ML20135E678

10CFR50.59 Summary Rept of Facility Changes,Tests & Experiments for Davis-Besse Nuclear Power Station,Unit 1, 941116-960602

ML20135E678
Person / Time
Site:	Davis Besse
Issue date:	06/02/1996
From:	CENTERIOR ENERGY
To:
Shared Package
ML20135E673	List: ML20135E671 ML20135E678
References
NUDOCS 9612110417
Download: ML20135E678 (170)

Text

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Dockst Numb:r 50-346 License Number NPF-3 Serial Number 2419

) Attachment 1 f

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10 CFR 50.59 Summary Report

of Facility Changes, Tests, and Experiments '

i for Davis-Besse Nuclear Power Station, Unit No.1 November 16,1994 - June 2,1996

(169 pages follow)

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a 9612110417 961202 PDR ADOCK 05000346 R PDR-

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i 10 CFR 50.59 Summary Listing initiating Document Safgfy Evaluation Iltig DCR 94-0077 95-0037 Revision of Plant Drawings DCR 94-0128 95-0052 Miscellaneous Drawing Changes DCR 95-0005 95-0022 Revision of Valve CD430 Position DCR 95-0022 95-0014 Resolution of Configuration Concerns on 125/250 V DC and 120 V AC System DCR 95-0044 95-0059 Revision of Q and EQ Status of MC56-1, Containment Recirculation Fan Motor DCR 95-0051 96-0001 Drawing Revisions to Reposition Valve SV4658 DCR 95-0059 96-0013 Revision of Makeup and Purification System Drawings DCR 96-0003 96-0030 Correction of Valve Position for AS156 DCR 96-0008 96-0027 Ex-Core Neutron Flux Monitoring Ch.2 Local Indicator Abandonment DCR 96-0011 96-0010 Revise DCC MCC Single-Line Diagram DCR 96-0014 96-0044 Revision of Plant Drawings DCR 96-0029 96-0053 Waste Gas Outlet isolation to Station Vent Control Valve, WG130

initiating Document Safety EvaluAllQD 1111R FCR 84-0051 96-0019 Utilization of the Low Level Radwaste Storage Facility to include Snubber Testing FCR 86-0272 Sup3 89-0116 R04 Replacement of Cyberex Inverters FPR 94-0190 95-0008 Replace Reactor Coolant Pump Monitoring Circuit Test Switches FPR 94-0264 95-0048 Change Normal Position for Valve WT105 FPR 94-0348 95-0051 Removal of Circulating Water Pump Transfer Switches FPR 94-0482 95-0004 Area Radiation Selector Panel C5781 Abandonment FPR 94-0524 95-0031 Add Hose Connection to Chemical Addition Tanks T49 and TSO FPR 94-0638 95-0039 Installation of Air Regulators on the Condensate Drain Pum,,s FPR 94-0716 96-0056 Fire Water Storage Tank (FWST) Repairs FPR 94-0744 95-0020 Abandon-in-Place Lube Oil Storage Room Sump l Pumps FPR 94-1229 95-0029 Wire Label Changes for 4KV Breaker AD113 l

FPR 94-1229 95-0030 Wire Label Changes for 4 KV Breaker AC107 FPR 95-0092 95-0007 Permanent Installation of Smoke Detectors in Room 114

l Initiating Document Safety Evaluation 1111e FPR 95-0130 96-0006 Abandonment of the Auxiliary Building Non-essential Humidifier i

FPR 95-0364 96-0008 Post Accident Sampling System Filter Removal 4

FPR 95-0433 96-0007 Replacement of Batteries and Chargers for

. Emergency Battery Lights FPR 95-5124, FPR 95-0071 Replacement of the 2 Relay in 4kv Switchgears )

95-5126 AC101 and AD101 FPR 96-0189 96-0048 Replacement of the Lunchroom A/C Unit l

Mod 91-0020 93-0004 R03 Pulling an OTSG Tube Segment, Sleeving an OTSG Tube and/or Pulling an OTSG Tube Mod 91-0029 95-0036 R01 Dry Fuel Storage Facility Mod 92-0013 Sup0 92-0069 R01 Replacement of Emergency Diesel Generator Air Start Check Valves Mod 92-0074 94-0084 Fix Power Supply Wiring for EDG Alarm Panels Mod 93-0016 94-0045 Resolution of Pressure Locking Concerns With Valves DH11 and DH12 Mod 93-0028 95-0035 Service Water and Component Cooling Water 4KV Breaker Interlocks Mod 93-0035 94-0014 Motor Operated Valve Modifications Mod 93-0041 95-0027 Modify ECCS Sump Pump Motors Starter Circuit

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Initiating Document Safety Evaluation Iltle l Mod 93-0051 95-0019 Removal of Unnecessary Loads in SFAS and SFRCS Mod 93-0060 94-0054 Repacement of Operators in Valves Mod 93-0070 95-0054 Addition of Six Baskets for Storage of Trisodium l Phosphate l

Mod 94-0009 95-0056 Remove Relief Valves SW10210, SW10211, and SW10212 i

Mod 94-0020 95-0042 Pressure Locking Concems for Valve RC11 1

Mod 94-0029 95-0046 Pressure Regulator PCV 1776 Replacement 6

Mod 94-0032 95-0070 R01 Containment Vessel Base Embedment Sand '

- Removal i

Mod 95-0001 95-0060 10RFO Fuel Repairs Mod 95-0011 95-0041 Resetting the Second Main Feedwater (MFW) i Pump Turbine in Mode 1 While Maintaining ARTS Trip Capability for Loss of MFW Pumps Mod 95-0012 96-0049 Removal of Unnecessary Wiring in C3615 and l C3616 Mod 95-0014 95-0066 Alternate Non-1E 480VAC Supplies for Containment Lighting Mod 95-0055 96-0004 Convert Room 302 to a Hot Shop j

Initiating Document Safety Evaluation Iltle Mod 95-0060 96-0022 RO1 Auxiliary Feedwater Pump Turbine Main Steam Heat Recovery Line 4

Mod 95-0062 96-0012 Rescale Steam Generator Startup Range Level 3

Strings Mod 96-0002 96-0018 Larger Motor Installation for Valve MU2B Mod 96-0003 96-0021 Replace Motor Operators for Valves FW601 and FW612

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Mod 96-0006 96-0026 Installation of a Vent and an Elbow on the High Pressure injection System i

MWO 1-96 0315-00 96-0037 Installing a Camera in the Annulus 4

MWO 7-91-0253- 95-0049 Temporary Disabling of all Service Water (SW) 14,-15,-16 Strainer Blowdown Valves PAT 94-0793 95-0034 Main Turbine Stop Valve Testing Above 85% Power PAT 95-1359 95-0057 Main Turbine Stop Valve Testing at 93 Percent Power Proc. DB-SS-04150 96-0052 Main Turbine Stop Valve Testing at 93% Power and Main Turbine Control Valve Testing at 96%

Power SCR 95-5004 95-0069 Raising the SFRCS High Level Trip Setpoint to 250 inches Startup Range SE 94-0071 R02 94-0071 R02 The Cycle 10 Reload Report, Core Operating Limits Report and End of Core (EOC) Tave Reduction

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l Initiating Document Safety Evaluation lit [g SE 95-0044 95-0044 Spent Fuel Cask Crane Functional Check Using Water Bags For Load SE 95-0064 95-0064 Cable / Hose Routing Alteration to Penetration P59

SE 96-0040 96-0040 The Cycle 11 Reload Rport and Core Operating Limits Report i

TM 94-0017 95-0021 Revise Make-up Water Treatment Chemical Supply Piping )

l TM 95-0011, -12, - 96-0009 Temporary Power Feeds I 14,-15,-16,-17 TM 96-0003 96-0005 Remove of the Spool Piece at FE11109 l TM 96-0012 96-0029 Bypassing SF1, Fuel Transfer Tube 1-2 Isolation Valve, Open Interlock I 96-0036 Installation of Video Camera for RCP 1-1 Pump TM 96-0016 Leak Monitoring for Cycle 11 TM 96-0019 96-0045 RO1 Fire Water Storage Tank Repairs UCN 94-147 95-0025 Channel Functional Testing UCN 95-010 95-0005 Main Steam Safety Valvo Seismic Qualification UCN 95-042 95-0026 Delete USAR Table 9.3-2, Station and Instrument Air Control Room Alarm Setpoint UCN 95-052 96-0077 Reactor Protection System Logic Drawing Update

l Initiating Document Safety Evaluation Iltle ,

UCN 95-066 95-0040 Emergency Diesel Generator Frequency Relays l

i UCN 95-068 95-0047 Revision of FHAR Description for Containment Air Coolers and Atmospheric Vent Valves l

UCN 95-072 95-0043 Combining Design and Plant Engineering Functions UCN 95-081 95-0050 Radiation Monitor Sample Flow Assembly Replacement UCN 95-096 95-0058 Non-rated Fire Barrier UCN 95-098 95-0061 Changing the Malfunction Analysis for the Chemical Addition System Described in USAR Section 9.3.6.3.2 l

UCN 95-105 95-0062 Elimination of Hose Houses Outside of the ,

Protected Area Fence l UCN 95-111 96-0011 USAR Steam Generator Sampling Update UCN 95-113 95-0068 Removal of USAR Figure 9.2-3 i

UCN 95-114 95-0065 USAR Description of CF7A and CF78 UCN 95-130 96-0002 Seismic Qualification of Replacement Components UCN 96-003 96-0014 Fire Protection Procedures Fequency Reduction Project UCN 96-004 96-0016 Revision to Vice President - Nuclear's Review of i Nuclear Quality Assurance Program Effectiveness l

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Inlilating Document Safety Evaluation Illig UCN 96-012 96-0017 Reorganizing Budgeting, Cost Control, Long Range

! Planning and Nuclear Projects Responsibilities UCN 96-016 96-0069 Changes to the Fire Hazard Analysis Report UCN 96-017 96-0033 Transfer of Quality Assurance Responsibilities l

UCN 96-019 96-0023 Clarification of Decay Heat Removal System Use i for Spent Fuel Pool Cooling l UCN 96-020 96-0024 Fire Detection in Main Steam Rooms 1 l

i UCN 96-022 96-0025 Removal of Plant Manager's Gai-Tronics Station and Deletion of USAR Figure 9.5-7 l

UCN 96-024 96-0020 Procedure Change to Pump Concentrated Boric Acid from the BAATs to the CWRT UCN 96-029 96-0031 Adding DB Supply Group References to and Eliminating Corporate Group Descriptions from USAR 17.2 UCN 96-030 96-0028 Positioning ECCS Room Cooler Outlet Valves Differently than Depicted on Figure 9.2-1 UCN 96-035 96-0035 Transfer of Quality Assurance Responsibilities UCN 96-038 96-0039 Aligning Breaker BF1194 Normally Open (Removing Power from Valve HP-31)

UCN 96-039 96-0038 FHAR Revision for Information Notice 92-18 Resolution

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jn[1[atina Document Safety Evaluation Iltle UCN 96-041 96-0047 Correction of USAR Figure 6.2-33, Containment Vessel Emergency Pump l

l UCN 96-046 96-0043 Reorganizing Engineering Responsibilities l l l UCN 96-052 96-0050 Evaluation of the Carroll Township Water Tower on the Operation of Davis-Besse UCN 96-059,96- 96-0062 Periodic USAR Review Discrepancy Corrections  !

060,96-061, 96-096 I l

1 UCN 96-062 96-0058 10RFO Periodic Review for USAR Sections l 8.3.1.2.14 and 8.3.1.2.20 UCN 96-072 96-0057 Revision of USAR Section 13.7 - Security UCN 96-076 96-0071 Component Cooling Water System UCN 96-077 96-0067 Updates to USAR Section 7.4, Systems Required for Safe Shutdown j UCN 96-078 96-0073 USAR Primary Chemical Addition Update UCN 96-079 96-0076 Combined Oil Drain Lines to the Reactor Coolant I Pump Oil Drain Collection and Correct the USAR I Description for the Drain Tank Configuration UCN 96-086 96-0059 Removal of Voltage Regulators from USAR Figure 7.4-1, Control Rod Drive Controls UCN 96-088 96-0061 Clarify USAR Sections 8.3.2.1.5 and 8.3.2.1.6 UCN 96-091 96-0060 Main Turbine Missile Generation Probability

l Inl11ating Document Safety Evaluation Iltig UCN 96-092 96-0074 USAR Sampling System Update i UCN 96-094 96-0088 Decay Heat RemovalInterlock UCN 96-095 96-0075 Cask Pit Slot Width UCN 96-099 96-0092 Reactor Pressure Vessel (RPV) O-Ring Hydrotesting i

UCN 96-100 96-0091 Update, Correct, and Reorganize USAR Chapters  :

8 & 9 for the Emergency Diesel Generators and l Station Blackout Diesel Generator  !

UCN 96-101 96-0070 Revise USAR Section 9.5.3, Lighting Systems, to Reflect the Correct Plant Configuration.

l UCN 96-103 96-0065 Service Water (SW) Flow Bypass Through a Spare 1 CCW Heat Exchanger Dunng Cold Weather UCN 96-112 96-0064 Changes to USAR Section 12.3.2.2.2, Counting Equipment for Radioactivity Measurement UCN 96-117 96-0072 Spent fuel Cooling and Cleanup System UCN 96-121,96- 96-0082 Revision of USAR Sections 9.3.6,11.2, and 11.3 122,96-123 UCN 96-125 96-0084 Condenser Tests and inspections UCN 96-128 96-0094 Revision of USAR Sections 9.3.3 and 9.3.3.1, Station Drains UCN 96-129 96-0085 To Correct Ventitation Dampers CV5024 and CV5025 Position

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Initiating Document Safety Evaluation Iltig UCN 96-134 96-0079 Condensate Demineralizer System Description ano Instrumentation l

UCN 96-145 96-0083 Station Computer System Sequence of Events Resolution Time UCN 96-169 96-0078 Penetrations Below Elevation 575 Feet UCN 96-170 96-0086 Revise USAR Figure 9.4-7 to Delete Station Heating Valve SH-385 UCN 96-172 96-0081 Use of Groundwater in the Local Area Around the Davis-Besse Nuclear Power Station (DBNPS) l l

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l Abbreviations Used:

DCR = Design Change Request FCR = Facility Change Request FPR = Field Problem Report MOD = Modification MWO = Maintenance Work Order PAT = Procedure Activity Tracking Number SCR = Setpoint Change Request SE = Safety Evaluation TM = Temporary Modification UCN = USAR Change Notice ]

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NOTE The following are summaries of the Safety Evaluations performed pursuant to 10 CFR 50.59 for the Davis-Besse Nuclear Power Station, Unit No. I from November 16,1994 through June 2, 1996. For a complete understanding of the safety evaluation, the reader must review the actual Safety Evaluation.

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SAFETY EVALUATION

SUMMARY

J FOR DCR 94-0077 (SE 95-0037, R.01)

TITLE:

4 l Revision of Plant Drawings CHANGE:

DCR 94-0077 revised Emergency Diesel Generator plant drawings to reflect the actual design of the generators.

i i REASON FOR CHANGE:

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! PCAQRs 94-0600 and 94-0712 identified discrepancies in plant drawings E-64B sh.1F, E-64B sh. 1G, M-180-4, M-180Q-13, and M-180Q-14, which are Emergency i Diesel Generator drawings incorporated by reference into the USAR.

) SAFETY EVALUATION

SUMMARY

The above-mentioned drawings were' revised to revise / add wire, contact, and relay j numbers, revise / add equipment designations, and revise references for location

> of contacts. The drawings were also revised to correct an incorrect reference to DS-1B detail, correct a CADD error involving a normally closed relay contact

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shown normally open, correct a spelling error, and remove a reference to a note.

The drawings were also revised to correct the incorporation of an incorrect DCN and a DCN which was incorrectly incorporated. The revisions to the I above-mentioned drawings were verified by inspection of drawings E-30B sh.20, j M-180-4, M-1800-13 through M-1800-18. Vendor drawings M-180-2 and M-180-3 were 4 superseded by M-1800-13 and M-180Q-14, respectively, in FCRs 81-0058 and 81-0067. Drawings M-180-2 and M-180-3 were listed in USAR section 1.5.3.10 as drawings incorporated by reference. All the above revisions to the affected

' dras4.ngs are either editorial (e.g. correct typographical errors) or administative (e.g. add wire designations, correct relay designations, etc.) and i 4

do not change the function of any piece of equipment. They are acceptable based j

i. on comparisons to other approved drawings and the intended operation of the j

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equipment as currently described in the USAR.

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SAFETY EVALUATION

SUMMARY

FOR DCR 94-0128 (SE 95-0052)

TITLE:

Miscellaneous Drawing Changes l l

CHANGE:

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DCR 94-0128 revised electrical drawings E-1042 and E-1043 (EDG load tabulation drawings) and E-31B. '

i REASON FOR CHANGE: l l

PCAQR 95-0117 identified drawing discrepencies on E-1042 and E-1043, concerning I the EDG loading of the Control Room Emergency Ventilation System (CREVS)

Condensing Units 1 and 2, MS3311 and MS3321, respectively. Drawing E-31B was revised to show the correct wiring configurations for the 52a contacts in Fil and F13.

SAFETY EVALUATION

SUMMARY

Components MS3311 and MS3321 are 10 HP motors which were being shown as STEP I l loads on drawings E-1042 and E-1043, but per drawing E-60B sheet 1, and vendor l drawing M-410-281 these loads are manually loaded. Per vendor drawing M-410-281, MS3311 (MS3321) will start only when the associated CREVS Fan is

, already running. Per drawing E-60B sheet 1, the CREVS Fans are manually loaded onto the EDGs. Therefore, drawings E-1042 and E-1043 (USAR Table 8.3-1) will be revised to show that MS3311 and MS3321 are manually loaded as well.

l Drawing E-31B sheets 4 and 9 were revised to show the correct wiring l configurations for the 52a contacts in Fil and F13. The wiring configurations

' l had depicted a single 52a contact for each 345 KV breaker. The correct '

configuration per drawings E-512 sheets 2 and 10, E-112, E-136 sheet 1, E-114, l and E-160 sheets 3 and 7 is three 52a contacts per breaker, one 52a contact per phase.

In addition to the drawing revisions noted above, similiar changes were made by DCR 94-0128 to drawings which are not referenced by the USAR. None of these drawing revisions have any effect on system operation. They are acceptable based on comparisons to other approved drawings and the intended operation of l the equipment as currently described in the USAR. The changes involved in DCR 94-0128 are safe.

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SAFETY EVALUATION

SUMMARY

FOR j DCR 95-0005 (SE 95-0022) l j

TITLE: l Revision of CD430 Position CHANGE:

This DCR changes the position shown for valve CD430, Turbine Bypass Desuperheating valve, on Operational Schematic OS 10, Sheet 2, and USAR Figure 10.4-11 from open to closed. l REASON FOR CHANGE:

This DCR revised the normal position for this valve on the Operational Schematic drawing and the USAR figure so that the configuration of the valve matches that in the field and in the procedures.

SAFETY EVALUATION

SUMMARY

The DCR changed the position for this valve on the USAR figure and the ,

Operational Schematic to the normally maintained position of valve. The j affected component's control logic is unaffected, and the valve's operation is j unchanged. Based on the above the action of this DCR is safe. l 1 I l

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! SAFETY EVALUATION

SUMMARY

I FOR DCR 95-0022 (SE 95-0014) l TITLE:

Resolution of Configuration Concerns on 125/250 V DC and 120 V AC System j

CHANGE:

Revised drawing E-7 and E-6 sheets 3 and 4 to correct configuration control

descrepancica.

4 REASON FOR CHANGE: j l

Walkdowns of the 125/250 V DC and 120 V Instrumentation AC System identified a drawing deficiency. The deficiency was that the neutral position of the I transfer-switch on the swing battery chargers DBClPN and DBC2PN was not shown on Drawings E-7 and E-6 sheets 3 and 4.

4 SAFETY EVALUATION

SUMMARY

$ Revision of Drawings E-7 and E-6 sheets 3 and 4 to include the neutral position on the transfer switch on the swing chargers has no effect on safety. The j swing charger will continue to operate as before, this is merely a correction

of the drawings to bring them into conformance with the field and the charger

as depicted on the Operational Schematic drawing. The neutral position is i already included in the applicable procedures.

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SAFETY EVALUATION

SUMMARY

FOR DCR 95-0044 (SE 95-0059)

TITLE:

Revision of Q and EQ Status of MC56-1, Contalwment Recirculation Fan Motor i

CHANGE:

DCR 95-0044 changed the Q boundary of the power supply for MC56-1, Containment Recirculation Fan Motor and reclassified MC56-1 as 'Q' and 'EQ' for circuit integrity.

REASON FOR CHANGE:

1 Containment Recirculation Fan Motor MC56-1 is a non-essential load which receives its power from essential Motor Control Center (MCC) EllB via breaker BE1169. This breaker's instantaneous unit can not be coordinated with the next upstream molded case breaker BE1120, which is the feeder breaker for MCC EllB.

As a result, failure of this load resulting in a high overload or multiphase fault may trip not only the load breaker (BE1169), but also the feeder breaker, BE1120. The loss of BE1120 will deenergize the essential MCC EllB and result in the loss of all loads powered from that MCC. MCC E11B provides power for d

several SFAS actuated components.

SAFETY EVALUATION

SUMMARY

MC56-1, Containment Recirculation Fan Motor, performs no active essential functions and is presently classified as 'non-Q'. It is a non-essential load which receives its power from essential MCC E11B via breaker BE1169. Thc next upstream molded case breaker, BE1120, is the feeder breaker for MCC E11B.

Since the breakers cannot be coordinated, and since during an accident analysis, any non-essential equipment is assumed to fail and can not be considered as the single failure, no credit can be taken for the various SFAS actuated containment isolation valves powered from MCC E11B. Revising the classification of MC56-1 to 'Q' and 'EQ' will eliminate a postulated failure mechanism for MCC E11B, 480 Volt Motor Control Center for Essential Equipment, and ensure its availability in the control or mitigation of accidents.

Preventive Maintenance Work Order 3-96-1328-01 is scheduled to replace MC56-1.

The reliability of MC56-1 and MCC E11B is not reduced. Reclassification of MC56-1 as 'Q' and 'EQ' will eliminate a postulated failure mechanism for MCC E11B and ensure its availability. There is no effect on plant security. There will be no increase in adverse effects from any hazards. No Radwaste Systems are affected by DCR 95-0044. This DCR will ensure the plant design is maintained as described in the basis of Technical Specification 3/4.8. 1 I

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SUMMARY

FOR DCR 95-0051 (SE 96-0001)

TITLE:

j Drawing Revisions to Reposition Valve SV4658 CHANGE:

DCR 95-0051 made changes to the schematic drawings (including USAR Figure 9.3-3a) to bring them into agreement with the operating procedure normal line-up.

REASON FOR CHANGE: 1 PCAOR 95-0885 identified SV4658, a solenoid operated 3-way valve in the Post Accident Sampling System (PASS) as being out of position. Valve SV4658 diverts PASS discharge flow to either the Reactor Coolant Drain Tank (RCDT) or to the Pressurizer Quench Tank. The RCDT is located in the Auxiliary Building. The ,

Quench Tank is located in the Containment Building. The revised line-up will '

show the system in its normal isolated, non-sampling line-up.  ;

1 SAFETY EVALUATION

SUMMARY

l Changing the line-up depicted in USAR Figure 9.3-3a from normal flow to the Quench Tank to showing normal flow to the RCDT will not have an adverse impact j on plant safety. Normal flow to the RCDT is acceptable. The RCDT is designed i to receive sample drains from the Reactor Coolant System. During an accident l SV4658 is aligned to the Pressurizer Quench Tank. This procedural step assures that the effluent is directed to the Quench Tank and not to the RCDT during accident conditions. Based on the above facts the drawing changes are safe.

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SAFETY EVALUATION

SUMMARY

FOR DCR 95-0059 (SE 96-0013)

TITLE:

Revision of Makeup and Purification System Drawings CHANGE:

DCR 95-0059 revised documents to ensure that valve positions in the Makeup and Purification System are consistent between drawings, procedures and the USAR.

REASON FOR CHANGE: l

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l PCAQR 95-0818 documented discrepancies between various documents pertaining to l the Makeup and Purification System.

SAFETY EVALUATION

SUMMARY

Opening the two outlet valves (MU223 and MU229) of Seal Injection Filter 1-1 has no impact on the filter since the inlet valves are already closed and there is no flow across the filter.

Closing the two inlet bypass valves (MU224 and MU225) of Seal Injection Filter 1-1 and Seal Injection Filter 1-2 does not adversely impact the filter since the valves are used to place the standby filter in service.

The P&ID and Operational Schematic valve position changes will not adversely impact the operation of the Makeup and Purification System. The change to USAR Figure 9.3-16 to depict correct valve position and configuration changes will not adversely impact the USAR Chapter 15 - Accident Analysis (Chapter 15.2.4 -

Makeup and Purification Malfunction) or other USAR analysis. The changes will  !

not adversely impact the malfunction analysis of the Makeup and Purification  ;

system, Table 9.3-9.

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SAFETY EVALUATION

SUMMARY

FOR DCR 96-0003 (SE 96-0030)

TITLE:

Correction of Valve Position for AS156 .

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CHANGE: j DCR 96-0003 revised documents to ensure that Auxiliary Steam valve, AS156, l position is consistent between drawings, procedures and the USAR. The documents were revised to indicate the valve normally closed.

REASON FOR CHANGE:

AS156, Caustic Dilution Water Heat Exchanger 1-1 Steam control Inlet Isolation valve, is normally only opened twice a month during demineralizer regeneration. 1 The valve is placed in a closed position after regeneration. ]

l SAFETY EVALUATION

SUMMARY

i This DCR documents changes to indicate the valve position change as closed will not adversely impact the operation of the Auxiliary Steam System. The proposed {

change to a USAR figure to depict the normal valve position will not adversely impact the USAR Chapter 15 - Accident Analysis or other USAR analysis.

It should be noted that the position of the valve described above is controlled by plant procedures.

Based on the above discussion it is concluded that the activity is safe.

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SAFETY EVALUATION

SUMMARY

FOR DCR 96-0008 (SE 96-0027) j f

TITLE: f I

Ex-Core Neutron Flux Monitoring Ch. 2 Local Indicator Abandonment CHANGE:

This DCR removed reference to the Gamma-Metrics ex-core neutron flux monitoring channel 2 local indicators from the Fire Hazard Analysis Report (FHAR). ,

REASON FOR CHANGE:

l The purpose of the DCR was to make permanent the change made by Temporary Modification (TM) 94-0013. This TM disabled the Gamma-Metrics channel 2 +250v power supply which feeds the local indication and the " power on" light on the signal processor (NY5875C), thereby abandoning the local indication in place.

This equipment is located in the No.2 Electrical Penetration Room #427. The

+250v power supply was disabled to eliminate the primary source of heat in the

. signal processor drawer to reduce the high rate of failure that has been i experienced with the signal processor internal components.

SAFETY EVALUATION

SUMMARY

The ability of the Gamma-Metrics system to perform its required

! capabilities is not impacted. This change, by disabling the power supply designated for the local indication, affects only the local indication of channel 2 and the " power on" light on the channel 2 signal processor, which in turn is also abandoned in place. Channel 1 local indication is maintained as well as all control room visual indication and all audible indication.

Therefore, this change has no negative effect on safety.

! By removing this +250v power supply, this change is expected to increase l reliability of the Gamma-Metrics system by reducing heat related failures of l channel 2 signal processor internal components. Based on this evaluation it is

! concluded that the change is safe.

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SAFETY EVALUATION

SUMMARY

FOR DCR 96-0011 (SE 96-0010) j l

s TITLE:

Revise DC MCC Single-Line Diagram I CHANGE:

1 DCR 96-0011 revised drawing E-6, Sheet 4, the single line diagram for essential l

DC Motor Control Center 2 (DC MCC2), to reclassify the +125VDC and -125VDC i fuses, downstream of the 500 ampere fuses, to non-1E. The 500 ampere fuses '

4 will remain Class IE. This made DC MCC2 consistent with DC MCC1. I j

REASON FOR CHANGE:

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) The change corrected the quality class of circuits downstream of the 500 ampere

fuses so that they are consistent with the design requirements.

SAFETY EVALUATION

SUMMARY

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' Review of the circuits on DC MCC2 shows that the circuits connected to the 500 ampere fuses (circuits D211 through D218) are not safety related. While l important to safety, Reactor Coolant Make-up Pump 2 Aux Oil Pump, circuit 2

D217, is not nuclear-safety related. In Memorandum NEN-88-10242, from  ;

Engineering to Licensing, it is stated that "It is unnecessary to provide j redundant support systems for each redundant Makeup train, therefore there is l no need to qualify the DC "startup" lube oil pump motors. In Report No. l 50-346/88023(DRS), the NRC stated "Since the failure of the main AC pump motor j would constitute a single active failure, the inspector agreed that qualification of the DC pump motor was not required." Therefore, reclassifying the D217 circuit is acceptable.

i The DCR did not change the fuse classification but redefined the IE/non-1E boundary. The 500 ampere fuses will remain IE and all the load fuses will be non-1E.

, IEEE Standard 384, paragraph 7.2.2.1 requires that the protective devices at J

1E/non-1E boundaries coordinate with upstream protective devices, so that the

, non-1E portion cannot degrade the IE portion. The installed fuses will ensure that " shorts in the non-1E side shall not degrade the 1E side." Fuses at these IEEE class 1E/non-1E boundaries will be batch tested to ensure compliance with IEEE Standard 384. l

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SAFETY EVALUATION

SUMMARY

FOR DCR 96-0014 (SE 96-0044)

TITLE:

Revision of Plant Drawings CHANGE: j This DCR revises drawings M-090 and M-172 to reflect the fact that the floor drains in Rooms 301 and 302 are covered.

REASON FOR CHANGE:

This DCR updates drawings to reflect the plant configuration and updates drawings to make them consistent.

SAFETY EVALUATION

SUMMARY

Covering of the subject floor drains was part of the original plant design. At the time it was decided to cover the drains, drawing C-211 was revised but drawings M-090 and M-172 were not. Flood analyses have been performed which do not include these floor drains and demonstrate that the Auxiliary Building is adequately protected from flooding.

Since the flooding analyses are not affected by this change, the DCR is merely a drawing update to reflect the plant configuration and make drawings M-090 and M-172 consistent with drawing C-211.

It is therefore concluded that the functions important to safety identified above are not adversely affected by this DCR. It is also concluded that the proposed action is safe.

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i I SAFETY EVALUATION

SUMMARY

I FOR

' DCR 96-0029 (SE 96-0053)

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TITLE:

I i Waste Gas Outlet Isolation to Station Vent Control Valve, WG130 i l 1

CHANGE:

Plant drawings and USAR Figure 11.3-1 were changed to reflect the the normally -

1 open position of WG130, Waste Gas outlet isolation to Station Vent control i valve.

i REASON FOR CHANGE:

? This DCR revises documents to ensure that Waste Gas Outlet Isolation to Station Vent Control Valve, WG130, position is consistent between drawings, procedures and the USAR.

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, SAFETY EVALUATION

SUMMARY

The valve position change will not adversely impact the operation of the Gaseous 4

l' Waste System since there is no minimum or maximum flow requirement when venting the Waste Gas Absolute Filter to the station vent. The proposed change to USAR

. Figure 11.3-1 to depict the normal valve position will not adversely impact the

} USAR Chapter 15 - Accident Analysis or other USAR analysis.

It is concluded that the proposed activity is safe and it does not constitute an i unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR FCR 84-0051 (SE 96-0019)

TITLE:

I Utilization of the Low Level Radwaste Storage Facility to Include Snubber Testing CHANGE:

The following changes were evaluated by this safety evaluation: relocating the .

snubber test equipment to the Low Level Radwaste Storage Facility (LLRWSF);  !

allowing the testing, disassembly / rebuild and storage of pipe or equipment I snubbers to be performed in the building; allowing the decontamination of tools and or minor equipment in the Trash Sorting Area; permit the opening of storage '

containers in the cell area; and, permit the transport of used process filters l to, and deposit in, storage / shipping container (s) located in a storage cell.

REASON FOR CHANGE: l Due to space allocation in the plant it was desirable to relocate the snubbe- i test machine to the LLRWSF. This allows direct access for slightly contaminated snubbers and provides more control over any designated Radiologically Restricted Area (RRA) setup for the test machine (currently the snubbers must be deconned and transported outside the RRA for testing / rebuild) l Also due to space allocation in the plant, it was desirable to relocate the temporary storage of used (radioactive) process filters from the Auxiliary i Building Train Bay to permanent storage in the LLRWSF cell area.

SAFETY EVALUATION

SUMMARY

]

The location for the snubber test machine is identified as the " sorting room" located off the corridor between the Auxiliary Building and the main storage areaofthegLRWSF. The largest snubber test machine has a floor loading of 3333 lbs./fg . However, the designed floor loading for the sorting room is only 250 lbs./ft (live load). A permanent base or rete pad will be installed to 2

distribute this load such that the 250 lbs./ft criteria is met.

The small snubber test machine has an a floor loading of 533 lbs./ft . The machine will be set on the base floor slab which is bearing on compacted fill material. With a slab thickness of 12 inches and the continuous bearing area under the slab, the distribution of the weight will be adequate such that there will be no adverse consequences to the floor or, in general, the LLRWSF. Also, due to the design of the machines, there are no significant operational loads when the snubber testing is being performed and therefore there will be no operational loading applied to the building structure.

[

Based on this, the building design is not structurally affected by the installation of the two machines.

The actual work activities (testing and rebuild of snubbers) will present radiological concerns for the LLRWSF. The work area will be designated as a restricted area to address any possible radiological concerns. The snubbers to be worked could be contaminated, typically by fixed surface contamination on the various snubber parts. The rebuild process does not involve any drilling, grinding or machining activities but will merely' involve parts replacement.

However, if airborne contamination would become a problem, the sorting room is included in the ventilation area of the LLRWSF, which includes HEPA filtration equipment.

Moving the spent filter storage / shipping container from the Auxiliary Building Truck Bay location to one of the cells in the LLRWSF will lower the overall station dose commitment by removing this source of dose from traveled areas.

The maximum activity being transported at any one time is assumed to be a Primary Letdown Filter which has been allowed to accumulate activity to the limit for shipment. An accident involving dropping the filter from the Transport cask would not result in consequences outside of the LLRWSF as the filter is dry, and_contains no gaseous or liquid radioactivity.

From the review, summarized above, it is concluded that relocating the snubber test machine and performing the work activities in the LLRWSF will have no significant adverse effects on any SSC's important to plant safety nor introduce any significant radiological hazards.

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i SAFETY EVALUATION

SUMMARY

4 FOR FCR 86-0272 sup 3, UCN 95-083,.UCN 95-108, UCN 95-109 (SE 89-0116 R4) l l

l TITLE:

j Replacement of Cyberex Inverters CHANGE:

Replace existing Cyberex System with the Solidstate Controls Inc. (SCI) inverters.-

REASON FOR CHANGE:

- The Cyberex inverters have had fault clearing problems. Faults on the branch circuits have caused the inverter DC input fuse to clear the fault before the branch fuse. They are also being replaced due to lack of vendor support in the maintenance of the existing Cyberex equipment.

SAFETY EVALUnTION

SUMMARY

This modification does not affect the operation of the Essential Instrument AC j l

Power Supply. System but will improve the reliability of the system. The addition of the Static Transfer Switch (STS) and essential alternate power {

supply for each channel through the CVT provides a more reliable Essential Instrument AC Power Supply in the event of fault on a branch circuit.

The replacement of the existing inverter / regulated rectifier /DC distribution panel units matufactured by Cyberex with new equipment manufactured by SCI will not affect proper operation of the Essential Instrument AC Power Supply System.

The new equipment is functionally equivalent to the existing equipment and fully qualified for C?. ass lE application.

The DC loading tables were calculated using an inverter efficiency of 71% and therefore are conservative since the inverters are rated at an efficiency of 80%. Also Civil Engineering has determined that the current installation is acceptable and will not adversely affect the ventilation of Room 428B.

The expected roon temperatures will not af fect operability of the equipment.

The installed fan has a 6% margin of capacity over what is needed to keep the i room at 104 degrees F on a 95 degree F day. Hence, the room will not exceed l

design temperature during LOCA.

l l This change is ccnsidered safe and does not constitute an unreviewed safety evaluation.

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i SAFETY EVALUATION

SUMMARY

FOR FPR 94-0190-701 (SE 95-0008)

TITLE:

Replace Reactor Coolant Pump Monitoring Circuit Test Switches  !

CHANGE:

This FPR replaced eight Reactor Coolant Pump (RCP) Monitoring Circuit test  !

switches in RC3601, RC3602, RC3603, and RC3604. These switches are used during calibration and functional and response time testing of RCP Monitoring Circuits for Technical Specifications 3/4.3.1, 3/4.3.2, and 3/4.3.3.

REASON FOR CHANGE:

On January 26, 1990, functional testing of RCP Monitoring Circuit I led to a plant trip from 73% power. It was later determined that the trip was caused by

" inadequacy" of the test switches. One of the corrective actions was to replace the test switches.

SAFETY EVALUATION

SUMMARY

The new switches are functionally equivalent to the previous switches, so there ,

is no effect on hazards, design bases, Technical Specification bases, single '

-failure criteria, or plant security.

The new switches are mounted the same as the old switches, but require less space. They have identical electrical ratings, and are available IEEE Class lE qualified, and are seismically equivalent. A Material Engineering Equivalency evaluation has been prepared.

The new switches are nickel-plated. Although nickel has a higher resistivity coefficient than copper, it is only used as a thin plating so the actual change in the resistance of the switch is insignificant. In addition, this is a l

current loop, and therefore is unaffected by slight changes in resistance. The '

improved material properties of nickel vis-h-vis copper will eliminate the j production of filings, and thereby reduce the probability of circuit failure.

This switch replacement is safe.

1 SAFETY EVALUATION

SUMMARY

FOR  ;

FPR 94-0264-702 (SE 95-0048) j i

l TITIS:

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Change Normal Position for Valve WT105 CHANGE:

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FPR 94-0264-702 modified the electrical controls for valve WT105, allowing both auto and manual modes of operation. This also involved changing the normal position for valve WT105 from " closed" to "open".

REASON FOR CHANGE:

Per the Ohio EPA regulations, increased sampling of once per hour is required during peak flow conditions in the water treatment system. The normal sampling frequency is once every four hours during non-peak flow conditions. Automatic operation of WT105 to fill the Fire Water Storage Tank created peak flow l conditions. Automatic operation of WT105 was being defeated by manually i operating WT105 in order to control when peak flow conditions occurred. This l allowed sampling requirements to be met, avoiding possible EPA l enforcement / fines. However, WT105 was failing/ leaking because it was not designed for frequent operation.

4 SAFETY EVALUATION

SUMMARY

This activity has no effect on the portion of the Makeup Water Treatment System i seismically installed to protect safety related equipment. This activity does affect the means of providing a water supply to the Fire Water Storage Tank and therefore indirectly affects the means of providing fire suppression capability.

This activity does not compromise safety because the Fire Water Storage Tank will no longer be isolated from its water supply. Once again, flow to the tank will be done automatically by tank level switches or manually by operator.

WT105 will be placed in the normally open position and will no longer be required to be manipulated to fill the Fire Water Storage Tank. This will prevent further failures of WT105 due to frequent manipulation. There is a low and high level alarm in the control room to alert the operating staff of the need to fill the Fire Water Storage Tank or that the Fire Water Storage Tank is being overfilled.

Based upon the above discussion, it is concluded that the change is safe.

SAFETY EVALUATION

SUMMARY

i FOR FPR 94-0348-703 (SE 95-0051) 1 1

l TITLE:

Removal of Circulating Water Pump Transfer Switches CHANGE:

Field Problem Resolution"(FPR) 94-0348-703 eliminated transfer switches BE31ATS and BF31ATS and returned the design to the original (pre-FCR 78-462)

! configuration.

REASON FOR CHANGE:

f These transfer switches are obsolete and have had past maintenance problems. l Transfer switch BE31ATS has been incapable of performing the transfer function j since April 1994.

SAFETY EVALUATION

SUMMARY

The four Circulating Water Pumps (CWPs) are arranged into two independent loops with CWPs 1 and 2 forming loop 1 and CWPs 3 and 4 forming loop 2. Each CWP has a discharge isolation valve. If one of the two CWPs in a loop trips, the valve for the tripped pump will close to prevent reverse flow from the operating pump, and the valve for the operating CWP will close to a throttled position to prevent pump run-out.

A modification was initiated in 1978 to address a potential concern following a loss of one of the two 13.8KV busses. A loss of either 13.8KV bus would result in a loss of power to one CWP in each loop. Power to the associated discharge isolation valves for these CWPs would also be lost, resulting in a failure of these valves to close. Facility Change Request (FCR)78-462 identified a potential concern with backflow through the tripped CWP, causing reverse rotation and possible damage of the pump and motor and possibly tripping the operating CWP due to low discharge pressure caused by the reverse flow. As a result of this concern, FCR 78-462 installed automatic bus transfer switches l BE31ATS and BF31ATS to automatically transfer the power supply to the discharge isolation valves to an alternate MCC should power from the associated 13.8KV bus be lost.

The design of the Circulating Water pump and motor (e.g. both have sleeve bearings installed) is such that reverse rotation will not result in damage.

While some backflow through the tripped pump will occur, it is not expected to be excessive due to the resistance of the pump itself and the lower pressure in the loop due to the throttling of the discharge isolation valve on the running pump. Since a plant trip will occur due to the loss of two reactor coolant i pumps, the heat load on the condenser will be minimal and well within the capacity of the remaining two CWPs even with some backflow through the tripped j pumps. A loss of one of the 13.8KV buses occurred in February of 1979, prior l

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to implementation of FCR 78-462. A review of the available data from this event indicates that power to the bus was restored after approximately ten minutes, although it could not be determined when the tripped CWPs were restarted. Vacuum was not lost'during this event. If backflow through the tripped pumps was excessive,.the running pumps would trip on low discharge pressure. If use of the condenser is lost due to this or any other event following a loss of either 13.8KV bus, plant stabi.3vation and cooldown would be accomplished via the atmospheric vent valves.

The USAR flooding analysis for a circulating water line rupture assumes the closure of the discharge isolation valves given a single failure of the ability to trip the CWP. This change will not affect this assumption since, if the pump _is running, power will also be available to the associated isolation valve. The USAR flooding analysis conservatively assumes the isolation valves remain open following the successful tripping of the CWPs to allow water levels to equalize between the channel and the condenser pit. This assumption is also unaffected by this change.

Based on the above, it is concluded that the elimination of automatic bus transfer switches BE31ATS and BF31ATS does not have an adverse effect on safety.

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i SAFETY EVALUATION

SUMMARY

l FOR I FPR 94-0482-901 (SE 95-0004) I

)

TITLE:

Area Radiation Selector Panel C5781 Abandonment i CHANGE:

This FPR formally abandoned in place the the Victoreen Area Radiation Selector Panel C5781, which was located in the control room.  !

REASON FOR CHANGE:

This panel had not been used by operations and no maintenance activities had been performed in years. Due to safety concerns of an unused but energized ]

cabinet this FPR was initiated.

SAFETY EVALUATION

SUMMARY

This change has no effect on RE8446 and 2Fet;7's ability to perform the safety function of stopping the Fuel Handling Area Ventilation System fans and diverting the fuel handling area exhaust flow through the Emergency Ventilation System (EVS). The portion of the circuitry that performs this function is unaffected by this change. This safety function is initiated from relay l contact outputs from the rate meter where the signal for the Area Radiation Selector Panel uses analog outputs from the rate meter. In addition, an isolated voltage / current transmitter is installed between the rate meters and the Area Radiation Selector Panel. This change will cease use of the signal from the analog output for the Area Radiation Selector Panel. Each of the radiation monitor channels feeding C5781 will retain its remote indication and i alarm function.

Based on this, there is no effect on the design basis as a result of this change. This work can be performed without removing the equipment from service. There is no increase in adverse effects from any hazards nor is there any effect on system reliability as a result of this change. Based on the above, it is determined that the change is safe.

l SAFETY EVALUATION

SUMMARY

FOR FPR 94-0524-901 (SE 95-0031) i TITLE:

Add Hose Connections to Chemical Addition Tanks T49 and T50 CHANGE:

1 FPR 94-0524 added a 1 inch coupling to Tank T49 and a " Quick Connect" along with a " Quick Connect" to an existing connection on Tank T50. This change also affected the text as described in the USAR in the method of adding chemicals.

REASON FOR CHANGE:

Prior to this modification, Chemistry had no hard connections available for pumping hydrazine or ammonium hydroxide (or alternate approved chemicals) into their respective tanks.

SAFETY EVALUATION

SUMMARY

The addition of the fill connections to the tanks provides for a safer (personnel safety) way of adding chemicals to the tanks. It has no effect on the secondary water chemistry. Revision of the description of chemical storage and handling and removal of the words in the USAR concerning the addition method and system piping failures make the description in line with the USAR ,

guidance for level of detail and more consistent with current descriptions. l These changes will have no effect on safe operation of Davis-Besse.

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SAFETY EVALUATION

SUMMARY

• FOR 4 FPR 94-0638-901 (SE 95-0039) i TITLE:

4 Installation of Air Regulators on the Condensate Drain Pumps

) 9?ifdGE:

1 The condensate drain pumps for the boric acid evaporators utilize manual

isolation valves for the control of the air supply to the pumps. FPR ,

I 94-0638-901 added air regulators and pressure indicators to these air supply j lines'to allow Equipment Operators to regulate air flow, i

I REASON FOR CHANGE:

Regulation of the air supply, when properly set, will decrease the potential for air blow-by past the discharge check valves. This air blow-by is j undesirable because the pump effluent eventually reaches the main condenser, j During these occurrences, small detectable increases in oxygen levels in condensate have been measured. i l

I SAFETY EVALUATION

SUMMARY

FPR 94-0638-901 will not adversely affect plant safety for the following

. reasons:

The piping in which the regulators and the pressure indicators are being installed is non-seismic. The new installation has received an Engineering Inspection Team (EIT) review which concluded that no new hazards are being created.

The regulators and pressure indicators are not being installed in a safety related section of the Station and Instrument Air System. The affected portion of the system is a branch connection off a standard air service station in room.

116. The demand on the air systems will not change as a result of this change.

The boric acid evaporators do not require these condensate drain pumps for

. operation. The evaporators have been run and were designed to run without the condensate drain pumps.

I The desired outcome of this installation is to reduce the potential for air blow-by past the discharge check valves for these pumps. This will eliminate the effect that operation of these pumps can have on the measured condensate j oxygen levels and any related general corrosion rates in the condensate system. j l

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SAFETY EVALUATION

SUMMARY

FOR FPR 94-0716-902 AND UCN 96-075 (SE 96-0056)

TITLE:

Fire Water Storage Tank (FWST) Repairs CHANGE:

The temporary repair defined in FPR 94-0716-902 is a design change to the FWST.

The changes to design are: the use of allowable design stress values that are higher than the allowable design stress values provided in the design standard; the use of a minimum wall thickness, 3/16 of an inch, that is less than the design standard for the tank, 1/10 of an inch; and the installation of reinforcing plates on the vertical weld seams of the first two rings of the tank. The higher allowable stress values will be 80 percent of yield strength for the bottom two rings and 88 percent of yield strength for all rings above the second as opposed to the design value of 50 percent yield strength.

REASON FOR CHANGE:

The FWST has a generalized Microbiological Induced Corrosion attack that has reduced the thickness of the plate and weld sections to less than the allowable minimum wall thickness based on the design standard for the tank, American Water Works Association (AWWA) D.100, delded Steel Tanks for Water Storage.

SAFETY EVALUATION

SUMMARY

The function of the FWST is to provide water for Appendix R fire suppression equipment.

Calculation C-ME-13.07-02 was performed to demonstrate the FWST is structurally adequate and, thereby, fully fit for service. The tank will hold the volume of water required and the reliability of the tank is unchanged for the duration defined by the FPR. This is based on values from the American Petroleum Institute (API) 653, Standard for Tank Inspection, Repair, Alteration and Reconstruction, which allows for higher allowable stress values for the tank material than was previously used.

These values are acceptable for the interim period because the FWST service conditions are similar to the service conditions of other industrial tanks that have been repaired in a similar manner, and have provided years of acceptable service.

Furthermore, the FHAR states that failure of the FWST does not result in a flooding hazard and does not result in a failure of the Fire Suppression System to supply required water to fire suppression equipment.

Based on the above, the change is safe and does not present an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR ,

FPR 94-0744-901 (SE 95-0020) {

TITLE:

Abandon-in-Place Lube Oil Storage Room Sump Pumps  :

CHANGE:

This FPR revises drawings to depict the abandon-in-place condition and the i disconnect of electrical service to Lube Oil Storage Room Sump Pumps. [

REASON FOR CHANGE:

This FPR revised drawings to show the current condition of the Lube Oil Storage  !

Room Sump Pumps.  ;

P SAFETY EVALUATION

SUMMARY

The Lube Oil Storage Room Sump system components do not' perform or affect any >

safety related function. Therefore abandoning-in-place and changing the  ;

position of valves from open to close will not adversely impact the operation of the Station Drain and Collection System. The change to USAR Figure 9.3-4 i depicts the position of valves from open to closed. The change of the valves from open to closed on the Operational Schematic and PEID make the drawings, USAR and procedures consistent. Instrumentation for sump continuous monitoring .

remains'in place and continues to be used. It provides the warning for using a {

portable pump to transfer the contents of the sump to the East Condenser Pit.

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SUMMARY

FOR FPR 94-1229-705 (SE 95-0029)

TITLE: 1 i

Wire' Label Changes for 4KV Breaker AD113 CHANGE. ,

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FPR 94-1229-705 made wire label changes on conductors associated with the '

number 2 Component Cooling Water (CCW) pump's 4KV breaker, AD113.

REASON FOR CHANGE:

PCAQR 94-1229 documented wire labeling discrepancies in several essential 4KV breakers.

SAFETY EVALUATION

SUMMARY

The changes made by FPR 94-1229-705 will not affect the safety functions of the affected SSCs.

Wire labels do not perform a safety function. Labels are not credited for equipment functioning during normal or emergency station operation. These labels show a conductor's "name" and provide a link between the actual field wiring and the wiring depicted on drawings. The conductor's name is used to facilitate construction, testing and troubleshooting.

Based on the above evaluation, it is concluded that changing wire labels associated with the number 2 CCW pump breaker is safe.

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SAFETY EVALUATION

SUMMARY

l FOR FPR 94-1229-707 (SE 95-0030) l TITLE:

Wire Label Changes for 4KV Breaker AC107 l CHANGE:

FPR 94-1229-707 made wire label changes on conductors associated with the l number 1 Service Water (SW) pump's 4KV breaker, AC107.

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REASON FOR CHANGE:

PCAQR 94-1229 documented wire labeling discrepancies in several essential 4KV breakers.

l SAFETY EVALUATION

SUMMARY

The changes made by FPR 94-1229-707 will not affect the safety functions of the affected SSCs.

Wire labels do not perform a safety function. Labels are not credited for equipment functioning during normal or emergency station operation. These labels show a conductor's "name" and provide a link between the actual field wiring and the wiring depicted on drawings. The conductor's name is used to facilitate construction, testing and troubleshooting.

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Based on the above evaluation, it is concluded that changing wire labels associated with the number 1 SW pump breaker is safe.

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I SAFETY EVALUATION

SUMMARY

FOR FPR 95-0092-901 (SE 95-0007)

TITLE:

Permanent Installation of Smoke Detectors in Room 114 CHANGE:

This FPR 95-0092-901 permanently installed the smoke detectors temporarily installed by Temporary Modification (TM) 93-0030.

1 REASON FOR CHANGE: ,

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Due to NRC concerns with the operability of the Thermo-lag fire barrier in Room 114 (Misc. Waste Monitor Tank and Pump Room), implementation of compensatory measures was required. In order to avoid a continuous fire watch, three Portable Detection System (PDS) detectors were installed and an. I hourly fire watch was established. The detection and the hourly fire watch  !

are necessary until the Thermo-lag fire barrier concerns are resolved.

In order to eliminate the high cost of maintaining the PDS, it had been <

determined that the existing Simplex detection system in Room 115 (ECCS Pump Room 1-2) could be extended into Room 114 until the Thermo-lag concerns are  !

resolved. TM 93-0030 temporarily installed three Simplex detectors in Room 114 in place of the three PDS detectors.

SAFETY EVALUATION

SUMMARY

2 The Simplex smoke detectors in conjunction with a roving fire watch, which is required due to the inoperability of TSI fire barriers, will provide the appropriate compensatory measures as required by the FHAR. Modification of the Simplex software for panels C5796B and C2720 to reflect the new fire zone for Room 114, FDZ114, will not affect the functions of these panels. The wiring for the permanent smoke detectors will be connected to the Simplex system at i detector DS8694M in Room 115. This will not affect the function of detector i DS8694M and the rest of the FDS. New 3/4 inch EMT conduit will pass through an existing negative pressure foam seal in the non-rated wall between Rooms 114

and 115. After the conduit is installed, the block-out for the conduit shall meet the requirements of drawing M-473B Detail HDE-4. This will maintain the negative pressure boundary between Rooms 114 and 115. The detectors and its  !

associated conduit have been evaluated to not be a Seismic II/I concern.

Installation of these detectors will not create an adverse environment.

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! SAFETY EVALUATION

SUMMARY

FOR \

FPR 95-0130-902 (SE 96-0006) l 4

4 TITLE:

d j Abandonment of the Auxiliary Building Non-essential Humidifier  ;

I CHANGE: )

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, This change abandoned the Auxiliary Building non-essential humdifer (S-41) in-place.

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REASON FOR CHANGE: j i l The Auxiliary Building Non-essential humidifier (S-41) was installed to control I the relative humidity of the air in the 603' level Chemistry Laboratories and Health Physics monitoring areas during the winter heating season. This was done for personnel comfort and is not needed for equipment operation. This j equipment has deteriorated due to buildup of deposits and corrosion from the moisture which is normal " wear" for this type of equipment.-Replacement parts

! are no longer available. The equipment has not worked properly for some time and replacement was deemed as unnecessary. Personnel comfort in the spaces is j acceptable without humidification.

i SAFETY EVALUATION

SUMMARY

l l ' The Auxiliary Building Non-essential Humidifier performs no functions important to safe plant operation. Abandonment of the Auxiliary Building Non-essential

' Humidifier will have no effect on safe operation of Davis Besse. The environmental' conditions for plant equipment as analyzed in the USAR remain unchanged.

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SAFETY EVALUATION

SUMMARY

FOR FPR 95-0364-901 (SE 96-0008)

TITLE:

Post Accident Sampling System Filter Removal CHANGE:

The change removed the Post Accident Sampling System (PASS) skid filter, F133.

REASON FOR CHANGE:

Filter F133 has been a source of recurring leakage and contamination. A thread sealant able to withstand operating conditions (high temperature) for extended periods of time is not available.

SAFETY EVALUATION

SUMMARY

The PASS is a nonsafety related system. The system is designed so that samples can be collected and analyzed without exceeding the dose guidelines of General Design Criteria (GDC) 19. A time motion study of sampling and analysis was performed. Based on this study it is concluded that the sample can be obtained and analyzed without exceeding GDC 19 dose guidelines of 5 REM to whole body, or its equivalent to any part of the body. This study and analysis was conducted prior to installation of the PASS filter and therefore does not account for, or take credit for, its existance in the analysis.

NUREG 0737 states: " consideration should be given ... for preventing blockage of sample lines by loose material in the RCS..." The location of the current filter at the PASS sample panel provides little protection from clogging of the majority of the system. The majority of the system piping is located upstream of the filter. A small portion of the PASS system is downstream of the current filter location. The probability that the portion of the system downstream of the current filter location would become clogged more quickly with the filter removed is less than the clogging that might occur with fine material clogging the filter media in F-133. Removal of the filter from the system should  ;

increase the probability of obtaining a sample from the PASS panel. l The PASS filter was added to the PASS skid as part of MOD 87-1193 as an enhancement. It shall be replaced with a section of straight tubing.

As stated above, the proposed change does not represent any affect on safety l and therefore the proposed action is safe. j i

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SAFETY EVALUATION

SUMMARY

l FOR FPR 95-0433-901 (SE 96-0007) l l

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TITLE:

1 l l Replacement of Batteries and Chargers for Emergency Battery Lights j j CHANGE:

I l FPR 95-0433-901 replaced the battery and charging module in the Teledyne Big Beam Emergency Battery Lights with a sealed maintenance free battery and a charging module from Dual-Lite. The FRAR stated that the battery charger is j capable of restoring the battery to full capacity in less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from l l 87.5% of the nominal battery voltage. The FHAR was revised to delete the 12 l hour time limit because the new charging modules, while having numerous improvements over the existing charging modules, do not have the output capacity to recharge the batteries in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Elimination of the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> recharge requirement has no impact on the equipment's ability to perform the required function, i.e. provide illumination for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

REASON FOR CHANGE:

The Dual-Lite charging module can automatically perform routine maintenance checks, providing valuable information on the condition of the emergency battery light unit without additional manpower requirements.

SAFETY EVALUATION

SUMMARY

Elimination of the requirement to recharge the batteries within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has no effect on safety, since no credit is taken for the Emergency Battery Lights  ;

after they provide illumination for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Hand-held battery powered lanterns j are available for any actions still left to be done.

The new charging modules are an improvement, providing battery self-testing and problem diagnostic features. The new charging modules have been proven to be reliable with nine years of industry service and no significant problems.

While the new battery and charging unit are heavier than the old battery and charging unit, an evaluation has been performed which verifies the new weight is acceptable.

The new charging modules draw more power but this is only on the recharge mode which would occur after the batteries have performed their safety function.

While there is the slight chance that the additional load may trip a lighting circuit or even a whole panel, this does not affect safety since the lighting circuits are not taken credit for.

Based on the above these changes are safe.

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SAFETY EVALUATION

SUMMARY

FOR FPR 95-5124-501 and FPR 95-5126-501 (SE 95-0071)

TITLE:

Replacement of the "2" Relay in 4kv Switchgears AC101 and AD101 CHANGE:

This FPR replaced the 7012PE relays with Agastat Model number E7012PE relays.

REASON FOR CHANGE:

In order.to create the two-step timing function.

SAFETY EVALUATION

SUMMARY

The safety function of EDG output breakers AC101 and AD101 is to connect each EDG to its respective essential bus to provide on-site standby power for safety related loads required to mitigate the effects of an accident combined with a loss of off-site power to safely shutdown the facility and maintain it in a safe shutdown condition. Each EDG output breaker has a "2" and "2X" relay in its closing circuit. The safety function of the "2" timing relay is to provide a time interval for energizing the "2X" relay.

Replacing Agastat Model 7012PE with E7012PE in AC101 and AD101 while continuing to use the originally qualified auxiliary switches does not adversely affect the safety functions of the affected SSCs. There are no apparent physical form, fit, or function differences between the Agastat Models. However, stricter quality control standards have been imposed on the 7000 Series models. Failures have not been recorded for the current "2" relays; therefore, failures should not be expected with the 7012PE replacement.

f The replacement relays have been determined to be seismically adequate for their proposed locations in switchgear units AC101 and AD101. The auxiliary switch contacts are configured single pole double throw and are enclosed in plastic.

They are used in both essential and non-essential applications and failure of these switches has not been a recurring problem.

Based on the above evaluation it is determined that replacing the existing "2" relay with an Agastat Model 7012PE while using the existing auxiliary switch is safe and does not constitute an unreviewed safety question.

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j h SAFETY EVALUATION

SUMMARY

FOR l FPR 96-0189-901 and'UCN 96-048 (SE 96-0048) 4 TITLE:

a

) Replacement of the " Lunchroom" A/C Unit CHANGE:

This.FPR removed the " Lunchroom" A/C(S65) from USAR figure 9.4-10.

l j REASON FOR CHANGE:

The unit was. originally installed to provide cooling to the lunch room i constructed on the south end of the Turbine Deck 623 foot level which is j considered as an extension of the office building. This room has since been j converted to Instrument Storage and various other uses which do not directly j support plant operation.

1 SAFETY EVALUATION SUMMAR :

The Air Conditioning Unit has no Safety Related or Important to Safety function.

The environmental conditions for plant equipment as analyzed in the USAR remain unchanged. Based on the above the proposed action is safe and presents no l

unreviewed safety questions.

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l SAFETY EVALUATION

SUMMARY

FOR j MOD 91-0020-00 (SE 93-0004 R03) l TITLE: {

l l

Pulling an OTSG Tube Segment, Sleeving an OTSG Tube and/or Plugging an OTSG Tube CHANGE:

This incorporates the actions associated with pulling an OTSG tube segment. The actions include removing of a segment of the tube from the OTSG, verifying the tube segment remaining in the OTSG is adequately stabilized and then plugging the tube sheet or tube segment with either a mechanical rolled plug (only if the tube to tubesheet joint of the tube segment remaining in the steam generator is unaffected) or a welded plug.

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REASON FOR CHANGE:

This is being done to repair tubes with leaks or defects.

L SAFETY EVALUATION

SUMMARY

The tube end will no longer be connected to the tubesheet following the l

machining process. The tube in the "as left" condition for weld plugs fulfills the requirements defined in OTSG Tube Stabilization Criteria, therefore, there are no adverse effects on vibration or stability of the tubes that will have welded plugs installed.

There is no concern that boric acid will be concentrated to create a corrosive environment for the upper tube sheet because the point of leakage is above saturation temperature for the secondary side. Therefore any leakage past the plug will immediately flash to vapor. During shutdown periods the risk of filling the tube with water such that a potential corrosive environment at the tubesheet exists is greatest during a wet lay-up. In this scenario there is no method once the fluid reaches the bottom of the tubesheet for the air to vent, I therefore the tube will fill only partially with fluid. i i

Based on the above the process of removing a segment of the steam generator tube j and re-establishing pressure boundary with previously discussed plugging  ;

techniques will not have any adverse affects on the steam generator or reactor coolant pressure boundary.

l l The proposed modification is considered safe and does not constitute an

! unreviewed safety question. _

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1 j SAFETY EVALUATION

SUMMARY

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FOR l MOD 91-0029 (SE 95-0036, R.01) 1 I i TITLE:

l Dry Fuel Storage Facility CHANGE: l Construct a Dry Fuel Storage Facility to temporarily store spent fuel assemblies.

l REASON FOR CHANGE:

l-Davis-Besse has decided to augment the spent fuel pool with the constructien of a Dry Fuel Storage Facility (DFSF). This DFSF will be utilized under the l generic rules of 10CFR72 Subpart K. The Dry Fuel Storage System to be used is the standardized "NUHOMS" System supplied by VECTRA TECHNOLOGIES, INC.

SAFETY EVALUATION

SUMMARY

This evaluation determines whether activities related to dry fuel sto? age under )

the general license involve any unreviewed safety questions. The review of j these activities will be addressed in the following General Areas: 'deavy Loads, Plant Modifications, and Dry Fuel Storage Activities.

Heavy Loads The Cask Crane has a design rating of 140 tons, which is considerably greater than the maximum load to be lifted, approximately 100 tons for the loaded transfer cask and lifting yoke. The cask crane has both procedural controls and electrical interlocks to prevent it from traveling over the spent fuel pool. Therefore, the transfer cask cannot be moved over the spent fuel in the pool.

A functional load lift was performed with the cask crane prior to its initial use for dry fuel storage activities. The load lifted was approximately 135 tons. Pre-load and post-load crane inspections were performed.

i The lifting yoke for the "NUHOMS" transfer cask has been designed (Reference l ANSI N14.6), fabricated, and tested as a single load path special lifting device. The VECTRA analysis that show the yoke design meets the heavy loads ,

requirements of NUREG 0612 and ANSI N14.6. The lifting yoke has been l successfully preservice load tested to 300% of design load.

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{ The yoke extension slings have been designed in accordance with the applicable requirements of NUREG 0612, ANSI N14.6 and ASHE B30.9. The yoke extension j slings will be proof tested at twice their rated load.

The transfer cask trunnions and trunnion sleeves have been designed to meet the i stress allowables for non-redundant special lifting devices of ANSI N14.6.

l The trunnions have successfully completed a one-time load test by the manufacturer at 150% of the design load.

i l Postulated Cask Drop i

l- The USAR does not discuss the possible radiological consequences from a cask

] drop accident. USAR Section 9.1.4.3 implies the maximum number of fuel

assemblies in a cask is ten. Since the "NUHOMS" cask contains 24 assemblies, J the offsite radiological evaluations are performed as a part of this review

using the guidance contained in Standard Review Plan 15.7.5; Spent Fuel Cask Drop Accidents. These resultant doses of this evaluation are less than the j offsite consequences given in the USAR Section 15.4.7 for a Fuel Handling l Accident Outside of Containment, and they are significantly less than the SRP 15.7.5 acceptance criteria values.

The ambient temperatures in the spent fuel pool area will be substantially 3 higher than the 0*F specified in Technical Specification of the "NUHOMS" i Certificate of Compliance, for transfer cask lifts of greater than 80 inches.

l The structural evaluation of postulated cask drop accidents in and/or near the i cask pit is evaluated. One calculation determined that the foundation of the l cask pit would experience minor damage due to the impact of a 125 ton cask j falling 47 feet without consideration for any deceleration effects from water.

The damage sustained would not affect the spent fuel pool, due to the cask pit being separated (three foot thick reinforced concrete wall) and isolable (watertight bulkhead) from the spent fuel pool. The base of the cask pit is 15 i foot thick reinforced concrete bearing directly on bedrock. The floor of the

cask pit is 6'-6" below the bottom of the spent fuel pool and both areas are

{ lined with " thick stainless steel plate. Damage to the cask pit liner plate

would cause water to flow into the leak chase channels. These channels are l connected to a piping system that is isolated by (normally) closed valves. The

water lost through leakage can be made up, in accordance with existing

procedures, to restore and maintain the cask pit water level. The analyses

have determined that there is no possibility of the cask falling into the spent l 2 fuel pool.  ;

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Postulated Shield Pluo Drop j i

A postulated shield plug drop onto the loaded Dry Storage Container (DSC) would j

! likely damage some fuel assemblies. However, the radiological effects are  ;

j clearly bounded by the postulated cask drop evaluation, where all assemblies l were assumed damaged and no water shielding was considered. l 1'

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Plant Modifications l The storage facility foundation, designed by VECTRA, is a non-Q, non-seismic design in accordance with SAR requirements.

A platform will be installed in the cask wash pit to facilitate dry fuel storage operation. The cask wash pit platform will raise the top of the transfer cask to Elevation 603' to facilitate the DSC closure activities in this pit. This platform has been designed as an augmented quality (AQ) structure for the weight of a fully loaded transfer cask and to be seismically rigid to preclude amplification of the floor seismic accelerations. All material, fabrication, and inspection requirements are in accordance with safety related specifications.

Lightning protection will be installed on the Horizontal Storage Modules (HSM) in accordance with applicable drawings.

A temperature monitoring system meeting the requirements of the NUHOMS Certificate of Compliance will be installed per the modification design report.

Dry Fuel Storage Activities All dry fuel storage (DFS) activities will be performed in accordance with written procedures. These procedures will ensure the safe loading (or unloading) of fuel assemblies into (or out of) the DSC. The procedures will delineate the actions necessary to maintain the fuel in a safe condition, control the radiological impact (dose and contamination (airborne / fixed)), and perform the required activity.

The DFS activities will interface with the following Plant equipment; Spent Fuel Pool and Handling Bridge, Emergency Ventilation System, AC Electrical Power, Cask Crane, Underground Plant Utilities, Demineralized Water, Station Air, Nitrogen, Station Drains and Helium.

In order to lower the transfer cask into the cask pit without submerging the cask crane main hook the water elevation in this pit will be lowered to l elevation 583'-0". In the unlikely event that the SFP gate would fail while l the cask pit water elevation is lowered, the water elevation in the SFP would l fall to approximately elevation 598'-3". This amount of water cover (approximately 21') is substantially more than the 9 feet (approximate) of ,

cover required to ensure adequate shielding as described in USAR Section 9.1.2.3.

l l The procedures used to perform dry fuel storage loading and unloading operations will specify the activities which will require the emergency ventilation system be operable. This is conservative in that no credit is taken for iodine removal by the EVS in the event of a cask drop accident.

Based on the above determination, the dry fuel storage activities, performed in

[ the auxiliary building, are safe and will not adversely affect the safety function of the spent fuel pool nor it's structural integrity. Also, these dry fuel storage activities will have no adverse effect on any other system, components, or structures. Therefore, an unreviewed safety question does not exist.

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SAFETY EVALUATION

SUMMARY

FOR MOD 92-0013 Supp. 0 (SE 92-0069 Rev 1)

TITLE:

Replacement of Emergency Diesel Generator Air Start Check Valves.

CHANGE:

Replaces Emergency Diesel Generator Air Start Check Valves DA24, DA25, DA38, and DA39.

REASON FOR CHANGE:

Replace carbon steel check valves with stainless steel due to moisture concerns.

SAFETY EVALUATION

SUMMARY

The replacements have a design pressure and temperature of 4100 psig and 400 degrees F as compared to the less than 2200 psig at 100 degrees F ratings for the old valves. This provides a greater safety factor and the replacement valves have a demonstrated history of reliability while the existing valves have a history of failure to perform functions of preventing reverse flow and allowing forward flow.

The change from carbon steel to stainless steel is acceptable because of the improved corrosion resistance properties and the acceptable design pressure / temperature. Galvanic corrosion is not a concern based on the acceptable experience with the stainless steel seats contained in the check valves being replaced.

The change from ASME Section III Class 3 to ANSI B31.1 for the check valves is acceptable based on a review of commitments relative to the Emergency Diesel Generator Starting System that revealed '?ledo Edison has not committed to any codes, standards, NUREG's or Regulatory J! des which invoke ASME Section III for these check valves. The valves must provide an acceptable level of quality and safety which is met by the replacement valves. It is noted that the piping between the air receiver and the diesel generator downstream of the first valves (normally open) is designed to ANSI B31.1. (The replacement check valve and the new flanges and fittings were designed to appropriate provisions of ASME, ASTM, and ANSI codes and standards. Seismic / stress analyses have been performed for the new piping configuration. These analyses concluded that the configuration, including the threaded and flanged connections, are adequate.)

Based on the above, the proposed change is safe and does not constitute an unreviewed safety question.

SAFETY EVALUATION

SUMMARY

FOR LIMITED MOD 92-0074 (SE 94-0084)

TITLE:

Fix Power Supply Wiring for EDG Alarm Panels CHANGES.

Limited Modification corrected an internal wiring error in EDG 1-1 Annunciator Alarm Panel 43 mounted in C3617 and EDG l-2 Annunciator Alarm Panel 44 mounted in C3618.

REASON FOR CHANGE:

This wiring error apparently had existed since the installation of the EDGs.

As documented in PCAQR 92-0265, Panalarm's replacement point modules (digital replacements for exici$ng analog circuit cards) fail catastrophically when installed into them vigrm panels.

SAFETY EVALUATION

SUMMARY

Correcting the internal wiring error in the EDG Alarm Panels does not affect the capability of the EDGs to perform their safety function. Although considered important to the operation of the EDGs, these EDG Alarm Panels do not perform or have a safety function. Additionally, the EDG Alarm Panels and associated inputs are electrically isolated from the EDGs and their control circuitry.

As the available quantities of the analog point circuit cards diminish, this activity will help ensure the maintainability of these EDG Alarm Panels.

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l SAFETY EVALUATION

SUMMARY

FOR I MOD 93-0016 (SE 94-0045, R01) l TITLE:

1 Resolution Of Pressure Locking Concerns With Valves DH11 And DH12 L

CHANGE:

For both valves DH11 and DH12 a vent was provided by drilling a small hole in the upstream disc of each valve wedge. This will continuously vent the bonnet to the piping line upstream of each valve thus eliminating the potential for pressure locking.

REASON FOR CHANGE:

Recent concerns expressed by the NRC have identified problems with the sizing and selection of motor operated valves, specifically with the mechanism of pressure locking of flexible wedge gate valves. Toledo Edicon has conducted reviews on the population of safety related gate valves. These reviews concentrated on those valves that had to open to perform a safety function and could be subjected to pressure locking. As a result of these reviews, actions were recommended to preclude the possibility of pressure locking of DH11 and DH12.

SAFETY EVALUATION

SUMMARY

For both DH11 and DH12 a vert path was provided for the bonnet via a small hole in the upstream disc of the valve wedge. This provides continuous bonnet venting to the piping line upstream of each valve. As such, bonnet pressure will be limited to the pressure upstream of each valve. This is a standard industry design for the resolution of concerns for pressure locking.

Drilling a hole in the disc for DH11 and DH12 does not affect any of the valves functions. Complete isolation is required only in one direction, the normal direction of flow, and isolation in this direction is not be affected.

Therefore, the containment ibolation valve function of DH11 is not impacted.

Isolation in the other direction, towards the RCS, is affected. This isolation would only be required for post maintenance testing following a repair or a replacement downstream of DH11 and DH12. The leakage out of the test boundary would be stopped by performing the test at elevated RCS pressures. The change to DHil~and DH12 is considered a change to the way these valves perform their isolation function and therefore it represents a change to the facility as described in the USAR.

l

! The venting arrangements described above will ensure that the pressure in the l bonnets of DHll and DH12 do not exceed RCS pressure. These actions ensure that

{ the thrust required to open DH11 (DH12) is not unpreoictably increased by pressure locking, thereby maintaining the required f.hrust within the capability of the actuator.

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SAFETY EVALUATION

SUMMARY

l FOR l MOD 93-0028 (SE 95-0035)

TITLE: f Service Water and Component Cooling Water 4KV Breaker Interlocks i

l CHANGE:

MOD 93-0028 changed the interlocks in the auto-close circuits for the number 1 and number 2 Service Water (SW) and Component Cooling Water (CCW) 4KV breakers AC107, AD107, ACll3 and AD113. This change involved substituting a relay contact which responds to the position of the number 3 pump's manual "CD" breaker in place of the previous breaker switch contact which responded to the position of the number 3 pump's auto-breaker. This substitution blocks an auto-close of the number 1 and number 2 SW and CCW pump breakers if the corresponding "CD" breaker is closed. Which means, if pumps 1 (2) and 3 were both aligned to a bus prior to an undervoltage, only the number 3 pump would be automatically loaded when the bus voltage was restored. The USAR was also revised to account for a particular CCW pump swap where the changes made by the I MOD will not prevent connecting more than one pump onto an EDG. ,

i REASON FOR CHANGE:

1 PCAQR 91-0334 documented an inconsistency between the USAR and the design of the auto-loading circuits for the SW and CCW pumps. j SAFETY EVALUATION

SUMMARY

Changing the auto-close circuit interlock was evaluated by considering each step involved in pump swapping. This evaluation revealed that the interlock change will prevent auto-loading two pumps onto one EDG. The evaluation also determined that the existing order of installing / pulling control power Fuses in the number 3 pump auto-breaker and closing / opening the "CD" nonauto-breaker must be reversed in these procedures. Reversing this order is required to j ensure that one pump is available for auto-closing.

Changing the interlock was accomplished by using a spare normally closed contact on existing General Electric HFA151 relays in place of the Westinghouse MOC breaker switches. General Electric type HFA relays are routinely used to provide interlocks and initiate trip signals in the Station's essential 4KV breaker control circuits. These particular HFA151 relays were purchased suitable for Class IE service and installed by FCR 84-177 Rev. C to resolve an Appendix R non-compliance. Using this relay to accomplish the proposed interlock will not invalidate the existing Appendix R analysis.

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The SW and CCW number 1 and 2 pump breakers are listed on the Seismic Qualification Utility Group (SQUG) " Safe Shutdown Equipment List" as requiring a relay evaluation. SQUG methodology was used to review the use of relay contacts to implement the proposed interlock. This review concludes that ,

I seismically induced contact chatter will not prevent the designed auto-loading of these pumps.

EDG 1-1 was used to evaluate the effect of loading two CCW pumps onto one EDG.

The effect on EDG 1-1 was determined by including two CCW pump motors rather '

than one CCW and one MU pump motor in the cumulative loads listed on the EDG load table. This shows'that EDG 1-1 will remain below its 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating 1 during all steps involving auto-loading. Administrative controls will continue to alert operations to this unconventional loading and require tripping the I I

nonselected pump. These controls will minimize the unnecessary EDG loading.

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SAFETY EVALUATION

SUMMARY

FOR MOD 93-0035 (SE 94-0014 R09)

TITLE:

Motor Operated Valve Modifications CHANGE:

HP2A, HP2B, HP2C, HP2D, AF3869, AF3871, DH1B, DHIA, MS106A, MS107, MS107A, MU2A, MU1B, MU1A, DH11, and DH12. FW601, FW612, SW2929, and SW2930 also described below.

REASON FOR CHANGE:

These valves were modified to meet the requirements of Generic Letter 89-10, Safety Related Motor-Operated Valve Testing and Surveillance.

SAFETY EVALUATION

SUMMARY

The closing valve control was changed from torque to limit control for HP2A, HP2B, AF3869, AF3871, DHlA, DH1B, MS106A, MS107, MS107A, HU2A, MU1B, MUlA, DH11, AND DH12. The advantages of these changes are as follows: Actuator stall torque capability is available during the valve closing stroke when the torque switch is bypassed, therefore the valves ability to close against high differential pressure is not limited by the torque switch; inadvertent torque switch actuation is eliminated during the portion of the valve close stroke when the torque switch is bypassed; torque switch setpoint accuracy does not need to be considered if the valve is controlled by the limit switch; and seating forces can be minimized as compared to torque seated control. The disadvantages are:

if the torque is bypassed, actuator motor burnout or damage to the valve would not be prevented if a locked rotor condition occurred during valve closing, and consistent seating forces may be difficult to achieve due to changing fluid temperature or pressure conditions.

HP2A, HP2B, HP2C, and HP2D's performance is enhanced during the last portion of the stroke when the torque switch is placed back in the control scheme. Also, the thrust rating for these valve's actuators are being increased. This rerating will allow torque switch setting to be increased for HP2B and HP2D without exceeding the actuator rating.

AF3869 and AF3871 will be allowed to be closed with the full motor capability, ensuring that they will close for a condition of design differential pressure with reduced voltage. The gearing in the valves' actuators will be replaced increasing the overall gear ratio to provide greater motor capability. The thrust rating for their actuators will also be increased to allow the valves to operate at design conditions without exceeding the actuator thrust rating.

DH1A and DH1B to be allowed to be closed without excessive seating forces.

Setup on limit switch control will allow the total thrust to be maintained less than the allowable while retaining the torque switch in the circuit as a back up to the limit switch. Reducing the seating forces will protect the valve and actuator from failure, and will limit the unwedging forces during valve opening.

MS106A, MS107, MS107A will be allowed to be closed without exceeding the valve or actuator continuous service limits. The limit control setup will minimize the seating forces for normal operation. The torque switch will be retained in the circuit to provide protection for the motor, the actuator, and the valve in the event a stall condition occurs while closing.

This modification will disconnect the motor brakes and remove the friction discs from FW601 and FW612. This will enhance their reliability by removing a possible failure mechanism without significantly affecting the seating forces.

The reliability of MU1A, MU1B, and MU2A is enhanced during the period of disc travel when the torque switch is bypassed and the performance is unchanged during normal operation for the last portion of the stroke when the torque switch is placed back in the control scheme.

Bypassing the torque switch will enhance the reliability of SW2929 and SW2930 by eliminating the possibility that the torque switch will interrupt their operation. Some protection for valve and actuator components is lost, but this i will not affect the ability of the valve to reposition in order to perform a safety function.

DH11 and DH12 closing direction control logic is modified so that these valves are controlled in the closing direction oy the limit switch. Limit switch control will allow them to be closed without exceeding any valve, motor, actuator or seismic limits. The limit control setup will minimize the seating forces for normal operation, but will allow the valves to close for design conditions. The torque switch will be retained in the circuit to provide protection for the motor, the actuator, and the valve in the event a stall ,

condition occurs while closing. l The above described valve control circuit changes, the disabling of the motor brakes, the gearing replacements and the rerating will enhance the reliability of the affected valves.

Based on the above discussion it is concluded that the proposed changes are safe and do not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR MOD 93-0041 (SE 95-0027)

TITLE:

Modify ECCS Sump Pump Motors Starter Circuit CHANGE:

This change modified the six (6) Emergency Core Cooling System (ECCS) sump pump motor starters from latching type to non-latching type. This modification also removed from each scheme a control relay and a latching auxiliary relay to simplify the control circuit.

REASON FOR CHANGE:

MOD 93-0041 was initiated as a result of PCAQR 92-0394. This PCAQR discussed the failure of the starter in biecker cubicle BE1115. The starter failure was due to the failure of a control relay manufactured by the Deutsch Co. There were two previous failures of this same control relay, and this type of Deutsch relay has a history of failures as documented in the PCAQR.

SAFETY EVALUATION

SUMMARY

The previous design of the ECCS sump pump motor control circuitry utilized a complex circuit of a latching type starter and miscellaneous control relays.

Modification 93-0041 Design Summary describes the control circuit design requirements in detail. The Design Summary shows that there are no requirements for using a latching type motor control scheme in this application. The Design Summary also proves that removing the latching scheme from these motor starter circuits will not have any adverse effects on the safety function of either the ECCS Sump Pumps or their associated MCC's.

Removal of the latching components deletes a source of control circuit failure, and therefore increases the reliability of the ECCS Sump Pumps.

The information provided to the operators in the control room by the pump running light will be more reliable as the contact that operates the light will be directly controlled by the starter.

The functions of the affected MCCs and the ECCS sump pumps are not being changed, therefore, the effect of hazards analyzed in the USAR remain unchanged.

This modification does not adversely affect the seismic qualification of MCC EllA or MCC F11D. The A200 series starters are seismically qualified to operate continuously energized.

This modification decreases the combustible loading, however, the change is not significant. There is no change to the Fire Hazard Analysis Review (FHAR), and there is no impact on the Fire Protection Program.

SAFETY EVALUATION

SUMMARY

FOR MOD 93-0051 (SE 95-0019, Rev. 01)

TITLE:

i Removal of Unnecessary Loads in SFAS and SFRCS CHANGE:

)

This MOD removed unuued electrical loads in the Safety Features Actuation System (SFAS) and removed local control pushbuttons for the Main Steam Isolation Valves (MSIV). The loads in the SFAS are the digital indicator and resistor originally used for calibration of the SFAS trip bistables and a transformed 60 Hz input to the timing circuit of the sequencer.

REASON FOR CHANGE:

The digital indicator and resistor were abandoned due to the requirement for use of measurement and test equipment (M&TE) that has been calibrated to a known standard. The transformed input to the sequencer was abandoned for an oscillator circuit. The MSIV pushbuttons are being abandoned due to the possibility that a failure of a pushbutton has the potential to put an MSIV in a half trip state or close the valve with a resultant plant trip if the right combination of two solenoids are de-energized. Also, a single failure of the no longer used MSIV 90% slow close pushbutton would result in the closure of the MSIV to approximately 90% open. The MSIV would then open slightly then close again to the 90% point repeatedly. The slow close pushbutton was originally used to satisfy ASME testing requirements. Industry experience determined these valves should not be partially stroked at power.

SAFETY EVALUATION

SUMMARY

The removal of the loads and the pushbuttons will enhance the operation of the systems by removing the possibility of the failure of unneeded and unused equipment. The SFAS loads have provided no functional use to the system for several years, yet have had connections to power sources that provide power to other pieces of equipment in the SFAS cabinets. The removal of these loads will ensure that the other components receiving power from the same power supplies as these unnecessary loads will not be degraded due to a failure of the unnecessary loads. The removal of the internals to the digital indicator was reviewed for seismic impact and determined acceptable. The removal of the resistor in the SFAS will allow the use of M&TE that has a verified tolerance.

The MSIVs' pushbutton removal will help ensure continuous power to the solenoids which actuate (close) the MSIV. The SFRCS is only indirectly affected by the changes to the MSIV pushbuttons due to providing the power source to the l solenoid valves. There is no adverse effect on the SFRCS by changing a normally

! closed pushbutton to a continuous wire.

l There is no adverse effect on safety.

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SAFETY EVALUATION

SUMMARY

FOR j MOD 93-0060 (SE 94-0054)

TITLE: l Replacement of Operators in Valves.

CHANGE:

Replaces operators in valves MS603 and MS611.

REASON FOR CHANGE:  !

To ensure that the valves can close against design differential pressure.

SAFETY EVALUATION

SUMMARY

This modification increases the reliability of operation of MS603 and MS611 by significantly increasing the available thrust to close the valve. Additionally, installation of the larger motor ensures that under degraded voltage conditions the operator is capable of developing full rated operator thrust. The change in the control scheme to limit control for closing enhances the reliability of these valves.

Increasing the motor capability will cause the potential stall thrust to be increased. The torque switch is retained in the closing control circuit to protect against motor stall. No failure of the pressure boundary is expected.

The effect of the additional weight of the replacement operator has been analyzed. All piping and supports remain within the allowable load limits.

Support 7EDD-61-H66 is being modified to limit seismic displacement.

There will be no adverse effects on the reliability of the Low voltage System due to the increased operating current of the larger sized motors for MS603 and MS611. All cabling and other distribution components are adequately sized to prevent overheating.

The increased motor size will have no adverse effects on the Low Voltage system in providing adequate voltage to the new valve operator or the other equipment.

Based on the above, it is concluded that the proposed change is safe and does not constitute an unreviewed safety question.

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l SAFETY EVALUATION

SUMMARY

FOR MOD 93-0070 (SE 95-0054)

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TITLE:

1 Addition of Six Baskets for Storage of Trisodium Phosphate I

CHIM15:

I This modification provides for the addition of six baskets for storage of l Trisodium Phosphate (TSP). The proposed changes will allow an increase in the f minimum and maximum boron concentration requirements for the Borated Water I Storage Tank (BWST) , and the minimum boron concentration requirement for the Core Flooding Tanks (CFTs).

REASON FOR CHANGE:

This change will allow flexibility for longer fuel cycles including the upcoming Cycle 11.

SAFETY EVALUATION

SUMMARY

The effects of the baskets on the concrete slab are negligible. The placement and restraint of the baskets will ensure that there will not be a interaction with the Core Flood Injection Line #2, Decay Heat Loop #2, Containment Air Cooler plenum or any other SSC's.

Since the bsstets are made out of stainless steel, they are not capable of i producing post-LOCA hydrogen gas, and thus will not add to the containment pressurization scenario.

The baskets and their attachments do not adversely affect the overall safety of the plant and do not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR MOD 94-0009 (SE 95-0056)

TITLE:

Remove Relief Valves SW10210, SW10211, and SW10212 CHANGE:

4 This modification removed the relief valves (SW 10210, 10211, and 10212) l installed at each Containment Air Cooler (CAC) Service Water return line between containment penetration numbers 9, 10, and 11 and isolation valves SW 1356, SW 1357, and SW 1358. Pressure Taps (PP 1389, 1390, and 1391) were installed at the locations where these relief valves were installed.

. REASON FOR CHANGE:

The thermal relief valves installed to protect the CAC's from overpressurization had a history of seat leakage.

SAFETY EVALUATION

SUMMARY

Pressure taps PP 1389, 1390, and.1391 will be used for performance testing of the CAC's and SWS. The taps will allow measurement of the service water pressure downstream of the CAC's and will be normally isolated by closed valves SW 1389, SW 1390, and SW 1391. The taps will not perform a function important to safety.

t This modification restores a portion of the Service Water 'Jystem to its original configuration. The modification therefore results in a configuration that was evaluated as part of the original plant design.

l The replacement of the relief valves with pressure taps does not affect the classification of this portion of the Service Water System as a closed system.

Also, the pressure taps maintain the SW pressure boundary integrity since they will comply with Specification M-200 and since their isolation valves will be i normally closed.

Removal of the relief valves and installation of pressure taps will not l

adversely affect the operation of CAC or SW systems. The removal of disabling devices and closing of valves SW 1389, SW 1390, and SW 1391 does not impact the system functions. If in the future it is decided to remove the disabling

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l devices from SW 66, SW 67, SW 70, SW 71, SW 74, and/or SW 75, there would be no effect on the system function. During normal operation, pressure protection of the Service Water System is provided by PSV 3962 and PSV 3963 which are set to relieve if system pressure is greater than 129 psig.

Based on the above it is concluded that the change is safe.

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SAFETY EVALUATION

SUMMARY

FOR Limited Mod 94-0020 (SE 95-0042) )

TITLE:  ;

l Pressure Locking Concerns for Valve RCll CHANGE:

This modification drilled a 1/e inch hole in the upstream disc of the wedge for RC11, Pressurizer Pilot Operated Relief Valve (PORV) block valve. The UCN associated with this modification also revised information contained in the USAR regarding RC11.

REASON FOR CHANGE:

Several industry pressure locking events indicated the need for a detailed i review of the susceptibility of safety related Motor Operated Gate valves for l pressure locking. Davis-Besse's initial and subsequent review of these concerns recommended that RCll be modified to alleviate the potential of pressure locking. j i

SAFETY EVALUATION

SUMMARY

Drilling a hole in the upstream disc of the wedge for RCll does not affect the Reactor Coolant Pressure Boundary or the sealing surfaces for the valve in the normal flow direction. If RC11 is closed to isolate flow sealing occurs on the j downstream disc of the wedge against the downstroan seat ring. Since only the  !

upstream disc is drilled, normal isolation wil:. occur. For flow in the opposite direction this is not true. Leakage will occur past the first disc of the wedge and then through the hole drilled in the other disc. For this reason the valve j provides complete isolation in the normal flow direction only. Since RCll is I only required to isolate in the normal flow direction, this is acceptable. The l hole is drilled in the location recommended by the manufacturer and is not i located in the section of the wedge that provides pressure boundary isolation.

As such, drilling a hole in the wedge will not have an adverse impact on plant i safety. The overall effect on safety is the elimination of the possibility of the pressure locking of nell.

The changes to RCll improve the reliability of the PORV relief path and the Feed and Bleed relief path. Elimination of pressure loc king concerns for RCll increases the probability that it will be able to perform its safety functions and that the pressurizer vent can be controlled as required by the operator during both normal operations and feed and bleed.

The changes to USAR Table 5.2-10 reflect that the wedge and the seat material are a grade CF3M material vice a grade CF8 material. The CF3M material has a lower carbon content and is more resistant to stress corrosion cracking. The existing valve stem is constructed from SA564 Type 630 material vice ASTM-A461 Type 630. ASTM-A461 does not have an equivalent SA grade material. Both of the newer materials are acceptable for use in ASME Section III components and were designed by the manufacturer for the use in the intended application. The

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ASME Section III Code Edition that the existing valve was built to has been reconciled against the Draft Pump and Valve code and was found to be acceptable.

USAR Table 5.1-1b lists the shop hydrostatic test pressure for components with the RC Pressure Boundary. The existing valve was shop hydro tested at 5500 psig. The table is being changed to reflect this pressure. Finally, ASME Section III does not require a volumetric examination for valves from 2-4 inch nominal pipe size. The existing valve received a surface examination only.

Meeting the requirements of later Editions of ASME Section III is acceptable for replacements per the requirements of ASME Section XI. The volumetric examination was a requirement imposed by the ASME Draft Pump and Valve Code for all forgings regardless of size. The exception in the later codes reflects the Since j limited value of a volumetric examination for the smaller valve sizes. '

this exception is permitted by the later editions of the Code it is acceptable and a note to reflect this is being added to Table 5.2-14.

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[ SAFETY EVALUATION

SUMMARY

FOR MOD 94-0029 (SE 95-0046, R01) l l

TITLE:

Pressure Regulator PCV 1776 Replacement CHANGE:

l The modification replaced the Reactor Coolant Drain Tank (RCDT) nitrogen gas pressure regulator, PCV 1776, with a non-venting style regulator.

REASON FOR CHANGE:

The former style was a venting regulator. The venting function of the regulator was determined to be a contributor to the Make-Up Tank sampling incident 12/21/93 (radioactive gas release in Auxiliary Building). The vent in the former regulator provided an escape path for the gaseous contents of the RCDT to the Auxiliary Building. Since the new regulator does not have a similar vent path, the likelihood of the contents of the RCDT escaping via the regulator is greatly reduced.

SAFETY EVALUATION

SUMMARY

The change has no effect on the function of the Reactor Coolant Drain Tank and Containment Vent Header System. The new regulator performs the same existing ,

function under the same technical requirements as the existing regulator. The l nitrogen blanket will be maintained as necessary during implementation of this change with a portable nitrogen bottle temporarily connected to the RCDT.

The change has no effect on the safety function of the Auxiliary Building Radioactive Area HVAC System. This change eliminates the connection (vent line) from the regulator to the Auxiliary Building Radioactive Area HVAC System, therefore, any effect would amount to reducing the load on the Auxiliary Building Radioactive Area HVAC System.

This change has no effect on the safety function of the Nitrogen Gas System.

The new regulator performs the same existing function under the same technical requirements as the prior regulator. The containment isolation valves in the Nitrogen Gas System are not affected by the change.

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SAFETY EVALUATION

SUMMARY

FOR LIMITED MODIFICATION 94-0032-00 (SE 95-0070 R01)  ;

TITLE:

Containment Vessel Base Embedment Sand Removal CHANGE:

This modification: 1) Permanently removed sand from the pocket area at the vessel base embedment; 2) Applied a protective coating on the steel vessel to prevent future corrosion; 3) Installed a check valve in the drain line to ECCS Sump #1 and adds a positive closure device in this line to establish a boundary  ;

for the negative pressure area in room 105.

i REASON FOR CHANGE:

To address the problem of accelerated corrosion on the steel containment vessel near the base embedment. Drainage problems and the presence of moist sand against the vessel wall have been identified as contributing factors to corrosion found on the vessel.

SAFETY EVALUATION

SUMMARY

Annulus Drains: ,

The removal of the sand will have a negligible effect on the seismic adequacy of the annulus drain piping within the Shield Building.

The removal of the sand will uncover the drain to Auxiliary (Aux) Building Sump l

1. 1 which will provide a potential flowpath past the negative pressure boundary. ,

The installation of the check valve in the Aux Building Sump line will allow  !

normal drainage but will not allow the reverse flow of air during EVS drawdown. ,

This check valve then re-establishes the airflow boundary.  !

Containment Vessel:

The surface preparation and coating application will be performed in accordance with existing criteria. The vessel integrity is not adversely affected by this work and therefore, the function of the Containment Vessel and the requirements of Tech Spec 3/4.6.1, Primary Containment, will be maintained.

The original design of the Containment Vessel included the embedment of the l vessel into a continuous concrete fill under the ellipsoidal bottom head. At the termination point of the concrete, a " transition zone" was included to j minimize the local shell stresses. The removal of the sand will have an effect ,

on the localized shell stresses, however the effects will be negligible.  !

The limiting analyzed transient for Containment Vessel stress is a Large Break Loss of Coolant Accident (LBLOCA). During a LBLOCA, the peak pressure of approximately 36 psi occurs at 60 seconds. Because of the dominance of pressure I

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1 induced stress over the other contributors, the analysis was performed only at the time of peak pressure. Due to the large difference in conductivity between steel and sand, and the high heat transfer rate from liquid water (or condensing steam) to the steel vessel, the presence or absence of sand will have minimal effect on the Containment Vessel temperature distribution at 60 seconds. At later times, when the removal of the sand could have a small effect, the overall stress will be lower due to a large reduction in pressure induced stress.

l It was also determined that with the sand removed the vessel wall flexibility would increase slightly allowing for a larger volume inside containment during a pressure transient. The effect on the Annulus volume would be partially offset by the removal of the sand volume from the annulus. The small incremental amount of change with respect to the total volumes of the containment Vessel and Annulus will be negligible. It was concluded that this will have an insignificant impact on existing high energy line break evaluations and/ or the drawdown time of the Emergency Ventilation System.

Shield Building:

This Mod has no effects on the Shield Building or the Shield Building safety functions.

Annulus Sand Pocket:

The removal of the sand does not adversely affect the drainage capability, but in fact will enhance it. The drains could be more susceptible to plugging of the perforated strainers by debris washing into the open pocket. However, due to the configuration of the strainers, the location of the two drains, the low volume of water discharged through the drains, and the controlled access of the annulus area, the probability of significant flooding of the open sand pocket is negligible.

Based on the above, this change does not adversely affect plant safety nor does it create an unreviewed safety question.

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l SAFETY EVALUATION

SUMMARY

FOR MOD 95-0001 (SE 95-0060)

TITLE:

10 RFO Fuel Repairs CHANGE:

The repairs are accomplished by the removal of defective fuel rods in a controlled process.

REASON FOR CHANGE:

Fuel repairs are performed to allow extraction of remaining energy from defective fuel assemblies that have no structural damage.

SAFETY EVALUATION

SUMMARY

Prior to the commencement of the repair process, a defective fuel assembly (FA) is raised onto a pedestal, which is inserted into a SFP storage cell. The top of the FA itself will then be covered by approximately 22 1/2 feet of water during the repair. Because of the raising of the defective FA onto the pedestal, during a fuel rod's extraction, 8 feet of water may result. The effect of this reduction in water depth for a single fuel rod is offset by a greater reduction in specific activity. That is, one fuel rod with 8 feet of biological shielding has a much lower specific activity which results in a lower radiation field than 208 fuel rods in an FA normally handled with 9 1/2 feet of water.

Based on a conservative thermal analysis by BWFC, fuel repairs shall be performed only after a minimum cooling period or time after reactor shut down of 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> and below or at the administratively controlled ambient Spent Fuel Pool temperature limit of 120 degrees F. This will ensure that for up to 4.2 w/o enriched fuel, no thermal limits are reached with the present pedestal design.

During the fuel repair the amount of water depth available for iodine removal

(" scrubbing") in the event of a rod break would vary depending on the location of the break. If it is conservatively assumed that the rod breaks after it has been fully extracted from the FA, ie. at the upper end cap, the available water  !

depth could be 8 feet. In such a case, the iodine scrubbing by the SFP water '

would be less than is assumed in the Fuel Handling Accident (FHA). However, this postulated rod breakage incident is less severe than the FHA analyzed in ,

the USAR.

]

The radiological consequences for dropping or breaking a fuel rod are bounded by the Fuel Handling Accident (Outside Containment). That accident assumes that all of the peripheral fuel rods in an FA, 56 out of 208, suffer mechanical damage to the cladding. All of the released noble gases are assumed to leave the

SFP and the iodine gap activity is released to the SFP. The FHA analysis shows that the environmental consequences satisfy the acceptance criterion of being less than the guidelines given in NUREG-0800, the Standard Review Plan.

Since the fuel repairs will be performed in the SFP with the Auxiliary Building's ventilation system operational and the Emergency Ventilation System (EVS) operable, any releases that result in EVS activation will be discharged through high efficiency charcoal filters to the station vent. Therefore, releases to the environment resulting from a broken fuel rod will be negligible.

In conclusion, there is no effect on safety, since this fuel repair is i controlled with proven procedures, performed by BWFC personnel with fuel handling experience, uses tested fuel handling tools, quality components, and repair techniques and hardware. The repaired fuel will meet all of the requirements of the BWFC Field Change Authorization (FCA). FCAs document compliance with the BWFC Quality Assurance Program for Fuel Design Control, l which ensures that the repaired fuel assemblies meet all fuel design l requirements. The modification consisting of repair by recaging and/or i reconstitution based on the above is, therefore, considered safe and does not I constitute an unreviewed safety question, j l

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- SAFETY EVALUATION

SUMMARY

j FOR MOD 95-0011 (SE 95-0041) l .,ITLE.

Resetting the Second Main Feedwater (MFW) Pump Turbine in Mode 1 While

' Maintaining ARTS Trip Capability for Loss of MFW Pumps.

I CHANGE:

Modify Plant to ensure that the Anticipatory Reactor Trip System (ARTS) will detect a loss of both main feedwater pump turbines (MFPTs) on a loss of one MFPT

if only one is available.

4 REASON FOR CHANGE:

When a Main Feedwater Pump Turbine (MFPT) is in a " reset" (not tripped) condition, the hydraulic control oil system is pressurized. The ARTS senses hydraulic pressure to determine whether when both MFPTs are tripped. If one MFPT is operating and the'other is not available but is reset, the ARTS loss of' both MFPTs trip will not be activated if the single operating MFPT should trip.

SAFETY EVALUATION

SUMMARY

This modification installed a test toggle switch associated with each of these main feedwater pump turbine oil pressure switches. The administratively controlled toggle switch simulates a trip condition to the logic when the respective main feedwater pump turbine is not tripped yet not providing flow to the steam generator, such as during plant startup.

The replacement switches will meet applicable requirements of IEEE Standard 279-1971 as specified in USAR sections 7.4.1.4 under which the existing ARTS was installed. When placed in the test position the switches would provide a continuous MFW pump trip status to ARTS. The switches would be placed in this position for a MFW pump turbine which is reset, but not available. Thus, loss of the opposite MFW pump would result in the desired ARTS trip. IEEE 279 requires visible indication of test bypass switch position. However, since the input test switches initiate (rather than prevent) actuation of a protective trip function, position indication is not required. Failure to reposition the switches after restoration to two MFW pump operation would have a potential for causing inadvertent reactor trips. However, administrative controls and existing locked cabinet doors will adequately control the switch position and ibnit- unauthorized or inadvertent switch repositioning. The correct operation  ;

of the switches will be periodically verified by successful completion of ARTS channel functional testing. Since multiple failures of~ components would be l required to impact the 2 out of 4 trip function, single failure analysis of the ARTS is not affected.

Based on the evaluation of operation with the ARTS MFPT control oil pressure switches isolated and vented (or use of a test switch trip signal) for an unavailable MFW pump is safe and does not involve an unreviewed safety question.

SAFETY EVALUATION SUMI'.ARY FOR j LIMITED MODIFICATION 95-0012 (SE 96-0049) l 1

l TITLE:

Removal of Unnecessary Wiring in C3615 and C3616. ]

l CHANGE:

The removal of extraneous electrical conductors in the train one and train two j emergency diesel generator protective relay cabinets, C3615 and C3616, respectively.

l REASON FOR CHANGE:

The unnecessary conductors in C3615 were discovered while attempting a tagout for control switch CS/AC110. During design development for Limited Hodification 95-0012, a review of the wiring associated with control switch CS/AD110 in C3616 revealed similar unnecessary wiring. The removal of the extraneous wiring affects USAR Figure 8.3-8 and Drawings E34B Sheets 6 and 12.

SAFETY EVALUATION

SUMMARY

The train 1 (train 2) extraneous conductor being removed, take 125V DC trip circuit control power from alternate source breakers ABDC1 (AACD1) pass it through the third normally closed contact pair on control switches CS/AC110 (CS/AD110) then end in an open circuit after sending the 125 V DC control "oltage back to breakers ABDC1 (AACD1). ,

Removing the extraneous wiring and revising the associated drawings will j facilitate any future work involving these switches by clearly showing contact  !

pair number three as spare. Also, removing the extraneous conductors foraing  !

the open circuit will limit the exposure to faults on the alternate source l breakers' trip circuits. I Based on the above discussion, it is determined that the proposed change is safe and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR Mod 95-0014 (SE95-0066) l

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TITLE:

Alternate Non-1E 480VAC Supplies For Containment Lighting CHANGE:

}'

The containment lighting is powered by redundant IE 480VAC sources. Limited Modification 95-0014 added the capability of supplying these lighting loads from 480VAC non-1E sources. These non-1E sources will be used during plant modes 5 or 6 (usually a refueling outage) when the normal source is deenergized for maintenance.

REASON FOR CHANGE:

During prior outages temporary modifications were performed to power containment lighting from non-1E sources. Due to safety conern that arose during the last outage it was decided a permanent change was required.

SAFETY EVALUATION

SUMMARY

The non-1E sources will not affect the existing IE 480V AC electrical coordination because the non-lE/lE connection is on the load side of a 1E i molded case circuit breaker and this 1E circuit breaker must be opened before the non-1E source molded case breaker can be closed.

Addition of the temporary power cable receptacles atop MCCs EllC and Fila has been reviewed to ensure that environmental and seismic qualifications have not

been degraded.

The molded case circuit breakers selected to connect containment lighting the non-1E source have been evaluated to ensure that any electrical faults are cleared prior to exceeding the withstand rating of electrical penetrations PAP 2P or PBP5D. Should these breakers fail, the trip settings of upstream breakers in unit substations E2, F2 or F6 will clear the fault prior to exceeding the penetrations' withstand rating.

Fire barriers will not be breached or blocked open by the temporary cable because the MCCs being connected are located in the same room. The extra fire i I

loading caused by the temporary cablas in rooms 304 or 427 is controlled as a permanent transient in accordance with the combustible loading procedure.

The length of temporary cable needed to connect the MCCs is approximately seventy feet. These cables will be routed overhead, secured to existing raceway / raceway supports. The routing will not interfere with the operation of any plant controls or with access to or egress from these areas. The temporary cable plugs and receptacles will be grounded to guard against electrical shock.

Based on the above evaluation, the changes made by Limited Modification 95-0014 are safe.

SAFETY EVALUATION

SUMMARY

FOR MOD 95-0055 (SE 96-0004)

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l TITLE:

Convert Room 302 to a Hot Shop.

CHANGE:

Modification 95-0055 converts Room 302 (Drumming Station) of the Auxiliary Building to a Hot Shop. Part of this modification involved converting some piping from Primary' Water to Demineralized Water. A common demineralized /

primary water line contained a flow element (FE2158).which was removed and replaced with a blank plate. The internals of check valve WM19 were removed and the normal positions of valves DW13, PW15 and PW55 were changed from

" closed" to "open". This allows Demineralized Water to flow from the Domineralized Water header through valves DW13, WM172, WM19 and PW15 to the former Primary Water header and then to the Hot Shop. The former Primary Water-header was isolated from the remainder of the Primary Water System by inserting a blank plate between two pipe flanges. This arrangement allows the seismic qualification of the line to be maintained without installing any additiona) pipe supports. Valves PW16, PW31, PW56, PW28, PW29, PW36, PW17, PW3, PW32, PW58 and PW34 were closed / verified closed and tagged to maintain the branch lines from the new Demineralized Water (former Primary Water) header isolated.

REASON FOR CHANGE:

This modification allows work to be perfomed on contaminated equipment within the Radiological Controlled Area.

SAFETY EVALUATION

SUMMARY

The Demineralized Water System continues to be able to detect a pipe break in the system, including a break in the piping converted from Primary Water to Demineralized Water use, and close valve DW6880.

This modification involves removal of some piping that passes through the wall between Rooms 301 and 302. This wall forms part of the Fuel Handling Area Negative Pressure Boundary. The penetrations will be resealed after routing of new conduits through them; therefore the pressure boundary integrity will be restored. While any of the penetration are breached the action statement for Technical Specification 3.9.12 will apply. Therefore, fuel handling activities and use of the crane over the spent fuel pool will not be permitted during the time the barrier is breached. This will ensure that the Emergency Ventilation System functions important to safety are not affected.

The USAR identifies that portions of.the Demineralized Water System have been ,

l upgraded to Seismic Class I and that an isolation valve is provided which  !

closes on low system pressure to protect from flooding in the event of a rupture of the non-Seismic Class I portion of the system. These features will j

remain intact following implementation of this modification. l

SAFETY EVALUATION

SUMMARY

j FOR MOD 95-0060 (SE 96-0022, R.01)

TITLE:

Auxiliary Feed Pump Turbine Main Steam Heat Recovery Line CHANGE:

During the 10th operating cycle, trap bypasses around Steam Trap (ST) ST133 and ST134 were opened to reduce disc tapping that has been occurring in MS734 and MS735. Modification 95-0060 diverts steam flow that is currently being bypassed to the condenser to feedwater heater E6-2.

REASON FOR CHANGE:

The additional steam flow through the valves lifts the discs off the seats, reducing the number and magnitude of the impacts.

SAFETY EVALUATION

SUMMARY

The new lines will be seismically installed in the AFW pump rooms. As such, the lines will not become seismic hazards for existing installed seismic class 1 components. Any breaks or cracks in these new high energy lines are bounded by the AFP room 6" line breaks evaluated in the USAR. Walkdowns were conducted in the AFP rooms to ensure that the new piping installation would not create any new targets or hazards of other types. Pipe whip was specifically reviewed to ensure that there were no opposite train targets. Safe shutdown circuits were reviewed to ensure piping breaks and the resulting jet-effects will not affect the opposite train of AFW.

The new piping will be designed and constructed to the requirements of ASME/ ANSI B31.1 Power Piping. The pressure and temperature rating of this line will be 4

the same as its main steam supply line. Valves and fittings in this line will meet these pressure and temperature requirements. The new 1 1/2" valves meet the requirements of ASME Section III for class 3 components and establish the "O" boundary for the new piping.

The penetration will be sealed to ensure that the fire, security, and flooding barrier integrity is maintained. Sinca the penetration will be grouted, there is no longer a requirement to have a penetration flood dam installed.

The new 1 1/2" piping will be insulated and will have an insignificant effect on the heat loading of the AFP rooms.

The modification will have no adverse effect on the safety functions of the affected SSCs and will be designed, constructed and installed in accordance with the established standards for these SSCs. It is, therefore, safe.

SAFETY EVALUATION

SUMMARY

FOR MOD 95-0062 (SE 96-0012)

TITLE:

Rescale Steam Generator Startup Range Level Strings CHANGE:

This modification changes the Steam and Feedwater Rupture Control System (SFRCS) Steam Generator Startup Range level instruments from the previously calibrated range of 0-250" to a new calibration range of 0-300".

REASON FOR CHANGE:

The change was required due to the increased fouling of the Steam Generators (SG). As the SG fouls, the indicated startup, operate and full range levels increase. This modification only changed the Startup Range transmitters and associated strings. The operate and full range transmitters were already calibrated for the complete range available from tap to tap, therefore, those strings did not need to be changed. Only the Startup Range had a reduced calibration span.

There are two types of otartup range transmitter strings to be considered, the Steam and Feedwater Rupture Control System (SFRCS) strings and the non-SFRCS strings. The non-SFRCS strings were not changed.

SAFETY EVALUATION

SUMMARY

)

The SFRCS low level string will still trip prior to the Tech spec values as determined by the revised calculation C-ICE-083.03-001. The new error when combined with all of the other errors results in an increased instrument error ,

of only 0.04". The " extra" 0.58 inch margin, included in the calculation by l Design Engineering, will be reduced by this change. Compared to the extra l margin added, 0.04" is not significant. The Safety Limit and the Tech Spec j values are still vell protected after taking into account the instrument errors and the remaining margin. The change is therefore considered acceptable.

The AFW system level control will still control at the required levels. Since the level control range will not be changed, there is no adverse effect on safety. The possibility of the level controller being off-scale high was reviewed and is considered acceptable.

I The Auxiliary Shutdown Panel (ASP) level indicators have no controlling function. Since there is no controlling function and the only time that the ASP is manned is post trip, when the level would be on-scale, not changing the range is considered acceptable.

The Post Accident Monitoring Panel indicators are similar to the Auxiliary Shutdown Panel in that there are no controlling functions from these indicators. These indications originate from the SFRCS transmitters and will be changed. There is a chance for additional error displayed by the

indicators. Similar to the SFRCS trip setpoint, the error is small and will be indistinguishable to the operator. Therefore, the change is considered acceptable.

The SPDS/ Plant Computer could be off-scale high. The SPDS/ Plant Computer will continue to read 250" even if the level is above 250". Since these are not used at 100% power for operator interface, there is no adverse effect by the indications being off-scale. When the levels have decreased below 250", the transmitters will respond normally. Therefore, not changing the range is considered acceptable.

The NNI/ICS/ SASS has a controlling function at 40". If the transmitters are off-scale high, the ICS will simply close the controls valves. With respect to instrument wind-up (instruments going into saturation), the boil down of the water in the steam generator will allow sufficient time for the instruments to come out of saturation. Therefore, not changing the range is considered acceptable.

For the non-SFRCS transmitters in general, being off-scale high has no adverse effect as demonstrated by putting the Steam Generators in wet lay-up. Not changing the range is therefore considered acceptable.

Based upon the above, there is no adverse effect on safety.

SAFETY EVALUATION

SUMMARY

FOR MOD 96-0002 (SE 96-0018)

TITLE:

Larger Motor Installation for Valve MU2B CHANGE:

This modification installs a larger motor on MU2B, the normally open Make-up and Purification letdown isolation valve, to increase the available motor operator output thrust.

REASON FOR CHANGE:

This modification ensures that there is sufficient margin between the available motor operator output thrust and the required operating thrust to close MU2B against design differential pressure.

SAFETY EVALUATION

SUMMARY

This modification increases the reliability of operation of MU2B by significantly increasing the available thrust to operate the valve. This is accomplished by installing a larger motor.

Increasing the motor capability will cause the potential stall thrust to be increased. The torque switch is bypassed in the closing control circuit until MU2B is > 97 % closed. If stall occurs, some yielding of the valve seat may occur but the integrity of the valve pressure boundary will not be affected.

The effect of the additional weight of the replacement motor has been analyzed.

All piping and supports remain within the allowable load limits. MU2B remains seismically qualified with the new motor installed.

There will be no adverse effects on the reliability of the Low Voltage System due to the increased operating current of the larger sized motor for MU2B. A review of the existing molded case circuit breaker and power cable for MVMUO2B concludes that the existing cable is adequately sized for the replacement motor and the existing molded case circuit breaker will easily pass the motor starting current.

The increased motor size will have no adverse effects on the Low Voltage System in providing adequate voltage to the new valve operator or to other equipment.

A voltage drop calculation has been completed indicating that sufficient terminal voltage will be available at MU2B during the worst case accident scenarios to ensure that the valve can be stroked.

Based on the above discussion, it is concluded that the proposed changes are safe.

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SAFETY EVALUATION

SUMMARY

FOR MOD 96-0003 (SE 96-0021)

TITLE:

Replace Motor Operators for Valves FW601 and FW612 )

CHANGE:

1 Modification 96-0003 replaced the existing 200 ft-lb, 3600 RPM motors on FW601 and FW612, the Steam Generator Main Feedwater isolation valves, with 300 ft-lb, 3600 RPM motors to increase the available motor output thrust. Since the motor replacement could significantly increase the inertial energy of the valve components after limit switch trip, Modification 96-0003 also increased the packing gland loading and live-load the valve packing.

REASON FOR CHANGE:

Modification 96-0003 was initiated to address the issue concerning the use of pullout efficiency versus run efficiency for motor operator gearbox performance when calculating the capability of motor operated valves.

SAFETY EVALUATION

SUMMARY

Engineering calculations were performed to evaluate the capability of the motor operators with a 300 ft-lb motor installed for FW601 and FW612. These analysis show that with the replacement motors installed, the motor operator thrust evaluated using pullout efficiency and with the increase packing friction will provide an adequate margin above the required thrust for FW601 and for FW612.

Replacing the existing motor with 300 ft-lb motor will increase the motor stall i torque. Calculations evaluate the effects of increased stall thrust and torque on the operator and the valve. With the replacement motor the stall thrust and torque is expected to be within the one time limits of the actuator components.

However, the estimated stall thrust significantly exceeds the one time limits of various valve components in the opening and the closing direction. If stall occurs while closing, it is possible that yielding of the stem or the yoke could occur. If the motor stall occurred while attempting to unseat FW601 or FW612, separation of the stem and wedge could occur. The possibility of damage to valve components if stall occurs is not a new condition created by this modification since the existing motor also can produce a stall thrust in excess of valve component limits. Also, for stall conditions to occur a valve or actuator malfunction must already have occurred. In either direction the possibility of additional damage to valve components will not affect the ability of these valves perform their safety function, which is to close.

Additionally, the stall thrust is within the ene time limits of all valve components that constitute the pressure boundary, therefore the integrity of the pressure boundary would be expected to be maintained if a stall condition occurred with the larger motors.

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The heavier 300 ft-lb motor increases the overall actuator weight and affects I the operator center of gravity. The changes in weight and center of gravity j cause the seismic thrust limit to decrease. The revised seismic thrust limit i

may be exceeded during the closing stroke and after the valves are in the '

closed position, depending on the differential pressure conditions. However, exceeding the seismic limit during valve closure or while.the valve is closed will not prevent the valves from performing any of their safety functions. t

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During normal operating conditions FW601 and FW612 are open so the seismic  !

limit would not be applicable. During valve stroke the seismic limit would <

only be exceeded if a significant differential pressure exists. This magnitude j of a differential pressure would only occur in a condition where the steam '

generator is depressurized. However, a seismic event is not postulated l concurrent with a line break event that would depressurize a steam generator, '

therefore under no postulated conditions will a seismic event cause a valve  ;

component failure during the valve stroke. Depending on the differential j pressure conditions during closing, the seismic thrust limit may be exceeded I while FW601 and FW612 are in the closed position. In this case a subsequent I seismic event could cause the limit to be exceeded. The specific seismic limit '

that would be exceeded is the allowable stress in the yoke arms. However, '

exceeding this limit will not affect the safety functions of the valves because these valves have no safety function to open and exceeding an allowable stress limit in the yoke arms will have no effect on the valves ability to maintain containment isolation if the valve is closed.

Replacing the existing 200 ft-lb motors with larger 300 ft-lb motors will increase the valve inertia that exists after limit switch trip. However, these valves are currently setup in a limit switch control scheme that minimizes the thrust that the valve and operator components are subjected to when seating.

Therefore, the total thrust will be maintained within the valve and operator continuous service limits for all operation. Additionally, since packing friction is essential in controlling inertzal overshoot for valves with this type of limit control setup, this modification will live-load the packing.

Live-loading of the packing will maintain the packing friction relatively constant over packing service life, but will not create any unexpected friction load on the valve that could impede closing operation.

The effect of the additional weight of the larger replacement operator has been analyzed. All piping and supports remain within the allowable load limits.

Technical Specifications specifies a response time of < 16 seconds for Main Feedwater Stop valves. However, this modification will not have a measurable effect on the stroke time of FW601 and FW612. There will be no changes to the gearing implemented by this modification and the new larger motor will be able to deliver a given torque with less of a reduction in speed thereby slightly decreasing stroke time at fully loaded conditions. The proposed live-loaded packing design will slightly increase the load at design conditions but the increased packing load only represents five percent of the stem thrust required at design conditions. Therefore, this added load will have a negligible effect on valve' stroke time.

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j SAFETY EVALUATION

SUMMARY

FOR j MOD 96-0006 and FPR 96-0006-001 (SE 96-0026)  ;

i TITLES- 1 l

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, ' Installation of a Vent and an Elbow on the High Pressure Injection System f

$ CHANGE:-

i l This modification installed a vent in the High Pressure Injection line  !

, downstream of valve HP2A and revised the vent by adding a forty five degree i elbow between the branch point at the 2 1/2 inch line and the first isolation ,

valve. The vent has two isolation valves (HP40 and HP40A) and a threaded pipe l

cap. 5 I' f REASON FOR CHANGE

i

, This modification is a corrective action to LER 96-001. l l SAFETY EVALUATION

SUMMARY

4 This modification provides a means to vent High Pressure Injection line 2-1 to )

ensure that the line is filled with water. This enhances the ability _to ensure  ;

that the HPI can perform its functions important to. safety. The vent follows (

all requirements set by ASME codes. It is identical to the other vents on the l other three HPI lines except its first isolation valve is at a higher elevation I than the adjacent tee branch, an elbow has been added after the second isolation valve to make it easier to attach a hose when using the vent, and it is oriented 45 degrees from vertical to avoid interference with an existing electrical conduit.

Based on the above it is concluded that the proposed action is safe and does not constitute an unreviewed safety question.

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( SAFETY EVALUATION

SUMMARY

l FOR MWO 1-96-0315 (SE 96-0037)

TITLE:

Installing a Camera in the Annulus CHANGE:

I Maintenance Work Order 1-96-0315-00 installed a remotely monitored and controlled video camera and lighting in the annular space between the Shield {

l Building and the Containment Vessel.

REASON POR CHANGE:

The purpose of this equipment is to allow remote observation of cables ,

associated with the safe shutdown of the plant in the event of a serious .

I station fire. The monitor will allow detection and response to a fire which could threaten the integrity of these cables. The particular cables affected l

are protected by Thermolag. This material has been found to be ineffective in l meeting.its design function. Therefore, until an alternate method of providing protection is implemented compensatory measures are required. This is provided i by routinely monitoring the area for fires. Because the Annulus is not accessible during power operation, a method of remote monitoring was developed. l SAFETY EVALUATION

SUMMARY

The penetrations passing between the containment and the Shield Building have been evaluated for acceptability during and following a seismic event.

. Installation of the camera was evaluated and found to not create any concern regarding the seismic design of the Shield Building or the various penetrations in the vicinity of the new equipment. Consequently, there is no effect on the seismic design of the plant.

The heat load added by the continuously energized lights will not cause a safety concern. There is no upper limit on the Annulus temperature. The only limit is on the lower temperature of the annulus to preclude the Containment vessel from reaching its design temperature of 30 degrees Fahrenheit. Consequently there is no safety concern from the added heat load in the Annulus.

.The ability to detect a fire to safely shutdown the plant is maintained by providing a method to monitor the Annulus. These cables are protected by Thermolag, which may not prevent the cables from being damaged due to a fire.

! By providing a monitoring system, damage is prevented by early detection and response to a fire. This has been reviewed and accepted by the NRC as an acceptable compensatory action. Fire Protection personnel have evaluated the additional fire load effect on the response of the Annulus to a fire and found j

that there is not significant impact on the fire area. Consequently, there is no

' effect on the safety of the plant during and following a serious station fire.

l Therefore, it is concluded that the proposed change is safe.

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SAFETY EVALUATION

SUMMARY

FOR ,

MWO 7 0253-14, 7-91-0253-15, and 7-91-0253-16 (SE 95-0049)

(:

! 'TITLEr '

f

-Temporary Disabling of All Service Water (SW) Strainer Blowdown Valves. I t i 1

CHANGE:

4 l

To allow the temporary closing of all the SW strainer blowdown valves (SW1379,  :

SW1380, SW 1381).

REASON FOR CHANGE: i The bonnet of the SW strainer blowdown valves will be removed so that the stem l material can be changed. ,

SAFETY EVALUATION

SUMMARY

The service water strainers are provided to filter any material which might plug the downstream components and heat exchangers. The strainer blowdown valves are ASME section III, Class 3 pressure boundaries and serve to discharge purging water from the strainer when the strainer is actu&ted due to high differential pressure. However, none of the SW strainers routinely actuate based on differential pressure, and it would require more than several days to develop-high differential pressure due to strainer clogging. The valves also open on high dicharge pressure to prevent lifting of the service water relief valves.

In the event of low service water flow demand, service water discharge pressure will-increase to the setpoint of the high SW pump discharge pressure switches.

These setpoints will normally cause the strainer blowdown valve to open before relief valves SW-3962 and SW-3963 actuate.

The diversion of water through the strainer blowdown valves provides minimum j flow protection for the SW pumps, while not requiring lifting of the relief valves. The SW pump shutoff head is sufficiently high that the relief valves will lift if required for minimum flow protection. It is noted that lifting of j the relief valves is reliable, safe, and adequate but is not desirable from a '

maintenance standpoint because the valves will wear and may develop increased seat leakage.

It is not anticipated that a high discharge presrure will occur during the period while the strainer blowdown valves are no: available because work will be l

performed during warm summer months, when system pressure is lower due to  !

increased cooling flows. Based on the above, the strainer blowdown valves are desired, but not currently necessary for minimum flow protection and the proposed work to be performed on SW-1379, SW-1380, SW-1381 is safe and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR PAT 94-0793 (SE 95-0034)

TITLE:

Main Turbine Stop Valve Testing Above 85%* Power.

CHANGE:

Raise the power level when periodically performing the Main Turbine Stop Valve (MTSV) Stroke Test.

.EEASON FOR CHANGE:

It is desirable to conduct this test at a higher power level, because it minimizes the required power reduction, thereby minimizing the RCS activity changes due to iodine spiking, which occurs during power reduction.

SAFETY EVALUATION

SUMMARY

i Raising the upper power limit on MTSV testing from 85% to 90% power will have no effect on the MTSV function of safety. The test will continue to verify that the valves will stroke closed. The steam-flow rate through the valve is not relevant to this capability. The power difference could affect the steam generator tube integrity due to the increased steam flow in the steam generator on the untested side. Recent MTSV tests at other B&W facilities were initiated from 95-96% power, experienced momentary steam flow surges of 6-8%. The MTSV tests initiated at DBNPS from 85% or slightly less have had steam flow increases of approximately 5.4%. This corresponds well with other plants' experiences.

It is expected that the steam generator will remain below the design steam flow, when starting from 90% power or less. This keeps steam flow within the design basis and provides significant margin to B&W's evaluated limit to prevent flow induced vibrations.

Of particular concern with regard to flow induced vibration is the effect of the collapsed internal AFW header on the steam velocities in the upper span of the tubes. It has been determined, through visual inspection and eddy current testing, that the gaps between the header and the nearest tubes is greater than the tube to tube gap. This means that the local radial or axial steam velocities will not be greater than in other regions of the tube bundle.

Therefore, the flow in the outlet gap is unchanged from the as-built condition.

The increase in the initial power level for the MTSV testing will have a slightly greater effect on the RCS temperature. The cold leg temperatures will drop slightly more opposite the side being tested and increase slightly more on the side being tested. This causes a larger temperature gradient entering the reactor. This will not impose any greater duty cycle on the fuel rod cladding than that caused by routine CRA and APSR positioning.

Based on the above it is concluded that the proposed change does not have an adverse effect on safety and does not constitute an unreviewed safety question.

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l SAFETY EVALUATION

SUMMARY

FOR l l PAT 95-1359 (SE 95-0057) i I

TITLE: ,

Main Turbine Stop Valve Testing at 93 Percent Power I l CHANGE:  ;

Perform a restricted change to procedure DB-SS-04150, Main Turbine Stop Valve (MTSV) Stroke Test, to perform this test one time at less than 93% power. ,

REASON FOR CHANGE: )

It is desirable to conduct this test at higher power level, because it ,

minimizes the required power reduction, thereby reducing the RCS activity l changes due to iodine spiking, which occurs during power reduction.

SAFETY EVALUATION

SUMMARY

Raising the upper power limit on MTSV testing from 85% to any power level below 93% power will have no effect on the MTSV function important to safety. The test will continue to verify that the valves will stroke closed. The steam flow rate through the valve is not relevant to this capability. Therefore, there is no effect on safety.

1 The change from 85% to less than 93% power could affect the steam generator tube integrity due to the increased steam flow in the steam generator on the untested side. B&W has evaluated the flow induced vibration potential in the steam generator, including the effects of the collapsed internal Auxiliary Feedwater Headers in the Davis-Besse Steam Generators. The 85% limit is based on a 15% assumed flow surge during MTSV testing which ensures that the maximum flow during the test does not exceed the design flow of a steam generator.

Based on two previous MTSV tests performed at Davis-Besse, a maximum feedwater flow increase of 6.5% is predicted for testing at 93% power resulting in a  !

maximum feedwater flow rate of 99.5%. Consequently, it is expected that the steam generator will remain below the design flow rate, when starting from less than 93% power. As discussed in a B&W letter, momentary steam flow surges above 100% equivalent power steam flow are acceptable for limited periods of time. This will ensure flow induced vibration does not occur.

Of particular concern with regard to flow induced vibration is the effect of  !

the collapsed internal AFW header on the steam velocities in the upper span of the tubes. It has been determined, through visual inspection and eddy current testing, that the gaps between the header and the nearest tubes is greater than l the tube to tube gap. This means that the local radial or axial steam velocities will not be greater than in other regions of the tube bundle. Also, the gap between the top of the header and the upper tube sheet has not changed from its original value. Therefore, the flow in the outlet gap is unchanged arom the as-built condition. This means the steam velocity is within the sfiginal design.

Dased on the above, it is concluded that the steam generator will be operated below its design capacity throughout the test and there are no localized flow restrictions which could cause an early onset of flow induced vibration.

Consequently, there is no effect on plant safety.

The increase in the initial power level for the HTSV testing will have a slightly greater effect on the RCS temperature. The cold leg temperatures will drop slightly more opposite the side being tested and increase sligntly more on the side being tested. This causes a larger temperature gradient entering the reactor. D&W has evaluated this condition. It was concluded this temperature tilt will not impose any greater duty cycle on the fuel rod cladding than that caused by routine CRA r.nd APSR positioning. Consequently, it is concluded that raining the MTSV test initial power level from 85% to less than 93% will not have an unsafe effect on the reactor.

The change from 85% *.a less than 93% power for HTSV testing and the changes in feedwater and stear. flow during testing will have no effect on the Main Feedwater and Mais Steam System functions important to safety.

Based on the above, it is concluded that the change does not have an adverse effect on safety.

SAFETY EVALUATION

SUMMARY

FOR  !

.DB-SS-04150 change 1, UCN 96-050, and UCN 96-053

{

(SE 96-0051 and SE 96-0052) l i

TITLE:

1 Main Turbine Stop Valve Testing at 93% Power and Main Turbine Control Valve j Testing at 96% Power '

CHANGE:

Change the power limitation of 85% during Main Turbine Stop Valve (MTSV) Testing and Main Turbine Control Valve (MTCV) Testing to 93% and 96% respectively.  ;

REASON FOR CHANGE:

l It is desirable to conduct these tests at a higher power level, because it minimizes the required power reduction, thereby minimizing the RCS activity l changes due to lodine spiking, which occurs during power reduction.

)

J SAFETY EVALUATION

SUMMARY

Raising the upper power limit on MTSV testing from 85% to any power level below 93% power will have no effect on the MTSV reliability. The test will continue to verify that the valves will stroke closed. The steam flow rate through the valve is not relevant to this capability.

Through tests performed at Oconee and Three Mile Island and evaluations performed by B&W, it is concluded that the steam generator will be operated below its design capacity throughout the test and there are no localized flow restrictions which could cause an early onset of flow induced vibration.

The increase in the initial power level-for the MTSV testing will have a small effect on the RCS temperature and the changes in feedwater and steam flow during testing will have no effect on the reliability of the Main Feedwater and l Main Steam Systems. The change from 85% to 96% for MTCV testing does not affect i the reliability of the Main Feedwater and Main Steam Systems because the transient flows remain within the system design basis.

As a MTCV is tested closed, High Pressure Turbine first stage pressure changes.

This change is sensed by the EHC first stage pressure feedback circuitry which compensates for a substantial portion of the load drop by opening the remaining control valves. This will result in a transient drop in generated megawatts of l I

less than approximately 3%. Closing a MTSV does not cause a significant power' transient because turbine flow is made up from the opposite OTSG. Therefore the transient resulting from control valve and stop valve testing is within the design basis and has no effect on RCS or Steam Generator reliability.

It is concluded that the proposed changes are safe and present no unreviewed safety questions.

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SAFETY EVALUATION

SUMMARY

FOR SCR 95-5004 (SE 95-0069)

TITLE:

Raising the SFRCS High Level Trip Setpoint to 250 Inches Startup Range  ;

CHANGE:

This Setpoint Change Request (SCR) raised the allowable SFRCS high level trip to 250 inches Startup (SU) range from the previous level of 240 inches SU  ;

8 range.

REASON FOR CHANGE:

This change was initiated to restore the operating margin lost due to fouling '

~in the steam generators.

SAFETY EVALUATION

SUMMARY

The design function of the SFRCS high level trip is intended to limit'the amount of carryover into the main steam line during a severe overfeed transient. For conservatism, the SFRCS high level trip is correlated with the loss of superheat at the steam generator exit during an overfeed transient.

RELAP5 was used to perform the steam generator overfeed analysis. The analysis consisted of three overfeed transients at 45%, 75% and 100% full power. In each of the overfeed transients it was assumed that the'MFW control valve failed open and the MFW pumps were at their high speed stops. This resulted in overfeed rates (percent of initial feed rate) of 411%, 231% and 155%,

respectively. In each of the overfeed transients, RELAPS predicted a loss of superheat at the steam generator exit at 240 inches SU range. This was the basis for the SFRCS high level trip setpoint being 240 inches SU range.

In the above analysis, RELAP5 was also benchmarked against plant data from an overfeed transient which occurred on November 16, 1980. The benchmark showed ,

that RELAPS underpredicted the increase in SU range and the level necessary to cause a loss of superheat during the overfeed transient. This underprediction of SU range was approximately 15 inches. Therefore, if RELAP5 predicts a loss of superheat at 240 inches SU range, the plant would not be expected to lose superheat until approximately 255 inches SU range. This is the basis for extending the setpoint to 250 inches SU range.

The analysis referenced above also showed that as steam generator levels increase due to fouling, the loss of superheat occurs at higher startup levels.

Therefore, the higher steam generator levels which have occurred since completion of the 1992 analysis will make the 250 inches SU level trip increasingly conservative.

During an overfeed event, the cold water in the downcomer cools the outer shell and produces compressive loading on the tubes. The differential temperature between the steam generator shell and the steam generate tubes is limited by procedure to 60 degrees F. Additional analysis was provided to show that the tube to shell temperature difference is acceptable for the worst case overfeed transients terminated at loss of superheat. This analysis shows that the 60 degrees F allowable differential temperature between the steam generator tubes and the steam generator shell will not be exceeded if the overfeed transients are terminated by an SFRCS high steam generator level actuation at a RELAP predicted 240 inches SU range. As discussed previously, however, the actual steam generator levels required to cause a losa of superheat during an overfeed transient will be approximately 15 inches higher than predicted by RELAP.

Therefore, the differential temperature analysis will still be applicable if the setpoint is increased to 250 inches SU range.

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SAFETY EVALUATION

SUMMARY

FOR SE 94-0071 R02 TITLE:

The Cycle 10 Reload Report, Core Operating Limits Report and End Of Core (EOC)

T Reduction.

ave t CHANGE:

The proposed maneuver consists of reducing the average core temperature by 7 degrees F in order to extend Cycle 10's Full Power capability.

REASON FOR CHANGE:

This maneuver will extend Cycle 10's Full Power capability.

SAFETY EVALUATION

SUMMARY

Plant limits are unaffected by the T reduction. Thernal loadings of RCS components were evaluated as having a, negligible effect. Thermal stratification in the surge / spray lines and attached piping was evaluated as having an insignificant effect on RCS fatigue usage factors relative to that incurred while operating at the nominal average temperature. The effects on OTSG operations were evaluated and the extent of present tube plugging still results in sufficient superheat to preclude moisture carry-over to the turbine. l Plant performance and control during off-normal conditions has been evaluated.

Plant performance will remain bounded by the transients previously evaluated in the USAR. The following characteristics will be accommodated by the ICS during operation with reduced RCS average temperature: a) the steam temperature will be reduced; b) feedwater flow will increase slightly; and c) steam generator liquid volume will increase slightly.

Additional analyses were performed to demonstrate acceptable fuel thermal and

, mechanical performance. Also the RCS average temperature reduction will be limited to 7 degrees F, a generic LOCA evaluation concluded that the existing LOCA analyses will remain bounding for an EOC T reduction up to 10 degrees F.

A generic evaluation for all non-LOCA events waS#also performed, with the i I

conclusion that previous analyses remain bounding for an EOC T y reduction also up to 10 degrees F.

All of the previously issued Cycle 10 Operating Limits were verified for acceptability for the 7 degree F leduction, therefore the conclusions stated in Section 7 of the Cycle 10 Reload Report also remain valid.

The generic part of the T reduction evaluation by FCF resulted in the new BTU limits curve which is cycie" independent and is, therefore, a change made during cycle 10 that can be maintained for future cycles. The new BTU limits curve is conservative while operating at a nominal T of 582 degrees F, since it produces an increase in margin to the BTU l!mit alarm.

Based on this evaluation of the effects on safety, implementation of the changes to Table 4 of the Core Operating Limits Report (COLR) and operation of Cycle 10, including an EOC reduction of T by a maximum of 7 degrees F and the concurrent use of new BTU limits #$as been determined to be safe and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR SE 95-0044 TITLE:

Spent Fuel Cask Crane Functional Check Using Water Bags For Load CHANGE:

Use water bags (140 tons) for Spent Fuel Cask Crane load, rather than rigid steel weights, during functional check.

REASON FOR CHANGE:

i The Spent Fuel Cask Crane used to move the Vectra spent fuel transfer casks for dry fuel storage, had not been tested to capacity since construction.  !

Originally, the lift test was performed by lifting an assemblage of steel l materials. Common practice in the industry now is to lift large, specially l

constructed water bags. 1 i

SAFETY EVALUATION

SUMMARY

In general, lifting of water bags is considered preferable to lifting solid objects. If a water bag should inadvertently be dropped, the impact force to any structures beneath would be reduced compared to dropping a solid object from the same height. However, since rupture or potential flooding of auxiliary ,

building areas were not possible in the original testing, this test is l considered to be a " test or experiment not described in the USAR". I The safety evaluation defined the quality standards of the water bag manufacture I to ascertain that the probability of rupture was low. Assuming that rupture of the bags or a dropped load should occur, the effects of water impact and flooding on plant structures and potential spread of radiological contamination l was evaluated. 1 i

Prudent compensatory measure; were also outlined to minimize the effects of dropped load or ruptured bags. The proposed Spent Fuel Cask Crane functional test will not reduce the margin of safety as defined in the basis for any Technical Specification because no margins of safety as described in the Technical Specifications are affected. Therefore, this functional test was determined to be safe and did not involve an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR i SE 95-0064 i

TITLE:

Cable / Hose Routing Alteration to Penetration P59 l CHANGE:

A

, The proposed activity w'ill temporarily use containment penetration number 59 to route cabling / hoses associated with outage work inside containment.

REASON FOR CHANGE:

l These hoses are needed for various tasks such as steam generator descaling activities.

1 SAFETY EVALUATION

SUMMARY

i Penetration 59 has blind flanges installed inside and outside containment providing redundant barriers in the full-strength design condition. When l opened, P59 connects containment atmosphere to Mechanical Penetration Room #3.

Subsequent leakage to outside atmosphere is assumed. Thus P59 may provide

direct access from containment atmosphere to outside atmosphere, unless properly sealed. During refueling a single barrier, not necessarily capable of

. withstanding full containment accident pressure, is adequate. The proposed j activity opens P59 to install cables / hoses, then seals the penetration at the j exterior flange using a single full-strength barrier, a specially fabricated Sealed Penetration Assembly (SPA). The SPA is designed to restrict leakage to less than 10 liters per minute at full containment design accident pressure.

The USAR analysis of the fuel handling accident inside containment assumes no 1

restrictions to the release of radionuclides to the environment, whereas installing the' SPA conservatively restricts such leakage. l i

)- If pressurized hoses are routed through P59, there exists a potential to create a leakage path, should the hoses become depressurized, drained and opened on both sides of containment. Hose connections will be fittti with manual valves at the exterior side of the SPA that are capable of sealing against containment design pressure. Prior to core alterations, the required ,

manual isolation valves are verified and installed in service hoses, and the l shift supervisor will verify that the operating crews for the associated temporary systems are cognizant that breaches in their systems will affect containment closure.

Removal of the 8" diameter closure cover from the blowout panel along the north wall of Room 303 creates an opening in the negative pressure boundary when the EVS is lined up on the Shield Building annulus and penetration rooms.

This sa opening has no effect on the function of the EVS in the Modes applicable.

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i Removal of the section of the missile barrier along the north wall in Room 303 affects the tornado hazard analysis for equipment in the room.

Removal of the affected missile shield sections was evaluated for acceptability, with the restriction that DH9B must be closed and depowered.

This restriction on DH9B is the only effect on the decay heat removal and containment spray systems. The normal condition of DH9A and DH9B in all plant modes is closed and depowered, to prevent loss of inventory from either the BWST or RCS, and possible loss of decay heat removal capability.

As described in the Fire Hazard Analysis Report (FRAR), Room 303 is within Fire Area AB, and Containment is Fire Area-D. It concludes that based on their design and low combustible load, a fire will not propagate through these openings.

( Removal of the 8" closure cover from the Room 303 blowout panel, the small l section of missile shield, and the 8" flanges on Penetration 59 were evaluated by Security to be within their criteria for acceptable openings.

Based on the preceding analysis, Alteration of Penetration 59 does not pose any undue safety concerns for the plant or personnel.

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SAFETY EVALUATION

SUMMARY

FOR SE 96-0040 TITLE:

The Cycle 11 Reload Report and Core Operating Limits Report J CHANGE:

The Cycle 11 core loading, as described in the Reload Report, consists of the following: 1 batch 9, Mark B8A Fuel Assembly (FA) with an initial uranium 235 (U-235) fuel enrichment of 3.38 weight percent (w/o); 40 batch 11 Mk BBB FAs, with an initial fuel enrichment of 3.77 w/o; 60 batch 12, Mk B10AZL FAs (The Mk B10AZL is a radial zoned fuel assembly design with 24 rods containing 3.79 w/o and 184 fuel rods enriched to 4.09 w/o); 12 batch 13A, Mk B10A FAs that contain

^

fuel rods with uranium enriched to 4.46 w/o; and 64 batch 13B, Mk B10A FAs that contain fuel rods with uranium enriched to 4.46 w/o.

There will be a total of three reconstituted fuel assemblies (FA) with one stainless steel rod each in Cycle li's core. The core also contains the following control components: 20 Extended Life Control Rod Assemblies, 33 standard control rods and 52 Burnable Poison Rods Assemblies, with concentrations of 1.4, 2.3, 3.0, 3.5 w/o boron carbide.

REASON FOR CHANGE:

This is to assure that operation of the core configured as defined in the Reload Report will not violate operating and safety limits.

I SAFETY EVALUATION

SUMMARY

The review of the effects on safety is to assure that operation of the core configured as defined in the Reload Report with power distribution controls delineated by the core operating Limits Report (COLR), will not violate operating and safety limits. )

The reference fuel cycle for Cycle 11 is Cycle 10. The nuclear and thermal hydraulics (design) analyses were based on duration of cycle 10 to 488 +/- 15 Effective Full Power Days (EFPDs), Axial Power Shaping Rods (APSRs) full withdrawal at 610 (+/- 10) EFPDs, an End-of-Cycle reactor coolant average temperature (T reduction of 7 degrees F and a planned power coastdown.

Cycle 11 was a$aiy) zed to 675 EFPDs, and the Cycle 11 operating limits and _

setpoints reflect that licensed design length, i l

The following subjects have been considered and have been found to be safe or have been dealt with accordingly: Power / Imbalance / Flow, Statistical Core Design methodology, control rod group designations, ELCRA manufacturing problems, leakage of the rodlets, significant operating anomalies, FCF physics and power escalation testing recommendations, evidence of several fuel defects, visual examinations of high power fuel rod fuel assemblies, condition of misplaced spacer grids, visual grid inspections, metallurgical tests, Reactivity Insertion

i Accidents, high rod exposures, nuclear design analyses, longer fuel cycle, excess reactivity, BOC hot full power boron concentration, Hot Zero Power, higher lithium and boron concentrations, orifice plates in the OTSGs placed in l the 20% open position, OTSG sampling / inspection by eddy current probe, l introduction of 3 long emmiter incore detectors, and Moderator Temperature l Coefficient limit.

Based on the above evaluations of the effects on safety the proposed action and operation of Cycle 11 has been determined safe and does not constitute an unreviewed safety question.

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i SAFETY EVALUATION

SUMMARY

FOR TM 94-0017 (SE 95-0021) 1 TITLE:

I Revise Make-Up Water Treatment Chemical Supply Piping j l

CIIANGE :

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TM 94-0017 will install hoses and valves to temporarily change the point of chemical additions which control Hake-Up Water turbidity and revise drawings to ensure that piping tie-in points are consistent in design drawings and procedures.

REASON FOR CHANGE:

These changes permit better control of turbidity of water going to the Clarifier. Turbidity upsets in the future will cause greater problems with EPA regulations. j i

SAFETY EVALUATION

SUMMARY

Adding hoses and valves, changing the position of piping tie-ins or the position of valves from open to close will not adversely impact the operation of the Water Treatment System, and has no effect on the safety of the plant. j 1

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SAFETY EVALUATION

SUMMARY

FOR TM 95-0011, 95-0012, 95-0014, 95-0015, 95-0016 AND 95-0017 (SE 96-0009)

TITLE:

Temporary Power Feeds CHANGE:

These TMs involve the routing of numerous jumper cables between essential and non-essential sources and loads used during the replacement of Channel 2 and 3 essential 120 VAC power sources.

REASON FOR CHANGE:

These TMs were needed to provide necessary temporary power feeds to required loads during the replacement of Channel 2 and Channel 3 essential 120 VAC power sources under the inverter replacement modification. (FCR 86-0272)

SAFETY EVALUATION

SUMMARY

Cable ; outings have been evaluated by Engineering. Seismic mounting, dynamic loading, static loading hazards for equipment and supports near the temporary cables as well as physical protection of the cables have been addressed by the evaluation. Added combustibles were addressed by the required administrative controls.

The additional voltage drop with any of the jumpers is negligible. This is based on the fact that the jumper lengths are relatively short.

The TMa do not create any new concerns as a result of electrical noise. The areas where the jumpers are to be located have either been tested for the effects of electrical noise or have been restricted from the use of noise generators such as radio transmitters. Additionally, no electrical noise issues were identified from the previous installation of the channels one and four temporary jumpers.

The jumpers in TMs 95-0011, 95-0012, 95-0014 and 95-0015 meet or exceed the minimum channel separation of 12 inches between any Q channels and non-Q channels. In TMs 95-0016 and TM 95-0017, the Q channel separation is not a concern since the SFAS channels 2 and 4 will be considered inoperable in Modes 5 and 6.

All battery loading has been conservatively calculated based on the heaviest loaded battery IP. The additional loads added by the TMs will not exceed the l

load limits of any of the batteries. The additional loads will not exceed the l

load limits of the battery chargers or the limits of the Emergency Diesel l Generator f2. All loads have fuses and are fault coordinated with the upstream devices no differently than the normal configuration.

SAFETY EVALUATION

SUMMARY

FOR

'TM 96-0003 (SE 96-0005) l TITLE:

Removal of the Spool Piece at FE11109 CLIANGE:

Temporary Modification (TM) 96-0003 removes the spool piece at FE11109 and installs a blind flange on the upstream pipe flange.

REASON FOR CHANGE:

This modification allows inspection / repair of ECCS room cooler #5 during upcoming scheduled maintenance. The installation of the blind flange allows Service Water flow through the Train 1 ECCS room cooler #4 and at the same time provide a means of isolation so that work can be performed on BCCS room cooler #5.

SAFETY EVALUATION

SUMMARY

This TM allows for removal of FE11109 and its associated spool piece upstream of ECCS room cooler #5. In their place, a three inch blind flange will be installed on the upstream pipe flange. The blind flange and bolting material will not affect the pressure boundary of the system since they will be manufactured to the requirements of piping specification M-200 and those of the Service Water system.

Installation of the blind flange in place of FE11109 and the spool piece has been evaluated under the Civil Engineering Calculation 84E, Rev C12 and found to be acceptable without degrading the seismic integrity of the affected piping.

To install the TM, an entry into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement will be required during which time SW256 will be isolated rendering ECCS room coolers #4 and #5 inoperable. This configuration is explicitly allowed for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by the respective Technical Specification 3.5.2. After the blind flange is installed under the TN, SW256 will be re-opened allowing room cooler #4 to be returned to operable service and the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement will be exited. ECCS room cooler #4 will maintain the ECCS room #1 temperatures below the normal operation and Design Basis Accident temperature limits with Service water intake temperature less than or equal to 75.3 degrees F. If the ECCS room cooler #5 fan breaker BE1136 is opened, ECCS room cooler #4 will maintain the ECCS room #1 temperatures within the limits with a Service Water intake temperatures of less than or equal to 78.5 degrees F. Operation department personnel will monitor this Service Water temperature.

The removal of FE11109 and the resulting loss of the ability to monitor Service Water flow rate in the three inch piping will not create an adverse condition.

The flow element is used to measure flow only during refueling outages for flow l

l balancing or during performance testing of the ECCS room cooler #4 and #5.

i FE11109 is used to measure flow to ECCS room cooler #5 which is to be out of service for maintenance therefore, it has no function during this maintenance activity. The flow element is not required during any accident scenarios.

The proposed TM does not adversely effect any equipment required for the safe shutdown of the plant.

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SAFETY EVALUATION

SUMMARY

5 FOR i

3 TM 96-0012 (SE 96-0029) i i TITLE:

! l Bypassing SF 1, Fuel Transfer Tube 1-2 Isolation Valve, Open Interlock CHANGE:

This temporary modification proposes to bypass the interlock by installing a f jumper around the interlock to allow refueling operations. I REASON FOR CHANGE:

SF1 is a manual gate valve which is used to isolate the fuel transfer }

canal / Spent Fuel Pool from the transfer tubes during normal operations. During l refueling operations SF 1 is opened to allow fuel transfer between the SFP and I the refueling canal. SF 1 has an interlock which enables the use of the fuel  !

transfer mechanism when the valve is open. This interlock is discussed in the  !

USAR under section 9.1.4.4. Currently the interlock on SF 1 is not working and when SF 1 opens, the fuel transfer mechanism will not enable.

SAFETY EVALUATION

SUMMARY

The importance to safety comes from the fact that with the interlock bypassed, SF 1 could without the proper controls be closed with the fuel transfer carriage in operation and potentially damage a fuel assembly. Therefore compensatory administrative requirements will be implemented to verify SF 1 is open prior to movement through the transfer canal. Specifically, SF 1 will be visually verified open approximately once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Also, SF 1 is in view of the '

crew on the SFP side during refueling operations which will allow additional oversight to prevent unauthorized operation of SF 1.

Based on the above evaluation the proposed change will not impact previous assumptions concerning refueling operations and does not constitute an unreviewed safety question.

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1 l SAFETY EVALUATION

SUMMARY

FOR  !

I TM 96-0016 (SE 96-0036) 1 l

IITLE:

Installation of Video Camera for RCP 1-1 pump Leak Monitoring for Cycle 11 l

CHANGE:

1 Temporary Modification 96-0016 provides a video camera, cooling unit, mirror, lights, and power pack in the containment building to monitor the potential '

leakage on reactor cooling pump 1-1 casing during cycle 11. One insulation panel was removed from the pump casing to provide a view of the area of concern.

REASON FOR CHANGE:

A potential leak was identified during the 10RFO inspection. Following heatup and reaching normal operating temperature and pressure, the leak was no longer observed and the decision was made to remove the camera equipment.

Although the camera equipment is not installed, and the RCP insulation is installed, the coaxial cable was installed as described below.

SAFETY EVALUATION

SUMMARY

Removal of four square feet of pump casing insulation (one panel) has been evaluated. The additiogal heat load in the containment building was calculated to be approximately 1E ETU/HR. The CACs shall maintain a maximum containment air temperature of 120 degrees F at the inlet 6f the CACs as per TS 3.6.1.5.

Each CAC operating in slow speed provides 75E BTU /HR heat removal capacity.

There is more than adequate CTMT heat removal margin capacity to accommodate the additional heat load due to the RCP insulation removal. The removal of the insulation panel from the RCP pump casing will not create support concern for the remaining balance of insulaticn.

Information concerning the effects of elevated temperature on electrical cable ,

was obtained from the applicable Electrical Equipment Environmental Qualification Packages. An evaluation of the effects on the cable indicates that the life of the cables closest to the heat pource would be approximately two years. Therefore, the cables should withstand the elevated temperature for the duration of the lith operating cycle.

The temperature of the local concrete wall area near the RCP has been evaluated as not to exceed 200 degrees F with the removed insulation. The structural l

integrity of the concrete wall will not be impacted.

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i The camera equipment selected has a high tolerance for heat and radiation. The l camera has a cooling unit to reduce the affect of ambient heat. The camera will be located in the snubber pit out of the direct shine of radiation from the pump. A mirror will be used to aim and focus the camera on the area of interest. The lights installed will have regular incandescent bulbs which are permitted in CTMT by the USAR.

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Electrical Design Engineering has evaluated the cable and penetration fault protection for the power penetration PAP 2PI which is to be used for the monitoring equipment power. No penetration fault concern exists. The power penetration is non-Q therefore no o separation concern exists.

The video signal cable will pass through Q coaxial penetration P1L1LI. It is acceptable to run this cable through this penetration. No separation concern exists.

The cables will be routed in CTMT along other non-Q cable conduits and will not create a concern of cable separation. The free running cables will be secured by tye-wraps designated for use in the CTMT. Therefore, the cables will not become loose and become a potential source of debris in the CTMT sump in the event of a LOCA.

Other circuits that could be affected are the RCP monitoring instrumentation.

This cabling is all non-nuclear safety related. Even though the cabling could be affected, there is no nuclear safety related concern.

Based on the use of the mounting details, no seismic restraining concerns exist with the monitoring equipment.

The monitoring equipment is mainly aluminum. Aluminum can become a hydrogen generator in the event of a LOCA. The added amount of aluminum equipment is small. No concern is created.

The cabling introduced by this TM is rated for high temperature application.

The cables are spaced out and not concentrated. There are fire detection alarms in the areas where the TM cables run close to permanent groupings of cables.

There is no fire loading concern.

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1 l SAFETY EVALUATION

SUMMARY

FOR TM 96-0019 (SE 96-0045 R01) l I.I.TLF3 :

Fire Water Storage Tank Repairs CHANGE:

TM 96-0019 temporarily supplies the suction of the Jockey Fire Pump (JFP) directly from the clearwell instead of the Fire Water Storage Tank (FWST) .

l REASON FOR CHANGE- 1 The FWST will be out of service for repairs. Other systems must be temporarily l l

modified to ensure that adequate fire suppression capability is available, j 1

SAFETY EVALUATION

SUMMARY

This TM will defeat the Diesel Fire Pump (DFP) automatic start feature on loss of FWST level. The DFP will still automatically start on low system pressure to perform its function. This condition is acceptable in the short term in order to support drain down of the FWST for repairs. While the FWST is out of service, the requirements of FHAR Operating Specification will be followed to l ensure adequate fire suppression capability is available.

The JFP discharge is normally recirculated to the FWST. This TM will direct the JFP discharge to a floor drain in the water treatment building. This drain leads to the Water Treatment Building Backwash Sump. This additional flow is well within capability of the installed sump pumps.

The JFP receives an automatic shutoff signal on low FWST level. This TM disables this interlock. In the unlikely event that the clearwell is drained, if available the Electric Fire Pump will start on low header pressure to provide Fire Protection. If the Electric Fire Pump is unavailable the DFP will start on low header pressure to provide Fire Protection.

This TM installs a backflow preventer in the discharge of the JFP. This is an EPA requirement to prevent cross contamination of the Clearwell. The single failure analysis for the Fire Suppression System postulates a failure of the JFP discharge piping. The backflow preventer is rated for 175 psig which exceeds the JFP discharge line design pressure of 135 psig. Therefore installation of the backflow preventer will not change the probability of an inservice failure of this line.

It is concluded that the proposed TM does not have an adverse effect on safety i

and it does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

i FOR

! UCN 94-147 (SE 95-0025) ,

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I Channel Functional Testing

< CHANGE:

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! This USAR change:

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a. . Revises the USAR specified test frequency to be consistent with Technical ,

- Specification Surveillance Requirements;

b. Describes SFAS and SFRCS actuated equipment which cannot be tested during reactor operation; l

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. c. Defines the boundaries of the systems covered by.the various test requirements, and-

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. d. Revises USAR Table 7.3-2 to accurately reflect the testing performed in the j j plant.

REASON FOR CHANGE:

During the course of an internal inspection of instrumentation and control j systems, a concern regarding the adequacy of Channel Functional Testing was

j. identified and documented in PCAQRs 94-0459 and 94-0462. In the evaluation of these PCAQRs, it was found that the USAR description of the testing's compliance with AEC Safety Guide.22 (now Regulatory Guide 1.22) did not accurately reflect the testing being performed. This was documented in PCAQR 94-0480. The resolution of this PCAQR is to ensure the channel functional testing of the SFAS and SFRCS complies with Reg Guide 1.22, and to revise the USAR description of the testing to match that being performed in the plant.

SAFETY EVALUATION

SUMMARY

General Design Criteria 21 requires that protection systems; such as the SFAS and SFRCS, be designed to allow inservice testability while the reactor is in operation. Regulatory Guide 1.22 describes NRC acceptable methods of periodically testing the protection systems' actuation functions.

.USAR Section 7.3.2.6 and Table 7.3-2 state that the SFAS logic is half trip

-tested biweekly and system tests are performed quarterly. The proposed revision removes.the need to separate the two half trip logic tests per actuation channel by a two-week interval. This USAR interval requirement increases wear and tear on SEAS actuated equipment which is operated to allow performance'of the Monthly Channel Functional Test. No additional safety margin is provided by separating the half trip tests, as both halves must be working simultaneously for the actuated equipment to receive an actuation signal. Exactly when operability is determined relative to the other half channel is not relevant, since the logic and output devices are independent up to the control circuits of the actuated equipment.

l The revision to Section 7.3.2.6 and Table 7.3-2 also reflects that the logic channels and output relays of the SFAS are included in the half trip test and the actuated equipment is tested, in general, quarterly. An explanation is being added to demonstrate that the various tests meet the Reg Guide 1.22 positions. Justification for not testing some of the SFAS output relays, actuation devices, and actuated equipment during power operation is provided in the USAR change.

A footnote regarding the method of testing the HPI and LPI pump breakers was removed from Table 7.3-2. The testing being performed does not require the pump breakers to be placed in the test position.

USAR Section 7.4.2.3.3 describes the SFRCS compliance with Safety Guide 22.

It was revised to reflect the testing which is being done in the plant, which meets the Safety Guide positions.

A sentence which describes the logic and output relay half trip testing twice per month was added to clarify which portf.ons of SFRCS are included in that ,

testing. A sentence which states the actuated equipment is tested in accordance with other Technical Specifications demonstrates how that portion of the protective system meets the Safety Guide position.

The list of actuated equipment which cannot be tested during reactor operation was expanded to include all such SFRCS actuated equipment. These valves are smaller than the valves previously listed and are of less significance to USAR analyses than the previously listed valves. They were added for completeness.

One exception exists in the testing of the SFRCS. The MSIV Bypass valves, which are maintained closed during reactor operations due to an interlock with i the MSIVs, have a unique circuitry due to being solenoid valves. This circuitry prevents verification that the output relays change state. Testing these output relays on a refueling outage frequency was previously justified in UCN 94-104 (Safety Evaluation 94-0044). These relays are used to close the valve. The logic is tested similar to the other SFRCS circuits. Testing of the output relays and the actuated equipment occurs when the valve is open and closed.

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! In conclusion, none of the changes proposed by UCN 94-147 have an adverse effect on safety, 4

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SAFETY EVALUATION

SUMMARY

FOR UCN 95-010 (SE 95-0005)

TITLE:

Main Steam Safety valve seismic Qualification CHANGE:

This UCN updated the description of the analysis method presently addressed in USAR Section 3.9.2.9.5 to include specific seismic frequency and acceleration values.

REASON FOR CHANGE:

The USAR was revised to reflect the revision to the Main Steam seismic qualification made by the valve manufacturer.

SAFETY EVALUATION

SUMMARY

The basis for the seismic qualification of the Main Steam Safety Valves is the use of seismic acceleration values which correspond to the first natural frequency of the valve and which are taken from seismic spectra data for the plant location of the valves.

A recent pipe stress reanalysis of the Main Steam piping determined that the first mode of frequency values for the Main Steam Safety Valves were less than the original seismic qualification frequency values. These lower frequency values represent a greater flexibility of the valve structure and may result in higher valve structure and piping stress levels. The original vendor seismic qualification used seismic acceleration values based on a higher first mode of frequency and were subsequently determined to be incorrect.

A reunalysis of the valves was therefore prepared by the valve vendor to  !

address the identified piping analysis first mode frequency value..This l reanalysis considered the seismic accelerations which correspond to the latest j first mode frequency value for each valve model and for each of the two plant l area locations, Area 7 and Area 8. ,

1 The seismic reen:3ysis of the Main Steam Safety Valves has no impact on plant safety becaura the valves remain seismically qualified.

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! SAFETY EVALUATION

SUMMARY

FOR UCN 95-042 (SE 95-0026)

)

TITLE:

Delete USAR Table 9.3-2, Station and Instrument-Air Control Room Alarm Setpoint  !

CHANGE:

l Deletion of USAR Section 9.3.1.6.1 and Table 9.3-2.

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REASON FOR CHANGE:  !

l The listing of switch setpoints is not consistent with the type of information contained in the USAR. Setpoints are maintained by other plant administrative l 1

documents.

l SAFETY EVALUATION

SUMMARY

The Station and Instrument Air System is not safety-related, with the exception of the Station and Instrument Air containment isolation valves. The system provides a reliable, continuous supply of dry, oil-free compressed air for pneumatic instrument operation control of safety-related and non-safety-related l

pneumatic valves, operation of pneumatic tools and other miscellaneous uses.

USAR Table 9.3-2, Station and Instrument Air Control Room Alarm Setpoint, lists devices and their setpoints that result in a control Room annunciator alarm.

The devices listed are aesociated with the three air compressors and the two air dryer skids. These components are not safety related. None of the Engineered Safety Features depend on the supply of Instrument Air for their operation.

These setpoints will continue to be controlled by the existing administrative Controls.

This change does not affect containment isolation valves in the Instrument Air system.

This change does not impact the safety functions of the Station and Instrument Air System described above or in Section 9.3.1 of the USAR and is considered safe.

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i SAFETY EVALUATION

SUMMARY

l FOR I UCN 95-052 (SE 96-0077) i i

TITLES.

Reactor Protection System Logic Drawing Update  !

CHANGE:

Replace USAR Figure 7.2-1 with design drawing M-536-1. This change also adds ,

the design drawing as a reference in section 1.5 of the USAR.

REASON FOR CHANGE: .

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These changes reflect "as-built" information regarding the Reactor Protection ,

System and to evaluate the differences between M-536-1 and the current Figure 7.2-1.

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SAFETY EVALUATION

SUMMARY

f The function of the RPS is to initiate a reactor trip when a sensed parameter f (or group of parameters) exceeds a setpoint value indicating the approach of an "

l unsafe condition. In this manner, the reactor core is protected from exceeding l design limits and the Reactor Coolant (RC) System is protected from {

j overpressurization. The differences between the figures are minor note  ;

j nomenclature and clarifications, additions of sensors and channel designations, i

! and where minor configuration input tie-ins are located.

l The change is a drawing change to reflect as-built information regarding the  !

Reactor Protection System. These changes have no effect directly or indirectly -

on the safety functions of the RPS or on the ability'of the RPS to perform its  !

safety functions. Therefore, there is no adverse effect on safety as a result  !

of these drawing changes and it does not constitute an unreviewed safety  ;

question.

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4 SAFETY EVALUATION

SUMMARY

) 'FOR UCN 95-066 (SE 95-0040)

TITLE:

Emergency Diesel Generator Frequency Relays CHANGE:

This UCN deletes references to frequency relays in USAR Section 8.3.

REASON FOR CHANGE:

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! USAR.Section B.3.1 states that relays connected to the potential and current transformers at the Emergency Diesel Generator terminals will detect generator frequency and provide a permissive interlock for closing their associated

, generator circuit breakers. PCAQ 92-0033 pointed out that the EDGs have no

! frequency relays.

1 SAFETY EVALUATION

SUMMARY

I 1

This UCN has been written to make the USAR consistent with actual plant design  :

and is considered " safe" for the reasons outlined below. Because EDG speed will <

continue to be controlled by the electro-hydraulic and mechanical governors, plant conditions and plant operation will be unchanged, and this UCN will not  ;

affect hazards or plant operating conditions.

Addition of frequency relays will not improve EDG reliability or performance.

The frequency relays would provide a permissive interlock for closing the EDG output breakers. Therefore, relay failure would have the potential to prevent an otherwise properly functioning EDG from supplying its respective bus. In this respect, the EDGs will be more reliable without the relays than with them.

The USAR states that frequency will not drop below 95% during the loading sequence. This frequency is controlled by the electro-hydraulic and mechanical governors, and no frequency relay is required.

Engine speed is monitored by magnetic pickup speed switch contacts (SS) set at various engine speeds, with respective alarms. In addition to these alarms and control logic, tachometers at the engine control panel and at a control room panel indicate engine speed, which is directly proportional to frequency.

The frequency relays do not control frequency, they merely prevent loading of an EDG which is not at full speed. Continued operation at a frequency which could cause damage would require failure of the electro-hydraulic and mechanical governors. Therefore, the frequency relays do not perform any safety function, and are not required.

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l I i SAFETY EVALUATION

SUMMARY

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FOR i

) UCN 95-068 (SE 95-0047) i 1

j TITLE:

Revision of FHAR Description for Containment Air Coolers and Atmospheric Vent J Valves l CHANGE:

1 Revised the description in the FHAR of the Containment Air Cooler (CAC) Fan, j the Steam and Feedwater Rupture Control System (SFRCS) and the notes for the i Atmospheric Vent Valves (AVVs) in the event of a fire in the Control Room or 4

Cable Spreading Room.

i l REASON FOR CHANGE:

] This change is based on Potential Condition Adverse to Quality (PCAQ) 94-0166 f' and PCAO 95-0022.

SAFETY EVALUATION

SUMMARY

)

i Revision of operation of the CACs does not affect safety since this change is applicable only in the event of a Control Room or Cable Spreading Room Fire, and i there is no change in operation for all other situations. This method of starting the CACs on low speed does not affect their operation. The only l difference is that the motor starter is engaged manually versus electrically.

l The revised operator action is to open cubicle door and start the CAC on low I speed by pushing the button on the latching starter. The revised action was 3 facilitated by changing the regular screws on the cubicle door to knobs.

i Revision of the FHAR to include the statements about manual actuation of SFRCS f prior to leaving the control room and subsequent local manual action does not affect safety since these actions are already described in existing operating s procedures. These actions have been evaluated during the development of the 1 applicable operating procedures to insure that they do not affect the safety of I the plant and do not create an unreviewed safety question. These actions are l being added to the FHAR to explain why the Atmospheric Vent Valves will operate

properly in case of a control Room or Cable Spreading Room Fire.

$ There are two immediate actions operators take in the Serious Control Room Fire procedure; trip the reactor and actuate SFRCS. The immediate SFRCS actuation is

expected to isolate the Main Steam System (MSS) by closing the Main Steam a Isolation Valves (MSIVs) and closing the AVVs. Local manual control of the AVVs is taken by Operators in later steps of the procedure.

i These changes do not affect the Plant Design Bases, Technical Specification bases, single failure criteria analyzed in the USAR, increase the effects from

any hazards, or reduce reliability and are therefore safe.

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l SAFETY EVALUATION

SUMMARY

FOR UCN 95-072 (SE 95-0043)

TITLE:

Combining Design and Plant Engineering Functions

' CHANGE:

In order to facilitate the combination of the Design and Plant Engineering functions, the USAR was changed to show the Design Engineering functions reporting to the Plant Engineering Manager.

REASON FOR CHANGE:

The change allows more efficient operation by reducing the number of l Inter-departmental interfaces required to complete activities. l l

SAFETY EVALUATION

SUMMARY

The change to the USAR showing the combination of the Design and Plant j Engineering functions under the Plant Engineering Manager has no effect on any SSCs or their associated safety functions. The proposed change is j administrative in nature and does not affect the operation of any plant l systems.

The proposed change to the USAR is safe.

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SAFETY EVALUATION

SUMMARY

1 FOR I UCN 95-081 (SE 95-0050) 1 l

TITLE: 1 Radiation Monitor Sample Flow Assembly Replacement I CHANGE:

l FPR's 94-0272-901, 94-O'407-901, 94-0656-901, and 95-0311-901 revised.the sample '

flow assemblies on radiation monitor skids RE4597AA, RE4597BA, RE4598AA, and RE4598BA as depicted on USAR figure 9.4-11A.

REASON FOR CHANGE:  ;

l The previous sample flow assembly were obsolete and unreliable.

SAFETY EVALUATION

SUMMARY

The replacement sample flow assembly improve performance and reliability by l a) Eliminating the 120V, 60Hz transformer from the electronics enclosure. 4e transformer is both a heat source and electrical noise source for the electronics housed in the' enclosure. b) Replacing the motor operated valve with an automatic metering solenoid valve. Because the valve is solenoid operated, failures due to mechanical degradation are greatly reduced. Noise from Ifmit switches and stepping motors is completely eliminated, c) New [

electronics include high temperature integrated circuits which will extend the i life of the flow assembly. Power requirement of the new electronics is i 15 VDC thereby eliminating the transformer in the present system.

The changes made by the FPRs increase both the performance and the reliability l of the affected Systems, structures, and components (SSCs) . The replacement ,

equipment will perform the same function as the existing equipment. The j change has no affect on safety. No functions important to safety are affected. >

The proposed change does not increase the adverse effects from any hazard.

The affected portion of the USAR is Figure 9.4-11A. The new flow control valve and flow sensing element are housed together in a single unit for each containment monitor (RE4597AA and RE4597BA) . They are currently shown ,

separately on Figure 9.4-11A. This change will show them as a combined unit on l Figure 9.4-11A. Station vent monitors RE4598AA and RE4598BA have only flow l indication. The flow control portion was disabled previous to this change. l This change will delete the flow control valve for RE4598AA and RE4598BA as i depicted on USAR Figure 9.4-11A.

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SAFETY EVALUATION

SUMMARY

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FOR UCN 95-096 (SE 95-0058)

TITLE:

' j I Non-rated Fire Barrier CHANGE: j j UCN 95-096 revised the Fire Hazards Analysis Report (FHAR) to reflect a l non-rated opening in fire barrier 427-N/428-SI. This barrier separates' fire area DF and area X. i i

f REASON FOR CHANGE:

s PCAQR 95-0699 identified that there was a combustible material, cork, used to

$ isolate a steel beam going through the wall, a fire barrier,.from the concrete-

! wall itself.

SAFETY EVALUATION

SUMMARY

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! Fire areas are bounded by barriers of fire-rated construction. Where there are

! openings that are not rated, Generic Letter 86-10, allows licensees to perform j an evaluation to assess the adequacy of the opening. This evaluation was done i in calculation C-FP-013.06-088 to address the gap in barrier 427-N/428-SI. .The j conclusion was that the fire barrier's ability to provide an acceptable level of fire protection is not reduced by the non-rated opening. The proposed

-change is considered safe.

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s SAFETY EVALUATION

SUMMARY

FOR UCN 95-098 (SE 95-0061)

TITLE:

Changing the Malfunction Analysis for the Chemical Addition System Described in USAR Section 9.3.6.3.2 CHANGE:

The Malfunction Analysis was modified to update the numbers associated with more reactive cores for required minimum boron concentrations for the Reactor Coolant System (RCS) in order to maintain a 1% delta k/k Shutdown Margin (SDM), the RCS boron concentration due to Borated Water Storage Tank (BWST) injection for contraction during cooldown, and the required volume of water to be injected from the BWST in order to obtain the required 1% delta k/k SDM boron concentration for Hot Standby, Hot Shutdown, and Cold Shutdown. The Malfunction Analysis was also modified by determining the SDM boron concentrations based on the definition of SHUTDOWN MARGIN in Technical Specification (TS) 1.13.

REASON FOR CHANGE:

These changes were required due to the increase in the minimum boron concentration of the BWST. The BWST minimum boron concentration was increased to allow flexibility in future reactor core designs.

SAFETY EVALUATION

SUMMARY

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1 This Safety Evaluation evaluates changing the assumption of not crediting l control rods to crediting all control rods except the most reactive. The requirements of the reactivity control system are to maintain the reactor subcritical from all operating conditions, control reactivity transients, and i preclude inadvertent criticality in the shutdown condition. A 1% delta k/k SDM l in all MODES ensures these requirements by definition. This assumption is  !

consistent with SRP Section 9.3.4, TS 1.13, and USAR accident and malfunction analyses. Therefore, this change will not adversely affect the safety of the plant and is acceptable.

The change also adds data which was not formerly presented for 280 Orgrees F and corrects the definition of Hot Shutdown (HSD) within the context of this analysis. TS 1.4, Table 1.1, defines HSD as RCS average coolant temperature 280 degrees F >T > 200 degrees F. This is a clarification of the information presented and thSE$ fore does not adversely affect the safety of the plant.

Based upon the previous discussion, changing USAR Section 9.3.6.3.2, Malfunction Analysis for the Chemical Addition System, to be consistent with the License Amendment 207, the Technical Specification definition of Shutdown Margin, and Standard Review Plan Section 9.3.4 will not adversely affect the plant and is considered safe.

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l SAFETY EVALUATION

SUMMARY

FOR  !

UCN 95-105 (SE 95-0062) l TITLE:

Elimination of Hose Houses Outside of the Protected Area Fence CHANGE:

Revised the Fire Hazards Analysis Report (FHAR) to indicate that no credit is i taken for hose houses outside the protected area fence and clarified that only hose houses inside the fence are stocked with hose and other fire fighting equipment.

REASON FOR CHANGE:

These changes were made to reduce the amount of surveillance testing and maintenance that is required to maintain the hose houses since there is little likelihood that they will be used. Fires in structures outside the fence will be fought by local offsite fire departments rather than the plant's fire l brigade. I SAFETY EVALUATION

SUMMARY

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The elimination of the hose and equipment from the areas outside the fence will l not affect plant safety because the hose houses are not important to safety.

The local fire department, who is relied upon to fight fires outside the fence, use their own hose and equipment brought with them on the apparatus.

The plant fire brigade is not expected to leave the protected area to fight a fire since there are no hydrants that will be used to protect structures outside the fence that are relied upon for safe shutdown. Previously hydrant 12 (located outside the fence, east of the Switchyard) was credited for protection of the Diesel Generator Week Tanks. Review of the area indicates that use of hydrants inside the fence would be quicker to access by the brigade than a hydrant outside the fence.

The proposed changes are considered safe.

SAFETY EVALUATION

SUMMARY

FOR  !

UCN 95-111 (SE 96-0011)

TITLE:

USAR Steam Generator Sampling Update CHANGE:

This change removed four secondary sampling locations from USAR section 9.3.2.1.2 that are not used during normal operations and clarified two sampling locations  ;

in USAR section 5.2.4.7.d.

REASON FOR CHANGE:

The sampling locations removed from section 9.3.2.1.2 were steam generator water and steam generator steam samples for each steam generator during normal l operation. The steam generator water samples are not taken above 15% power since the sample is essentially feedwater. Below 15% power the steam generator secondary side samples are available and used. The steam sample points are just ,

upstream of the main steam line non-return valves MS209 and MS210. The steam ,

sample collection was curtailed due to personnel safety concerns and the  ;

inability to obtain a meaningful sample from that location.

The changes to section 5.2.4.7.d clarified the locations used to determine if '

primary to secondary leakage is the cause of a loss of primary inventory during l normal power operations.

SAFETY EVALUATION

SUMMARY

The steam generator water and steam locations are still physically available, but they are not used during normal operation as currently noted in the USAR sections affected. The steam generator water samples are only available below

• 15% power. Neither the position nor operation of the containment isolation valves for the associated steam generator water samples are affected. l The steam sample obtained just upstream of the main steam line non-return valves MS209 and MS210 have not been used since the sample obtained provides no valuable information. Additionally, the steam sample was a personnel hazard due to the high energy in the line. Various other sample points such as final feedwater, moisture  !

separator drains, condensate, and steam jet air ejector provide equivalent information on steam generator chemistry quality and primary to secondary leakage. j i

Section 5.2.4.7 changed to sample the plant secondary side for radioactivity ,

rather than the steam generator secondary side as stated, when attempting to I identify steam generator primary to secondary leakage. The plant secondary system provides better information at power on primary to secondary leakage than would be ,

available from sampling steam generator water. Above 15% power, the steam I generators act as flash boilers, therefore water sampled from the bottom of the steam generator downcomer is the same as final feedwater. Condensate and steam jet air ejector samples provide better information on condensable and non- i condensable fission products introduced via a primary to secondary leak.

As stated above, the change does not represent any effect on safety and, therefore, the action is safe.

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i SAFETY EVALUATION

SUMMARY

FOR j UCN 95 113 (SE 95-0068)

TITLE:

1 Removal of USAR Figure 9.2-3 I

J CHANGE:

)

This USAR Change Notice removed USAR Figure 9.2-3, Makeup Water Treatment System from the USAR.

! REASON FOR CHANGE:

USAR Figure 9.2-3, Makeup Water Treatment, contained non-safety related i information not required to be in the USAR.

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) SAFETY EVALUATION

SUMMARY

i

! The text of USAP Section 9.2.3, Makeup Water Treatment System, provides the l

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information required to be included in Davis-Besse's Safety Analysis Report, per Regulacory Guide 1.70. This USAR figure provided additional detail that

' was not required to be in the USAR and provides no safety related information.

l A portion of the MWTS is routed through Service Water Pump Room and is d

classified as Seismic Class I. This portion of piping is analyzed to determine the environmental Offects of a postulated rupture on safe shutdown. Since this piping is not being changed by this change and is not depicted on this USAR 1

Figure as seismic class I, there is no effect on the hazards analysis.

The other function of the MWTS important to safety is to provide water to the

fire water storage tank. This function is described elsewhere in the USAR and i

FHAR, therefore eliminating the depiction of this function from a figure in the

! USAR does not remove the description of this function from the USAR and FHAR.

j The MWTS does not perform any functions that are important to safe plant

, operation. Credible failures of the system will not compromise safety-related

systems, nor will it prevent the safe shutdown of the plant.

i Based on the above discussions, it is concluded that these changes are safe.

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SAFETY EVALUATION

SUMMARY

FOR UCN 95-114 (SE 95-0065) k, TITLE:

j USAR Description of CF7A and CF7B l 1

CHANGE: j l

USAR Section 6.3 described relief valves CF7A and CF7B as having an actual capacity of 1536 scfm nitrogen (47 gpm water). This UCN replaced the valve manufacturer specific data with the design basis requirements for the capacity of CF7A and CF7B.

1 REASON FOR CHANGE:

\

A B&W Specification established the procurement requirements for the original valves. This specification requires a minimum capacity of 650 scfm to assure I that excess capacity would be available and system operation would not be limited by relief valve capacity. The USAR identified the system relieving capacity requirement as approximately 375 scfm to protect the core flood tanks from overpressurization while filling with the make-up pumps. Operational procedures no longer permit filling from the make-up pumps, therefore the current system relieving capacity requirement of CF7A and CF7B has changed.

SAFETY EVALUATION

SUMMARY

The capacity of CF7A and CF7B must be suSficient to prevent overpressurization from both (1) inadvertent overfilling of the core flood tanks and (2) excessive leakage of CF30 and CF31. The maximum allowable leak rate for CF30 or CF31 permitted by Technical Specification le 5 gpm. Of the two cases the limiting requirement is that CF7A and CF7B must be able to pass the flow that occurs during an inadvertent overfilling of the Core Flood Tanks. A Mechanical Calcution documents the maximum fill rate when filling the Core Flood Tanks (CFT) from a High Pressure Injection (HPI) pump. This value is 226 scfm (approximately 34.3 gpm water). If each valve can pass this amount of flow, the tanks will be protected.

i USAR Table 6.3-5 lists a value of ~375 scfm as the system relieving capacity requirement when filling with the makeup pumps. As mentioned above, the 2 make-up pumps are no longer used to fill the core flood tanks. In fact, operational considerations make it desirable to avoid using the make-up pump for this service. A 1993 PCAQR documents overpressurization of the core flood tank filling line when improper valve sequencing was used. Components downstream of a pressure reducing orifice reached pressures that approached l their hydrostatic test pressures. An evaluation of these conditions revealed that, had the make-up pumps been used for filling, the hydrostatic test pressures of these components would have been exceeded. As such, filling the CFTs from the make-up pumps, is not recommended and Table 6.3-5 is being revised to reflect that the CFTs are filled using the HPI pumps. The revised value is 226 scfm to reflect the current system relieving capacity requirement,

,i eliminating inconsistency with the discussion of this issue on page 6.3-36.

SAFETY EVALUATION

SUMMARY

FOR UCN 95-130 (SE 96-0002)

TITLE:

Seismic Qualification of Replacement Components CHANGE:

Clarification is added to USAR Section 3.10 stating that equipment replacement parts are qualified in accordance with the appropriate standard (IEEE 344-1971 as a minimum) and to the acceleration levels required. A description of the analysis or test procedure are contained in the applicable test report.

REASON FOR CHANGE:

It has become necessary to buy replacement parts for equipment which are not.

certified to the test program described in'Section 3.10. j i

SAFETY EVALUATION

SUMMARY

USAR Section'3.10 describes the seinmic qualification programs for the major types of Category I Instrumentation and Electrical Equipment (e.g. Emergency Diesel Generator, 4.16 KV Switchgear, Batteries and Racks, etc.).

A statement is added to USAR Section 3.10 which clarifies that the qualification programs described in this section are for subsequent qualification would be in accordance with the appropriate standard, at least IEEE 344-1971, to the acceleration levels required for the application. Details of the test program are contained in the applicable test report.

The replacement parts are seismically qualified for their application to the appropriate standard.

Based on the above these changes are safe and do not constitute an unreviewed safety question.

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4 SAFETY EVALUATION

SUMMARY

FOR UCN 96-003 (SE 96-0014) 4 3

TITLE:

Fire Protection Procedures Frequency Reduction Project i CHANGE:

This UCN allowed for extending the frequency of performance of surveillance

i inspection and testing activities performed for fire protection related 1

eM &"* .ystems.

RL, SON FOR CHANGE:

c- s.-

f. The Fire Protection Surveillance Engineering Evaluation was performed to i document the evaluation of the past performance and intrinsic value of the 4

surveillance tests and inspections. This evaluation provides the technical bases for changing and extending the surveillance frequencies beyond those in FHAR and in some cases beyond those in the National Fire Code. The changes and extensions of the surveillance frequencies are based on past performance of i

the equipment / systems, industry practice, and the value attributed to the i tests / inspections.

SAFETY EVALUATION

SUMMARY

An engineering evaluation was performed to document the evaluation of the past l

i performance and intrinsic value of the surveillance tests and inspections.

This evaluation provides the technical bases for changing and extending the l surveillance frequencies beyond those in the FHAR and in some cases beyond

, those in the National Fire Code. It is the supporting basis for this safety evaluation.,

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} There were several factors that went into the decision process on what changes could be made. A selected set of plant procedures was reviewed to determine if j there were available surveillance frequency extensions that could be adopted, based primarily on the past performance of the equipment, without affecting the

• reliability or availability of the equipment. The decision to extend surveillance frequencies and by what amounts was a subjective one based on the preponderance of the past performanca data and the experienced engineering judgment. Our station's property insurance carrier is Nuclear Mutual Limited (NHL). NML allows the determination of inspection and testing frequencies to be made considering the past performance of the equipment / systems. Finally,

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the fire detection and alarm system was upgraded via MOD 91-0046 and the new

! system provides self monitoring for several of the functions previously

< evaluated by the performance of surveillance testing.

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The Fire Protection Surveillance Engineering Evaluation has concluded that the changes to be implemented by this UCN are technically justified and will not adversely affect the reliability and availability of the fire protection equipment and systems, safety of the plant or the ability to achieve and I maintain safe shutdown, i

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-004 (SE 96-0016, R. 01) l l

TITLE:

Revision to Vice-President Nuclear's Review of Nuclear Quality Assurance  !

Program Effectiveness CHANGE:

This change modifies the basis of the Vice President-Nuclear review of Nuclear Quality Assurance Program effectiveness from periodic (or annual) to continual and transfers responsibilities for advising cognizant management of Nuclear Quality Assurance Program effectiveness to the Manager-Quality Assessment from the Director-Nuclear Assurance.

REASON FOR CIIANGE:

These changes are intended to improve the description of the review of Nuclear Quality Assurance Program effectiveness and is intended to assign a quality assurance functional activity to the senior Quality Assurance management individual.

SAFETY EVALUATION

SUMMARY

i The changes do not affect the safety function of any SSCs and als considered to be safe. The changes are solely administrative as they only modify existing management review activities and pertinent USAR 17.2 Nuclear Quality Assurance Program description requirements are not being reduced.

SAFETY EVALUATION

SUMMARY

FOR UCN 96-012 (SE 96-0017)

TITLE:

Reorganizing Responsibilities for Dudgeting, Cost Control, Long Range Planning and Nuclear Projects.

CHANGE:

USAR Section 17.2.1.4, Toledo Edison Nuclear Group, and USAR Figure 17.2-1, Organization Chart, will be altered to split the responsibilities of 1) materials management and 2) budgeting, cost control, long range planning and nuclear projects. The Manager - Nuclear Support will be responsible'for the materials management activities and the newly created position of Manager - DB Business Services will be responsible for budgeting, cost control, long range planning and nuclear projects activities.

REASON FOR CHANGE:

This will facilitate the reorganization of the Nuclear Support Section and will permit the eventual transfer of the materials management activities to the newly formed Strategic Business Unit Service Group.

SAFETY EVALUATION

SUMMARY

The Director = Engineering & Services will continue to be responsible for the materials management activities, and the planning activities of cost & resources and nuclear projects. The proposed change does not affect the safety function of any SSCs and does not affect the operation of any plant system. The change is solely administrative as it revises the DBNPS site organization. All j functions continue to be performed. The technical qualifications necessary to operate the DBNPS continue to be provided by the Toledo Edison nuclear organization. As required by Technical Specifications, the new organizational structure provides well defined lines of authority, responsibility and communication. The staff qualification and staff training requirements of the Technical Specifications are not changed by the proposed reorganization. It is j therefore concluded that the proposed change is safe and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

l FOR 1 UCN 96-016 (SE 96-0069)

TITLE:

l Changes to the Fire Hazard Analysis Report CHANGE:

The FHAR was changed to revise information on carpet testing, delete a duplicate cable listing and update room information for fire detection. l REASON FOR CHANGE:

These changes were made to update information inadvertently omitted in the FHAR, i

to delete redundant information and to reflect test acceptance criteria, rather than actual test results.

SAFETY EVALUATION

SUMMARY

The proposed changes do not impact safety because: the updating of the location of smoke detection in the chase above room 324 to agree with other plant documents is safe because the additional detector means faster detection of a fire than previously described; deletion of the cable entry is acceptable since l the cable is not ussd for ohutdown and protection is not required; and the l

acceptance criteria used for carpeting reflects the highest standard for floor l finishes in the Life Safety code. The proposed changes are considered safe and do not constitute an unreviewed safety question. 1 I

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SAFETY EVALUATION

SUMMARY

FOR  ;

UCN 96-017 (SE 96-0033)

TITLE: ]

I Transfer Quality Assurance Responsibilities CHANGE: )

Transfer of the responsibility, quality assurance reviews to assure that requirements associated with implementation of the Nuclear Quality Assurance 4 Program are properly translated into' implementing procedures, from the Director

- Nuclear Assurance to the Manager - Quality Assessment and the Manager -  ;

Nuclear Safety.& Inspection.

ne. -

REASON FOR CHANGE:

This change is made to be consistent with SE 95-0017 and LRC Log 1-3573.

SAFETY EVALUATION

SUMMARY

The proposed changes will not reduce the effectiveness of any program, will not reduce any oversight or reviews, will not eliminate any activities, will not i reduce any responsibilities, and will not add any non-quality assurance functions to a quality assurance section.

The proposed changes do not affect the safety functions of any SSCs and are l considered to be safe. The changes are solely administrative and pertinent USAR {

17.2 Nuclear Quality Assurance Program description requirements are not being I reduced.

Based on the above the change is considered safe and does not constitute an l unreviewed safety question.

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i SAFETY EVALUATION

SUMMARY

FOR UCN 96-019 (SE 96-0023) 5 TITLE:

Clarification of Decay Heat Removal System Use for Spent Fuel Pool Cooling 4

i CHANGE:

1 i The purpose of this UCN is to clarify the USAR discussion regarding use of l_ the Decay Heat Removal (DHR) system to provide cooling to the Spent Fuel ,

4 Pool (SFP).  !

4 l REASON FOR CHANGE:

This UCN is meant to clarify that the use of the decay heat removal system is d

only necessary if the spent fuel pool cooling system is unable to maintain spent 4

fuel water temperature within required limits.

1 i SAFETY EVALUATION

SUMMARY

Reviewing the various sections of the USAR that discuss spent fuel pool I

cooling, two conclusions can be made with respect to the USAR description )

i for the design and operation of systems which provide SFP cooling:

}

j- a. Varying SFP temperatures can be realized given the several combinations of potential spent fuel loadings and the number of SFP pumps and heat

exchangers in operation. The maximum referenced SFP water temperature t is 155 degrees F. This value bounds abnormal operations when a full core

off-load is in the SFP. For core off-load conditions, a temperature of 140 degrees F is considered to be acceptable, whether maintained by the

{ SFP cooling system or the DHR system.

i i b. The decay heat system is available as a backup to provide additional

] cooling when abnormal heat loadings are placed in the pool. ,

i l With the USAR and SER information as summarized above and with flexibility in

} operating conditions, it has generally been interpreted that use of the decay l heat removal system is only necessary if the SFP cooling system is unable to maintain SFP water temperature lower than desired. The changes associated with l this UCN clarify this to avoid any possible misinterpretation.

t If the SFP cooling system can maintain a sufficiently. low water temperature, j the decay heat system can then be used for other plant functions needed to

! support outage activities. It is, however, necessary to maintain the j capability to align the decay heat removal system to the SFP in a timely manner should the need arise.

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i An engineering calculation provides an estimate of the SFP heatup rate due to a total decay heat load of 12E6 Btu /hr. Ratioing the calculated heatup rate of -

5 degrees F/hr to a total heat load of 30E6 Btu /hr, the heatup rate would be -

13 degrees F/hr. If the pool is being maintained at 140 degrees F, this allows for over an hour to align the decay heat system to the SFP prior to the pool reaching 155 degrees F. This estimate confirms that sufficient time does exish to align and operate the decay heat system to the SFP if it is being utilized for other outage functions and is fully capable of performing its heat removal function. Adequate alarms and monitoring is available to provide indication to operators to align the decay heat removal system.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-020 (SE 96-0024) i l

TITLE:  ;

)

Fire Detection in Main Steam Rooms CHANGE:

The UCN revised the wording in the FHAR so that no credit is taken for the fire detectors in rooms 600, 601 and 602.

I REASON FOR CHANGE:

This UCN was' issued in preparation for deleting fire detectors from the Main Steam Rooms.

' SAFETY EVALUATION

SUMMARY

Fire detection is located in various rooms and spaces around the site for the protection of the Davis-Besse facility. In some rooms this was required to meet regulatory requirements and in others it was done for property protection using good engineering judgment. The fire detectors in the Main Steam Line Rooms are not required for Appendix R compliance.

The temperature detectors can provide indication of a steam leak in these rooms, however this is not a function important to safety.

The detectors are located in fire area DH. The rooms in this fire area all relate to equipment a,ssociated with the main steam piping and containment purge.

The remaining rooms do not have detection installed. The fire hazards in the area are negligible. Because of the noise and heat in the rooms with the plant running, and the conditions that could occur during a plant trip, personnel do not routinely work in these rooms. Thus the fires typically associated with cutting, welding and grinding as well as with transient combustibles are much less likely to occur there than elsewhere in the plant. The sources of fire ignition, are small, the amount of material to burn is small and the volume of the rooms is large. A fire that did manage to start would be limited in size and not impact the ability of the plant to safely shut down. l l

The proposed changes are considered safe.

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SAFETY EVALUATION

SUMMARY

FOR j 4

UCN 96-022 (SE 96-0025) l l

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I TITLE:

S Removal of Plant Manager's Gai-Tronics Station and Deletion of USAR Figure 9.5-7 f

i CHANGE:

This UCN removes USAR Figure 9.5-7 from the USAR. USAR Figure 9.5-7 is a block 4 diagram, showing the different types of Gai-Tronics stations, and the types of channels connected to each type of station.

REASON FOR CHANGE:

The Gai-Tronics station was removed,from the plant manager's office via a d

maintenance work order. A PCAQ was written when it was discovered that the field wiring varied from what was depicted on plant drawings.

The special capabilities of the plant manager's Gai-Tronics station are not used by the plant manager, are not taken credit for in any analysis, and are not discussed in the USAR. Therefore, there is no need to have such a station, and in the absence of such a station, there is no need to uniquely identify it in the USAR. In addition, it was determined that the USAR intra-plant communication discussion does not need to include a block diagram of the normal plant communication system. l SAFETY EVALUATION

SUMMARY

The Gai-Tronics Communications System is not important to safety, and there is no effect on any safety-related power source. Therefore, there is no effect on safety.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-024 (SE 96-0020)

TITLE:

Procedure Change to Pump Concentrated Boric Acid from the BAATs to the CWRT.

CHANGE:

To pump concentrated boric acid from the BAATs through existing lines that do not have heat tracing to the CWRTs.

REASON FOR CHANGE:

To reduce the amount of concentrated boric acid that will be needed to borate I

the RCS during cooldown. l SAFETY EVALUATION

SUMMARY

This procedure change pumps concentrated boric acid from the Boric Acid Addition Tanks (BAATs) through existing lines that do not have heat tracing to the CWRTs. After the transfer of the desired volume, the line is flushed with demineralized water.

Two systems are affected by this change. The Boron Injection System which j ensures that negative reactivity control is available during each mode of facility operation and the Clean Liquid Radwaste System which handles liquid wastes such that the estimated releases comply with 10 CFR 20 and 10 CFR 50.

The proposed changes do not affect the functions important to safety of either of the affected systems. The transfer of concentrated boric acid from the BAATs to the CWRT does not affect the ability to transfer boric acid to the Make-up and Purification system or the Decay Heat Removal System during any mode of operation. The change in system operation of the clean Liquid Radwaste System does not increase the release of radionuclides to unrestricted areas and maintains releases as low as reasonably achievable. The transfer of concentrated boric acid through the existing pipes will not result in an )

increase in any hazards.

1 There is a possibility that the concentrated boric acid could solidify in the piping that is not heat traced. The piping used in this transfer, however, is i not important tv safety because it is not Q, AQ, or nonsafety related equipment j j

that supports safety related SSCs and is not used during normal, abnormal or emergency operations for the mitigation of any accident or malfunction.

This change in operation will have no adverse affect on the safety functions of the affected SSCs, therefore, the proposed change is safe and does not constitute an unreviewed safety question.

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l SAFETY EVALUATION

SUMMARY

FOR 3 UCN 96-029 (SE 96-0031) j I

TITLE:

Adding DB Supply Group References to and Eliminating Corporate Group Descriptions from USAR 17.2.

CHANGE: j

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The proposed changes are as follows:

1) The " Nuclear Support" Section is being renamed "D-B Supply."

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2) Descriptions of non-Davis-Besse corporate hierarchy positions and corporate groups providing services to Davis-Besse are being eliminated. j REASON FOR CHANGE:

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The reasons for these changes are as follows.

1) It is consistent with corporate re-organizations.

2) They are deleted because the existing descriptions are unnecessary. The l responsibilities are matrixed to site Directors. The responsibilities are being l merged into their responsibilities as outlined in Section 17.2.

SAFETY EVALUATION

SUMMARY

The proposed changes do not affect the safety function of any SSCs and are considered to be safe. The changes are solely administrative and pertinent USAR 17.2 Nuclear Quality Assurancs program description requirements are not being reduced as evaluated in accordance with 10 CFR 50.54(a).

This does not constitute an unreviewed safety question.

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l SAFETY EVALUATION

SUMMARY

FOR ,

UCN 96-030 (SE 96-0028) )

TITLE:

Positioning ECCS Room Cooler Outlet Valves Differently than Depicted on USAR Figure 9.2-1.

CHANGE:

Allow either the motor operated service water discharge isolation valves (SW5425, SW5424, and SW5421) for an ECCS Room Cooler, or the manual bypass valves (SW89, SWll3, SW121) to be positioned differently than depicted in the USAR Schematic, if it has been verified that sufficient flow will be available.

REASON FOR CHANGE:

It is sometimes desired to temporarily reposition these valves during power operation for maintenance or other evolutions.

SAFETY EVALUATION

SUMMARY

Currently both the MOV lines and the manual bypass lines must be open in order for a cooler to be considered operable. If the piping is adequately clean of microbiological corrosion and sediment build-up, this is not required. In this case, one of the two discharge paths could be closed, if necessary.

In the present fouled piping condition, adequate flow rates are verified by test (differential pressure across the cooler and flow measurements) and evaluation.

The testing (and evaluation, if required) is capable of determining whether the MOV discharge line, the MOV bypass or both lines are required to be in service to achieve the required flow.

If the flow rate available to an ECCS cooler (adjusted for design basis accident conditions) assures that the conditions of the calculation are satisfied, then sufficient flow is available to meet all required safety functions of the cooler. Verification of adequate flow may be accomplished by testing, calculation (s), or engineering evaluation (s). Accordingly, a brief note will be i added to USAR figure 9.2-1. This note will state the technical equivalent of: l "ECCS room cooler discharge valves may be positioned differently than depicted, l provided that adequate flow is assured under design basis accident conditions without operator intervention."

It is'sometimes desired to temporarily reposition these valves during power operation for maintenance or other evolutions. This USAR change would allow I

.either the manual valve (s) or automatic valve (s) to be positioned differently than depicted in the USAR schematic during normal operation, if it has been verified that sufficient flow will be available.

Periodic testing of ECCS room cooler coils is able to diagnose increased fouling before the functionality of the cooler is at issue. If the cooler discharge piping is sufficiently clean, opening both outlet paths from a cooler has little effect on flow and is not necessary.

This change is safe and does not constitute an unreviewed safety question.

l SAFETY EVALUATION

SUMMARY

FOR I

UCN 96-035 (SE 96-0035) 1 TITLE:

l Transfer of Quality Assurance Responsibilities j

, CHANGE:

l This USAR change transferred additional quality assurance responsibility to the Manager - Quality Assessment from the Director - Quality Assurance which includes responsibility for the Approved Vendors List (AVL).

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REASON FOR CHANGE:

This change has been made to be consistent with the transfer of duties already processed in SE 95-0017 and NRC Log 1-3573.

SAFETY EVALUATION

SUMMARY

The proposed changes will not reduce the effectiveness of any program, will not  ;

reduce any oversight or reviews, will not eliminate any activities, will not reduce any responsibilities, and will not add any non-quality assurance functions to a quality assurance section. The proposed changes do not affect the safety function of any structures, systems or components and are considered to be safe. The changes are solely administrative and pertinent USAR 17.2 Nuclear Assurance Program description requirements are not being reduced.

Based on the above, the change is considered safe and does not constitute an unreviewed soloty question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-038 (SE 96-0039)

TITLE:

Aligning Breaker BF1194 Normally Open (Removing Power from Valve HP-31)

CHANGE:

Motor operated valve HP-31, High Pressure Injection (HPI) Pump 1-2 Recirculation Stop Check Valve, which is normally open will have power removed by changing breaker BF1194 from normally closed to normally open. Procedural steps were added to direct operators to include breaker closure as part of other emergency core cooling system (ECCS) breaker alignments performed post-accident, restoring power to HP-31 prior to the need for valve closure.

REASON FOR CHANGE:

Given the Appendix R valve control circuitry concerns identified in Information Notice 92-18, 480V MCC F11E essential breaker BF1194 which is currently normally closed will be changed to be normally open. This will remove power from the motor operator for valve HP-31.

SAFETY EVALUATION

SUMMARY

The operators are currently directed by procedure to close HP-31 from the control room prior to aligning LPI to HPI and taking a suction from the emergency sump. With power removed, this action from the control room would not l be possible. To restore this capability prior to the need for valve closure,  !

the operators will be directed to enable the motor operator locally.

A time-motion study by Operations indicates that one minute and forty five  !

seconds of additional time are required to close BF1194. The associated additional dose to the operator performing this action was estimated to be 300 mrem. The additional 300 mrem dose to the operator closing BF1194 added to the previously estimated 1.71 rem, the total operator dose for performing breaker i closures is increased to approximately 2.0 rem. This total remains within the l GDC guideline of 5 rem and below the 10 CFR 20 annual dose limit.

In the limiting case there is approximately 37 minutes until the BWST reaches the level for which a transfer to the emergency sump is initiated. The additional one minute and forty five seconds required to close DF1194 beyond the three minute time to close in the decay heat valve breakers represents only a minor increase in response time when compared with the time available prior to the need for swapping to the emergency sump. With clear procedural guidance and sufficient time available to perform this action, there is adequate assurance l that HP-31 will be available without a meaningful increase in the likelihood of l valve malfunction or failure.

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I With the proposed arrangement, there will be no control room position indication j for HP-31 until power is restored. Without power, however, there will be no active means for the motor operator to spuriously close. In addition, HP-31 is ,

part of the administrative locked valve program.

HP-31 depowered, it can no longer spuriously close due to a fire as was previously possible in several areas of the plant where the valve was relied upon. The depowering thus obviates the need for manual actions to re-open the I valve due to fires in these areas. The need to protect the HP-31 circuit will no longer exist since a fire will no longer be able to cause the valve to close.

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l SAFETY EVALUATION

SUMMARY

FOR UCN 96-039 (SE 96-0038)

TITLE:

l FHAR Revision for Information Notice 92-18 Resolution  !

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CHANGE: i l

This Safety Evaluation reviews several changes to the safe shutdown analysis in the FHAR to assure safe shutdown can be achieved following Appendix R fire induced MOV failures. l I

REASON FOR CHANGE.

The existing Appendix R Safe Shutdown Analysis addresses spurious operation of MOVs due to hot shorts but assumes these valves can be manually repos'tioned.

The additional concern identified by the Information Notice is for a valve required, for example, to be open for safe shutdown which spuriously closes and  ;

experiences internal valve or operator damage preventing the valve from being reopened. Recent evaluations have determined that several motor operated valves currently used in the safe shutdown analysis for Davis-Besse could be damaged due to the scenario discussed above.

SAFETY EVALUATION

SUMMARY

l The current FHAR analysis requires the restoration of RCP Seal Injection, Seal Return, and RCS Letdown within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the fire. The RCPs are assumed to have been tripped at the start of the scenario and are not restarted. The i requirement for restoring RCP seal injection and return was based on testing and limited operational experience when the FHAR analysis was completed in 1990 with the new N-9000 Seals installed during 6RFO. Maintaining RCP Seal integrity and preventing gross leakage is required for RCS inventory control post fire. In the N-9000 Seal, each of the three stages is capable of withstanding full RCS pressure so that loss of Seal Return does not result in seal failure due to the loss of seal staging. The Individual Plant Examination (IPE) concluded, based on the testing and design of the N-9000 Seals, the seals will not experience gross leakage following a loss of ceal cooling and return provided the RCP's are tripped and the plant is subsequently cooled down. Any leakage that might occur would be within the capacity of the Make-up System. A maximum RCP seal leakage of 25 gallons per minute per RCP was used for loss of Seal Cooling, Seal Return, and Seal Injection in the Station Blackout scenario.

This leak rate is well within the capacity of the Make-up System. Based on the IPE and the SBO evaluations, restoration of RCP Seal Injection and Seal return 1 is not required for the plant to achieve safe shutdown following an Appendix R ,

fire. This eliminates the requirement to reopen motor operated Seal Return I valves and Component Cooling Water Valves after an Appendix R fire. I The requirement to restore RCS Letdown was based on the need to control RCS i inventory and increase boron concentration. The current natural circulation  !

cooldown rate is 10 degrees per hour. This cooldown rate and the requirement for Make-up flow to replace the RCS volume lost due to contraction allow for

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the injection of sufficient BWST water without Letdown. The addition of water i- from the BWST to replace the RCS contraction volume will result in a boron concentration adequate.to maintain the required one percent shutdown margin.

Based on the above, and the elimination of the requirement to restore RCP Seal Injection, .the requirement to restore Letdown following an Appendix R fire can be eliminated. This eliminates the requirement to reopen Letdown motor operated valves and component Cooling Water valves.

Closure of SW1399 is required to isolate Service Water from the TPCW heat exchangers to prevent flow diversion and possible runout of the Backup SW Pump. Manual isolation valve SW45 will be utilized for a fire in the Service Water Pump Room. Manual Valves SW54, SW55, and SW56 are located at the TPCW Heat Exchangers and will be utilized for a fire in the control Room / Cable Spread Room. These valves can be manually closed to support restoration of SW j flow to required components following an Appendix R fire.

Closure of AF3869 is required to isolate AFW flow from train one to Steam Generator 2 to prevent flow diversion and overfilling concerns. Manually closing motor operated valve AF599 will provide the required isolation function in the time frame allowed in the Appendix R safe shutdown analysis.

l The circuit design Aw AF599 prevents spurious valve operation for any hot I short which could occur in the circuit assuring the availability of this valve following an Appendix R fire.

l Valve MS107 is required to open to provide main steam to operate AFPT 2. An alternate flow path for providing steam to AFPT 2 requires opening motor l operated valve MS106A and manual valves MS 728 and MS733. MS106A will not be damaged due to spurious operation by a fire in the fire area which could potentially cause a hot short resulting in the inability to open valve MS107.

The alternate flow path can be established within the time frame allowed in the Appendix R safe shutdown analysis.

Valves MU6419 and MU6421 are required to open to provide a flow path from Make-up Pump 1 to the RCS. An alternate flow path from Make-up Pump 1 to the l RCS requires opening motor operated valves MU 6408, MU6409, MU6420, and j l

MU6422. The MOV's required for the alternate flow path will be able to be l reopened following a spurious closure with the motor operator stalled due to I the hot short condition. This alternate flow path can be established within the time frame allowed in the Appendix R safe shutdown analysis.

! Valve DH63 or DH64 is required to be closed to prevent diversion of RCS l inventory and to allow the Make-up Pump to maintain suction from the BWST l following the initiation of DHR cooling. If DHR Train I were placed in j operation with valve DH64 open, RCS inventory could be diverted through the HPI Pump 1 and Make-up Pump 1 minimum recirculation lines, and the auction ,

line to the Make-up Pump would be pressurized from the DHP preventing the l addition of BWST water to compensate for RCS contraction due to the continued cooldown. Initiation of DHR cooling does not occur for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

after the fire. Damage to valves DH63 and DH64 is limited to the operator.

Based on the above, it is concluded that the proposed changes to the FHAR will

, not have an adverse effect on safety.

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) SAFETY EVALUATION

SUMMARY

FOR UCN 96-041 (SE 96-0047) l TITLE:

Correction of USAR Figure 6.2-33, Containment Vessel Emergency Sump CHANGE:

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USAR Figure 6.2-33, Containment vessel Emergency Sump was corrected to reflect j 1

the actual location of the access hatch and ladder.  !

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' REASON FOR CHANGE:

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! A PCAQ identified a document discrepancy between the USAR Figure and Design Drawings for the location of the access hatch and ladder.

4 SAFETY EVALUATION

SUMMARY

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• The change in personnel access hatch and ladder location within the Containment j Vessel Emergency Sump as documented by USAR Figure 6.2-33 does not change the l required safety functions of the sump or the missile shield structure nor does 1 it affect the functions of the sump or the flow of fluid for the Decay Heat 4 Removal System and Containment Spray System.

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! The USAR and Technical Specifications discuss the Containment Vessel Emergency l Sump without specific reference to the access hatch or ladder locations. This 4 UCN will not have an effect on the Containment Vessel Emergency Sump and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

j FOR

) UCN 96-046 (SE 96-0043)

] TITLE: ,

Reorganizing Engineering Responsibilities- '

4 i CHANGE: ,

This UCN realigns Engineering into Design Basis Engineering and Plant l Engineering and changes the procedural responsibilities of the Manager - Nuclear l 4 Engineering and the Manager - Plant Engineering. ,

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+ REASON FOR CHANGE:

} This restructuring of the Engineering responsibilities serves to levelize the engineering staff manning and responsibilities and also more clearly align engineering responsibilities based on current DBNPS processes.

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SAFETY EVALUATION

SUMMARY

1 i The proposed change does not affect the safety function of any SSCs and does not

affect the operation of any plant system. The change is solely administrative l' as it revises the DBNPS site engineering organization procedural l j responsibilities. All functions continue to be performed.  !

j' The technical qualifications necessary to operate the DBNPS 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.

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! Commitments to ANSI N18.1-1971, Selection and Training of Nuclear Power Plant l, Personnel and ANSI N45.2.ll-1974, Quality Assurance Requirements for the Design

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of Nuclear Power Plants, continue to be met as described in USAR Table 17.2-1.

4 The NQAM and design control procedures continue to identify responsibility for l performing design activities and describe responsibilities and methods for i complying with specified requirements for design activities.

. It is concluded that the proposed change is safe and does not involve an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-052 (SE 96-0050)

TITLE:

Evaluation of the Carroll Township Water Tower on the Operation of Davis-Besse CHANGE:

The evaluation of the impact of the Carroll Township Water Tower was added to the USAR.

REASON FOR CHANGE:

This evaluation was conducted to determine the impact the construction of the Carroll Township Water Tower would have upon the station operation, particularly the operation of the meteorological tower.

SAFETY EVALUATION SUHMARY:

The Meteorological Monitoring System is used to collect weather data and used to ensure that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of a routine or accidental release of radioactive materials to the atmosphere. It has no direct nuclear safety function.

It is recommended to locate the sensors at least ten obstruction lengths away from the obstruction to minimize its influence. The tower is only seven heights away, however, due to the shape of the tower, this distance is more than sufficient.

Radiant energy should not be a problem since the large amount of water in the tank should remain at or near the temperature of the air. No convective heat effects are expected due to the small surface profile and the distance.

The water tower will have ho effect on the safe operation of Davis-Besse and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-059, UCN 96-060, UCN 96-061, AND UCN 96-096 (SE 96-0062)

TITLE:

Periodic USAR Review Discrepancy Corrections CHANGE: l Revise USAR description of: 1) Testing frequencies and methods for Emergency Core Cooling Systems (HPI, DHR, and CF) to consolidate and clarify information;

} 2) Makeup and Purification System operation permitting letdown flow control l

valve configuration to maximize RCS cleanup through the demineralizers during  :

cooldown; and 3) Normal makeup tank operating pressure range to reflect current RCS hydrogen concentration limits.

REASON FOR CHANGE:

I Review, evaluate and resolve d(screpancies noted during the USAR review.

SAFETY EVALUATION

SUMMARY

1) Testing frequencies and methods:

USAR Section 3.D.1.33 discusses HPI and LPI testing but is unclear on the l statement that testing is performed with the minimum recirculation lines. The deletion of discussion of the use of minimum recirculation flow lines for HPI i and DHR testing will have no effect on the operation of the DHR or HPI system j during testing. Section 6.3.4 accurately discusses the periodic testing of DHR 1

and HPI pumps. , I USAR Section 3.D.l.33 discusses CF testing. The discussion of testing the core flood system is accurately described in Section 6.3.4. The test method described in Section 3.D.1.33 was used during initial testing. The CF tanks can

! be refilled using the MU pumps but the HPI pumps are normally used.

USAR Section 6.2.2.6.3 requires a leak test be performed of the containment vessel emergency sump isolation valves (DH9A and 9B). This is in conflict with USAR Table 6.2-23 which clearly states the valves are not required to be tested for 10 CFR 50 App J. Any leakage through the emergency sump valves from the BWST back towards the containment would be identified during refueling outages.

Discussions on the containment emergency sump inspections performed each refueling outage for debris and degradation of the emergency screens will be j added.

USAR Section 6.3.4 describes the testing performed on DHR. The current description does not discuss performance of DHR pump ASME testing while the DHR pump is cooling the reactor vessel in an outage. This method is acceptable and allows the test to be performed without recirculating to the BWST.

USAR Section 6.3.4 discusses testing intervals for the DHR system and components. The section also states that testing is performed in accordance with Technical Specifications which either state the methodology explicitly or references the ASME code. Table 6.3-9 discusses the testing intervals and methodology again. This table is redundant and is not required. Deletion of the table will have no adverse impact on the DHR testing discussed in the USAR.

USAR Section 6.3.4 states that all LPI valves not normally in its safety position must be able to reposition within 30 seconds. However, valves from the BWST to the RCS are normally in their safety position. A statement will be added to the USAR stating all valves are in the safety position during normal operations. This clarification will have no adverse impact on the operation of the system.

2) Makeup and Purification System Operation USAR Section 9.3.4 discusses the operation of the letdown system. Normally all three parallel letdown flow control valves are not opened at the same time.

Opening all three valves at normal operating pressure would result in undesirable flow and pressure on the letdown piping. It also states that during RCS cooldown, the RCS pressure reduces and if desired both remotely operated valves can be opened to maintain letdown flow at 140 gpm. This maximizes RCS cleanup through the demineralizers. At the point where both remotely operated are full open and the letdown flow is less than 140 gpm, the manual bypass valve can also be opened to maintain flow up to 140 gpm. The USAR will address that all three parallel letdown flow control valves can be opened if desired during reduced RCS pressures to maintain letdown flows up to 140 gpm. Flow will be maintained less than the maximum design letdown flow of 140 gpm.

3) Makeup Tank Operating Pressure:

USAR Table 9.3-7 shows the normal makeup tank pressure as 15-35 psig. The RCS hydrogen concentration was changed from 15-40 cc/kg to 25-50 cc/kg. The increased hydrogen concentration affects makeup tank pressure. The normal pressure range in the makeup tank is 25-45 psig with the increased hydrogen concentration. The makeup system piping is designed for much higher pressures than 45 psig and the makeup tank is designed for 100 psig. Makeup tank overpressure relief is set at 88 psig with a relief capacity of 450 gallons per minute. The change in pressure has no adverse safety significance for the makeup system.

As described above, the changes are considered safe and do not constitute an unreviewed safety question.

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i SAFETY EVALUATION

SUMMARY

FOR ,

UCN 96-062 (SE 96-0058)

TITLE:

10 RFO Periodic Review for USAR Sections 8.3.1.2.14 and 8.3.1.2.20 CHANGE:

The USAR provides a brief description on how the labeling of raceways reveal the function and service of these components. The term raceway is used to include cable tray, wireway and conduit.

This change revises the de1cription by using the terms conduit, cable tray and wireway in place of raceway and clarifies numbering, labeling and color code conventions.

The USAR describes the routing of Class 1E circuits with emphasis on the I separation distance between essential redundant circuits and the separation distance between essential and non-essential circuits and the administrative responsibility and control provided to assure compliance with this criteria.

These administrative responsibilities and control reflect the controls in place during construction between Toledo Edison and the Plant's Architectural Engineer.

These administrative controls are being revised to accurately reflect the l current review and approval process of raceway layout drawings and the current involvement in cable pulling and inspection by Toledo Edison QC.

REASON FOR CHANGE:

Review, evaluate and resolve discrepancies noted during the USAR review.

I SAFETY EVALUATION

SUMMARY

IEEE 384-1974 provides the physical separation requirements for circuits and equipment comprising or associated with the Class IE systems and equipment. One l specific requirement in the document is that exposed Class 1E raceways be marked in a distinct permanent manner at intervals not to exceed 15 ft. Color bands

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were used to mark exposed Class 1E circuits and an absence of a color band denotes a conduit containing a non-Class 1E circuit. It should be understood that only partial conformance to IEEE 384-1974 is acknowledged in USAR Section 8.1.5 because complete adherence was not possible since plant design pre-dated the issuance of that document.

Currently, raceway layout drawings receive two levels of review and approval: ,

checking by an engineer and a approval on these drawings. The present USAR I description requires four levels of review for raceway layout drawings. The current design process ensures that sufficient review is provided to preserve the independence of redundant safety cables.  ;

These reviews'are conducted by qualified engineers. Thus, although the layout ,

drawings only receive two levels of review, any design change affecting raceway. '

in seismic category 1 structures receives sufficient review to ensure the independence of redundant safety cables in maintained and to preclude the loss of safety cables due to a design basis event.

Currently, QC reviews Maintenance Work orders (MWO) which install essential electrical cable. The Plant's cable pulling procedure requires that the cable bend radius and cable pulling tension be continuously monitored and documented by the cable installers. After completing the pull, the cable installers also verify cable continuity by meggering and document this reading.

QC remains involved with essential cable pulls through reviews of the MWo.

Chapter 17, Section 17.2.10.1 continues to require that QC utilize the direct in-process inspection of work activity when quality can not be verified during inspection of the end product.

These practices are in conformance with the NRC accepted Quality Assurance program as described in USAR Section 17.2.10 and, as such, do not constitute a reduction in commitment to the Quality Assurance program.

Based on the above evaluation, it is determined that the changes are safe and do not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-072 (SE 96-0057)  ;

TITLE:

Revision of USAR Section 13.7 - Security CHANGE:

This USAR change updates Section 3D.2.17 regarding how Davis-Besse takes into consideration Safety Guide 17 for development of security measures. It also removes overview descriptions of security measures from USAR Section 13.7. The security plans will simply be referenced in the text.

REASON FOR CHANGE:

The USAR change eliminates redundancy and identifies security measures at j Davis-Besse comply with Federal Regulations.

SAFETY EVALUATION

SUMMARY

l' The proposed change affects how information relative to Security is discussed in USAR Section 13.7. Security measures at Davis-Besse were developed with guidance from Regulatory Guide 1.17 and ANSI N18.17. Regulatory Guide 1.17 was withdrawn by the NRC on May 21, 1991.

As discussed in the USAR and not in the Security Plans, the following has changed: Ottawa National Wildlife Refuge personnel are no longer required to notify security personnel upon entering and leaving the marsh area in the Owner controlled Area as identified in USAR Section 13.7.1.1, Physical Security Design.

No significant hazards to the public health and safety will result from not requiring notification of security personnel upon entering and leaving the marsh area nor does it constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-076 (SE 96-0071) )

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TITLE:

Component Cooling Water System l CHANGE: 1 l

l This UCN proposes changes to Sections 1.2.8.3.3, 9.2.2, and 9.4.2.1.2.5 to eliminate redundant or unnecessary information and to clarify CCW operation during plant cooldown and following a Decay Heat Removal (DHR) System Cooler tube rupture.

REASON FOR CHANGE:

1 The changes consist of removing redundant or unnecessary information from the I USAR and clarifying current text.

SAFETY EVALUATION

SUMMARY

This UCN revises information regarding the CCW System discussed in various sections of the USAR. These revisions delete redundant or unnecessary information and clarify system operation during a plant cooldown or following a DHR Cooler tube rupture. There are no changes to the operation, design, or procedures associated with the CCW System. This UCN does not remove any information required by Regulatory Guide 1.70 to be contained in the USAR.

l All of the changes are acceptable because they are either not required in the l USAR, stated in another section, or clarifying operations during plant cooldown or following a DHR Cooler tube rupture. There are no changes to the operation, design, or procedures associated with the CCW System.

Based on the above, it is concluded that the changes are safe and do not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-077 (SE 96-0067)

TITLE:  ;

l Updates to USAR Section 7.4, Systems Required for Safe Shutdown ,

CHANGE:

I Revise USAR Section 7.4, Systems Required for Safe Shutdown, to reflect the J current plant confiauration. The changes revise the configuration and operation l

of: Auxiliary Feeowater System (AFW); Main Feedwater System (MFW); Main Steam

System (MS); Steam and Feedwater Rupture Control System (SFRCS); Steam Generator (OTSG); AFW system valves

MS5889A,B; MS106,106A,107,107A; AF3869,70,71,72; SFRCS cabinets: C5761A, C5792A; MFW - SG delta P switches PDS2685A,B,C,D; PDS2686A,D,C,D; Main Steam Isolation Valves: MS100,101; Reactor Coolant Pump Monitor cabinets: RC3601,2,3,4; Auxiliary Feedwater pumps; Auxiliary Feedwater Pump Turbines; SG drain valves: MS603,611; MFW 6 top valves FW601,612; Startup and MFW control valves: SP7A,7B,6A,6B; and MFW block valves: FW779,780. (Safety Evaluation 96-0067 is available for a more detailed discussion of the changes).

REASON FOR CHANGE:

Review, evaluate and resolve discrepancies noted during the USAR review.

( SAFETY EVALUATION

SUMMARY

None of the functions of the Systems, Structures, or Components (SSC) are affected by the changes. The changes made to the USAR are to either clarify or correct the USAR contents. No changes are being made to the plant or the way it is operated and maintained. The changes make it agree with' actual plant configuration and required plant design.

The changes made are to make the description of system operation clear, correct l and in agreement with design basis documents. The revised USAR text has no l effect on any accident initiators. The affected systems are all still capable of performing their accident mitigation functions as required by the USAR.  !

l Equipment reliability is not affected by any of these changes. The affected j l systems are all still capable of performing their accident mitigation functions '

l as required by the USAR. No new accident types will be created by the changes l described. No new failure mechanisms are being introduced. The changes will j not reduce any margin of safety as defined in the basis for any Technical I Specification. The requirements of the Technical Specifications and bases are unaffected by these changes.

The changes are considered safe and do not constitute an unreviewed safety j question.

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SAFETY EVALUATION

SUMMARY

FOR 1

UCN 96-078 (SE 96-0073)

TITLE:

USAR Primary Chemical Addition Update ,

CHANGE:

The change includes describing the addition of hydrogen peroxide to the reactor coolant system via the primary chemical addition system and rewording the usage of hydrazine for dissolved oxygen control.

I REASON FOR CHANGE:

This brings the USAR into agreement with current practice on the addition of supplemental chemicals to the Reactor Coolant System (RCS).

SAFETY EVALUATION

SUMMARY

a 1 Hydrogen peroxide addition to the RCS accelerates the natural process of radiolytic breakdown of water in the RCS. During shutdown periods, gamma radiation from the core causes radiolytic disassociation of water molecules in the RCS to form dissolved oxygen and peroxides. The materials of the Reactor Coolant System and appurtenances are not affected by normal radiolytic products.

1 This forced oxidation of the RCS induces a quicker and more complete oxidation of metal contaminants in the RCS to permit cleanup of these soluble corrosion products by the Makeup Purification system demineralizers. Hydrogen peroxide is added after decay heat cooler outlet temperature is below 140 degrees F to allow use of the purification demineralizers for clean up. Operation of the demineralizers at higher temperatures could damage the resin. Clean up of the activated metals is faster since the contaminants are oxidized and released

quickly, rather than gradually as oxygen migrates through the RCS.

Hydrazine addition to the RCS is unnecessary and ineffective during refueling conditions or when the RCS is open to air. Corrosion rates of RCS materials are extremely low at low temperatures. Hydrazine should not be added when the RCS is non-critical and below 200 degrees F.

The proposed change is considered safe and does not constitute an unreviewed i safety question.

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l SAFETY EVALUATION

SUMMARY

FOR UCN 96-079 (SE 96-0076)

TITLE:

Combined Oil Drain Lines to the Reactor Coolant Pump Oil Drain Collection Tanks and Correct the USAR Description for the Drain Tank Configuration.

CHANGE: l The collection system was built such that two RCP's drain via a common drain l

line to each drain collection tank and the USAR states that each RCP oil pan i drains individually to the drain collection tanks.

REASON FOR CHANGE:

4 The collection system was built such that two RCP's drain via a common drain

, line to each drain collection tank. j SAFETY EVALUATION

SUMMARY

The function of the RCP Oil Drain Collection Tank is to collect oil leakage from the Reactor Coolant Pump Motor upper and lower oil systems as required by 10 CFR 50, Appendix R. Each tank has the ability to hold the complete volume of one RCP motor. The drain tanks are not affected by this change.

The oil collection system was found to be acceptable following approval of an exemption to 10 CFR 50, Appendix R required the collection system to hold the contents of the entire lube oil system inventory. The design as installed permitted the tanks to hold the contents of only one RCP. This was found to be acceptable following NRC review.

The use of a combined drain line to the oil drain collection tanks vice individual lines is acceptable because the ability of the system to perform its function is not impacted. The oil drain system will still operate and allow the oil drain collection tanks to collect the oil from one RCP motor.

This activity is considered safe and does not constitute an unreviewed safety question.

I SAFETY EVALUATION

SUMMARY

FOR UCN 96-086 (SE 96-0059) l TITLE:

Removal of Voltage Regulators from USAR Figure 7.4-1, Control Rod Drive Controls.

CHANGE:

This USAR change deleted the depiction of voltage regulators on the Control Rod Drive (CRD) power supply trains of the CRD system on USAR Figure 7.4-1. ,

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REASON FOR CHANGE:

1 The existing figure is incorrect as identified in PCAQR 96-1174. As described in the PCAQR, the voltage regulator is shown in the wrong location on one of the )

l power trains. Drawings E661 sheet 1,2, E662, E4 sheet 2, E763 and USAR Figures

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8.3-15 show the power feed for the CRD system going to the cabinet for the motor generator set (the voltage regulator) not the CRD trip breakers. This only applies to the Main Bus as identified on USAR Figure 7.4-1. The Main Bus and the Secondary Bus have different configurations for the voltage regulators.

Showing the voltage regulators on the Main and Secondary Buses does not add any useful information to the figure. These regulators do not affect the ability of the CRD system to perform its functions important to safety and they do not add any additional quality to the USAR.

SAFETY EVALUATION

SUMMARY

The Control Rod Drive System provides a rapid insertion of the Control Rod Assemblies into the fuel core (reactor trip) when directed by the RPS, ARTS, DSS, or the manual reactor trip switch. The Control Rod Drive Control System also provides a means of reactivity control by monitoring and controlling the motion and position of the group and individual Control Rod Assemblies. I The control Rod Drive (CRD) System performance requirements are still being maintained in the USAR. There are no effects on safety and implementing this change to USAR Figure 7.4-1 will not constitute an unreviewed safety question and is safe.

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SAFETY EVALUATION

SUMMARY

FOR-

) UCN 96-088 (SE 96-0061) j 4

l TITLE '

Clarify USAR Sections 8.3.2.1.5 and 8.3.2.1.6 i i j CHANGE:

In USAR Section 8.3.2.1.'5 it was stated, "The [rectifierj output voltage is I

) maintained 3 volts higher than the station battery voltage." Three volts is an

inappropriately exact value, and that the practical intent of this statement is i that the rectifier output voltage be maintained "at least 3 volts" higher than I the battery voltage.

1 4 The UCN further clarifles Section 8.3.2.1.5 by describing the purpose of the minimum of 3 volts difference between the rectifier output and the battery voltage. The UCN also clarifies that the battery and/ or battery charger may become the source to the essential inverter upon rectifier failure.

t i UCN 96-088 adds eight indicating lights associated with the DC Power System

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to'USAR Section 8.3.2.1.6 which are not presently listed in this USAR section.

Four of these indicating lights are associated with the status of DC MCC 2

disconnect switches that feed the essential inverters.

l REASON FOR CHANGE:

! Review, resolve and evaluate the above discrepancies noted during the USAR j review.

SAFETY EVALUATION

SUMMARY

1 l - The function important to safety for both the essential 125VDC and essential instrumentation 120VDC systems is to supply independent and redundant power to major safety related systems and components.

This change affects the description of the systems and not the systems l

themselves. Revising the quoted setpoint for the essential rectifier output l l voltage from "3 volts" to "at least 3 volts" is safe because the equipment is

designed for operation in this range. The original "3 volts" value was intended as a minimum to ensure, with sufficient margin, that the coupling diode between i the reserve supply and the essential inverter would remain reverse-biased. j j

Therefore, voltage differentials of "at least 3 volts" increases the margin of i i confidence that the coupling diode remains reverse biased. l I

Describing the purpose of the value of "at least 3 volts" in the USAR text, as l

j well as declaring the battery charger as a possible DC supply to the inverter 1 upon rectifier failure, is safe because it is simply an improvement of the USAR text description for a proper plant design, as already depicted in USAR Figure

. 8.3-25. This' holds true also for adding descriptions of eight indicating lights q

associated with the station DC Power system to USAR Section 8.3.2.1.6 as shown i

) on drawing E-38B and M-320-63.

- As described above, the changes are considered safe and do not constitute an j unreviewed safety question.

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l SAFETY EVALUATION

SUMMARY

FOR l UCN 96-091 (SE 96-0060) i i TITLE:

I Main Turbine Missile Generation Probability l

CHANGE:

USAR Section 10.2.5 was* revised to include a discussion of the calculation used at Davis-Besse to ensure that low pressure turbine rotor inspections and main turbine valve tests are performed at appropriate intervals to maintain the probability of generating a turbine missile within acceptable limits.

REASON FOR CHANGE:

The conclusion of USAR Section 10.2.5 is an excerpt from NUREG-0136 in which the l NRC staff concluded that the probability of such an event was acceptably low.

After this section was written, stress corrosion cracking in the keyway region of shrunk-on wheels became an issue for Westinghouse and General Electric turbines. At the request of the NRC, GE developed in the 1980's a probabilistic calculation method for determining the likelihood of generating a turbine missile. 1 SAFETY EVALUATION

SUMMARY

The probability of generating a turbine missile is calculated for each low pressure turbine rotor based on periodic volumetric inspections of the keyway l and bore. The primary wheel failure mode is assumed to be fracture due to the l presence of a stress corrosion crack in the keyway near the bore of the l shrunk-on wheel. A second, independent wheel failure mode due to simple ductile failure during overspeed is also included. This calculation depends on the testing interval of the main turbine stop valves, control valves and combined intercept valves.

Based on the above, it is concluded that the proposed change is safe and does not present an unreviewed safety question.

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SUMMARY

FOR UCN 96-092 (SE 96-0074) l TITLE:

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USAR Sam,) ling System Update CHANGE:

The proposed change clarifies and updates chemistry sampling requirements and specifications for condensate, feedwater, spent fuel pool and reactor coolant in the USAR. Structures, systems, components affected are condensate, main feedwater, spent fuel pool and reactor coolant. The components and systems remain intact, but the chemical species sampled and chemistry specifications for the systems will be updated and clarified in the USAR.

REASON FOR CHANGE:

The proposed changes clarify and update chemistry sampling requirements and specifications for condensate, feedwater, spent fuel pool and reactor coolant in USAR to be consistent with current industry standards.

SAFETY EVALUATION

SUMMARY

The proposed change has no adverse affect on safety. No functions important to safety are affected. The proposed change does not increase the adverse effects from any hazard. The safety function of the systems involved is not being affected. All of the sample locations previously sampled in the affected systems are still available. The chemical species sampled and the specifications are being updated to current, improved industry standards represents an improvement in monitoring for conditions or chemical species which could be deleterious to the systems involved.

The improved recognition of said conditions represents an improvement in safety, therefore the proposed actions are safe and does not constitute an unreviewed safety question.

l SAFETY EVALUATION

SUMMARY

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FOR UCN 96-094 (SE 96-0088)

TITLE: 1 Decay Heat Removal Interlock i

CHANGE:

Revise USAR sections discussing the RCS cooldown and heatup process, the Decay j Heat Removal System, and Interlock setpoints. I REASON FOR CHANGE:

Review, evaluate and resolve discrepancies noted during USAR review.

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SAFETY EVALUATION

SUMMARY

The change to Section 7.6.1.1.1, compliance with IEEE 279-1971, Margin Between Operational Limit and Onset of Unsafe Condition, to reflect the maximum allowed RCS pressure value for Decay Heat Removal System operation permitted by DB-PF-06703. The Decay Heat Removal System initiation occurs at RCS pressure of up to 270 psig (previously 260 psig). The lowest design pressure rating of the Decay Heat Removal System piping is 300 psig. This provides a margin of 30 psig from the lowest design pressure rating (previously 40 psig). A positive margin between the pipe design rating and the normal maximum operating pressure still exists. The NRC neither required nor described any other specific margin value.

I This change has no effect on safety.

The changes to Section 7.6.1.1.2 more fully describe the Decay Heat Removal System isolation valve (DHil and DH12) interlock design, including the setpoints of the interlocks. The description also now includes the deadbands of the devices and explains why, once the valves are opened, the automatic closure function is protected by the suction line relief valve DH-4889. The description

, also explains the purpose of removing power from the valves except when they are l being repositioned as a part of plant heatup or cooldown. l l

l Section 7.6.1.1.2 previously discussed the fact that there are no manual  ;

bypasses for the interlocks. Experience shows, however, that the open permissive signal cannot be obtained for some plant cooldown configurations.

This involves the pressure drops of the RCS relative to the sensor locations.

The actual pressure of the RCS at the RCS to DH interface is at an acceptable value. Consequently, procedures allow jumpering the permissive relays if plant l conditions are such that opening the valves is acceptable. The explicit procedural controls over the process ensure the plant remains safe through the process and ensure the jumper removal. Because of these administrative controls, this change has no effect on safety.

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This change also reflects operation of DH11 and DH12 prior to raising RCS pressure. The valves are repositioned closed using remote manual control rather ,

than relying on the automatic closure signal from the interlocks. This was evaluated in Technical Specification Amendment 159. The closure signals are still present, however, and SFAS control is not needed. None of these changes affect plant safety.

The changes do not affect the plant equipment, plant procedures, or the plant a design basis and have no effect on safety. The proposed changes are safe and do not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-095 (SE 96-0075)

TITLE:

Cask Pit Slot Width CHANGE:

The value for the cask wash pit gate slot width is changed from 24 inches to 36 inches. USAR Section 9.1.2 identifies the value of 24 inches for the cask pit slot width which is incorrect. The correct width is 36 inches.

REASON FOR CHANGE:

The value is being changed to correspond with as-built measurements of the cask pit slot width.

SAFETY EVALUATION

SUMMARY

The change of 24 inches to 36 inches in Section 9.1.2.2 is based on the as-built dimension of 36 inches identified in plant drawing C-249. It is the intent of this section to identify the existence of the slot, the ability to transfer fuel through it and the ability to isolate it. The dimension of the slot has no impact on the section's intent. The dimension of this slot in the USAR text has no effect on the safety functions identified above., The change is considered safe and does not constitute an unreviewed safety question.

SAFETY EVALUATION

SUMMARY

l FOR .I I

UCN 96-099 (SE 96-0092) l

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l TITLE: ,

Reactor Pressure Vessel (RPV) O-Ring Hydrotesting.

i CHANGE:

Eliminates reference to'using the pressure tap in the annulus between the RPV ,

head to flange 0-Rings for hydrotesting the outer 0-Ring seal after head  !

closure.

REASON FOR CHANGE:

i The change updates the USAR to reflect that'the pressure tap is used only to  !

monitor for inner O-Ring leakage. ,

SAFETY EVALUATION

SUMMARY

i USAR Section 5.4.2 currently indicates that the pressure tap between the RPV l head 0-Ring in used to hydrotest the outer O-Ring seal after head closure. Such 3 a hydrotest is not performed, required nor desirable. The O-Rings are designed i to seat / seal against a differential pressure in one direction only. They will j seal only against high pressure on the inside of the 0-Rings. ,

Performing a hydrotest by injecting pressure in between the 0-Rings.would subject the inner 0-Ring to a differential pressure in the reverse direction for l which it is designed, potentially damaging the inner 0-Ring.  !

The outer 0-Ring is designed as a backup for the inner 0-Ring. If the outer  !

O-Ring were also to leak, it would be identified as part of the verification of '

compliance with the Technical Specifications. Therefore, there is no need to independently verify leakage with a hydrotest. Since there is no requirement to  ;

perform an outer 0-Ring hydrotest, and as such a test may degrade the inner l O-Ring, it is safe to eliminate the implication that such a hydrotest is l performed. l 5

The proposed change is safe and does not constitute an unreviewed safety question. ,

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-100 (SE 96-0091)

TITLE Update, Correct, and Reorganize USAR Chapters 8 & 9 for the Emergency Diesel Generators and Station Blackout Diesel Generator.

CHANGE:

The changes update, correct, and reorganize information, creating a newly proposed section which discusses the major components of the exhaust system and operation of the turbocharger, adding new references to already existing tables, and incorporating the FSAR AEC Questions and Answers appropriately. (Safety Evaluation 96-0091 is available for a more detailed discussion of the changes)

_ REASON FOR CHANGE:

These changes create an easier to understand description of the EDGs and SBODG, and keep the level of detail consistent with that required.

SAFETY EVALUATION

SUMMARY

The safety function of the EDG System is to provide onsite standby power sources for safety-related loads required to mitigate the effects of an accident combined with a loss of offsite power and to safely shut down the power plant and maintain the plant in a safe shutdown condition. A second safety function is to supply power to the required safety-related loads on loss of offsite power not accompanied by an accident.

The function of the SBODG is to provide power, for a minimum duration of four hours, to equipment needed to cope with a station blackout as required by 10 CFR 50.63. In addition to the safety-related loads, the EDG and SBODG are capable of supplying power to certain nonsafety-related loads for certain plant conditions.

None of the functions are affected by the USAR changes. The changes made are to either clarify and correct minor technical errors, remove unnecessary redundancy, remove unnecessary detail, or to reorganize the text to improve the readability of the document. No physical changes are made to the plant or to the way it is operated and maintained. The changes make to make the description of the EDG and SBODG clear, correct and in agreement with design basis documents.

No accident initiators are affected by this change. The progression and severity of accidents discussed in the USAR are not changed. The proposed change does not affect equipment reliability or the ability of the equipment to operate within acceptable or design limits. The affected systems are all still capable of performing their accident mitigation functions as required by the USAR. This change does not add any new failure mechanisms or accident types.

No new failure mechanisms are being introduced. The change will not reduce any margin of safety as defined in the basis for any Technical Specification. The requirements of the Technical Specifications and Bases are unaffected by these changes.

Based on the above, implementing this change to USAR Chapters 8 and 9 does not constitute an unreviewed safety question.

SAFETY EVALUATION

SUMMARY

FOR UCN 96-101 (SE 96-0070) l TITLE:

Lighting Systems CHANGE:

Revise USAR Section 9.5.3, " Lighting Systems", to reflect the correct plant configuration.

REASON FOR CHANGE:

Review, resolve, and evaluate discrepancies noted during the USAR review with: ,

1) Cooling tower and switch yard lighting; 2) outdoor Security System lighting; j l 3) AC/DC Emergency lighting; and'4) hand held lighting.

SAFETY EVALUATION

SUMMARY

Normal Station and emergency lighting systems are not required to mitigate any I l design basis events. The normal station and emergency lighting systems provides normal, security, and emergency lighting to allow for proper operation of the I l

power plant and for personnel evacuation.

The reliability of the cooling tower and half the switchyard lighting is i unchanged and there is a slight decrease in reliability of the other half of the l l

switchyard lighting because the switchyard distribution panel has no backup l

source of power should there be a loss of power on bus C2. This reduction in l reliability is acceptable because this function is not important to safety and  ;

operations personnel have hand-held lights available should actions be required in the unlighted portion of the switchyard, i

l The type of lighting used in the Security Lighting System is not important as long as the minimum illumination level of 0.2 footcandles is maintained per 10 l CFR 73.55. The AC/DC emergency lighting is normally supplied by an AC source i and transfers to a DC source on a loss of its normal AC source. Finally, the hand-held lights revision deletes references to interim measures being taken until permanent 8-hour battery-powered emergency lights were installed. This brings the section up to date with the actual, correct plant configuration. l Based on the above, the changes are safe and do not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-103 (SE 96-0065)

TITLE:

Service Water (SW) Flow Bypass Through a Spare CCW Heat Exchanger During Cold Weather I

CHANGE: f During cold weather, this change will allow water-to be manually diverted through the spare Component Cooling Water heat exchanger.

REASON FOR CHANGE:

When SW is colder, the heat removal capability per pound of coolant is increased. This causes the operating flow demands and flow requirements to be  ;

reduced for design basis events. This can cause an undesirably low SW pump flow, undesirably high SW pump discharge pressure, and high differential pressure at various components. This change will enhance this system's operation in cold weather. It will reduce available flow to all SW components on the affected loop in the event of a SFAS actuation.

I SAFETY EVALUATION

SUMMARY

In order to enhance the systems cold weather operation it is desired to increase the overall flow of the SW pumps, especially of the SW pump supplying primary loads. This will be accomplished by manually diverting (or bypassing) water through the spare component Cooling Water (CCW) heat exchanger. Establishing a fixed bypass flow of this type will reduce available flow to all SW components on the affected loop in the event of a Safety Features Actuation System (SFAS) 4 l

actuation.

It is intended to make conservative allowance for " trading" a small reduction in SW flow to various heat exchangers for a lower SW inlet temperature. SW flow versus SW temperature requirements were determined for the CCW heat exchangers, i

.the ECCS room coolers, the CACs and the CREVS water cooled condensers.

conservatism was assured by use of design basis (or larger) heat loads, conservative operating conditions, and/or by conservative calculational  ;

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techniques.

All components will continue to function as previously analyzed. During an SFAS actuation, although SW flow demands are increased, limits on allowable SW ,

temperature for flow bypass operation will ensure that heat removal capability  ;

J will be maintained. The SW system will continue to operate as required by analysis in all conditions with adequate thermal capacity.

Based on the above evaluation and calculations the change is considered safe and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

l FOR l

UCN 96-112 (SE 96-0064) i I

l TITLE: l l

1 Changes to USAR Section 12.3.2.2.2, " Counting Equipment for Radioactivity Measurement" CHANGE:

USAR section 12.3.2.2.2 was revised to replace the description of obsolete ,

equipment with a description of the new ec;uipment used to count radiological measurements.

REASON FOR CHANGE:

This USAR change describes the counting equipment used for radiological measurements as it exists today.

SAFETY EVALUATION

SUMMARY

As state of the art equipment advanced, the equipment described in the USAR was replaced. This replacement was also predicted upon reduced numbers of spare parts being readily available and maintenance being simplified. This USAR change brings the equipment description up to date.

4' Based on the above, the change to the USAR regarding maintaining state of the

. art counting equipment is safe and does not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR l UCN 96-117 (SE 96-0072) 1 TITLE:

Spent Fuel Pool Cooling and Cleanup System CHANGE:

The following changes are proposed to Section 9.1.3: 1) Revise the wording to a

show that the demineralizer and/or filter can be used. 2) Revise the wording to show that increasing the water quality sampling frequency is done as needed. 3)

Revise the wording to not limit the use of the cleanup system. 4) Revise the j wording to be less specific on what operator actions will be taken on a high or I

, low level alarm in the spent fuel pool.

REASON FOR CHANGE:

To review, resolve, and evaluate discrepancies noted during the USAR review with: 1) operation of the SFP cooling cleanup system; 2) clarify water quality ]

sampling frequency; and 3) revise description of operator actions to be taken on I receipt of SFP high or low level alarms.

l SAFETY EVALUATION

SUMMARY

]

The spent fuel pool cooling and cleanup system is a non-safety system which functions to remove the decay heat from fuel stored in the SFP, and also maintains the quality and clarity of the SFP water, fuel transfer canal water and contents of the borat'd water storage tank. j The filter and demineralizer are designed euch that one can be used even if the other is not being used. Restricting the,use of the filter or demineralizer to only when they are both in use could reduce the quality of the water due to normal maintenance on the units requiring them to be out of service.

Allowing the frequency of sampling to be more tran weekly does not need to be prescribed, sampling can be increased as needed based on other considerations.

Radiation surveys, water clarity, etc. would provide indications that additional samples should be taken.

Continuing to run the cleanup system with water meeting quality standards helps keep it from ever falling below standards. l The discussion of operator response to high or low levels in the SFP can be less specific because alarm procedure DB-OP-02003, provides specific guidance for the operators to respond to the situation based on the conditions and the status of the plant.

Therefore, the above changes to the USAR are safe and do not constitute an unreviewed safety question.

SAFETY EVALUATION

SUMMARY

FOR UCN 96-121, UCN 96-122, UCN 96-123 (SE 96-0082)

TITLE:

Revision of USAR Sections 9.3.6, 11.2, and 11.3.

CHANGE:

Revise USAR Sections 9.3.6, 11.2, and 11.3 to correct inaccurate information and

! provide clarification to better reflect the actual plant operating characteristics and design. The changes update, correct, and reorganize the l information discussing the Chemical Addition System and the Liquid and Gaseous Radwaste Systems. (Safety Evaluation 96-0082 is available for a more detailed discussion of the changes),

i, REASON FOR CHANGE:

Review, evaluate and resolve discrepancies noted during the USAR review.

SAFETY EVALUATION

SUMMARY

Chemical Addition System:

The changes have corrected the function of the chemical addition system as described in 9.3.6. This includes boration of the Reactor Coolant System in the event of a loss of the BWST due to a tornado. Also included are changes in the equipment data on USAR Table 9.3-13 to incorporate actual information from the currently installed equipment.

The Technical Specification Bases B3/4.1.2 discusses that either the BWST or the boric acid addition system is required for boron injection to ensure the ability to provide adequate shutdown margin. USAR Section 3.8.1.1.5 discusses that the BWST does not require protection from potential missiles generated from a tornado. Therefore, the boric acid addition system is required in the event the BWST is damaged as a result of a tornado.

The boric acid mix tank code has been changed from "ASME Section VIII" to "N/A".

This is acceptable because the tank is vented to atmosphere and would not be required to be manufactured to the ASME Code for pressure vessels. The system requirement for the tank is that it be manufactured with material resistant to boric acid corrosion. The tank is make of stainless steel which is resistant to boric acid.

The boric acid addition tanks' volume has been changed from "6893 gal." to "7600 gal." This change has no effect on safety because the total volume of the tanks is greater than the technical specification required volume. The larger tanks are more conservative in that they allow additional boric acid to be stored for use to provide safe shutdown capability for the plant.

The boric acid pumps' discharge pressure has been changed to the actual design of the pump. This changes the quantity of head from "140 ft." to "180 ft."

which is more conservative. The Technical Specification requires the pumps to  !

have a minimum flow rate of 25 gpm. The increased head adds margin to the pumps and ensures the 25 gpm minimum is met.

Liquid Radwaste and Ga:seous Systems:

The Liquid Radwaste Sfstrm's function as described in USAR 11.2 is to collect, e

store, process, and dirpose of liquid radwaste generated by the Reactor Coolant System and supporting systems located in the auxiliary building. No safety functions are provided by these systems.

The Gaseous Radwaste System's function as described in USAR 11.3, is to collect, )

hold, and, reuse or dispose of radioactive gases generated by the station. No  ;

safety functions are provided by this system.

USAR Chapter 11.2 and 11.3 contain descriptions of the design and operation of various plant SSCs that were intended to ensure that gaseous and liquid releases for the plant would be as low as reasonably achievable. As such, all operations associated with radioactive materials are based on criteria and procedures ensuring that the operations can be performed in accordance with 10 CFR 20 and the "as low as reasonably achievable" standard. In practice the plant operates within the guidance of the Davis-Desse Offsite Dose Calculation Manual (ODCM).

The overall level of radioactive releases from the plant is monitored as specified in the ODCM. Demonstrating compliance with 10 CFR 50 Appendix I and 10 CFR 20 using the ODCM guidance ensures that there is no effect on safety.

The manufacturer's standard fulfills the functions required by these pumps and does not create an adverse effect on any SSCs required for a safe shutdown. The manufacturer's standard fulfills the functions required by these pumps and does not create an adverse effect on any SSCs required for a safe shut down.

The proposed changes are safe and do not constitute an unreviewed safety question.

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-125 (SE 96-0084)

TITLE:

Condenser Tests and Inspections CHANGE:

Section 10.4.1.1 states that the condenser is designed to the ASME Power Test Code for Steam Condensing Apparatus, PTC 12.2. Section 10.4.1.5 states that performance testing is conducted in accordance with the ASME Power Test Code.

These two statements are deleted and Section 10.4.1.5 which states that the condenser is leak and pressure tested is replaced with statements that the condenser waterboxes were pressure tested during construction and that leak testing is performed as needed.

REASON FOR CHANGE:

The statements are deleted because the ASME Power Test Code for Steam Condensing Apparatus PTC 12.2 is a test code only with no relevance to condenser design.

The Power Test Code is used for acceptance testing of a new condenser to demonstrate compliance with contractual requirements if requested by the owner.

Also, the replacing statements clarify that pressure testing was associated with l initial construction and that leak testing is an inservice activity which is performed as needed. j SAFETY EVALUATION

SUMMARY

j Deletion of the ASME Power Test Code from USAR Section 10.4.1.1 has no impact on safety and does not affect condenser reliability. Deletion of the reference to the ASME Power Test Code in Section 10.4.1.5 is acceptable because performance testing of the condenser is not important to safety and does not affect condenser reliability.

Also, there is no requirement for periodic pressure testing or leak testing of the condenser.

These changes do not affect condenser reliability and have no impact on safety, nor do they constitute an unreviewed safety question.

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SUMMARY

FOR UCN 96-128 and PCAQR 96-1276 (SE 96-0094) b 4  ?

TITLE:

Revision of USAR Sections 9.3.3 & 9.3.3.1, Station Drains.

CHANGE

• Revise USAR Section discussing Station Drainage system to reflect the current plant configuration. The changes revise the configuration and operation of the j i Station Drainage and Sump System (SDSS) including: the Storm Sewer System; the i Settling Basins; the Condenser Pit Flood Pumps; the Transformer Collection Tank; i the Marsh Transformer Vault; the ECCS Rooms Sump Pumps; and the Water Treatment

Building Sumps. I i

- REASON FOR CHANGE:

l Review, evaluate, and resolve discrepancies noted during the USAR review.  !

l SAFETY EVALUATION

SUMMARY

s i .

Other than the Containment Isolation function and the ECCS Sump Pumps, the

? Station Drainage and Sump System (SDSS) serves no nuclear safety related ,

i function, nor does it serve any function important to the safe operation of the  !

plant. The SDSS does provide a containment isolation function for portions of l

, the system passing through the Containment boundary. The safety function of the I containment isolation valves is to isolate the Containment from the outside l j environment in the event of a loss-of-coolant accident (LOCA).

i r I The ECCS Sumps collect water leakage from various parts of the Auxiliary I

Building. The ECCS Room Sump Pumps and associated piping and controls provide a I safety-related function of removing water that accumulates in the ECCS rooms as the result of flooding and/or leakage which may affect the operation of safety related equipment during normal plant operation and postulated accident 4 conditions.

i The SDSS will continue to perform their functions as designed. The SDSS will

} continue to prevent the accumulation of liquids throughout the plant and route j these liquids to the environment in accordance with Federal and State regulations. The SDSS will continue to protect essential equipment from

. flooding as analyzed in the USAR. The containment isolation function for 3 portions of the SDSS passing.through the containment boundary is not affected by these USAR changes. The ECCS Rooms Sump Pumps will continue to perform their safety function of removing water from their respective rooms which may affect 4

the operation of safety related equipment during normal operation and postulated accident conditions as analyzed in the USAR.

)

I Based on the above, it is concluded that the proposed action is safe.

f I

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-129 (SE 96-0085)

TITLE:

a To Correct Ventilation Dampers CV5024 and CV5025 Position CHANGE:

Revise Section 9.4.2.2.2 and 3D.2.13 to accurately reflect CV5024 and CV5025 as being normally open to allow EVS to automatically respond to a high radiation condition in the spent fuel pool area. CV5024 and CV5025 are the dampers that cross. connect EVS to the other Auxiliary Building Ventilation Systems.

REASON FOR CHANGE:

This corrects the USAR to be consistent with as-built conditions within the plant.

SAFETY EVALUATION

SUMMARY

The EVS fans have an automatic start from the spent fuel pool ventilation radiation monitors. With CV5024/CV5025 open, EVS will draw from the spent fuel pool area and CV5017/CV5018 will remain closed. Upon a SFAS actuation, the EVS fans will start, CV5024 and CV5025 will close, allowing CV5017 and CV5018 to open. This aligns EVS to draw down the negative pressure area through #4 Mechanical Penetration Room.

This function is periodically tested by surveillance tests. Therefore with CV5024 and CV5025 normally open, SFAS will place the dampers in their safe position, closed, to allow EVS to draw on the Negative Pressure Area. The dampers have no other automatic function.

The basis for Technical Specificatico 3/4.o.12 states that the requirement for EVS to service the storage pool area to e..sure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. UCN 96-129 modifies the USAR to reflect the design of the system to allow the EVS to service the storage pool area. Also the measurement of response time at the specified frequencies provides assurance-that the RPS, SFAS, and SFRCS action j function associated with each channel is completed within the time limit assumed i in the safety analyses.

This change is considered safe and does not constitute an unreviewed safety question.

SAFETY EVALUATION

SUMMARY

FOR UCN 96-134 (SE 96-0079)

TITLE:

Condensate Demineralizer System Description and Instrumentation.

CHANGE:

In Section 10.4.6.2, Condensate Demineralizer System is described with three  ;

demineralizers in operation with one in stand-by. The demineralizer tubes are l replaced resulting in improved condensate polishing. Also the "End of Run" signal described in Section 10.4.6.4.d is deleted.

REASON FOR CHANGE:

The change in system description does not affect the system reliability as the new style tubes optimize the resin coating. Also, the "End of Run" signal does exist for other plant demineralizers but not on the condensate demineralizers.

SAFETY EVALUATION

SUMMARY

FCR 87-0012 and MOD 89-0075 replaced the demineralizer tubes, resulting in improved condensate polishing. With all demineralizers in-service, absolute retention of ion exchange resin and other contaminants are optimized. The l

change does not affect the system reliability as the new style tubes optimize the resin coating.

The exhausted resin is handled by either backwashing to the Backwash Receiving Tank or the Condensate Polisher Demineralizer Holdup Tanks directly. If the l resin's radioactivity level warrants, the resins would be pumped to the i Condensate Polisher Demineralizer Holdup Tanks rather than the settling basin.

This change provides additional information of how exhausted resin may be dealt with if warranted.

An "End of Run" signal does exist on other plant demineralizers, but not on the condensate demineralizers. Chemistry monitors the system performance and determines when the resins are exhausted. Based on this, the "End of Run" signal can be deleted from the USAR. l 1

l The USAR implies that a Condensate Demineralizer System high differential l pressure alarm is interlocked with an open position alarm of the Condensate r System Bypass Valve, CD 751. If the differential pressure exceeds 40 psid, l annunciator 13-2-A comes in. One of the operator's supplementary actions is to verify that CD 751 is open. Since this is an action and not an alarm, the CD 751 open position alarm description can be deleted.

The Condensate Demineralizer System performs no safety function. Furthermore, failure of the system will not compromise safety related systems nor will it prevent the safe shutdown of the plant.

Based on the above, the proposed change is considered safe and does not constitute an unreviewed safety question.

l

l l

1 SAFETY EVALUATION

SUMMARY

l I

FOR l

UCN 96-145 (SE 96-0083)

TITLE:

l Station Computer System Sequence of Events Resolution Time  !

CHANGE: l The current USAR section 7.10 for the Station Computer System states that the j sequence-of-events file records points with a one millisecond resolution for storing abnormal occurrences. The plant computer system clock resolution is actually 5 milliseconds.

REASON FOR CHANGE:

This will correct the USAR wording to reflect the as built condition of the l plant.  !

SAFETY EVALUATION

SUMMARY

The plant computer system provides operators with plant data to assist them in l plant operations. It does not perform any process control function nor is it 4

required for safe shutdown of the plant. The plant computer is not a safety system but is considered important to operations.

There is no detrimental effect on safety as a result of this change. The five millisecond resolution is adequate for determining sequence of events and no a reduction in acceptability of data gathering will occur. This five millisecond resolution was discussed in Serial letter No. 1000 to the NRC for " Post Trip Review-Data and Information Capability" which states "The time of occurrence listed with the event is based on computer clock time and recorded to the nearest five milliseconds."

No adverse conditions will be created by this change, therefore, it is safe and does not constitute an unreviewed safety question.  !

4 d

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SAFETY EVALUATION

SUMMARY

FOR UCN 96-169 (SE 96-0078) 1 i

i TITLE:

j Penetrations below Elevation 575 Feet d

CHANGE:

i This change identifies that all electrical duct banks that enter safety related

, buildings below Elevation 575 feet as having waterproof membrane.

1-REASON FOR CHANGE:

' Currently the USAR inaccurately identifies three duct banks, (E-25, E-26, and E-27.all located at El. 577 feet) as being enveloped in waterproof membrane.

SAFETY EVALUATION

SUMMARY

I Although concrete is a solid, it is not a watertight / waterproof barrier. Water i can seep into and through the cracks and voids which are common in concrete.

} Therefore, when the electrical duct banks are located at or below the

! groundwater table, they should be protected to prevent the intrusion of water into the concrete. This will ensure that water does not migrate into the i conduits and into the structure.

The highest recorded static water level on Lake Erie is 573.51 ft. (recorded j

~

June 1973). In the vicinity of the site, the groundwater gradient is approximately 2 ft/mi. Since the site is approximately 1/2 miles from the Lake, the maximum groundwater table can be expected to be approximately 574.5 feet, (based on historical high lake level.)

Requiring all electrical duct banks with conduit elevations at 575 feet or lower,-that enter safety related building from the outside, be protected with a waterproof membrane provides adequate protection from gro2ndwater infiltration into the station. The normal high groundwater level of 574.5 feet is below the elevation in which the electrical duct banks are protected with waterproof membrane.

In addition, the time periods during maximum probable water levels are short lived and will not have a noticeable effect on the duct banks.

The proposed change is considered acceptable and will not constitute an unreviewed safety question.

i

SAFETY EVALUATION

SUMMARY

FOR UCN 96-170 (SE 96-0086)

TITLE:

Revise USAR Figure 9.4-7 to Delete Station Heating Valve SH-385 CHANGE:

Deletion of valve SH-385, Warehouse H&V Unit Heating coil Outlet Isolation Valve, from Figure 9.4-7.

REASON FOR CHANGE:

This valve does not exist in the field.

SAFETY EVALUATION

SUMMARY

The Station Heating System does not perform any functions which are important to safety. The function of the Station Heating System is to transfer hot water to various unit heaters, heating coils, and baseboard heaters located throughout the plant as required to maintain space temperatures during cold weather and plant shutdowns.

SH-385 would appear to be designed to be an isolation valve on the outlet of the Warehouse H&V Unit S15 Heating Coil, perhaps for maintenance; however, SH-386 is there to provide the same function. SH-385 serves no safety function.

No physical changes are proposed. This change is made to ensure consistency with design drawings and as-built configurations of the plant. The Station Heating System, the Office Building HVAC System, and the Warehouse H&V Unit S-15 will continue to perform their functions as designed.

It is concluded that the proposed action is safe and does not constitute an unreviewed safety question.

i 1

i

SAFETY EVALUATION

SUMMARY

FOR UCN 96-172 (SE 96-0081)

TITLE:

l Use of Groundwater in the Local Area Around the Davis-Besse Nuclear Power I Station (DBNPS)

CHANGE:

The USAR provides a description of the current use of the groundwater in the local area around DBNPS. The information presented in the USAR was a result of I

investigations performed for the initial safety analysis for DBNPS. These I sections of the USAR will be revised identifying that the information contained in the section may be slightly different than that currently, therefore, the  !

information should be considered historical, j

REASON FOR CHANGE: I I

It is not practical to continually update these sections to reflect minor l changes in the water usage on a regional basis and maintaining the figures that correspond with the water usage is not practical and of little value.

1 This section has not been kept current nor is it pertinent to maintain these  !

sections current. It is not practical and the information is of little value to update these sections to reflect the water usage around the area.

SAFETY EVALUATION

SUMMARY

I l

Groundwater is not used for station operation. There are no station activities that will significantly affect the flow of groundwater. In addition, the NRC i concluded that in the event of an accidental liquid radwaste spill, leakage to ]

the groundwater aquifer is improbable and radionuclide concentrations at the J beach wells well below the limits of 10 CFR Part 20. The water quality is poor and the municipals obtain their source of water from Lake Erie.

Large withdrawals of water will be obtained from Lake Erie and not the aquifer.

In addition, the groundwater will continue to be monitored for contamination by the Radiological Environmental Monitoring Program (REMP), reference USAR Section 11.6. j Therefore, there is no adverse effect on the station and the proposed change does not constitute an unreviewed safety question. l I