data -f CHANGE: .j i Section 5.5.11 of the USAR states that the quench tank cooler is designed to -l cool the quench tank contents from 2600F to 2200F within 4 hours following a steam release UCN 93-035 changes the performance data to indicate the quench tank cooler will cool the contents to 1200F vice 2200F in 4 hours. j REASON FOR CHANGE: Correct information in the USAR pertaining to the quench tank cooler. f SAFETY EVALUATION

SUMMARY

USAR section 5.1.7 states that during normal plant operation, water in the quench tank remains at or below 1200F. i USAR Revision 4 added performance data for the quench tank' cooler to section 5.5.11. This information states that the cooler is designed to cool j liquid in the quench tank from 2800F to 2200F in 4 hours following a steam release. Normal operating temperature of the quench tenk liquid as' stated above is 1200F. A Bechtel data sheet provides the approved specifications for the quench _ tank cooler as issued during plant construction. This data sheet notes that the 3 " time required to cool water from 2800F initial temperature to 1200F final temperature shall be 4.0 hours." ) I Correcting the performance data for the quench tank cooler will have no effect on plant safety. This USAR change will allow the USAR to accurately reflect j i the design conditions of the quench tank cooler. i 1 l

.i i i SAFETY EVALUATION

SUMMARY

FOR l UCN 93-041 (SE 93-0046) l l TITLE: I tiltigation of Potential Critical Crack of an Auxiliary Feedwater Line in the Annulus. l l CHANGE: .j r Revise USAR Section 6.2.1.3.3.d to reflect preferred operational method of f mitigating a potential critical crack of an auxiliary feedwater line-in the shield building-containment annulus. j i REASON FOR CHANGE: ? For operational consist'ency, it is desired to mitigate a potential critical l crack of an auxiliary feedwater (AFW) line in the shield building / containment annulus in a similar manner to other steam generator steam leaks. l SAFETY EVALUATION

SUMMARY

j i The present USAR description of actions taken for this event delineates initiating AFW to the affected generator, an orderly reactor shutdown and subsequent 1000F/hr cooldown. In the operationally preferred method, in the event of a critical crack'in the i Auxiliary Feedwater line between the Steam Generator and the Auxiliary Feedwater isolation valves (AF599/AF608), a high temperature alarm due to reverse flow from the Steam Generator will be initiated in the control room. After verification that a steam leak exists, the reactor will be tripped, and Auxiliary Feedwater initiated via manual initiation of SFRCS. After confirmation of which is the affected side, the operators will take actions to i ensure that all feedwater flow has been. terminated and the MSIV closed on the .{ affected steam generator, and will then open the Atmospheric Vent Valve (AVV) l to vent steam to the atmosphere. The resultant steam generator blowdown will .{ reduce, and eventually terminate, critical crack leakage into the annulus, j l I The annulus temperature and pressure response to this alternate mitigation method has been examined and found to be bounded by a break in the Steam Generator Blowdown (SGBD) Line in the annulus. Details of the SGBD line break ) are being incorporated into USAR Section 3.6.2.7.1.15, with the associated analysis detailed in calculation C-NSA-000.02-006. l .1 .i Additionally, electrical penetrations in the vicinity of the AFW flow lines were reviewed to determine if any equipment associated with these penetrations is necessary for mitigation.of the critical crack. No necessary mitigation equipment associated with these electrical penetrations was identified. The equipment necessary to mitigate the consequences of this break are not adversely affected. l , I

1 k ( I { l SAFETY EVALUATION

SUMMARY

l FOR UCN 93-052 (SE 93-0047) I TITLE: ) Revision to USAR Section 12.0 j CHANGE: i Revise the USAR to reflect the requirements of the revised 10CFR Part 20 i Standards for Protection Against Radiation. REASON FOR CHANGE: j i The revised 10CFR20 prescribed new administrative and programmatic controls for radiation protection programs. The USAR is being revised to reflect these i changes. SAFETY EVALUATION

SUMMARY

There are no effects on safety due to these proposed changes. These proposed changes continue to prescribe the required regulatory dose controls, limits and j monitoring methodologies which result in personnel dose being controlled and monitored as required by 10CFR20. Following is a brief discussion concerning the included changes: ] I The expected annual dose rates table has been deleted along with the sentence mentioning the table. Annual doses are monitored in accordance with 10CFR20. j Personnel dose is kept as low as reasonably achievable and forecasted each year i in the dose projection for Davis-Besse. Inclusion of this table in the USAR is redundant. I I The critical organ concept mentioned in 12.2.5.4 was eliminated by Revision 1 to IOCFR20. Access to radiologically restricted areas or exposure to airborne adioactive materials is no longer a consideration for a whole body count. i r This monitoring change occurred with the revision to;10CFR20. Administrative i requirements ensure that whole body counts are taken in accordance with .) regulatory requirements. The remainder of the changes to 12.2.5.4 are due to changes in 10CFK20 and are editorial or clarifying in nature. l Routine entries to containment, at power, have been in use at Davis-Besse_for .j several years. Regulatory requirements are not solely concerned with airborne j radioactivity. Due to the change in requirements.that occurred with Revision 1 to 10CFR20, this requirement in 12.2.6 has been deleted. The remainder of the changes to 12.2.6 are due to changes in 10CFR20 and are editorial in nature. -j i l The change to personnel monitoring in 12.3.3 allows the use of electronic dosimetry at Davis-Besse. This dosimetry combines the functionality of all= l dosimett 7 presently described in the USAR. This change clarifies the expectations for that dosimetry. The remainder of the changes are editorial in nature and reflect changes dealing with Revision 1 to 10CFR20. l. 1

,_~ l l l The duties of the Manager -' Radiation Protection vere clarified and the ~ l - unnecessary detail of the supervisory levels was eliminated. ' Administrative ~ procedures specifically detail those programs and the requirements for.them. l' This level of detail in the USAR is unnecessary. s 1 .i ' ? ' i t l l 1 i ~ l 1 ) 1 L I I ) i i 1 a S' V is +*um e..w.s ee, e r-1.i -w y w --w, .e .e.> w--g em.-

q J ) SAFETY EVALUATION

SUMMARY

i FOR UCN 93-066 (SE 93-0054) TITLE: Changes to High Energy Line Break Analysis for PCAQ 92-0195. i CHANGE: ) J The subject USAR change represents a nearly complete re-write of the High -i Energy Line Break (HELB) analysis 1or the following USAR Sections; Section 3.6.2.7.1.5 (Main Steam to Auxiliary Feedwater Pump Turbine), Section 3.6.2.7.1.6 (Main Feedwater), and Section 3.6.2.7.1.15 (Steam Generator Blowdown). 1 REASON FOR CHANGE: it has been documented in PCAQR 92-0195 that several of the previous HELB analyt is which were performed to determine environmental conditions (i.e. pressure and temperature) following postulated HELBs outs'ide containment used a faulted technique which caused heat transfer rates to be over-predicted. This resulted in an under-prediction of compartment temperatures for postulated HELBs which involved the release of superheated steam. The error was most pronounced in cases where the steam release rate was sufficiently low that heat i transfer to walls and ceilings could be effective in reducing compartment temperature. SAFETY EVALUATION

SUMMARY

Based on the piping stress analysis for the MFW piping, double-ended breaks and critical cracks are postulated only in rooms 303.and 314. Similarly, based on-the piping stress analysis for MS-AFPT piping, double-ended breaks and critical cracks are postulated in rooms 237, 238, 401, 500, 501, 601, and 602. In addition, critical cracks (only) are postulated.in rooms 400 and 404. For Steam Generator Blowdown (SGBD) lines, breaks are postulated in all rooms in the auxiliary building through which the SGBD lines are routed. Mass and energy release rates (blowdown) from breaks and cracks were reviewed. The original blowdown calculations were revised if postulated single failures could result in a more adverse blowdown or the original calculation appeared to be inadequate for other reasons. The blowdown energy in the revised calculations results in longer event duration and/or higher peak temperatures than the previous blowdown. A diagram of the existing overall compartment model which was used for-temperature analysis is contained in a USAR Figure. The compartment and flow path calculations were revised so that cumulative effects would be addressed. Therefore, all relevant auxiliary building compartments and flow paths were recalculated. This has resulted in a revised figure, as well as revised volumes in Table 3.6-6 Summary of Compartment Pressurization. Previous text in the af fected USAR sections provided some information regarding the flow areas j of selected flow paths and failure pressures of selected doors and blowout panels. This information has been deleted from the USAR text. j i i i )

L 1-I E In the reanlayzed cases where a.line break of full' cross sectional area is involved, sufficient information is available to the control room personnel to detect the break and take manual actions from the control room within ten minutes. This is in accordance with previous USAR analysis except for the case of the SGBD line break, where thirty minutes was previously assumed. Ten minute response time is-adequate because the control room personnel will l have noise, fire alarms, and other diverse indications (including sump flow) which will nearly immediately indicate the presence of the break. The computer code used in the compartment reanalysis, PCFLUD is a one-dimensional personal computer based compartment analysis code, which provides for heat transfer into walls, ceilings, (n components. The code is-based on NRC guidance as provided in NUREG 0588 and gives results similar to [ l Bechtel's COPRA code which were used for other USAR analysis. While the L one dimensional limitation of these codes prevents prediction of potential hot j opots within compartments, the lumped volume approach is consistent with.the previous analysis. PCFLUD is expected to give higher lumped volume temperature results than would be obtained using best estimate codes, especially where superheated -team release is involved. USAR Section 3.6.2.'7.1.1 describes the COPRA code. The UCN associated with this evaluation appends a short description of PCFLUD to the description of the COPRA code. l The results of the reanalyses show higher temperatures in compartments where-restricted amounts of superheated steam is released. This was anticipated as a [ result of correction of the non-conservative heat transfer error as noted in i PCAQR 92-0195. The relevant temperatures are being included in revised Table 3.6-11. Results Summary - Peak Temperature in Each Room for Each Break. l As a result of using more conservative qualification temperature profiles. l equipment required to mitigate postulated HELB events should theoretically be less likely to malfunction in an actual harsh environment, improving plant safety. In most cases, pressure results are similar to results of previous analysis. s This is anticipated because both the code (PCFLUD) and the models are substantially similar to the original work which was performed for pressurization studies. In all cases, pressure results are within the structural capacity of the associated compartment as-listed in the existing USAR Table 3.6-6. Table 3.6-6 is modified to' include new results-. The temperature column is being deleted because the temperature values presented in Table 3.6-6 came f rom fonner pressurization studies. These temperatures were not used for equipment qualifications and appeared to conflict with the l temperatures which were presented in Table 3.6-11. The new analyses replace both the previous pressurization and previous temperature analyses.

_~ I t SAFETY EVALUATION

SUMMARY

FOR DB-FF-10119 (SE 92-0034) TITLE: i Disabling Relaying During Full Load Testing of the Station Blackout-Diesel-Generator 'i CHANGE: The full load testing of Station Blackout Diesel Generator 3 (DG3) to determine the magnitude of neutral current will require the instantaneous ground relay (associated with the generator's feeder) to be disabled. Additional 4.16 KV bus relaying (which cannot physically be set any higher) will also require l disabling to ensure that the testing does not cause these relays to suspend the test prematurely. j REASON FOR CHANGE: i Post-modification testing of the DG3, installed under MOD 89-0109, has found that a large magnitude of neutral current is generated when DG3 is operated in j parallel with a second grounded source. Further testing has found that.this current has primarily a 3rd harmonic component which appears to be generated j by the Station Blackout Diesel Generator's less than perfect sinusoidal waveform. The resultant neutral current has caused the ground relaying j (50GS/AD213) associated with the D2 bus'to operate during the testing and trip-the generator. Since the relaying trips the generator prior to reaching its maximum operating limits, the magnitude of neutral current cannot be = quantified. The maximum neutral current needs to be known in order to determine the course of action that Toledo Edison must take in correcting the' operational difficulties associated with running the Station Blackout Diesel Generator in parallel operation. l; SAFETY EVALUATION

SUMMARY

i The disabling of the 50GS/AD213, 51GS-2/D2 and 51GS-1/D2 ground relays and the i Iowering of the 51G/DG3 pickup, performed in order to quantify the neutral current generated by the DG3, will not have any adverse effects on safety. { The relays being disabled are associated with the non-essential 4.16 KV bus D2. During the test, both essential buses C1 and D1 vill be aligned to transformer AC and therefore the essential buses will not be affected by any protective device operation that may occur as a result of excessive neutral current generated by the Station Blackout Diesel Generator. Both Emerge.ncy Diesel Generators will be operable during the duration of the test. i Disabling the three groundfault relays will not eliminate the groundfault-protection of DG3 or of the non-essential 4.16 KV bus D2. Sufficient groundfault protection will remain to protect the distribution equipment from-either prolonged neutral current or from a groundfault. i t b -n

~ 1 i r l t Each feeder from D2, with the exception of DG3's feeder, will retain its j instantaneous groundfault relay (50GS) set at 5 Amps. The neutral overcurrent relays at both DG3 (51G) and Transformer BD (51!1) will j also remain operational and are each set sensitive enough to provide complete j groundfault protection for all 4.16 KV distribution components. Its important to note that during the test, the level of neutral current will i i be monitored so the operators can intervene if a high magnitude of neutral l current persisted. i l The neutral current that is generated by the Station Blackout Generator will l not adversely affect the 13.8 KV bus B. This circulating ground current is limited to the 4.16 KV Bus D2 as a result of the delta-wye transformer connection of Bus-Tie Transformer BD. The delta connected primary (13.8.KV) of i this transformer effectively blocks the flow of circulating ground currents ~ [ because there is no ground connection. The disabling of the relays will not increase the chances f or a malfunction of l either the AC (CT sensing circuits) or DC (trip circuits) control circuits. Realignment of Bus D1 to its alternate source, Transformer AC, will not degrade any distribution components. Supplying two essential and one non-essential 4.16 KV buses is well within the ratings of the 12/16 MVA Bus-Tie transformer. l This configuration is a requirement of the plant's design bases. In the event that a safety features actuation signal is received during this line-up, there i is sufficient capacity to start the required equipment important to safety from i the offsite source and maintain adequate terminal voltage to ensure that the { equipment operates satisfactorily. l 1 i i l

= s SAFETY EVALUATION

SUMMARY

FOR DB-SP-03372 and DB-SP-03377 (SE 93-0020) TITLE: Deletion of Requirement for Replacing Orifices During Makeup System Surveillance Procedures CHANGE: The maheup system surveillance procedures DB-SP-03372 and DB-SP-03377, Quarterly Makeup Pump 1 (2) and Valves Inservice Test and Inspection (RCS Pressure Less Than 150 PSIG). require that rest ricting orifice R0-5989 be-replaced by RO-HP1 prior to performing the test. Revisions are proposed to remove this requirement from the procedures. REASON FOR CHANGE: During shutdown plant conditions the makeup pumps are not permitted to pump into the RCS, primarily to prevent overpressurization of the RCS. Consequently, the recirculation line normally used for HPI testing is used to flow test the MU pumps. However, due to the differences between the HP1 pumps and the MU pumps, RO-5989 is replaced by RO-HPl. Orifice RO-HP1 places a smaller orifice in the line so that the MU pumps cannot achieve runout ~ conditions during the test. Orifice RO-5989 is reinstalled prior to the next HPI pump test. The change out of the restricting orifice creates unnecessary burdens on plant personnel and causes radiation exposure. Therefore, the requirement to change the HPI recirculation line orifice is to be deleted. SAFETY EVALUATION

SUMMARY

Deleting the requirement to replace RO-5989 with RO-HP1 to support MU pump testing when the RCS is depressurized to less than 150 PSIG will have no effect on plant safety. This is because neither the MU6P system nor the HPI system has any' primary safety function during the aforementioned plant condition. While both systems can be used to supply barated water to the RCS, other sources also exist. It is noted that while the pump is in the~ test lineup, it is incapable of supplying water to the RCS, so that a boron dilution accident is not credible. Therefore, the flowrate during testing, as affected by the orifice has no affect on plant safety. It might be argued that there is an increased potential for running out and possibly damaging a MU pump due to this change. This does not affect plant safety, since there would still be multiple sources of borated water available-to the RCS in the event damage did occur. Because the test is conducted with~ manual control, there is complete operator attention focused on the system during the test. The heightened operator awareness helps ensure.that damage to the MU pumps will not occur. The plant technical specifications ensure that the plant will not enter a condition requiring both MU pumps to be operable prior to any damage being addressed.

l t i = In'the unlikely event that both the HP1 pump 2 and a MU pump are both discharging through the recirculation line, there 'is no potential for the HP1 i pump to lose its recirculation flow with R0-5989 installed. This is because 'l both pumps are centrifugal pumps. They will share providing the flow through J; the recirc line, which ensures neither pump will be operated at shut off head. t Therefore, it is concluded that not installing RO HP1 during makeup pump surveillance has no adverse effect on plant safety. t 5 } i i i I i i 1 t i 'I l- 'I i 1 i i a 2 a k 4 .... -.}}