ML20114D356
| ML20114D356 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/29/1992 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2080, NUDOCS 9209080136 | |
| Download: ML20114D356 (87) | |
Text
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t 9 CENTERCOR ENERGY Donald C.- Shelton 300 Madison Avenue Toledo, OH 436520001 '
. Vre President Nuclear -
- Davis Besse - (419)249 2300 t Docket Number 50-346-License Number NPF _i Serial Numb'er 2080 August?29, 1992 c .-
United States Nuclear Regulatory Commission -;
Document Control Desk Vashington, D.'C. 20555 Subj ec t : 1991 Annual 10 CFR 50.59 Report of Facility Changes, Tests and Experiments Gentlemen:
The Toledo Edison Company-hereby submits, pursuant to 10 CFR 50.59(b)(2), the 1991 Annual.10t CFR 50.59 report of facility changes, tests and experiments for Davis-Besse Nuclear Power Station, Unit:1.
Those-changes,ctexts and experiments identified via the safety reviev ,
Lprocess during the reporting-period offJanuary 23,-:1991- through January 22, 1992 are. enclosed. Attachment-1 provides an executive
' summary of those changes,' tests and experiments contained in the enclosure. -The' attached-safety evaluation summaries do not-involve an unreviewed safety question.
-If you- have any further questions concerning this matter, please contact-Mr. R. V.'Schrauder, Manager-Nuclear Licensing, at
"(419)-249-2366.
,- Very 1ily yours, j) -
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r-Attachment L' cc: A.~B. Davis, Regional Administrator, NRC Region III J. B. Hopkins, NRC Senior-Project Manager
- V. Levis,'DB-1 NRC Senior Resident Inspector Utility hadiological. Safety Board LO40022 92ovosota6 920sa, -
PDR opeianng companaes - R ADOCK-o5000346 PDR
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i Cleverond Bectne mammating - / -
li tedeco Edison / ,
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p Docket Number- 501346
-License Number NPF'3-Serial Number:2080-
' Attachment
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Page-1 ATTACHMENT I 10CFR50.59
SUMMARY
SHEET Number' Title DCR 90-0017 Revision of USAR Figure 8.3-2 DCR 90-0036 Installation of Agastat Time Delay Dropout Relays on. Valves HP2A, 2B, 20, and 2D
-DCR 90-0134 EVS Air Flow Diagram DCR 90-0163 Revision of Design Dravings Associated with the Sevage Treatment Plant LDCR'90-0178- Revision of Drawings Associated with Station Air DCR.90-0368 Update Drawings to Reflect "As-Built" Status DCR 91-0012 Revision of the' EDG Loa Tables ar.d Various Elementary Dravings i DCR 91-0013 Change in Condenser Pit Sump Flow Path DCR~91-0026 Addition of Pressure Indicator and Isolation
-Valve on the Flash Tank DCR 91-0036 Re-rate High Pressure' Injection Discharge Line DCR 91-0045 Removal of SV-5423 and SV-96 i
DCR 91-0074 Correction of' Motor Data-of the Diesel Oil Storage Tank Transfer Pumps and Fuse Size:for HCC D1 and D2 DCRu91-0097 and Laundry Facility Connections and Operation of LTH 91-0013 the'DBNPS-Vet Vash Facility
[ FCR.86-0272 Replacement of Two Cyberex Inverters FPR 77-0334-001 Revision of USAR Figure to Show "As-Built" ;
Configuration FPR 82-0148-001 Concentrate Storage Tank Filter FPR 91'-1485-901 Replacement of Cantainment Smoke Detectors in Fire Detection-Zone 317
-MOD 87-1063 Installation of a Three Vay Electro-Pneumatic Asco Solenoid Valve in the Condensate System l
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r g Docket Number 50-346 License-Number NPF-3
_ Serial Number 2080-l L ' Attachment Page 2 _j HOD 87-1167. Main Generator _ Sequential Tripping ;
HOD 87-1315' Replacement of Service Vater Valves DJ-1424, -
1429. and 1434 H0D 88-0067 Electric Fire Pump Controller Replacement .
H0D.88-0105 . Removal of Nitrogen Cover Gas From the Boric ,
Acid Concentrates Storage Tank H0D 88-0188 Replacement of Alison Smoke Detectors in the
-Auxiliary Building and Annulus Space i
. MOD 88-0206 Replace. Evaporator Condensate Drain Pumps P167-1 and P167-2 HOD 88-0227 Installation of Synchronism Check Relays for *
-Breakers AC101 and AD101 MOD 89-0025 Installation of a' Vet Pipe Sprinkler System-in Service Building #6 MOD 89-0124 Radiation Effluent-Honitors H0D 90-0006 _SFAS Shutdown Bypass H00 90-0014 -Chlorine Detector Jumper H0D 90-0019 Replacement of Containment Vacuum Breakers HOD 90-0036' Backup Protection _for Electrical Penetration Assemblies L~
t H00 90-0059- Service Vater System Modifications Redundant-Steam Generator Level Indication-
. -HOD 90-0078 H0D 90-0079 Removal'or Abandonment of Seismic Restraints on the Reactor Coolant System Piping and Equipment MOD 91-0016 Repair of Fuel Assemblies H0D 91-0019 ' Removal of a Snubbet From the-Main Feedvater System.
'PCA0R'90-0463 Evaluation of Containment Air Coolers Operating
'Vith Less Than Design Service Water Flov 7-l1 PCAOR 92-0086 Isolating Two Turbine Bypass Valves During Normal Plant Operation
Docket Number 50-346 License Number NPP-3
-Serial Number 2080
-Attachment PageL3~
500.90-3021 Removal of Intake Canal Vater Temperature Measuring = Instruments
' SCC _91-3000- _ Connect Backup Heteorological System Transformer SE 91-0024 Performance of DB-NE-03213, Moderator Temperature Coefficient Heasurement by Boron-Svap SE.91-0060 Graduated Filter Replacement Program SE 91-0067 Temporary Changes to Decay Heat Removal System Annunciator Alarms During Plant Shutdown SE 91-0075 Refuel Reactor-SE 92-0003 - Increased Radioactivity Levels in-the CCV-System-UCN 89-012 Revision of USAR Description for Shield-Building Annulus Heating System UCN 89-129 Distribution and Collection of Personnel-Dosimeters
.UCN_90-042' Integrated Control System (ICS) USAR
- Description UCN 90-093 Incorporation of Fire Hazards Analysis Report into the USAR UCN 91-015 Replacement of-Engineering-Assurance's Day-to-Day;0uality Reviev.of Procurement
-Documents with an Additional 4and Separate Procurement Engineering Technical and Quality Reviev UCN-914042 ' Reducing Setpoint for Decay Heat Pumps Suction Temperature: Alarm as Referenced in the USAR-
'UCN 91-043- Revision of USAR Description of the-Administrative ~ Dose Conttol Guidelines
.UCN 91-044 Revision of USAR Description _of Counting Equipment for Radioactivity Measurement
, UCN.91-055 Removal of Makeup Pumps and: LPI Crossover Piping and Valves from ECCS Equipment Reliability Considerations UCN 91-056 Power Range Nuclear Instrumentation Calibration
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Docket Number 50-346:
-License Number-NPF-3~-
1 Serial Number 2080
-Attachment Page 4 Programmatic ~ Controls and a Yearly Surveillance
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UCN 91-058
__ in Lieu of Biennial Reviev v.
tK:N 91-073 Capability of Siphon Breaker to Prevent Vater Loss from the Spent Fuel Pool '
UCN 91-078 On-site Meteorological'Heasurement ProFram UCN 92-002 Deletion of General Material Inspection Checklist UCN 92-007 Centerior Management Reorganization UCN 92-023 Change visator and vehicle control from Owner
-Controlled Area to Protected Area in USAR
' Section 13.7 UCN 92-025 Removal of Shift Technical Advisor (STA) Log UCN 92-034- 480 Volt Auxiliary Outdoor Distribution System UCN 92-036 Change in'Section Title from Radiological Control to Radiation Protection UCN 92-051 Receipt Inspection Process-DB-OP-00005,- Revision 2 to DB-0P-00005, Operator Logs and
-Rev, 02 Reading Sheets P
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SAFETY EVALUAT1011 SUM!iARY FOR DCP 90-0017 (SE 91-0057, R01)
TITLE:
Revision of USAR Figure 8.3-2 CHANGE:
On USSR Figure 8.3-2, this DCR will add EE303 to 487 volt unit substation E3, will add BF303 to 480 volt unit substation F3, and revise the load description for 11F203 (480 Volt unit substation F2).
REAS011 FOR CHANGE: _
This DCR corrects a drawing discrepancy identified by RFA 90-1585.
SAFETY EVALUATION
SUMMARY
The affected Structure, Systems, and Components are not safety-related. No physical changes will be made under this DCR. This DCR merely corrects paperwork discrepancies between existing design documents. Therefore, there will be no effect on safety.
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SAFETY EVALUATION
SUMMARY
FOR- ,
DCR 90-0036 (SE 90-0079) 1TLE:
Installation of Agastat' Time Delay Dropout Relays on Valves HP 2A, 2B, 2C, and 2D
~ CHANGE:
This Document Change Request (DCR) makes permanent the "As-Built" installation of Agastat Time Delay Dropout Relays, and associated' wiring in_the valve-control circuit-installed by'TMs ~ 89-0022, 89-0023, 89-0024. and 89-0025.
This circuit modification will allow valves HP 2A,.2B, 20, and 2D to continue to travel to their full open position upon receipt of a Safety Features Actuation' System (SFAS); Level 2 actuation signal and a concurrent Loss of offsite Power.
REASON FOR CHANGE:
The modifications were required.because, with the previous configuration in the
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event of a SFAS Level-2 actuation and a concurrent Loss of Offsite-Power,.the HP1' valves would receive an actuation signal 5 seconds after the Emergency Diesel Generator (EDG) output-breaker shuts but due'to the SFAS Sequencer design the actuation signal would be removed 3 seconds later. The circuit did not have a seal-in contact and'ther'efore the opening relay coll vould have
-dropped out and the valve motor would have stopped. . Seventeen seconds later
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the-actuation signal.would be re-applied which would then allow the valva-to finish' stroking full open. This' scenario was unsatisfactory because Technical Specification 4.3.2;1,3 and. Table 3.3-5 require the' valve to be fully open wittiin 30 seconds of an actuat.on signal" including logic, diesel and sequencing response times. When ccmbined, these times-exceeded the allowable value.
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SAFETY-EVALUATION-
SUMMARY
The addition of.an.ngastat-Time Delay Dropout Relay and. associated wiring _to the valve control-circuits'does not adversely-affect the safety function of the 3 High Pressure -Inj ection Syste.a.
The operation of these valves during normal operation will not be affected in-
~any way byLthe. relay' additions. The added relay and contact are not a part of the normal control room or remote valve control circuits.
'These relays enable the circuitry to open the valve fully within Technical Specification required' times:in the event of an SFAS Level 2 actuation with or i without a-concurrent-Loss of Offsite Power.
l l Alliother functions and indications of these valve control circuits. remain as before. The added relay does not aff ct normal (non-SFAS) control of the valve.
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f The-Agastat relays are environmentally and seismically qualified,-and are
+ seismically' mounted in the MCC. Installation of the_ relays and relocation of
.some. components-does not adversely ~affeet the safety function of the motor control centers.
The'effect on total: stroke time of the valve will be to reduce it significantly in the event of a SFAS-Level 2 actuation concurrent with a loss of offsite power. - The stroke time of' the valves themselv s when actuated by SFAS will- be longer by the-time it takes to pick up the Age est but that time is on the order of 50 milliseconds, an inconsequential amount.
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SAFETY EVALUATION
SUMMARY
FOR; DCR 90-0134 (SE 91-0040 R01) ,
, TITLE s -
EVS Air Flow Diagram-C_ HAllGE:
Revision of USAR_ figure.6.2-43 to correctly show the air flow path between rooms for EVS operation
' REASON ~FOR CHANGE:
USAR Figure 6.2-43. currently-shows that the Make-up Pump Room has two vent' paths to the Decay' Heat Cooler. Room. This figure is revised to show that one
-of_the vent paths.is the ECCS' Room 1.
SAFETY EVALUATION
SUMMARY
-The proposed drawing changes will correctly depict the vent paths between rooms served ~by Emergency Ventilation. No hardware is affected by these changes.
These changes will enhance system understanding by eliminating conflicting
-information. The reliability of the-EVS is unaffected.
LThe EVS is required to, draw down the Containment Annulus and portions of the Auxiliary building to the design' negative pressure within-13 minutes to support_
the USAR accident analyses. The calculational model used to demonstrate compliance with this_ requirement is--not based on the EVS vent area diagrams shown-in USAR Figure 6.2 43 or'the P& ids. Therefore, these-drawing changes do not affect--the accident analysis.
The air flowL paths inLthe Auxiliary Building were modeled to support environmental qualification analyses which-predict temperature and pressure conditionsffollowing vArlous high energy-line breaks--(HELB), Thisimodel was 1hased on~ layout drawings and: plant =walkdowns and is consistent with the actual J
plant configuration. Therefore',lthe changes to USA; Figure'6.2-43 and the
'P& ids'do not affect the EQ analysis, f
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1 SAFETY EVALUATION SUtHARY-FOR 1 DCR 90-0163-(SE 90-0163)
TITbE
- Revision of Design Drawings Associated with the Sewage Treatment Plant CHANGE
Clarificatibn of existing drawings to reflect plant configuration and to eliminate confusion caused by the current numbering scheme.
- REASON ' FOR CitAf1GE :
-The' drawing discrepancies are related to the 1982 assimilation of'the Davis-Besse Units 2/3 sewage-treatment facilities, and the subsequent abandonment of the original Davis-Besse Unit 1 sewage treatment' plant. In particular.-identification numbers were used in both sewage treatment plants, Land this has led to incorrect equipment names being associated with this equipment, related sewage treatment plan _ equipment, and the associated feeder
-breakers. This -DCR- clarifies existing labels and drawings.
- SAFETY- EVALUATION SMD4ARY:
This safety evaluation is only required because the revision of drawings (depictedfas USAR. figures is considered--a change to the facility-as described in the=USAR. Revising plant drawings as described above will enhance the performance of plant personnel and protect personnel from inadvertently working e on energized circul's.
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SAFETY EVALUATI0tl
SUMMARY
FOR:
. TITLE:
Revision of Drawings Associa(ed with Station Air CRA!!GE -
fRevisiontof'USAR figure to show deletien of pressure indicators (PI) 5390 and
= 5391.
- REASOff FOR CllANGE:
These. gauges were intended to monitor the station air supply to the air driven spent resin transfer pumps but- have not been installed due to interference -
. problems. These gauges have,been-evaluated-as-unnecessary'and the drawings are
- being' updated to reflect the "A3-Built" plant condition. ,
SAFETY EVALUATION SURIARY:
.-The proposed drawing ~ and USAR figure updates will have no adverse af fect :on the Station Air / instrument AirH(SA/1A) function either directly or indirectly. .The removal of the two1PIs will not .af fect any . design basis f or the SA/IA system.
-The gauges.have_not:been installed or needed and their removal from-the plant
.will have no. adverse affectcon safety.
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SAFETY LVALUAT10!! SUtil!ARY i F0k l DCE 90-03t8 (SE 90-0144)
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T 1 'I 1.1: t tf pd a t e Drawings to hellect 'As-lullt' 9tatus
- Cll AtlGE
Revision of USAR figutes to correct discropancies found during walhdowns.
FFA50!1 FOR CllAtlGE:
Drawing discrepancies as.aciated with drawings 101 miccellaneous sy9tems were
.ibnetved and documented in accordance with Toledo Edison Procedure flE5-140
- Funct ional Valkdown and llameplat e Dat a Cn11ec t ion . Theme drawing de l e ie' .es, were evaluated on t he component /syst em basis to det ermine-a) what drawing changes, if any, would be tequired to ensure consistency wit h the design basis documant s, and b) i f t he dire t epancies would compr omise plant nafety or operation.
SAFETY FVALUAT1011 EUltttARY:
Revising the inst rument numbert, showing the correct equipment numbet and location of Local control statlon, local hand evitches, delet ing/ capping t he ununed renduits on electrical layout drawings does not adversely effect the safety furc 'ons et the components or f unc t ions of any of t he af f ec ted t.y stems.
The change oposed correctly depict the As-ruilt configuration of the pinnt.
Each d:awi. hauge proposed has been reviewed against the design basis tot the syst em and , been accert ained to be correct . Thetefore, the proposed changes deuc t Ibed above have no et f ect s on t he snf et y of the platit .
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f SAFETY EVALUATIO14 SUtttARY [
roR .
, TITLE:
Itevision of the EDG Load Tables and Various Elementary Drawings CilAtlGE:
Update Electrical Drawings E-1042 Sheets 1 and 2, _and E-1043 Sheets 1 and 2 (USAR Figure 8.3-1). load tabulation drawings for Davis-Bease's Emergency Diesel Generatots (EDG). ,
t ItEAS0!! FOR CilA!JGE:
Revise EDG load table to: 1) Include the load on the Emergency Diesel Generators as a result of transformet's load and no-load losses: 2) Include :
recalculated brahe horsepowers ( tilps ) for the driven equipment powered from !
each of the EDGs 3) Document on the EDG Load Tables the potential tor the containment Spray Pumps and the Containment Air Coolers to be loaded on the EDGs in Step _1 of the sequence: 4) Document on the EDG Load Tables that the EDGp. Air c ompressor could be running-in the first sequence steps and 5) Correct 111scellaneous Typographical Errors. 1 Elementary Drawings E-34B Sheet 13 (USAR Figure 8.3-9) illustrates part of the essential 4.16 KV switchgear feeder breaker relaying schematics. This drawing is being revised to correct typographical errorn in referenced equipment numbers.
SAFETY EVALUATIO!1 SUl@iARY:
There will be no adverse effects on safety ar a result of the proposed EDG load
--t able - revi sions . 11 0 physieni changes to the plant-are performed.- The proposed changen provide accurate horsepower requirehents for each of the EDGs connected i loads..in effect *as-building" the document. ,
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-The revjsed load tabulations do not result in any cumulative loadings which exceed the EDGs rating. . The total cumulative loading identified for each of the Emergency Diesel Generators actually decreases as a result of the proposed t revisions. Therefore, no increase in the chances for an EDG failure will tesult from any tevised Individual loadings.
The documentation of the Containment Air Coolers (CAC).and Containment Spray Pumps-(CS) starting in;the first step results in no adverse' impact on plant >
, cafety. The CAC contributes'34.1 KW-and the CS contributes 161 KW for a. total-
- of 195;2 IN' additional- load starting in Step 1 of the sequencer. The sum of l .
'this additional load, withLthe previously identified Step 1 loads results in a st ap -. load well below the EDGs performance capability. 0f withst anding a single ll ster. load of 110% of' rated EDG load _(EDG rated at 2600 IN) .
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- s. There vill be no adverse effects on safety as a result of the proposed .l l elementary drawing changes. The changet, proposed for the elementary drawings l depicting the EDGs breaker control and the 4.16 KV switchgear relaying are to !
revise typographical etrots r. sly. There is no increase in hazards as a result i of the revisions proposed. The changes to the plant drawings, including USAR Figure 8.3-9 increase drawing consistency and will theref ore, have no adverse
- i. - elfett on--safety, ;
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t SAFETY EVALVATIOll SUll!!ARY FOR DCR 91-0013 (SE 91 00$1) {
TITLE:
Change in Condens< <
Sump Flow Path !
CH AllGE : j DCR 91-0013 will change the system lineup to reflect a new normal line-up
_ routing the turbine building condenser sumps to the settling basin.
REAS0!J FOR CHAllGE Due to the increase in the secondary side radioactivity levels DB-0P.06272 Station Drainage and Discharge System, has been revised to normally al.gn the
- tutbine building condensate pit numps to the settling basin.
SAFETY EVALUATIO!1 SUlillARY:
The original line-up for the turbine building condenser pit sumps was through all. interceptors into the station storm sewer system which drains by gravity ,
int o t he Toussaint River. Primary to secondary leakage in the steam generators bas-increased activity _ levels in the fluid collected in the condense. pit -
numps. -The Toussaint River is not an authorized radioactive effluent discharge path, therefore, the condenset pit sump fluid flow path has been changed to discharge into the setting basin. The settling basin effluent is discharged ,
thtough the collection box to Lake Erie which is an authorized radioactive effluent release path. .
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SAFETY EVALUATIO!1 $UlMARY FOR DCR 91-0026 (SE 91-0073)
TITLE:
Addition of Pressure Indicator and Isolation Valve on the Flash Tank Cil AllGE :
I Revision of USAR figure to show the addition of
- pressure indicator and isolation valve on the-flash tank. !
REASoil FOR CHA11 gel-Document Chango Request (DCR) 91-0026 updatet. U!iAR. Figure 10.1-2 to show the addition of Pressure Indicator (Pil 2712 and isolation valve AS271' JC. This gauge is int ended to rnonitor pressure in the Flash Tank (7113) . The gauge has
. been evaluated as necessary for performance of DB.00-00100, Boric Acid Evaporation Operating Procedure and the drawings are being updated to reflect
'as-built
- conditions.
SAFETY EVALUATION SU141ARY:
The. proposed addition of PI 2712 will have no adverse affect on the Flash Tank '
or Steam Trap Hender. This P1 is installed on a test connection specificclly used to monitor pressure in the Flash Tank. The addition of the P1 will not change the f unction 'of the Flash Tank or t he Steam Trap Header therefote. At will have no adverse affect on safety.
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I SAFETY EVALUATION SUlRiARY FOR
- DCR 91-0036 (SE 91-0053, R01)
, TITLE:
Re-rate liigh Pressure Injectior Discharge Line l CllANGE l
Re-rate the HPI discharge line to a value greater than ' e Makeup System l pressure. This is being accomplished by re-rating the piping between valves ,
HP-2A, B, C and D and valves HP-22 and 11P-25 f rom 2000 psig and 2000F to 2600 psig and 2000F and the piping between valves HP-2A, B, C and D and valves ,
HP-48. HP-49, llP-56, and HP-57 from 2000-psig and 2600F to 2500 peig and 6500F. ;
The piping f rom !!U-6421 and MU-6422 to the HPI discharge line was also re-rated '
to 2500 psig and 6500P for consistency between Specification M-200 and Drawing M-602, REASON FOR CHANGE:
- PCAQ 90-0695 was generated following observation that the HPI piping upstream of valve HP-2B was pressurized to approximately 2250 psig due to leakage of Makeup System-injection past valve HP-2B. This pressure exceeds the present design rating of 2000 psig for the piping involved.
SAFETY EVALUATION SUlRiARY:
Implementation of this DCR will not affect plant safety.
This is a paperwork only change. No physical changes will be performed. The
- affected components were purchased with adequate requirements (i.e., 1500 lb.
class valves, thickness and material of piping, etc.) that this piping run ,
could.have originally been rated at_the new valuesi however, the_ approach used' at the time of construction was to rate the piping at a value dependent on the shutoff head of the associated pump. _ Leakage from adjacent higher pressure systems was apparently not considered when originally rating this piping.
The ability of individual components (pipe, flanges, valves, fittings) to withstand a pressure / temperature of 2500 psig/6500F for piping between_ valves llF2A, 2B,-2C, and 2D and valves HP-40, ilP-49, HP-56 HP-57, MU-6421 and MU-6422 has been analyzed and is documented in Calculation C-HE-65.01-108. The ability of the piping-between valves llP-2A, 2B, 2C and 2D and valves HP-22 and HP-23 to withstand a pressure / temperature of 2600 psig/2000F has also been analyzed.
Calculations demonstrate that the material is suitable for the new design pressure / temperature.
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A " view of the original hydrontatic tent tevealed that Train 1 of this piping was tented at 3889 psig while Train 2 was tested at 3812 psig. Since the lowet of these exceeds the hydtostatic test p r e ta u r e required by ASitE Section 111-1971 f or the new des ign pr e s sut e / t etnpe r atur e , t he hydr ost atic t est s also demonst rate the adequacy of the r evised design r at ing f or t he piping bet ween valves llP- 2 A, t C and D and valves llP-2 2 and 11P- 2 3. The piping betwee llP- 2 A , 11 , C and D and valves 110-40, llP- 4 9, 11P-56 IIP- 5 / , 11U- 6 4 21 and !!U-6421 has been included within t he boundaty of the hydt os t at ic test of the Reactor Coolant System. The Reactor Coolant System hydtortatic test is petformad at a pt r>s sut e cennistent with a design rating of 2500 psig/6500T since thin is the design t at ing of the teac t or Coolant system.
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i SAFETY EVALUATIO!! SUtiliARY FOR DCL 91-0045 (SE 91-0059, RO1)
T_l '!_L E :
Femoval of SW-5423 and SW 96 CH AllGE t Revisjon of USAR figure to reflect the t ernoval of valves SW-96 and SW-5423.
SV-5423, a discharge valve Ior ECCS Roem Coole C31 03, was temoved itom set vh e by Tli 89-0050. This Tit was init iated since a two inch bypass line around SW-5423 drains to a redundant Service Water headet for all five ECCS toom coolern. Continuoun flow of Setvice Watet is maintained through this parallel return path to avoid developing microbiological 1y induced corrosion nnd t he nerumulation of sediment in the coil.
SW-96 is an isolation valve f or SW-5423 pe tmitting it s temoval f or maint enance.
With SW-5423 removed, SW-96 is no longer sequired and ..as t 'e rno v e d .
PEAS 0!! FOR CHAllGE:
Valven SW-96 and SW-5423 wete temoved from the Service Wat er System and replaced with sections of pipe via Temporary !! edification 09 0050. DCR 91-0045 tevises the drawinp , the USAR and other documents.
S AFETY EVALUATIO!! SUtil1ARY:
Implementation of this DCR will not affect plant safety.
Thin is a paperwork only chat.ge. 11 0 physical changes will be performed. The physical changes were performed via Temporary Modification 89-0050 wnich -
included Safety Evaluation 89-0290. Safety Evaluation 89-0290 concluded that -
the Temporary Modification did not create an unreviewed safety question based on the followings the ECCS Room Cooler would continue to maintain a suitable room temperature because t he cooling wat er flow would be limit ed by the size of the inlet piping the operation of t he room cooler blower would not be affectedt the sein~ic capacit y of the piping would not be degraded; the integrity of the sine would not be degraded, therefore the change of a line br eak prevent ing the flow of Service Water to the Engineered Safety Features components was not increased; the ECCS Room Cooler could still be isolated in the event of a Service Wat er line break t the temoval of the Limitorque actuated valve removed a potential failute mode f or Service Wat er flow t ht ough t he room coolet: and the removal of Valves SW-5423 and SW-96 did not affect the safety function of the Service Water System or ECCS Room Coolers.
The assettions and conclusiens of Safety Evaluation 89-0290 temain valid for indefinite use of the system as modified.
SAFETY EVALUATIOll SUl!!iARY FOR DCR 91-0074 (SE 91-0076)
TITLE:
Cottection of Ilotor Data of the Diesel 011 Storage Tank Transfer Pumps and Fuse Slze for itCCs D1 and D2 CilAllGE :
Document Change Request (DCR) 91-0074 revises the motor rating of the diesel oil storage tank transfer pumps (11P195-1, HP195-2 ) and the fuse sizes for the 125/250V DC HCC, D1 and D2 voltmeters._ USAR Table 8.3-1 and USAR Figures 8.3-23.- 8.3-24,-8.3 46 and 8.3-47 are being revised to reflect changes in rating associated with motots and fuses, ptAGoll FOR CilANGE USAR Table 8.3-1. USAR Figure 0.3 23 and USAh Figure 8.3-24 are being revised to show the revised motor rating for HP195-1 and MP195-2 is 2HP (1.5 kW).
Motor Data Sheet 7749-H-74A-8. Vendor Manual H-74-35. Design Specification H-74A, and field walkdown data agree the MP195-1 and HP195 2 are 2HP (1.5 kW) motors.
USAR Figure 8.3-46 and USAR Figure 8.3-47 are being revised to show the revised fuse size, as presently installed, f or HCCr. D1 and D2 is 3 amperes instead of 10:amperen. The 3 ampete fuse vill adequately supply the load (i.e..
voltmeters) whlie providing appropriate protection for the circuit and coordination with the panel fuse.
SAFETY EVALUATION SUltMARY:
-DCR-91-0074 will have no impact on the safety function-of the Emergency Diesel Generator System or the 125/250V DC Power System.
There is no effect on the plant or an increase in hazards,-because DCR 91-0074 is.a " Drawing Change only' document. The change to plant drawings and the USAP increases drawing consistency and will 1. ave no adverse effeet on safety, and ate In conformance with the plant design basis and will not increase the radiological ~ consequences of an equipment malfunction.
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SAFETY EVALU.ATIO!! SUM 1ARY FOR DCR 91-0097 (SE 91-0063. R02 and SE 92 0041) and Til 91-0013 (SE 91-0030)
TITLE:
Laundry facility Connections and Operation of the DE!JPS Wet Wash Facility Cil AllGE :
Til 91-0013- will provide electrical, ws.ter (demineralized or domestic), fire protection, telephone, and drainage connections for the new laundry facility.
The purpose of saf ety evaluations 91-00ti3 and 92-0041 are to evaluate potential offnite dose consequences from operation of the wet wash facility and j demonstrate that any potential releases of radioactivity would not exceed i regulatory requirements. j REAS011 FOR CliA!JGE - ,
The existing Protective Clothing (PC) laundry facility utilizes freon as the cleaning agent'to clean the PC. Due to environmental, economic, and disposal concerns -this method of laundering PC is no longer desirable or prudent. The current pref erred method f or- laundering PC utilizes detergent and water an the !
cleaning media and hot air dryers for drying.
- SAFETY EVALUATIO11 SUletARY:
The temporary connections to the systems af fected will not af fect the safe operation of the stationt nor will they affect the ability to safely shutdown the plant. The temporary modifications are not located near any equipment important to safety which could be affected by the failure of any component
. installed for the temporary modification.
v The addition of the connections to the systems affected will not change the -p design basis'of any. interfacing-systems.
The security. requirements for the connections are met through the use of l
existing penetrations. ,
Radiological Control;has evaluated the radiological consequences for operating the Vet Wash Facility. Both liquid and-gaseous effluent pathways were analyzed. The pathways were evalvated in accordance with Offsite Bose Calculation llanual methodology as required by Technical Specifications. -
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Any leakage ftom hoses, pipes, et c . , wit hin the haundty Facility is contained by cut bs within the Facilit y. If gross leakage occurs, the Facility is designed to drain to the Radiologically Conttolled Area where it i s intercepted by f loor dralna connected t o t he 111scellaneous Waste Drain Tank (t!UDT) . 1he
- capacity of the ifWDT is lat ge enough to accept all leakage f rom t his f acilit y.
' The facility 18 designed to launder cont aminated clothing using watet item the Demineralleed Water System. Demestic Water System or Recycle System. When the Facility is operated with Deminetalized or Domestic Water, laundty wash liqueur is pumped. 111tered and directed to the Detergent Waste Drain Tank (DWDT). A back-flow device is installed in the Demineralized and Domestic Water 11ake-up line to prevent potentially tedioactive contaminated water from entering the Dernineralized and Domestic Water System 9. Additionally, disconnects are installed on the laundry washers which tequire the laundry operator to physice.1]y connect to or disconnet t from water supplies to prevent potential
- system cross contamination. The additienal demand placed upon t he hlquid Radwaste System is within the-design basis of the system. The Pacility is nottnlly operated in Recycle Flode where the laundry wash liqueut is pumped through a (11t ration syst em and directed back to the Recycle Tank. The Recycle Tank pr ovides make-up wat.cr f or the washers and established a closed loop system to minimize liquid tadveste generation. The flitration system removes oil and particulate mattet to 1 micton from the laundry wash liqueut before it in dischar ged- to t he DVDT or Recycle Tank.
Sample points are provided in the filtration system to petmit measurement of r adiological and biological gr owth in the wash liqueur when the system is operated in the Recycle Mode, t o opt imize system opetation and perf ormance.
The filt er s generated are considered radwaste and will be processed and disposed of in accordance with Plant procedures. The filtration skid is equipped with a check valve to prevent back flow of water from the DVDT to the haundry Facility environment. The total curies released as a result of an additional 3.5ES gallons per yeat of liquid processed is 1.9E2 curies. The resulting of f site dose for- the additional activit y released is approximat ely 7 percent of the Technical Specification limit.
The levels of radioactivity contained in the wash water would not adversely impact of f site' dose even if the wat er was released to the environment via the storm sewer nystem.
Once it has been determined the HEPA units have been compromised dryet ,
operation shall be terminated until the filters are replaccd. The operation of the wet wash facility will not adversely affect any margin of safety nor impact offsite dose.
The Facility is further protected with a fire suppression system capable of suppressing any fire in accordance with flational Fire Protection Association
~
(ilFPA) standards.
- The water supply, fire supptession supply, and-drcia It'ae have had a seismic II/I evaluation and comply wit h t he criterien for this evaluation.
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SAFETY EVAltfAT10tl SUlU!ARY FOR l FCR 86-0272 (SE 89 0241 R02) l TITLE: ;
Replacement of Two Cybetex Inverters CllA!1GE :
FCR 86 0272 ptoposes to replace the four Cyt+ rex inverters YVJ, YV2, YV3, and YV4 with Solid State Controls, Incorporated ; SCI) inverters which provide Class IE 120V unintet ruptible power to the safety related systems such as Reactor Ptotection System (RPS), Safety Feature Actuation System (57AS). Steam and Feedwater Rupture Control System (SFECS), and various essential 120V AC
- instrument loads.
FOR 86-0272, Supplement 0, was originally planned to replace all four >
inverters, however, due to the limited duration of the 6RFO, it was proposed to proceed with only partial implementation of this modification during 6RFO.
Due t o the partial implementation of this FCR. two Cyberex inverters YV2 and v YV3 will be in operation during the Seventh fuel Operating Cycle. FCR 86-0772, t Supplement 1. describes the modifications being petformed on those two Cyberex invertern and its impact on safety.
REAS011 FOR CilANGE:
The primary reason for the replacement of the Cybetex invertets with the SCI invertern is the inability of the Cyberex inverters to clear-the short circuit faults on branch circuit feeders. The concern is that in the event that a Cyberex inverter fa! - to clear the fault due to a fire and goes in current >
11miting, and a lost, d offsite power occurs soon thereafter, there is no source of 120V AC Essential Instrumentation Power left for the Distribution Panels Y2, Y2A, and Y3. This is because the present source of alternate power to these panels is from Panele YAR and YllR which are non-safety related and with the loss of offsite power, there 19 no alternate source of_. power such as Emetgency Diesel Ceneratots available for these panels.
SAFETY EVALUATION SU141ARY:
The proposed interim modification will.not have any adverse effect on any of the cafety related systems. The electrical equipment needed to implement this
. modification is purchased fluclear Class 1E and will be installed seismic Category I and therefore, will-not effect the safe functioning of the other safety related systems,
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SAFETY EVALUATI0ft SUIMARY FOR FPR 77-n334 001 (SE 92-0001)
- TITLE:
Revision of USAR Figure to show "As Built
- Configuration CllAl1GE :
Removal of valves and pressure indicator from decay heat drawing to show the "As-Du11t* condition. l REASON FOR CilA!1GE:
The refueling canal. drain line and drain pump were installed per FCR 77 0334 in -
order to drain the deep end of the refueling canal to the Borated Water Storage Tank (BWST). The-FCR was interrupted while'it was being implemented, and certain steps were deleted. .Dil 106, the drain valve for the refueling canal drain pump, was not: Installed, and a drain plug was used in its place. Dit 9839 and PI 9839. the root valve and pressure indicator on the discharge of the
- refueling canal drain pump, were also not inst alled.
- Although Dil 106. DH 9839, and PI 9839 were'not installed in the plant they were included on plant. drawings and documents.
. FPR 77-0334-001 will remove DH 100 Dil 9839, and PI 9839 from applicable plant doctunentation to show the "As-Built * - condition.
SAFETY EVALUATION SUtMARY:
The specific components that will be deleted from plant drawings and documents by FPR 77-0334-001 are all associated with the non-safety related refueling
- canal drain line. DH_106 was-to have been the drain-valve-for the refueling canal drain pump. It was never installed and_a drain pl,g was used in its
- place, The drain valve is required to be verified closed in DB-OP-06023. Fill, !
Drain, and Purification.of the Refue13ng Canal, _ _The drain plug as installed is-adequate for the-pump.and piping, and the use of a drain plug'in piece of a
- valve will have no effect on safety.
DH 9839:and P1 9839 were intended to provide local indication of the discharge
_ pressure of the refueling: canal drain pamp, but-this indication is not necesnary to safely _ operate the pump. If required temporary pressure indicators can be installed in the drain line at vent or drain connections to monitor discharge pressure.
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SAFETY EVALUATIO!! SUlfhARY FOR FPR 82-014R 001 atul UC11 92-00 3 (SE 92-0010)
IIIL.E :
Concenttate Storage Tank Filtet CllAtlGE Reviso USAR Figate 11.2 2 to refleet FCR 82-0148 REASOli FOR CllA!!GE:
Facilit y Change Request (FCR) 82-0148 installed a particulate filter on the -
discharge of the Concentrate St ora ge Tank ttansfer pump using stainless steel hoses. Thin was regarded sa a temporary inst allation with the hoses t o be replaced by piping and heat traced. The purpose of FFR 92-0148-001 and UCli 92-003 is to update documentation to indicate that the stainless steel hose installation is permanent.
SAFETY EVALUAT1011 SUMitARY:
The proposed document changes incorporate the Concentrate Storage Tank Filtet
, into the configuration control system. Additionally, a local pressure gage nhown on some drawings is deleted. This gage does not exist in the field and is not necessary for filtet operation. 11 0 hardware is affected by these changen. The reliability of the Clean Liquid Fadwaste ISstem 19 unaffected.
The use of the stainless steel hose on a permanent basi- is acceptable because the hone is constructed flom allowable material and is compatible with the se rvice conditions of 115C-13. The installation is non-seismic, and an EIT walkdown conducted for the original FCR concluded that there were no hazards from this installation. The lack of heat tracing has been compensated for by -
procedural requirements to flush the hoses and illter following use.
/
l SAFETY EVALUAT10!1 SUlillARY FOP FPR 91-1405-901 (SE 91-0077)
TITLE:
Replacement of Containment Smoke Det ect ors in Fire Detection Zone 317 Cil AtlGE i Replacement of 26 Alison smoke detectors with a new Pytotronics model in Containment FDZ 317. Addit iona lly . FDZ 317 Pyrotronics smoke detectors will be supervised by a Pyrot ronics zone module located in Fire Alatm Panel C4720A.
RFASoll FOR CJ1311GE:
The Alison 11odel A-12000-2S radiation hardened ionization smoke detector, which is used exclusively within the Davis-resse containment, is obsolete and the site han limited spares in stock. Several Alison r,moke detectors monitoring Fire Detection Zones (FDZ) 317 and 410 have failed. Lased on the number of smoke detectors assigned to monitot FDZ 317 (26 dc : tots) and the fact that these detectors are accessible without scaffolding, it was decided to explote replacement of the aging Alison smoke detectors with new Pyrott onics ionization smoke detectors and bases.
SAFETY EVALUATIOll SUllMARY:
The changes proposed by MWO FPR 91-1485-901 will not adversely affect the tunct ion / operation of the Station's fire detection system.
o The Pyrotronics model D1-6 detectors are compatible with the Station's Pyrotronics System 3 tite alarm panels (C4720A). The twenty-six smoke detectors deeigned to monitor FDZ317 remains within the maximum allowable (thirty detectors) for a Pyrotronics zone module.
o The Pyrotronics model DI-6 smoke detector has demonstrated te11 ability as this detector it currently being used to monitor other fire detection zones in the Auxiliary Building.
o Investigations revealed that model DI-6's were tested per UL-268 (Section 43 - variable ambient temperature test) which requires that a detector operate for its intended sJgnaling performance when tested in ambient temperatures of 00C (320F) and 490C (1200F). Review of inlet air g temperatures for Containment Air Coolers 1, 2 and 3 show that the highest
? average temperature, which was recorded for the air inlet of a Containment Ait cler was 117.90F. Thus, it is concluded that the avetage ambient tem,crature of FDZ 3.17 will remain within t he UL tested limit s of the DI-6 detector.
, s USAR Table 7.2-3 reports a relative humidity of 80 percent in the containment vessel. This humidity is within the published limit (831) for the D1-6 detector.
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i o The electrical equipment qualification giveu a 40 year normal operation
- dose of 2.54 E4-(.073 rad /hr) for Room 317. Based on exposure to gamma i radiation and an ' approved sensitivity decrease of two times
- the Pyrotronics model DI-6 detector could survive in excess of 78 years.
l Thus, the expected dose in-Eoom 317 will not result in premature failute
- of these smoke detectors.
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h o Based on personal obw rvation, the weight, material and construction of ihe Alison model 12000-2S and the Pyrot ronics model DI-6 with DB-4TS base l- are comparable. Thus, the DI-6 smoke detector would be enveloped by the j existing seismic 11/1 evaluation for Room 317
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SAI'ET Y EVALUATIO!! SUM!iAP.Y yon
'10D 8 7-100 (SE 90-0149) r TITLE:
Installation of a Three Way Elect ro-i neumat ic ASCO Solenoid Valve in t he Condentate System C. i_l A.t._lG C Thin !!adification 87-1003 installs a three way electro-pneumatic ASCO solenoid q valve (SV-578A) between t he actuator of Condensate Recirculation Valve CD-578 and itn exicting solenoid valve SV-579. The new solenoid valve vili se poweted
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from panel DAP (non-es9ential 125 VDC) and will not be part of the valve /condensat e pump logic REASO!! FnR _CitA!1CE :
Thin $h.alfication will enable CD-578 to fall open on losn of powet. Ptenently, 00 *,78 lailn open ,n loss of air ut coottol signal. The condennate tecirculation valve CD-578 in not safety telated and in non-q.
SAFETY EVALUATI0ll_ SU1111APY:
The new solenoid valve being added by liodificat ion 87-1063 will be notmally enetnized and poweted by panel DAP. Undet normal operating conditions, the new nolenoid valve will be transpatent to the operation of CD-578. On loss of power on panel DAP. the new solenoid valve will simultaneously isolate the normal actuator alt- supply and vent the actuator of CD-578 allowing the valve to fall open, t her eby, providing minimum recirculation flow on low condensat e flow conditlens and protecting the condensate pumps from overheating.
The system and components dir ectly af f ect ed by 11odification 87-10u3 are the -
Condensate System, Inst rument Alt System and the 125/250 VDC lion-essential Power Distribut.lon System, nene of which perform a safety function. Thetefore, from a satety function standpoint, the plant in not affected.
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SAFE 1Y EVALUAT10!3 SUlt!!ARY FOR liOD 87-1167 (SE 89-0138) j i
TITLE: I l
liain Generator Sequential Tripping Cll A!1GE :
I l
Install a separate reverse power telay to help detect the flow of power into the tutbine-generator. When the leftover steam in the turbine has been dionipated in feeding power to the-grid and power statts flowing from the !
transmbolen line into the generator resulting in it s motoring, the reverse power relay in energized t hus t r ipping the 345 KV power citcuit breaker which
-isolates t he turbine-gener ator f rom t he tr ansmisalon line. This modification eliminates the risk of the turbine-generator overspeed. ,
< PEAsoll FOR CllAl1GE:
This enodification deals with the improvernent in itain Generator sequential t ripping circuit based on the General Electric Company Technical Information Letter (TIL) 886. The recommendation in General Electric Company (TIL) deals
- with the modification of the existing circuit in order to reduce the risk of
- turbine-F,enerator overspeed.
~ 04FETY EVALUATIO!! SUNIARY: .
The changes made ander this modification to the Anti-110toring circuit of the Turbine-Generator do not affect the safety of the plant, The changes performed reduce the risk of damage to the turbine-generator due to overspeed if the #
345 KV circuit breakers are opened premstutely before the residual steam in the turbine has been expanded and exhausted to the condenser.
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4 SAFETY EVALUAT1011 SUtMARY FOR MOD 8 7 1315 (SE 90-0028, t02)
TITLE:
Peplacement of Service Water Valves SW-1424, 1429 and 1434 CilAllGE Peplat e er.10 t ing Se t vice Wa t e r Valve s SW-1424, 6W-1429, and SW-1434 with
!!elow-Janie9 bury QLM-tall Valves.
FILAsoft FOR CHA!1GE:
Service Wate System buttetfly control valven SW-1424 1429, and 1434 have hnd a bl9toty of fallute problems at the Davis.Pesse ILclear Power St at ion. The valven in current use ate 16-luch Class 150 butterfly valves of syrunet rical disc desi;.,a f urnished by ITT llanunel Dahl Company. The replacement valves, llelen-Jamesbury QLii-tall Valves, a r e tno t e suit able for flow cont r ol ove r the operatlug iange of the seivice water supply to the component cooling watet heat e x c h an r,e t 9 .
S AFETY EVALUATIO!1 SUlitiARY:
The proposed modifitat ion will not affect the fonetions of the setvice Water, Component Cooling Water. Wast e Gas Decay and Exhaust . or the Station Al t / Innt t ument Air Systems.
The replacement fieles-Jamesbury ball cont rol valven utill e parallel perforated plates in the ball flow opening t o stage the pressure drop as the flow panses through. An the valve opens, flow is forced through the holes in the perf oratad attenuator plates. The plates eteate a frictional path, where each plate and the seating orifices teduce the pressure step by step. This prevents -
excessive velocity generat ion, lowers the noise level, and minimites cavitation and eroclon. As the opening angle is increased, resistance decreases as the flow bypasses the plates. The QLit ball has an increased f tictional path in the first chamber between the attenuator plates. This deolgn provides precise control of high ptessute differentinis while allowing high flow when fully open.
The tieles-Jant stury QLH ball valves ate not currently available as AS!!E fi- s t amped valves . Therefote, the valves will be procuted and installed under the provisions of 11EC Generic Letter 80-09.
l The replatement CCW heat exchanger outlet control valves are designed to provide the-requited Snvite Water flow over the entire notmal range of operation. The roplu ement valve / actuator assemblies are designed to iall open ,
upon loss of air and/or power t o any of the actuator except power to the limit !
switchen. Receipt of an SFAS level 2 signal to the valve control loup will '
cause the valve to fail to the open position to provide for maximum tooling of CCW. *fhe service water system flow calculations were reviewed with respect to the replacement of the CCW llest Exchangers output Control Valves and modification of the piping. Comparison of the performance data for the !
replacement valves and the values und in the flow esiculations indicate that the net change in system pressute drop will not significantly change the system flow balance during emergency operation. ,
The existing valven may be closed using air utored in one of the accumulators in the ovent IA is lost. The teplacement valves do not have this capability.
The capability to close SW 1424 SW 1429 and SW 1434 using air stored in the accumulator tanks is not a requit ed f unct ion f or t he t eplacement valves . The replacement valves can be manually closed using a handwheel override after a lons of IA.
The installation of the new valves and the modif1 cations to the Service Water piping downstream of the CCW heat exchangers has been evaluated by calculation and determined to be acceptable.
Rerouting of CCW piping 1*-!!BC-39, 1\*-ItBC.E22. and relocation of drain valve CC-21 have no effect on CCW system function. The piping changes have been analyzed by calculations and determined to be arceptable.
lleles-famenbury has analyzed the valve / actuator assembly to determine the seismic loading the valve can withstand. The results of the analysis were compared to the. loading imposed on the valve by the piping system and han been det etmined t a be acceptable. ,
Relocation of PDIS 2002, terouting of the presuure sensor tubing and rerouting of the in9ttument signal conduit, will not affect the function of the switch.
The PDIS, modified tubing, and conduit will be installed Seismic Class 1. .
The new instrument air supply line will tie into the IA system downstream of normal closed isolation valve 1A 530. IA 530 is on the supply header for the spent. fuel pool pump room. The new IA piping will be installed Seismic 11/1 t o preclude any new seismic hazard to existing equipment. The new instrument air supply for the CCW heat exchanger outlet control valves will not adversely affect instrument air system operation as adequate capacity axists in the system and in the supply header to the SFP pump room to add tae new demand.
l
SAFETY EVAttfAT10!1 SUlt!!ARY FOR
!!OD 88-0067 (SE 90-0125) l TITLE:
Elettric Fite Pump controllet Replacement CHAllCM The purpose of liodification 88-0067 is to replace the existing Lexicon controller which statts the Electric Fire Pump (EFP) when the ftre supprension system pressure drops to 120 PSI. In addition, the interlock to stop the EFP on. Fire Water Storage Tank.(FWST) low level will be eliminated to rneet the j tequirements of 11FPA-20 Section 7-5.4. j REASO!1 For, CHANGE:
The cut tet1t Lexicon controller is obsolete And replacement parts ate not available. The new EFP cont roller snects the tequirements of NFPA-20. UFPA-20, '
Section 7-5.4 allows autetaatic shutdown af ter automatic start, only after statting causes have returned to normal. The FW$T may empty before system prensure returns to notmal. Therefore, NFPA-20 does not allow automatic i shutdown on FWST low level.
SAFETY EVAttfATIO!1 StrigiARY:
The only design change being made under 110D 88 0067 is to eliminete the automatic shut down of ' the. EFP on FWST low level to meet the- requirements of flFPA-20. 11FPA-20 does not allow automatic shutdown of the EFP unless the system returns to normal pressure, In the existing configuration, once the EFI has been started, it will automatically shutdown when the water in the FWST teaches a level of three feet as sensed by level switch Ls1051A. The deletion of=the FWST low level shutdown of the EFP requires that-the EFP 'oe manually shutdown at--local-panel C3024 aftet starting.
The new controller meets the. requirements of NFPA-20 and 11FPA-70. The-new controller is a like-for-like replacement for the old controller. Therefore, no adverse effects on saf ety is created with the installation of the new EFP controller.
During a-tire situation, the EFP will automatically start on low system pressute and pump water as designed until manually shut down or until the FWST empties. If the FWST empties and the EFP is not shutdown, damage may occur to '
- the pump and/or motor. However, the EFP will have completed its required function and the Diesel Driven-Fire' Pump will have st arted.
- Level Sw3tch LSIL31A creates a potential failure mode which could prevent the EFP from performing its tequired function. -The remoral of the FWST low level
' interlock to EFP improves the reliability of the Fire Suppres, on System in-that the EFP will operate with a failure of LS1051A and therefore, does not have'any adverse effects.on safety.
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$ AITTY 1;V A1,UAT1 Ol1 SUl U1Al;Y l' O R 110D 8 0 - 010 5 (SE 40-0132) 11 T11:
temoval of IJittogen Cover Gan Ftom the 00: 1< Acid Concentrates Storage Tank Cll Af1Gr i Remove t he tii t t ogen cove r gan f rom t he l'ut ic Acid Concent rates Stotage Tanh.
[ F A T,011 ThE CHAllGE:
The scope of this modiifration i t. to temove the nit togen cover gas f 1om t he 1:ot ic Acid Concent rates St orage Tank This is to be achieved by isolat ing t h" inlet t.ontrol volve PCS 19n7, and by removing PCV 1910 and replacing it uit h a iInnged vent pipe, meking it en at mosphe t it tank. Pelief Valve pSV 1908 vill be npated in plare. This action is being taken to simplif y t he system and eliminate a maintenance ptoblem when liquid carryovet into the tank's covet gan diccharge header cauren the formation of bot on t t y n t a19 on PCV 1910.
SAF ETY F_ val.UAT10!1 SUlittARY:
1he removal of the nitrogen cover gas has been evaluated and has been determined that the nitrogen covet gas is not tequired for 1) protection against development of an explosive mixtutet 2) control of the concent ration of dionolved oxygen in t he concentrated boric acid and 3) prevention of evaporation of the contents of the tank.
Thetefore, the conclusion tan be drawn that the removal of nitrogen cover gas from the BACST will not create any adverse conditions or degrade the performance of the system.
The vent ing of the tank to r oom atmosphet e was also evaluated and determined that is doeri not create any adverse conditions or degrade the performance of the syrtem. This van det ermined since i stent ial radioactive discharges are routed to locations where propet processing has been verified.
d
SAFETY EVALUATIOll SUMliARY FOR
TITLE:
1 Replacement of Alison Smoke Detectors in the Auxiliary Building and Annulus l Space j CHAliGE :
Replace Alisen smeke detectors, installed in Auxiliary Building and Annulus wit ,
-- Pytotronjes Model DI-0.
REASoli FOR CHAtlGE:
- The 'Alison Model 1200-RAD radiation hardened ionization smoke detect or, is ;
f obsolete and there are no. spares in stock, To obtain spare Alison detectors. Plant liodification 88-0188 replaces the ;
A1190n detectors currently installed in the Auxiliary Building and Annulus Space with Pyrotronics Model DI-6 ionization smoke detectors. The Alison
- detectors vill be used f or spare parts f or the Alison detectors in the containment.
SAFETY EVALUATIO!J
SUMMARY
. The changes. proposed by liOD 88-0188 will not adversely affect the function / operation of the Station's fire detection system.
pyrot ronics model DI:.6 are compatible with the Station's Pyrot ronics System 3 fire.alorm panels.
i The Model DI-6 has demonstrated reliability as this detector-is used in-other
- Station fire detection zones.
The effects of radiation on the 11odel DI-6 have been evaluated. This evaluation concluded that the Model Dl-6 will have a minimum service life of -
11.4 years.
. The Model DI-6 has been tested per Uh-268. This test requires detector
- operation in' ambient temperatures of 1200F. This temperature is above or equal to'the expected normal operating temperatures of the affected areas.
' The. replacement detector head and base assembly will not overload the existing
- supports; .The combined weight of detector head and base assembly is approximately one-third the maximum' design weight of the existing supports. ,
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l SAFETY EVALUATIoll SU1111ARY !
FOR 110D 88-0206 (SE 90-0122) i TITLE:
Replace Evaporat or Condensate Drain Pumps P167-1 and P167-2 CHAllGE : 7 The existing Condensate Drain Pumps which are centrifugal pumps are being .
.teplaced by two ptessure powered pumps. The installation will require the following additional changes be-made: 1) the supply and discharge piping will be reconfigured in order to accommodate the new pumps: 2) a pump vent i connection will be provided: 3) two check valve bodies (valve internals !
previously temoved.by FOR 77-0373) associated with Steam Traps ST98 and ST99 in the boric acid. addit.lon rooms will be teplaced with isolation valves: and
- 4) finally, a connectton itom the St ation Air System will be installed t o provide the motive force for pump operation.
REAS011 FOR CHA!1GE:
The existing Condensate Drain' Pumps P167-1 and P167-2 are obsolete. Recurring maintenance problems and difficult repairs necessitate their replacement.
SAFETY EVALUAT10ll StritliARY:
~ Modification 88-0206 will not adversely affect plant safety for the following reasons o The new pumps will be installed in the same location as the old pumps. ;
Failure of the new piping would not create any new hazards. The postulated water and steam environment is bounded by failure of the existing pumps and piping.
o The-revised piping connections will be seismically analyzed and supported where necessary'to protect safety related piping, conduit or equipment ,
during'a design bases seismic event.
o A slight increase in humidity caused by the pump vent is expected during pump' operation and wil1~not adversely impact the operation of any safety related equipment, o The air requirements' for the new pumps is less than 20 SCFM. This is well-within..the capacity of the Station Air System. The Instrument Air System will not be affected. _
o Breakers 89 and 71 on MCC 21D and 210 will be spared out. This raduces the loading on the motor cont rol centers.
o the new pumps will be functionally equivalent to the old pus The 10 psig Condensate System will perform.as designed, r
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SAFETY EVALUAT10!i SUlillARY FOR 110D 68-02 2 7 (SE 91-0008)
TITLE:
Inut a11at ion of Synchronism Chm k Relays fot Eteakets AC101 and AD101 CH At1GE :
Plant Modification 88-0227 adds synthretdsm (sync) chetk telays to prevent snanual cloning of the emergency diesel generator output breakers (AC101 and AD101) 11 the voltages on either side of these breakers are not in a ptedetermined f requerity , phase , and magnitude t elat ionship.
NOTE: Synchronism check telays were only installed on EDG 1-1 R.EASON FOR Cil AllGE :
This modification implements Eabcock and Wilcox ownets Gtoup recommendation TR-144-TES. This recommendation directs members to evaluate hardware which will teduce the likelihood of extensive diesel neneratot damage as a tesult of patalleling out of sync.
SAFETY EVALUATION SUtiliARY:
Adding the sync-check relays will not adversely affect the safety function of the EDGs or the output breakers.
The fnc-check relay permissive contacts are not being connected in series with contacts which automatically close the EDG output breakers on loss of essential bus voltage. Thus, the st ate of the sync-caeck relay permissive contacts cannot affect auto closing of these out put breaket s. The new sync-cheth telay is provided with features which, if selected, will defeat the normal sync check -
function. Selection of the HLDD (Hot Line--Dead Bus) feature permits loading the EDG wit h c bus at less than or equal to 50 percent voltage. Thus, manual loading of a " Dead Dus", as backup t o auto closing, is pet._itted when HLDB is selected.
Installation of these syne-check relayn in the elect rical control and relay boards will not degrade the seismic I category classification of these cabinets. Relay " contact bounce
- will not. tesult in undesirable circuit operation as sync-check relay cont ac t s are permissives, normally isolated from the breaker's closing coil.
1
._ _... _ _ _ _ ._ . - . _ . _ ._ _ ._ _ _ _ _ _ _ - ~ . _ . - - . _ __ _ _ _ _ m SAFETY EVAtt1AT102: StHliARY FOR HOD 89-0025 (SE 90 0150)
TITLE:
Installation of a Wet Pipe Sprinkler System in Service Building #6 CHANGE:
Modification'89 002$ will provide for the installation of a wet pipe sprinkler system for Service Building #6.
REASoll FOR CHAllGE:
- The Fire Water Distribution Loop was installed with a 6" f eed rnain leading into the building ending in an isolation valve and blind flange. Installation of the new service-Building #6 sprinkler system will includ** Installation of non-seinmic supports, piping, sprinklers. and alarm devices downstream of the existing 1, 'ation valve.
SAFETY EVALUATION SUM!iARY:
The installation of a sprinkler system in Service Building #t> is not required ;
to ptotect safety related equipment or to ensure safe shutdown capability. The system is being installed to comply with the Ohio Basic Building Code and was l recommended by the American Nuclear Insurer.(ANI) to. enhance Davis-Desoe fire prot ect len capabilitles.
- Inst allation of. the Service Budiding #6 sprinkler system will add an additional water discharge point from'the existing station fire water distribution system.
Tle-in to the distribution.syntem was performed during-the initial installation of the undergtound piping therefore the integrity of the distribution piping will not be affected by this modification. Additionally, the water distribution. system contains sectionalizing valves such that failura or ,
operation of the Service Building #6 piping system.can be isolated, if e
- necessary..to maintain operability of suppression systems important to' safety.
Th'ere-1s no equipment important'to safety located in Service Building 16, therefore r there are no adverse effects associated with spra; cr-flooding caused by sprinkler system actuat ion or pipe f ailure. J The new sprinkler system'1n Service Building 16 'is hydraulically designed .to
, ~ provide discharge densities required by NFPA 13 Code. The system demand is well within the capability of the existing fire water distribution system. ,
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SAFETY EVALUATIO!1
SUMMARY
FOR MOD 89-0124, 0011 90-096 and UC!l 91-024 (SE 91-0035) ;
II.l.L.E 8 >
Radiation Effluent Monitors ,
CilAllGE :
i Update of radiation detector sensitivities and ef ficiencies. !
' REAS0!! FOR CHAllGE:
The detector efficiency for RE 18221h Vaste Gas System Discharge Radiation !
!!onitor, is being revised due to installation of a new detettor under ',
HOD 89-0124. In addition._ radiation detector sensitivities / efficiencies listed in USAR Table'11.4-1 are bein6 updated with new data obtained-from the vendor.
SAFETY EVALUATION SUl9tARY:
Radiological Controle has evaluated the new detector efficiencies a.d '
sennitivities obtained from Victoteen. This data is traceable to the flational Institute of Standards and Technology. The new data differs from the original ;
efficiencles and sensitivities listed in USAR Table 11.4 1. Therefore, this ;
evaluation includes the impact on radiation monitor alarm / trip setpoints.
Setpoints were evaluated per off-site Dose Calculation Manual (ODCH)
- methodology..
'The new detector efficiencies will be used to calculate setpoints for effluent streams in accordance with ODCH methodology tot ensure releases will be terminated prior to exceeding off-site limits of 10 CFR 20 should a malfunction of-a;1iquid or gaseous waste system occurs and.enseres alarm / trip setpoints are ,
within the limits of 10 CFR 20 'as required by Technical Specifications 3.3.3.9 and-3.3.3.10.
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SAFFTY EVALUAT1011 SUll!!ARY FOR Mop 00 0n06 (SE 91 0011)
T_I TLlj SrAS Shutdown Eypass Cil AllSE :
A shutdewn bypaso capability will be installed in all four channels of the SFAS. The bypans will be administratively centrolled by . locked SFAS cabinets, hey switthes to "EllAELE" the bypass, and key switches to allow " S ET / P.Er ,T " of the bypass. One annunciato window will be provided to inf orm th- uret at or s of the bypaso condition of any SFAS actuation channe19. _
PEAE0ll FO R __Q_ll Al1GE :
In order t o prevent spurious saf et y-syst em act uat ions in Mode 5, this modif icat ion will add a chutdown bypass capability to the SFAS. A separate bypass foi each piece of SFAS ac t uat ed equipt.._ at will be pcovided.
SAFETY EVALUAT10!1 SUM 11ARYt The surveillance cards and display panels ate procuted 'Q", lE, and seismically qualified. The installation of the cards and thw panels in the cotilnet s will also be seismically quvlified. This will ensure that no new hazards will be int roduced by the ins;.n lation of the equipment.
The reliability of the actuation equipment will remain the same. The latch type trelays were specifically chosen to ensure that the bypass could not be inadvesteutly initlated. The requirement to have the key switch enabled be!ute the latch type telays have power ensures no unwanted telay state change.
The portion of the SFAS being modified for oypass capability is not required to perform any saf ety f unction during 11 ode 5. The Technical Specifications opec tf y these operability crit eria. The SFAS shutdown bypass will teduce the possibility of an inadvertent trip of the SFAS system when it is not required to provide a safety function in Mode 5. Activation of the shutdown bypass tequires access to the SFA9 cabinets and activation of two key switches. These cabinet s ate locked and ' 't h the door keys and the switch keys are under administ rative con' The SFAS data lights which indicate the bypass status and/or its citeuit components perform no safety funetion. .
In order t o ensure compliance with Technical Specification Surveillance Requirement 4.3.2.1.2, opetability of the logic for the SFAS shutdown bypess will be verified at powet during the monthly Channel Functional Ter,t. While per f orming t he periodic tests, the proper operati on of the STAS data lights verifles that no inadvertent Shutdown Bypass is initiated.
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SAFETY EVALUAT10!1
SUMMARY
FOR 110D 90-0014 and it! 89-0034 (SE 89-0264)
TITLE:
Chlorine Detector Jumper CliAllGE :
Chlorine detectors AEL: 3 A and B, located outdoors in the Chlorine Detector Block llouse and AESM - E, located in the Con.rcl Room Air Conditioning i Equipment P,oom 603, wi . , each 1. ave a jumpet installed by Temporary Modificat inn
( T!!) 89-0034.
REAS0tl FOR CHAllGE:
The installat ion of electtical jun pers will prevent the unnecessary isolation of the Conttol Room and subsequent loss of the Control Room Normal Ventilation System due to a spurious chlorine detector trip.
SAFETY EVALUATIOll
SUMMARY
2 The chlorine hazards analyced under USAR Chapter 15, the chlorine storage tank car and the chlorine cylinders used for water treatment, have been removed from site, The gaseous chlorination system has been replaced with a liquid sodium hypochlorite system by TH 87-03S5 and justified by SE 87-0196. Due to the removal of gaseous chlorine " rom site, License Amendment 134 was approved which deleted Technical Specification 3/4.3.3.7, Chlorine Detection Systems. Safety Evaluation 87-0294, which supported this revision, concluded that there were no sources of chlorine near Davis-Besse that would pose a threat to control room habitability requiring automatic isolation of the control room ventilation syntem by the chlorine detection system.
f Installing a jumper across the norma 11y closed chlorine detector contacts will mainta n the control power circuit to the isolation dampers independent of the chlorine detector state. Disabling the detector will not prevent tt> closure of tho isolation dampers due to a high station vent radiation alarm, SFAS Level 1 signal or manual actuation from the local or Control Room Switch, s
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1 SAFETY EVALUATION
SUMMARY
TITLE: .
' Replacement of containment Vacuum Breakers CHANGE This modification replaces the Containment Isolation Valves ~for the containment Vacuum Breaker System _(CV5070-CV5079).
The replacement valves are eccentric disc soft seated butterfly valves.
REASON FOR CHAUGE:
The existing Fisher model 7620_ butterfly valves are obsolete _and require offsite repair. Local' Leak Rate Test-(LLRT) failures'during the sixth
-refueling outage necessitated blanking off-two vacuum breaker containment penetrations.
SAFETY = EVALUATION SUIMARY:
-Review cf the replacement valves concluded that the stresses generated by the cantilevered vacuum breaker penetrations _on the containment vessel do not
- Jncrease; 'The_ existing analysis bounds the use of the replacement valves and is_therefore acceptable.
5An. analysis was conducted of the operation of the Vacuum Breaker System with the replacement valves installed. The analysis concluded that the 0.5 psid
. external pressure differential will not be exceeded for an accidental initiation of-containment-wpray. -Any combination of' existing and replacement valves--is=also bounded. The replacement valves are seismically qualified and
-have been reconcil' edias suitable replacements for.CV5070-CV5079.
The: replacement valves will be used_with'the existing valve operators. There are no cont rol- or instrt. mentation modificatici.s required, 'The new valves meet .
a>more stringent seat leakage criteria than the existing valves and have replaceable ' elastomer seats._ The existing operators. provide sufficient torque ,
to operate the new valves-under all design conditions and the_ closure times
- will_not-be affected-by this modification.
4
- i SAFETY. EVALUATION
SUMMARY
-TITLE:
Backup-Protection for Electrical Penet' ration Assemblies CHANGE:
Plant Hodification 90-0036 adds backup fault protection (fuses) to 21 power
-circuito which are supplied from the: Station's 480V distribution system and
. pass through.an Electrical Penetration Assembly (epa).
y n 9HANGE:
ci< s selected'for backup fault protection if-the secondary protective 4, 480Vcswitchgear breaker, Molded _ Case Circuit Breaker (MCB).or current exceeds the current carrying' capabilities of the electrical s feedthrough assembly.
j .ALUATION
SUMMARY
The weightfof the backup fault protection fuse assembly (i.e. f u r,e s , Raychem insulation and connecting hardware) will not adversely-affect the seismic response of the affected MCCs.
Normal ambient-temperatures in the affected areas'and the insulating effects of the Raychem covering- have been considered in selectir.g the ampere rating of these fuses. The fuse ampere rating ensures that the MCB will operate first to-clear all faults with a magnitude below the magnetic ' instantaneous) trip region of the breaker. -In the instantaneous region there is a possibility that the1 fuse mav operate first, but-this is not a concern,-since the faults in-this cregion are a' result of a significant circuit - f ailures . Thus, . unneces sary- powe r-interruptions -to-safety related equipment will not -occur as a result of these new. fuses, Accident environment has also'been considered.- The fuse assembly will be evaluated to ensure that these assemblies are environmentally qualified for the applications they will see.
As. discussed above, adding the backup fault protection. fuses will not adversely
'ffect the safety functions of EPAs, Class 1E MCCs, or.any safety related' a
Lequipment supplied by these Class 1E HCCs.
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SAFETY EVALUATION StHMARY FOR MOD 90-0059 (SE_91-0026)
TITLE:
Service Water' System Modifications CHANGE:
This 60D provides'for the installation of six isolation valves, six flow I elements, and;several removable p3 ping spool pieces into the Service Water
-System.
REASON FOR CHANGE:
Valves will_be provided to allow isolation of the six inch supply and return headers for the-ECCS Room Coolers'and the supply headers for the Auxiliary Feedwater Pumps. The. installation of these valves will limit the-impact of work on the Service Water piping off of the ten inch heauers and will allow for
-nome maintenance activities to be conducted in 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> accion statements during normal plant operation.
Venturi elements will be installed-in the two ten inch Service Water supply
' headers, three orifice plate flow elements will be installed in the six inch suppiy-headers _to the ECCS Room Coolers and one orifice plate flow element will be installed inlthe three= inch Train I supply to ECCS Room Cooler No. 5. The
~ installation-of these' flow elements will enhance Service Water System balancing and quarterly pump Ltesting.
-Spool pieces will-also be installed in the supply headers-to the ECCS Room LCoolers'and the Auxiliary _Feedwater System and in the ECCS Room Cooler return piping.J The' installation of these spool pieces will increase.the capability to conduct repairs / inspections 1of some Service Water piping and. components during normalLoperating cycles with limited: Impact on' plant operation.
! SAFETY EVALUATION
SUMMARY
-The portion of the Service Water System required'for-. emergency operation was
. designed to ASME Section III Clacs13 and Seismic Category I requirement - The piping system changes required to implement this modification-were also designed _tof.hese_ requirements. This includes reconciliation of:the existing
-piping seismic / stress analyses. This reconciliation concluded that the stress
- levels in the piping systemLdoinot exceed-ASME Section III allowables.
.Therefore, the above listed modifications do not impact the structural capability of the Service Water Oystem to perform its intended function.
The addition of several. flanges in the Service Water System piping will not
~ increase the probability of pipe _ rupture because flange & connections are F -allowed by the original design code and, as stated above, the piping seismic / stress analyses.have been reconciled to include-installation of the flanged spools..
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i 5
1 The permanent addition of isolation valves.and flow elements introduces addition'al pressure losses in the_ Service Water System piping to the engineered safety features-components. However, these addit ional- losses are negligible when compared to existing-system piping.1 fittings and components and therefore
< will have negligible affect on system operation,or flow balance, calculations show that installation of these components will not reduce Service Water flow to engineered safety features components below their design basis requirements.
The isolation valves installed in the Service Water supply headers for the
_ Auxiliary Feedwater System will be maintained and verified open as part of the locked valve program to-insure availability of the backup Auxiliary Feedwater supply.
The isolation valves installed in the ECCS Room Cooler supply and return headers.are not required to be locked because they-do'not meet the' locked valve criteria. Inadvertent closure of one of these valves is not considered a credible malfunction due to normal administrative controls in place to verify valve position and because of their location in portions of the Service Water System which are in service during normal plant operation, 9
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SAFETY EVALUATIO!1 SIMiARY FOR
TITLE:
Pedundant Steam Generator Level Indication CHAllGE :
This MOD will add two indicators to provide redundant steam generator (SG) level indication for the p1. ant operators. These indicators will be powered f rom essential power sources diverse from the present SG level indicators' powe: nources. Precently, the power for SG-1 level indication is derived from channel 1 power sources and SG-2 3evel indication is derived irom channel 2 -
power sources.
REASoll FOR CHANGE:
Reg, Guide 1.97, requires redundant Steam Generator (SG) level indication from an alternate power source. Steam and Feedwatet Rupture Control System (SFRCS) channels 1 and 2 will supply the indication since the power sources, Y1 for channel 1/SG-2 level indication, and Y for channel 2 /SG-1 level indication, are diverse from the existing indicator power sources. The MOD will install the two new Dixson level indicators on the Post Accident tionitoring (PAM) panels.
SAFETY EVALUATION SIM1ARY:
There is no adverse effect on safety. The indications will be derived from qualified isolated outputs from the SFRCS, therefore; there will be no fault derived downstream of the isolators that will affect the SFRCS. There is no modification to the SFRCS upstream of the isolators, The Dixson jndicators will be located in the PAM panel. The indicators are -
qualified and will be mounted seismically. The addition of an indication in the PAM panels has no adverse effect on safety. The conduit connecting the '
SFRCS output to the PAM panel will be installed seismically.
There is no degradation of reliability and no hazards created by the addition of indication of steam genetator level to the PAM panel.
M
SAFETY EVALUATION SUl!!!ARY FOR
!!OD 90-007 9 (91-0046)
TITLE:
Removal or Abandonment of Seismic Restraints en the Reactor Coolant System Piping and Equipment CH AtlGE :
This modification involves the permanent removal or abandonment in place of seismic restraints (snubbers) on the teactor coolant system piping and equipment.
The 20" Grinnell Hydraulic snubbers at the base of each steam generator (3 per generatet) will be disabled by unpinning them from the steam generators and abandoning them in-place.
REASON FOR CHAIFJ3:
Use of ASME Code Case N-411 permitted the deletion of certain snubbets in the RCS system.
SAFETY EVALUATION SUlt!ARY:
In 1985 Toledo Edison requested approval from the Nuclear Regulatory Commission (NRC) to use ASME Code Case N-411, altetnative damping values for responue spectra analysis of Class 1. 2. and 3 piping, and the approval was obtained. The code case allows the use of increased damping values in seismic response spectra analysis with the ultimate result of lower seismic loads on restraints and equipment. Under this Mod, the use of the code case permitted the permanent deletion of the 19 snubbers.
The use of Leak-Before-Break (LBB), exclusion of dynamic effects from postulated pipe rupture, was also pursued. Per 10CFR50, App. A, General Design Criteria 4, the exclusion of the dynamic effects associated with poatulated pipe ruptur events is permitted if the utility gains approval from the NRC of the analyses th at shows that the probability of a piping rupture is very low.
That analyses was perf ormed by BW and approved by the NRC f or the BW Ownet 's Group. As part of the approval utilities were requested, and Toledo Edison has submit ted, information that indicates that the Davis-Besse Nuclear Power Station, Unit 1 RCS Leak Detection Systems meet the inter,t of Reg. Guide 1.45, RCPB leakage detection systems. The underlying basis for the LBP approach is that detectable RCS leakage will develop via a pipe crack and that once detected, corrective action can be taken prior to the postulated cat atrophic tupture. Based on fracture mechanics theories and analyses, sutticient time f or corrective action exists between the c rack formation ar.d the point at which rupture will occur.
During 1990, the Class 1 stress analyses f or t he reactst cenlant system was revised by BW utillaing the two alternate criterion and LBB.
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SAFETY EVALUATION SUltiARY FOR MOD 91-0016 (SE ..-0061)
TITLE:
Repair of Fuel Arisemblien CilANGE :
Repair fuel assemblies based on results of ultrasonic testing.
RF;A,50H FOR CHANGE:
During the seventh tefueling outage (7 RFO) Davis-Besse will perform ultrasonic -
testing (UT) of the fuel assemblies (FAs) that will be teused in cycle 8. The goal la to identify fuel rods with defective cladding. Once identified, a decision will be made regarding repair of those FAs. Fuel repairs (replacement of defective fuel rods with dummy rods) may be performed to reduce the expected iodine activity during the next operating cycle.
SAFETY EVALUATION SUll1ARY:
Replacement of fuel tods with dummy rods in both processes does not affect the assembly's structural ability to withstand normal handling or its performance during a seismic event. The FA's structure is determined by its upper and lower end fittings and the guide tubes that separate them. Fuel rods and dummy rods are contained by eight spacer grias located along the length of the guide tubes. The dummy rods have haen designed so that they will be firmly captured over the remaining design life of the fuel assembly.
The stainless steel ;35 304) material used for the dummy rods is a standard FA material, suitable for the reactor or Spent Fuel Pool (SFP) environment. The thermal expansion of the SS pin in the radial direction will be about three times more than a circaloy-4 clad fuel pin. That expansion will compress the spring stops on the spacer grids about .002 inches more than the fuel rode.
That compression is within the elastic range of the spring stops, and will not cause any set of the spring stops.
By design, there is no adverse mechanical effect translatable to grid springs that hold fast the surrounding fuel pins. Therefore, there is no inc re a s e in the potential for fuel defects due to grid fretting.
The axial effect of differential temperature expansion between the SS pin and the zircaloy guide tubes was originally analyzed by BWFC over a tempurature range of 70 to 6000F, An engineering evaluation for Davis-Besse, with a fuel design temperature limit of 6500F, indicates a net growth differential of only 0.33 mils (3.3 E-4 in.) from that previously analyzed. Results of this evaluation showed that clearance is maintained between the dummy steel rod and the upper end fitting (UEF) under reactor operating temperature and irradiation conditions. There are no interferences anticipated over the remaining length of the assembly's life time, and there will be no rod bowing potential.
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i The radiologica'l consequences offdropping-or breaking a fuel rod are bounded by the~ Fuel Handling Accident Outside Containment (FHA-0C). The FHA-0C analysis-shows that_th_e environmental consequences satisfy the acceptance criterion of being less than the. guidelines given in-10 CFR-100.
Reactivity control during repairs has been evaluated. . Technical Specification of < 0.95.
3/4.9.13.requiresthatstorageoffuelbemaintainedwithaK(kiscriterion Engineering Evaluation RE-91-01, dated 06-18-91, assures that' will:be maintained.
Concerns 1over possible cladding scratches, resulting f rom removal and insertion of fuel rods were. initially addressed by. testing and were later resolved with field experience. Scratches are potential cladding stress risers, tiene of the
. tested rods incurred scratches deeper than 0 002 inches , which is the acceptance criterion for a new FA, Insertion force _will also be monitored and limited administratively. Field experience verified that irradiated rods are-more scratch resistant than new rods, since the rods' cladding will have been irradiation-hardened and also will have crept down.
Iri conclusion, there is, therefore, no effect on safety, since this fuel repair is controlled with proven procedures, . performed by BWFC personnel with fuel handling experience, uses tested fuel' handling tools and quality components, and uses repair. techniques and hardware justified.
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l SAFETY EVALUATIO!! SUl! MARY FOR MOD 91-0019 (SE 91-0029)
TITLE:
Removal of a Snubber from the Main Feedwater System CHAtlOE :
i Hodification 91-0019 removes support (snubber) EBD-12-SR43 of the Main Feedwater system. This Grinnell hydraulic snubber is embedded in a foam filled wall penetration between Rooms 304 ac.d 310 of the Auxiliary Building at Elevation 599 feet.
REASON FOR CHANGE:
This instal.lation requires additional manhours to reseal and inspect the penetration each time the snubber is removed for maintenance. Due to this and general accessibility problems, Hechanical Maintenance requested it be deleted.
SAFETY EVALUATION
SUMMARY
Pipe Stress Analyeis Problem 42B has been reanalyzed with Support EBD-12-SR43 deleted and the piping meets code allowable stress values. All the remaining supports have been reviewed and are acceptable for the new support loadings.
After the removal of the snubber, Penetration 304-E2-078/310-W1-089 will be resealed par detail LDF-1 to maintain the required fire rating for CMU wall 3457.
Based on the above, the modification to be performed will have no adverse -
effects on safety.
SAFETY EVALUATION SUlttARY FOR.
pCAQR 90-0463 and UCN 92-038 (SE 90-0104)
TITLE: b Eva]uation of Containment Air Coolers Operating with Less Than Design Service Water Flow
' CHANGE:
This evaluation provides technical justification for the ability of the Containment Air Coolers (CACs) to perform their safoty related functions, given
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service water (SW) flow to the coolers is-'less~than the nominal design value.
The.overall ability of the service water system to perform its other safety functions 1s-also evaluated.
REASON FOR CHANGES.
As a result of Service Water System (SV) fouling the flow of SW to the CACs is reduced. System testing has shown that present service water flow to the CACs .
during design basis operating ~ conditions would be a minimum of 1150 gpm with service' water.to=the component ~ cooling water heat exchanger throttled to approximately 8000 gpm. Consequsntly. the design heat removal rate at a containment saturation temperature:of 2640F (with 850F service water) is
. reduced to 70,7.million btulhr and therefore.the-ability of_the CACs to perform their safety functions has to be assessed.
SAFETY EVALUATION
SUMMARY
Containment' analyses were performed incorporating both revised CAC performance curves and:theilower: containment spray temperature.
Results show both peak containment _ temperatures ~and pressures _are lower than the USAR bounding
. analysis values. This result is expected, due.to containment sprays having a significant impact on containment response earlier in the transient than CACs.
with.the greater:heatLremoval'of' lower temperature' spray water compensating for effects of reduced CAC_ performance before the coolers have their greatest impact.onothe transient.
The1 analytical results have been reviewed with the determination that the USAR cont'ainment response curves remain bounding for equipment qualification
- p rof ile s '.
Therefore.-the containment responsi to a LOCA at full power with the previously
In addition to considering the effects of' reduced service water flow to the
- Caco,-the possibility that'other service water safety functions are also
- degraded is. addressed below for completeness.
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Service water l flow to the 1) CCW heat-exchangers has been tested and has been
. verified =as meeting'the design basis flow requirement; 2)' ECCS. room coolers fl,
_12, and #4 is sufficient. such that each can maintain its room temperature within allowable envitenmental limits. 'Only one ECCS toom cooler is required to be functional In each ECCS_ pump room to declare the corresponding ECCS pumps
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operable; 3). control room EVS condenser units has'been tested and verified as
= meeting at least'the design basis: flow requirements: 4) control roon EVS condenser units have been tested and verified as meeting at least the design basis' flow requirements: 5) hydrogen dilution system blowers have been verified-
- during surveillance testing. This testing is considered-adequate to demonstrate acceptable service water _ flows. Olote that service water to these
- blowers is not required until at least 20 days following a LOCA at which time the other service water loads would be relatively low). >
Service' water also provides_the seismically qualliied backup water supply to the AFW system. The only design basis event where service water-supplied AFV has been credited jn the SER is_a design basis seismic event. Although not discussed in the-SER,-there is a design basis accident where non-seismic category I equipmert (which would include the condensate storage. tanks) cannot be taken credit fort that event-is a LOCA. The only LOCA which currently requires AFW-flow is a ' category 2" small break (.003 ft2 - .02 ftt), for which AFW provides secondary water levels to_ maintain a condensing surface for boiler-condenser heat-transfer. -While heat transfer .in this mode is cyclical in nature, an' estimation of AFW flows needed to maintain steam generator level at the 10 ft. SFAS level-2 setpoint can be made. -Additionally, any flow reduction to the CACs during periods of-service water-supplied AFW is considered acceptable, as this flow is providing direct removal'of reactor decay-heatTenergy which is then not available for-release to containment.
Therefore, service water supply to the AFW system is considered adequate.
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!L Given the Labove, itL is concluded that -the USAR containment response to a LOCA L remains bounding even with less-than-design service water flows'to the CACs, and that the service water _ system remains capable of performing its safety functions.
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. SAFETY EVALUATION
SUMMARY
FOR PCAQR 92-0086 (SE 92-0012)
- TITLES.
- 1solating Two Turbine: Bypass Valves During Normal Plant Operation CHANGE: ,
This. evaluation provides_ technical justification for routine power operations
- with two-turbine bypass valves (TBVs) isolated, j REASON FOR CHANGE:
Recent problems have been experienced with the operation of two TBVs (one per steam generator), resulting in higher than desired post-trip cooldown rates.
Normal plant-operation with.all TBVs (three" per steam generator) is preferred, as TBV capacity minimizes steam releases directly to the atmosphere. However, considering the potential negative effects of sp rious TBV operation, particularly post-trip, it may be-prudent-to conduct toutine power operatione
. with'two TBVs. isolated-. The purpose of this safety evaluation is_to evaluate the-safety _ significance of isolating one TBV per steam generator.
SAFETY EVALUATION SUF&tARY:
Normally, six valves-are available,'having a combined relief capacity of 251
= ratedi steam flow. 'Each individual valve has-a relief capacity of 5% rated
- ster.m itow. - During power operations the_TBVs remain closed, unless utilized to handle electric load changes. For this purpose, the USAR states the plant '
control system-is designed;to accept a 102 step load rejection without steam
. generator _ code safety valve or TBV action. With TBV action, plant design allows for a 40% load reduction or a turbine trip-from 40: loadivithout steam generator' code safety' valve actuation.
' Total TE" capacity _was_also a consideration when determining an appropriate-arming. threshol'd. for : the- Anticipatory Reactor Trip _ System (ARTS) for turbine trips._ yCompared to the generic B&W analysis ori which the current 'value of 45!
is-based.: Davis-Besse has approximately 102 of full power op_erational margin
't . due;to the-generic analysis.not crediting all of tne total available plant
. specific _ bypass capacity -(including the atmospheric vent- valves and first bank of steam generator: code safeties) . With four of six TBVs available, bypass capacity-is1 reduced by:approximately 82 of rated steam flow -which is within-Lthe-102 operational msrgin. -Therefore, it11s estimated that the current: arming threshold of 452.is still acceptable.
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Use of the TSVs is also discussed in USAR Section 15.4.2, Steam Generator Tube Rupture. The TBVs are described as being used to aid in reactor coolant system couldown efforts by directing steam to the condenser rather than directly to the atmosphere. Prior to isolation of the affected steam generator, it is estimated that four TBVs vould provide sufficient cooling capacity to remain within the USAR analysis. During the longer-term cooldown of the RCS to enable actuation of the decay heat removal system, the USAR analysis assumed a 100 degree per hour cocidown rate is initiated and maintained. With one TBV isolated per steam generator. If additional cooldown capacity is desired during any actnal plant cooldown the isolated TBV on the unaffected generator should be unisolated and utilized.
Additionally, it should be noted that the radiological consequence evaluation associated with the tube rupture accident assumed that all fission products leaking irom the RCS went directly to the atmosphere (not taking credit for iission product " scrubbing" of steam directed to the condenser). Therefore, a reduction in total TEV bypass flow capability will not affect radiological consequenc es calculated for this accident. and remaining steam relief capability exists to ensure a plant cooldown.
With two TBVs isolated, a total of four will remain functional during plant operation. Since the TBVs have no safety function, there is no adverse affect on piant safety associated with the valve isolations. The main effect on plant operation will be to limit the ability of the plant to handle load rejections during power operations and to in; rease the post-trip fraction of steam flow which is handled by the st eam generator code saf eties and atmosphe ric vent valves. If additional plant couldown capacity is desired, the TBVs can be unisolated and utilized.
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SAFETY EVALUATION SU!?1ARY FOR SCC 90-3021 (SE 90-0146)
TITLE:
Removal of Intake Canal Water Temperature Measuring Instruments CHAllGE :
)
Removal of temperature indicator TI-843 and associated temperature sensor and transmitter.
REAFO!! FOR CHANGE:
I' The censor and transmitter are located in the intake canal entry from the lake.
The transmitter requires excessive maintenance, and is a maintenance burden 4 because of its location. It is not required for operation. Alternative C
sensor, TE-738 and computer point T-413 are available to measure the Ultimate Heat Sink temperature which is needed to satisfy the Technical Specification requirement, 3/4.7.5.
SAFETY EVALUATION SLM1ARY:
The affected components are not safety related, nor do they interface with any safety related structure, system and component. Therefore the removal of these components will not have any effect on safety.
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SAFETY- EVALU5 TION - SUHFLARY - j FOR i SCC 91-3000 (SE 91-0023)
TITLEt Connect Backup Meteorological System Transformer CllANGE : I l
A backup meteorological system was installed and is currently powered from a-tcmporary connection at the microwave tower. This Simple Configuration Change (SCC) connects the backup meteorological system power supply transformer to the existin3_ nearby 480 volt feeder from the Route 2 residential service line.
REASON FOR' CHANGE:
When maintenance-is being performed on BYS, power is lost to the meteorological system. Following *he change it will be possible to remove l5 f rom service.
without disabling the entire meteorological system.
3AFETY EVALUATION SUNIARY:
'All the changes are to equipment outside the plant at the microwave tower and this_ equipment serves-no safety related functions.
As currently' designed the meteorological monitoring system is vulnerable to a
. single' failure ;; BYS. Even scheduled preventative maintenance at BY5' removes
-power from thele. ire-meteorological monitoring system. In the new design. .BYS can be Lremoved from service without disabling the backup meteorological
. manitoring system. 1The configuration-of other loads will be unaffected. The (new design also reduces theiltad on BYS. The noxmal-load on X3006 is.
increased:-however, the notmal load is significantly less than the load connectedLwh' lternate power is required. During. implementation,_ temporary connections 1.be used:to limit the length of power outages to less-than-the capability A the installed battery backup. .
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SAFETY EVALUATIO!1 SUlt!ARY FOR SE 91-0024 TITLE:
Performance of DB-!!E-03213, 11oderat or Tempe rature Coefficient Measurement by Boron Swap CH Af1GE :
The purpose of this safety evaluation is to justify a ditferent, wider power range til f rom Heat Balance Power deviation for use during the performance of DB-11E-03213, Moderator Temperature Coefficient Measurennent by Bo on Swap.
REASON FOR CilAl1GE:
The moderator coeffJcient of reactivity is measured by plant test procedure D D - 11E - 03 213 . This test determines the moderator coefficient by measuring the reactivity change produced by a Reactor Coolant System (RCS) average temperature (T change. The power range nuclear instrumentation (111) could deviate from fl E ) Balance power by more thar. the normally allowed amount during the T , change due to reactor vessel downcomer temperature effects. If this occurs,~the test would normally have to be interrupted to calibrate the powet range Ills .
SAFETY EVALUATION
SUMMARY
The power range NI calibration curves specified in DB-PF-06703 require that a power range NI channel be declared inoperable if it is indicating below the heat balance power. This is unnecessarily restrictive, as discussed belou for the associated trip functions.
liigh Flux Trip The design overpower (112.0 percent RTP) is equal to the maximum core power level that could be attained should a high flux trip occur.
Since the design overpower includes a 2.0 percent RTP allowance for the steady-state difference between the NT power and the heat balance power, the NI power could indicate two percent below Heat Balance power without violating design assumptions.
The power range NI calibration curves shown in DB-PF-06703 specify that recalibration is directed by the shift supervisor when the power range NIs are indicating above the heat balance power. This will remain unchanged.
Flux /AFlux/ Flow Trip Function:
The design analysis for the flux / flow portion of the flux /6 flux / flow trip function is baseo on a one Reactor Coolant Pump (RCP) coastdown with a maximum power of 108 percent RTP. The design power of 108 percent RTP includes a 2 percent RTP allowance for steady-state fl1 calisration. Therefore, the
P proposed change to the power range N1 calibration procedure is bounded by l existing analyses and will not affect t_he flux /Aflux/ flow trip function.
Anticipatory Reactor Trip System (ARTS)t ARTS arminn is based:en the power level indicated by the power _ range NIs and is required above 45 percent RTP per plant Technical Specifications. As stated
!- previously,'the. proposed changes are only applicable during performance of DB-NE-03213 when the reactor power in greatet than 70 percent RTP. ARTS will i
not_be affected by the proposed power range NI calibration changes because ARTS will remain in an armed condition if the power range NIs are indicating i 2 percent RTP below the heat balance power. If deviation fro.7. the moderator l l coefficient test procedure is required, the normal power range NI calibration
! curves will apply.
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I I SAFETY EVALUATIO!! SUl!!iARY FOR SE 91-0060, R02 TITLE:
Graduated Filter Replacement Program CilA14GE t Filter replacement to support the Source Term Reduction Program.
REASotl FOR CHANGE:
As part of the Source Term Reduction Program (STRP), Davis-Besse is in the -
ptoress of implementing a " Graduated Filter Replacement Program" (GFRP) for finer filtration in the !!akeup and Purificat ion System, the Spent Fuel Pool cooling System and the Clean Liquid Radwaste System. The GFRP targeted final removal ef ficiency of 0.2 micton absolute would be achieved in a three step process. Specifications currently identifies the filtration ratings as eit her
1 mic ron" or "10 or 3 mic ron
- absolute . Thu T!! is being written to allow the use of a range of filters as recommended in the GFRP. The filter material for the finer microu sizes is the same (epoxy coated glass fibers) as presently used.
SAFETY EVALUATION SUl! MARY:
The lunctions of the Makeup and Purification System, the clean Liquid Radwaste System and the Spent Fuel Pool Cooling System will not t> e adversely affected by the change in filter ratings. The " clean" delta pressure specified for the new finer micron rated filters is the same as for those currently in use.
Additionally, since the high differential pressure setpoints will not be changed, the " dirty" delta pressure is unchanged. Therefore, the system flow -
rates and balances will not be affected by the proposed changes.
More frequent change-out of the affected filters will not significantly increase personnel exposure because filter change-outs are controlled activities performed applying good ALARA practices. The use of finer filters witl provide for an overall exposure reduction by lowering the amount of radioactive particulates which collect in crua traps throughout the primary system piping.
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' S A F': / EVALUATION
SUMMARY
FOR SE 91-0067 TITLE:
Temporary _ Changes to Decay' Heat Removal' System Annunciator Alarms During-Plant ,
Shutdown-CHANGE:
The proposed activity will consist of the. temporary readjustment of selected decay heat _ removal: system' annunciator alarm setpoints and the modification of an annunciator alarm input logic in order to provide more useful and appropriate alarms during shutdown conditions.
More specifically, the praposed activity will readjust the setpoints of the
' 1. Decay Heat-(DH) high flow annunciator
-2. -Dh-low flow annunciator-
- 3. DH cooler outlet high temperature annunciator. ,
. REASON FOR CHANGE '
The current temperature and low flow setpoints are specifically listed in the USAR for normal Decay Heat Removal and Emergency Core Cooling Systems (ECCS) operations, During Modes 5 or 6 and core offloads, plant conditions can vary Leignificantly from normal operation thus potentially reducing the effectiveness or usefulness of-various alarms. Therefore. . the- temporary adjustment of the
- ref erenced alarms will provide more appropriate alarms during the shutdown plant conditions.
SAFETY EVALUATION
SUMMARY
DH Low-Flow-Annunciator During the 7RF0 and subsequent outages, the installation and removal of steam
= generator nozzle dams or other possible _ maintenance activities will require that DH.flowrates be. reduced below the existing alarm setpoint.(2800 GFM). The reduced flowrates:are required to, address _UPSH and vortexing concerns. Du rin'g periods of-' reduced DHR flowrates such as required:daring the performance of-maintenance'at low RCS levels of 18* and 14"'above the.RCS hot leg centerline, the existing setpoint of~2800 gpm will cause the alarm to be " locked in" in the a alurm state thus. causing ~a nuisance alarm and eliminatin6 its_usefulness.
Therefore,'in order to provide an effective alarm if DHR flow is lost during
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reduced: inventory conditions.'the operating train low flow alarm will be set at~
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less than the established operating flowrate.
DH' lligh Flow - Annunciat' or During reduced inventory _ conditions the DH high flow annunciator setpoint fot ,
the operating train :wil1~ be readjusted (decreased) to below curve " Predicted Total 1DH FlowLvs.' Reactor Water. Level". This will thus allow the alarm to
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provide an- indic_ation of a. condition which could result in the loss of adequate
- DH' pump ;!PS!!.
_ Dil Pump Breaker Status Input to the DH Low Flow Annunciator The DH low-flow annunciator alarm is interlocked with the corresponding DH pump switchgear to prevent initiation.of the low flow alarm when the pump is not in operation. However, during 'anutdown. operations if the DH pump is lost due to the breaker opening, the low flow alarm wou1J thus not be received. This change.will ensure a lov flow alarm will be received in the event a tunning DH
~ pump is lost. The change will be implemented on both DH pumps during shutdown and restored prior to Mode 4. This viii cruse the non-operating putnp annunciator to be in . constant alarm during shutdown, DH Cooler outlet Temperature _
A high temperature. fro'a tha decay' heat removal cooler is alarmed to signal a loss of cooling capability in the respectiva cooler. The DH cooler outlet temperature annunciator setpoint will be reduced from the existing setpoint of 2000F to.1400F,;following RCS.cooldown to less than 1400F. At the current 7 setpoint of 2000F the alarm would not realist.fcally be reached after the plant is cooled down and depressurized. The reduced alarm setpoint will improve ;he ability- to provide an indication of a loss of cooling water to the DH coolert, and will provide an indication of a slow or abnormal heatup.
The readjustment of alarms is temporary and is used during plant conditions where normal DH flow and t emperatures should be readjusted to more accurately
- reflect plant conditions. Readiusting the flow and temperature setpoints during Modes 5 and 6, is.more conservative than using the current setpoints.
All of the referenced changes will be restored to the normal operating values prior to plant startup and entering Mode 4, Y
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l SAFETY EVALUATION SUrMARY FOR Cycle 8 Refueling (SE 91-0075)
TITLE:
Refuel Renctor CHANGE.
Cycle 8 reload REASON FOR CHANGE:
Cycle 8 reload includes the implementation of the Cycle 8 Core Operating Limits -
Report (COLR) which is referenced in the Davis-Besse Technical specifications.
The nyeie 8 COLR, and the Reload Report on which it is based, were developed with NRC approved methodology by the B&W Fuel Company (BWFC). The cycle 8 core loading, as described in the Reload Report, includes a repaired batch 9 assembly and 64 batch 10 fuel assemblies that are of a Mark B8A design, which contain uranium dioxide enriched to 3.69 w/o U-235.
This batch 10 reload also incorpo ates minor fuel rod design changes, such as modified upper and lower end caps and a teduced pre-pressurization.
SAFETY EVALUATION SU1 MARY The reference fuel cycle for cycle a is cycle 7, and nu.' ear and -
thermal-hydraulics analysis were based on duration of cycle 7 to 400 Effective Full Power Days (EFPDs). Cycle 8 was analyzed to 479 EFPDs. This included a verifi~ation that the cycle 7 Reactor Protection System (RPS) limits, operating limits and setpoints were valid for cycle 8 to 479 EFPDs also, There have been no facility modifications that have affected the cycle 7 RPS setpoints, so it f is concluded that maintaining the same setpoints for cycle 8 would have no -
effect on safety.
The control rod group designations for cycle 8 also remain unchanged from that of cycle 7's, as reflected in the design analyses by BWFC, Furthermore, there have been no operating anomalies during cycle 7 which would affect safety or fuel performance during cycle 8.
There was an indicatior ,f some fuel defects in the last quarter of 1990.
Consequently, ultrasonic Testing (UT) of fuel assemblies (FAs) to be returned for cycle 8 operations was parformed. Results indicated that a total of 5 rods in 5 FAs had either f ailed or were questionable as to integrity.
The ensuing repair or fuel-defect-recovery campaign caused a modliication to fuel assembly and replacement of a defective Batch 9 assembly and 3 Bate'.. 8B assemblies and their in-core symmetric partners with batch BA substitutes.
BWFC analyzed the effects of the repaired FA and the new core loading on the Reload Report and mene'tvering analyses. The analy sis addressed the ef fect of the repair on performance parameters such as reactivity, pewer peaking, margin-to-nucleate bciling end mechanical design.
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Cycle 8 design verification indicated that there were insignificant effects and that previous analyses wete bounding and conscrvative. From a Thermal-Hydraulles standp int, the teconstituted FA, NJ0542, has over 100% DNBR margin during normal operation. Dutine a transient, if DNB based Limiting Safety System Settings are approached, the reconstituted FA would still have over 402 DNBR margin rela +1ve to the core's limiting FA, Fuel mechanical design changes were also reviewed. The minor design changes for batch 10 (Mark BBA) included tha redesign of the uppet end cap to a more grippable configuration. The boti a end cap also was changed to a bullet-nose shape. Neither of these changes nave a significant net effect on an assembly's form loss coefficient and, therefore, there is no thermal-hydraulic effect or impact on safety.
The fuel rod pre-pressurization for batch 10 has also been further reduced. -
The design change was verified by BWFC to have no effect on creep rates and resulting p*'.let-to-cladding strains. The reduced initial rod pressure will allow a highet burnup capability since it will allow the fuel rod to remain less than system pressure at higher burnups, as presently required by the Standard Review Plan, flUREG-0800. Verified as also meeting that criterion, the maximum fuel rod burnup at End-of-Cycle (EOC) 8 for batch 8 fuel is predir >d to be 46,840 MWD /HTU.
In summary, the fuel repair / replacement campaign during 7 RF0 resulted in a modified core nattern, i.e., one that was different than originally analyzed.
The changes were (a) the replacement of eleven fuel assemblies, (b) the use of eight fewer Burnable Poiscn Rod Assemblies, and (c) the insortion of a stainless steel filler rod in FA NJ0542. The total effects of these changes on the analyses performed for the cycle 8 Reload report and Core Operating Limits Report wete evaluated as to impacts on mechanical, nuclear and thermal design, as well as transient, reactiv4ty control and radiological analyses. In all cases, the modified cycle 8 design meets the requirements and acceptance criteria specified in 10 CFR 50 Appendix A. 10 CFR 50.46 and the NRC Standard -
Review Plan, NUREG 0800, so that safe operation of cycle 8 is assured. -
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l SAFETY EVALUATION ~SUINARY FOR SE 92-0003 TITLE -
. Increased Radioactivity Levels in the CCW-System CHANGEi Increase. levels of radioactivity in the CCW system.
REASON FOR CHANGE: .
The' purr se of this' safety evaluation is to assess the potential onsite-radiological' consequences, chemical consequences and offsite dose consequences from operating the Component Cooling Water -(CCW) system as a radioactive system Edue to in leakage _ of reactor coolant.
SAFETY EVALUATION
SUMMARY
Toledo Edison'has evaluated the radiological consequences-for operating the CCW
- system with low;1evels of radioactivity. All components and piping were
>physicelly walked down in the Auxiliary Building to identify any leaks. All leaking components areLtagged as potentially contaminated and the water directed to floor drains. These drains are routed to the Miscellaneous Waste ,
Drain Tank-(MVDT),.a receiving = tank for liquid radioactive waste. The only ,
components'which were outside of a radiologically controlled. area and had the-1 Lpotential to drain to_the storm sewer are the emergency diesel generators' water; cool.ing jacket. The drains'in the diesel room have been-labeled to preclude-personnel from draining any components.containing potentially
-radioactive material to'the storm sewer system. In addition, a tive lip beeniinstalled.around the emergency diesel generator room drm_a= to prevent
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_ minor =CCW: leaks fromLenter.ing the storm sewer system.
The . radioactivity levels-pf the CCW have not = increased external exposure-radiation levels-from_any-components. ' Surveys are-conducted:as part of the Radiological Control ~(RC)-surveillance program.
. nBoric acid leakingfinto the CCW will-reduce the pH below the pH control band.
With~the addition of ammonium hydroxide, the pH can adequately be controlled.
"The other-parameter.affected-by_ boric' intrusion into_the CCW is dissolved solids. ;With the present' leak' rate, the 25 ppm limit for dissolved solids is exceeded. Borate is not a harmful corrodent such as chloride or- sulfate.
~Therefore; when the111mit of 25 ppm dissolved-solids-is exceeded as a result'of boric. acid,_lt is not severely detrimental to the_CCW piping and components.
- With thefpresent leakage. the cation conductivity is approximately13 pS/cm
- Cation conductivity asfhigh as'2% pS/cm is permitted.
i Boric' acid 1n the CCWr is - corrosive to piping and ' components, particularly 7
carbon steel' surfaces. Stainless steel corrosion coupons in the1CCW have indicated corrosion rates less than 0.05 mpy less than 0.2 mpy is' optimum.
i Based on maintaining cation conduct ivity less than 25 pS/cm (even wit h dissolved solids exceeding 25 ppm), and the pH between 8.5-10.0, the present botic acid in the CCW is acceptable. The bcric acid ccncentration in the CCW should be minimized during cycle-8 using neriodic feed and tieed, or demineralization, and the intrusion stopped prior to cycle-9.
Although t here is no indication of any CCW 1eakage into service water presently, periodic samplins and analynis is performed. In addition, a radiation monitor located on the service water discharge header, would detect any gross radioactive material in the service water.
Two pathways were analyzed for the worst-case accidental release of CCW; Training Center pond via storm sewer system from the emergency diese generator floor drains and a release inside the Radiologically Restricted Area (RRA) to floor drains resulting in a Miscellaneous Waste Monitor Tank (MWMT) release.
The calculated doses for t ? worst case accidental release of CCW are below the 10 CFR 100 design criteria limits. In addit lon, the activity concentration of the Training Center Pond would be less than the 10 CFR 70 Appendix B, Maximum Permissible Concentration Limit a The release of entrained noble gases from CCW leakage would contribute an insignificant traction of the Technical Specificction limits for offsite doses.
Radiological surveys will be periodically performed in the CCW system areas to monitor for any detectable airborne activity.
There have been few radiological control problems associated with operating with the existing CCW cource term, No offsite releases to the storm sewer drains have been detected. No further evaluation is warranted until the CCW activity term increases by a factor of 2 in fission product gross activity as determined by Chemistry radioanalysis. There arf to adverse chemistry concerns which aren't addressed by current chemistry practices. Radiological Control, Chemistry and Systems Engineering are monitoring the performance of tne CCW and SW systems.
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SAFETY EVALUATION =SU1 NARY FOR UCH 89-012-(SE 89-0086) -
-TITLEr Revision of USAR Description for Shield Building Annulus Heating System CHANGE:
Revision of: USAR to reflect "e s-bui~ t" ' condition of the plant.
REASON FOR CHANGE:
.The following~ discrepancies-were noted.between.the. actual Shield Building Annulus Heating System and that indicated in paragraph 3D.3.1, page 3D-47, of the Updated Safety Analysis Report.
Installed USAR Number of Heaters 12 8
- Capacity of each.
Heater:(Btu /hr)- 153,585 -245,400 Engineering-confirmed'the' adequacy of the actual installed heaters. The calculation concludes that the annulus can be maintained at.82.40F with design minimum exterior temperature of -100F.
!= In'the original design' process, .the heaters were changed from larger units heated by-steam to. smaller electric units. The steam-units were never ordered-and never installed but were included in the USAR.
~
SAFETY EVALUATION
SUMMARY
LThe lowest acceptable service temperature of_the containment is 300F. - With
~
colder' temperatures, the containment could become brittle and its capabilityLto withstand stress would be degraded. . calculations conclude that the'12
. installed 1 electric heaters are able to ma'intain the, Annulus volume at a
-temperature ofE82.40F which provides a;52.40F margin. The calculation L conservatively does.not.take credit.for the heat generated within the containment vessel. >
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SAFETY EVALUATION SUltMARY i FOR !
UCN 89-129 (SE 92-0021)
TITLES-Distribution and Collection of Personnel Dosimeters (TLDs)
~
'-CHANGE:
Eliminate discussion of distribution-and collection on TLDs from USAR I Section 13.7.2.1. Industrial Security: Access to Station.
REASON FOR CHANGE:
USAR Section 12.3.3 details the use of TLDu.
SAFETY EVALUATION SUl@!ARY:
- The n.onitoring of. individuals for compliance with -10 CFR 20 limits is required
- in Section 12.3.1, TLDs as a preferential means of exposure monitoring are detailed in USAR-Section 12,3.3. The implementation of this exposure monitoring requirement and means are detailed.in various procedures. These procedures discuss distribution, retrieval, and storage of TLDs.
-Describing the.-location for distribution-and collection of TLDs in the USAR is unnecessarily restrictive as the. intent of the USAR is to ensure personnel are-monitored when required by regulation. The requirements for monitoring as described in Section-12 are sufficient to meet the intent to monitor personnel.
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i SAFETY EVALUATIO!1 SUltliARY FOR UCf1 90-042 (SE 92-0023)
TITLE:
Integrated Control System (ICS) USAR Description CHAi!G E :
USAE change 90-042 updates USAR Section 7.7 to reflect that the Lead Frequency Control (LFC) and Autor.atic Dispatch System (ADS) are not used at Davis-Besse REASOft FOR CHA!!GE:
These features were part of the original I egrated Control system (ICS) design as used in fossil plants, however, they were never utilized at Davis-Besse.
The LFC circuit was disabled undet FCR 81-0116 and later removed under FCR 86-0372.
SAFETY EVALUATION SUMilARY:
The Integrated Control System (ICS) provides core reactivity control (by control rod insertion and withdrawal), and controlled core heat removal at power and decay heat removal after a reactor trip (by control of the 9tmospheric vent valves, turbine bypass valves, main and startup feedwater vm'ves, main feed pumps and main turbine). These functions are important to safe plant operation but are not Safety Functions required for safe shutdown or to limit site boundary doses to the limits of 10 CFR 100 during accident conditions.
The function of the ICS 19 not affected by UCN 90-042 as the LFC and ADS -
features have never been used at Davis-Besse. UCN 90 '42 reflects hardware -
changen that were previously made under FCR's 81-0116 and 86-0372 that disabled and removed circuitry that was never used. Since the function of ICS remains unchanged, UCN 90-042 has no effect on safety or ICS functions important to safe plant operation.
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SAFETY EVALUAT1011 SUM!iARY FOR i Pull 90-093 (SE C1-0025) i i
_T I_T L E : ,
Incorporation of the Fire Harards Analysic Report into the USAR ]
Cll AllGE :
This saf ety evaluation is f or USAR Change i.otice (UCll)90-093 which i incorporates the Fire Hazards Analysis Report (FHAR). containing the relocated ;
Fite protection Technical Specifications, into the USAR Ly reference.
This change alsoi makes several fire protection program clasifications such as ;
title changens corrects compensatory actions for water curtain deluge systems: '
tevisen electric and diesel fire pump ~ total head values clarifies the number of and use of fire detect!on local alarm fire panels: revised description of fire detection used in the cable spreading roornt and Appendix 'R" exemptione 1 that were -pt eviously approved by- the 11RC, REASoll- FOR CHA14GE: ;
in accordance with fluclear Regulatory Congnission Generic Letter 'GL) 86-10 i
- 1mplementatitsn of Fire protectiot. Fequirements", dated April 24, 106, the ;
Fite P.totection Technical Specifice*. ions are being relee.ted itito uw FHAR.
The GL also indicates that the FHAR should be incorporated into the USAR.
This change also updates the description of the Fire Protection Program described in the USAR.
SAFETY EVALUni10N SUl%rtY:
incorporation _of the_ filar by ref et ence into the 'USAR will ensure that-future changes to the fire hazards analysis receive appropriate safety evaluations and j
t- Je-is no adverse effect on safety.
The addit.un of the liRC-approved exeroptions descriptien is an administrativa ,
action since _ these exemptf ors have previously been reviewed by the 11RC and.
therefore, there is no adverse effect on safety.
4 In tiiat the water curtain deluge systems function as a fire barrier, revision of USAR' Sietion 9.3.1.1 was made to more appropriately t eflect that- t he I compeonatory measures associated with fire barriers would be utilized instead of the.,e associated with the s ~ inkler system. Therefore, tuere is no adverse
~
- 3' I .effect en jafety, i l
l Clarifying that nine of the local-fire alarm panels are utilized for the HVAC l- duct smoke detectora does nc+ knpact the operation or function-of the remaining local lite alarm panels. The e, there is no adverse effect on the fire protection program or on 5.afety. ,
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i Revising the description of the method of fire detection used in the cable }
opreading room t o t ellets the use of area ionization type detectors rather than !
. linear thermal detectors in the cable trays does not impact the ability of the l plant to detect and respond to a fire in the cable spreading room, Therefore. t
, there is no advetse effect on the fire pt ot ect ion progr am or oa saf ety. !
t l Revising the electric and diesel fire pump total head values to confotm with [
! the design-specific nion values does not affect the function or capability of
- t he purnps. The design specification values have been ut111 red in the design {
and 3ubsequent modifications of the various fire prot.ection components (e.g., i i hose stations, nprinklers, etc.). -
, y i The ' remaining changes made under UCli 90 093 are editorial in nature and have no t l' adverse effect on safety. !
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SAFETY EVAltfAT10ll Stf!&lARY Tok !
11011 91 015 (SC 91 0019 R01) l 1J7LEt ;
Re pl a c ernent of Engineering Assurance's Day-to-Day (In-Line) Quality Review of Procurement Docurnents with an Additional and Separate Procur ement Engineering '
Technical and Quality Rev.'ee .
C..H AllC E. : -
' Revine USAR Section 17.2 4,3, to eliminate Engineering Assurance's in-line quality r eview of procut etnent document s.
REA50!! FOR CllAllGE:
The purpose of this change is t o eliminate the r edundancy of what is now an unneeennary thi.rd level of teview.
S AFETY EVAltfATIOll Slff!!!ARY:
Paesent.ly, Engineering Assurance performs en in-line quality review of Q C (comme clal grade dedication) and limited 1 ( American Society of liechan.ical :.
Enginear [ASME) at:i Fire Protect lon) procurement packages. Since August of 1990 Procurement Engineering has proceduralized an additional technical and quality level review priot to supervisor review and approval. This proposed change will replace Engineering Assurance's in-line review responsibility with l Pr ocurement Engineering'n proceduralited t echnical- and quality review: -
supplemented with periodic Engineering Assurance procurement document 888eBSment8.
The proposed change does not directly_or indirectly affeet any structures, systems,-or components.
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i SAFETY EVALUATIOli SU11 MARY FOR U011 91-042 (SE 92-0019 R03)
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Reducing Setpoint for Decay Heat Pumps Suction Tetnperature Alarm as Ref erenced j in the USAR '
l C!!A!1GE l
The proposed activity consists of teducing the setpoint for Decay Heat Pump cuct ion temperature annunciator alarm 3 3-H(1) from 3200F, currently l telorenced in USAR.
REASoli FOR CllAllGE:
i
. TI)is change will result in a tuot e conservative temperature als m setpoint and ;
make the USAR consistent-with the historical netpoint- for the annunciator alarm.
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$AFETY EVAL.UAT1011 SUI!!iARY:
The Dil Pumps suction temperature annunciator provides an alarm to signal the control room operators that the suction temperature is above normal.
This UCll 91 042 is changing referenced value from 3200F to the value of 3150F, 2 This change will make the USAR setpoint consistent with the Set point Index and ;
the field settings of-the instrument. This change will be more conservative with respect to th' caximum piping temperature limit of 3500F and will thus continue to provide appropriate indication of an above normal suction tempetature, i
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SAFETY EVAltfAT1011 Sulf!!ARY FOR UCll 91-043-(SE 91-0037)
. TITLE Revislou of USAR Descriptico of the Administrative Dose Control Guidelines CilAllGE s
. Revise description of the ptocess and control responsibilities for the administ1atIve dose cont 01 guidelines, g A5011 FOR Cil/dlGE:
This change replaces the spet;fic dose levels and approval ptoceso fot controlling dose leveln with a more generic description of the administrative guidelines f,sr controlling dose levels. ,
- SAFETY EVALtfATIOli SUlMARYi Deletion of the specifj guidelines in the USAR does not impact plant safety <
since conformance to the t egulatory requiremente are st.111 met.
Administ ratively_ cont r olled -dose levels are-still established to prevent 2 personnel from exceeding the dose limits of 10CFR20.
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I SAFETY EVALUATIO!1 $UlltARY 1
FOR I
U011 91-044 (SE 91 0048, Rol)
TITLE: r Revision of USAR Description of Counting Equipment for Radioactivity l
!!*asutement i CHAllCE : !
Revision of the USAR to describe the radioactJve counting equipment as it currently exists. <
_REA5011 FOR CllAllGE:
This USAR chat,e brings the equipment description up to date. The equipment replairement han improved t he r eliabillty, maint ainability, and cost-
- effectiveness of the gamma spectroscopy system. . This change describes t he system as it exists today. As-State of the art advanced, the-equipment i described in the USAR was replaced. This replacement was also predicated upon spare parts being readily available and maintenance being simplified.
,SAJETY EVALUATION SUMitARY:
The equipment replacement has improved the reliability, maintainability, and cost af fectiveness of the ganuna apectroucopy system, therefore, there is no adverse.effect on the function'of the system.
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i SAFETY EVALUAT10ll
SUMMARY
POR UCN 91-055 (SE 91-00$2) ,
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TITLE .
Removal of Hakeup Pumps and Lp1 Crossover Piping and Valves from ECCS Equipment kallabillt y Considerations Cll AtF'.E s Removal of the Makeup Pump *, (with associated auxiliaries) and the LPI Crossover ;
piping and valves-(in the auxiliaries list for LPI) from the USAR table of r Eni;1nect ed Saf ety Featuirs (ESP) equipment and the Aux 111atio required to support the ESP function of each piece of equipment-.
.REAS0tl FOR CilANGE:
The Makeup pumps and the LPI Crossover valves and piping were incorrectly included in USAR section of required ESF equipinent.
SAFETY EVALUATION _ SUll!!ARY: ,
Removing the Makeup pump and its associated auxiliaries from the list of ESF equipment in Section 6.3.2.11 of the USAR will have no effect on safety because the Makeup pumps have no'nafety function assumed in the SAR analysen.
peleting the LPI Crossover piping and valves f rom the list of auxiliaries ,
required for proper LPI. function will have no offect on safety. The primary ESF funct. ion of the LPI System, injecting water into the Reactor vessel from the BWST or Contairunent Emergency Sump is not af fected by the LP1 crossover piping. with one exception. The single exception is a Core Flood Line break with an assumed single failure on the other LPI t rain. Due to the insert in the Core Flood Line nozzle this break falls into the small break qize, therefore this scenario has been analyzed using only one Core Flood Tank and
. one !!PI pump. The results of the analysis were acceptable and are, presented in i USAR Section 6.3.3.1.3. -
I None of the other safety iunctions of the LP1/DilR system listed above are affected hv the LPI Croswover valves and piping. This feature has been added
- ito t he design to increase the system's flexibility to provide abundant _
core ,
cooling in any raostulated scenario however it should not. tie included in a list
-of required ESF equipment.
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i SAFETY EVALUATIO!1 SUl24ARY I FOR UCll 91-056 (SE 91-0056) ,
TITLE: ;
Power Range fluelear inst rumentation (!!I ) Calibration [
CilAllCE : ,
I Document additional details concerning power range til calibration.
REAS0!! FOR CHA!1GE: l The power range nuclear instrurnentation (til) channels do not measure absolute power icvel. Because of this, the power range 111 channels munt be periodically
- calibrated to maintain the indicated 111 power level within an acceptable range of the heat balance power. The basis for power range 111 calibration is. j teviewed in this safety evaluation. t SAFETY EVALUATIOli SUMt4ARYr Analysis of Transjent Induced til Errot s:
Preliminary Safety Concern-(PSC) 7-70 was initiated by Babcock & Wilcox and quuntionn!t.he accuracy-of the assumed 2.0 2RTP transient induced neutron .
measurement uncertainty included in the design overpower. This uncertainty could reach'13 IRTP for transients that result in an overcooling of the RCS coolant =in the reactor vessel downcomer region.- This-is because the reduced downcomer temperature increases the shielding effect and lowers the neutron leakage measured by the power range detectors. Since the transient neutron measurement uncertwinty could be larger than 2.0 %RTP, actual core power could
_ exceed the assumed 312.0 !RTP design value prior to initiation of a high flux reactor trip. A B&W evaluation of the temperature ' induced neutron measurtiment errors was conducted for Davis-Besse cycle 2. Results demonstrate that the DilBR penalty associated with power levels greater'than 112.0_%R7P is offset by the benefielal effect of the lower RCS temperature at the core inlet. Because of this, Df1BR margin is maintained for power levels up to 123 1RTP.
For cycles subsequent to cycle 2, PSC 7-78 was evaluated on-a cycle specific basisi The cycle 6 analysis of PSC 7-78 with crossflow methodology eliminated the need for further cycle specific analyses.
Iti summary. the resolution of PSC-7-78 shows that the transient induced neutron ;
measurement error that occurs during overcooling transients is acceptable in:
terms-of plant safety. Thus. _ no change to the original ill calibration specification-is needed.
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Flux / Flow / Imbalance Trip function:
The de<;ign analysis f or t he flux / flow pottion of the flux / flow / imbalance trip functton is based on a one teactot coolant pump (RCP) coastdowa with a maximum power of 108 1RTP. The derly,n power of 108 1RTP includes a 2 !RTP allowance lot s t eady- s t a te ill calibtations Therefute, the powet range til detectors may be allowed to indicate up to 2 percent below the heat balance powet and sti.ll ptenetve the assumptican uced in the de91r,n basis analysis.
Antirlpatory Reactor Trip Syntem (ARTS):
The powet t ange til det ect or s provide power level indication to ARTS. According to a Davis-Eesse calculation, the ARTS setpoint includes a 2 petcent allowance f or ill calibrat ion. Theteiote, the power tange til detectors may be allowed to indicated up to 2 percent below the heat balance power and the ARTS net point talculation will temain valid. -
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SAFETY EVALUATIOli SUM!iARY FOR UCll 91 058 (SE 91-0054)
TITLE:
Prov,rammatic Controls and a Yearly Surveillance in Lieu of Biennial Review CilAllGE :
Davis.Besse currently perionns a biennial review of plant procedures as requit ed by our conunittnent to A11SIi Alls 3.2 - 1982 Section 5.2.15. The proposed change will climinate the need for the required biennial review. This review will be replaced by many current programmatic controls which provide assurance that procedures are maintained current and with a yearly Quality Assurance surveillance of randomly selected plant procedures. This surveillance will provide assurance that plant procedures are_being maintained current.
REAS011 FOR CllA11GE:
The purpose of the char.ge is to eliminat e an unnecessary level of review of .
plant procedures. . This review is redundant to the many current programmatic controlo which provide assurat.ce that procedures are maintained current, This
_ change also provides for a yearly surveillance of plant procedures to provide >
added assurance that procedures are maintained current.
SAFETY EVALUATIOli
SUMMARY
As noted alove, the proposed change is an administrative programmatic change and does not affect the safety function of any structures, systems, or comportents.
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SAFETY EVALUATION SUFEtARY FOR UCN 91-073 (SE 91-0079)
TITLE:
capability of Siphon Breaker to-Prevent Water Loss from the Spent Fuel Pool CHANGE:
Revision of USAR to reflect additional analyres performed on the siphon breakers in the Spent fuel Fool.
. REASON FOR CHANGEt
' PCAQR 88-1001 evaluated the capabilities of the siphon breaker to prevent water loss-from the Spent Fuel Pool (SFP) in the event of a postulated pipe break in the seismic class 11 portion af the SFP Cooling and Cleanup Syste: (SFPCCS) piping. This safety evaluation provides the basis to clarify the information related to siphon-breakers in USAR section-9.1.3.11.
SAFETY EVALUATIOil SUF&iARY:-
The evaluations-performed ao a part of PCAQR 68-1001 conr.luded that the most limiting break in the SPFCCS seismic class 11 piping is a 6" pipe' break at SFP heat exchanger 1-1 discharge at 587'-6". It is noted that it is not required to postulate breaks in piping which in seismically supported. Therefore it can be conservatively assumed that with the present size of the siphon breaker the SFP water level following the break will siphon no lower than the break location of 587'-8". The top of the fuel tacks is at an elevation of 577'-8".
- Based on this assumption, the water level abcve the fuel racks following the break will be approximately 30 feet. The active fuel is approximately eight j inches below the fuel racks.
The HRC Safety Evaluation Report (SER) (Reference.6) considers that nine feet of water above spent fuel elemente provides a sufficient water cover over spent fuel daring postulated pipe' failures. This safety evaluation criteria is met for the limiting break of SFPCS 6" seismic class II piping. The =1 phon breaker.
provides protection against pipe breaks in the 2.5" seismic class II piping in the spent ~ fuel pool cleanup system which is located below 587'-6" elevation.
There are no seismic class.11 pipes greater than 2.5": connected to SFPCCS below 587'-6' elevation. The outlet piping does-not need a siphon breaker because it ;
terminates well above.the 587'-6* 1evel. The draire piping which is connected to the outlet' piping is isolated by a normally closed valve. Any temporary lines used_in the fuel pool will be evaluated for anti-ciphoning protection.
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SAFETY EVALUATION SUM!!ARY TOR UCN 91 078 (SE 91-0083 R01)
TITLE:
On-site Meteorological Measurement Ptortam
$^,U.G5 8 UCl1 91-07f1-19 a tewtite of USAR Section 2.3.3 to eliminate unnecessary detail and to permit design improvements.
REASoll FOR CHA!GE:
L
!!ost of the meteorological monitoring system equipment has become obsolete and' ,
- requires replacement. Davis-tesse has installed a backup system which meets Technical SpecifAcation and Regulatory culde 1.23 requirements and is redesigning the primary system to take advantage of modern technology and
- commercially available equipment. This will change equipment vendors, model numbers, and operating chatacteristics, and the methods by which this information is made available to the Control Rooms all of which are described in the USAR.
GAFETY EVALUATION SUM 11ARY:
This UCl1 incorporates the requirements of Regulatory Guide 1.23 by reference, and adds site-specific information required by the USAR standard format documents. In addition, the perf ormance criterls--in USAR-Table 2.3-8 are revised to be equivalent to the Regulatory Guide 1.23 requirements. Extraneous information such no model' numbers and vendor da*.a is deleted.
. The Meteorological Monitoring system collects data which can be used to predict the dispersion of radioactive releases from the plant. In the event of an [
accident, the Emergency plan uses meteorological monitoring system data in conjunction with-release information to estiraate radiological risks to the public. Because this system bas no control over plant operation nr radioactive releases, the proposed change will have na effeet on the radiological consequences of any event.
The UCN will not-reduce the margin of safety as defined in the basis for
- Technical- Specifications because t he change is editorial ir, nature, and does not affect any margin of safety. Required Meteorological Monitciring Ssytem capability is ensured by continued compliance with Technical Specification Sections 6.9.2(c) (Speclul Reports), and 3.3.3.4 (Limiting Conditions for_ '
Operation), and Tables 3.3-8-(Instrumentation) and 4.3-5 (Surveillance Requirements), as well as the commitment to Regulatory Guide 1.23 mentioned above.
1 The affected Systems, Structures and Components are not safety related and have no. direct impact on any system which is important to safety.
Physical changes
- which-wlll be permitted by this UCN constitute enhancements.
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. i SAFETY EVALUATIO11 SUlt'ARY ?
FOR UCl1 92 002 (SE 92-n007)
T,ITLE
!>eletion of the General M % 1al Inspection Check.ist (G!!IC)
CilAllGE r Procedute Ell-DP-00070 is being revised to delete the use of the Gtneral Mat eris! Inspection Checklist ( Gl!IC ) . In its place, the Data Ansignment Sheet (DAS) will be used to specif y ry special st orage or handling requ! rements. ;
USAR Table' 3 7.2-1 was also revised to state that procuretrent docur.ents. Instead i- of the G!1ic, are to be used when specifying if special storage or handling is required for equipment.
REAS0!J FOR CilAllGE:
The purpose of the change to E!1-DP-00070 is to eliminate the GMIC form to make
.- more efficient use of the Data Assignment Sheet in order to increase the efficiency of the-procurement process.
- This safety evaluation .is written to justify modifying the method of specifying
~ special storage or handling requirements as is described in USAR Table 17.2-1. 1 SAFETY EVALUATION
SUMMARY
r ,
' This change is an administrative, programmatic change which does not affect the ,
quality of any parts either in the p.lant or warehouse and does not directly or indirectly affect the safety functien(s) of any structure, system, or component.
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SAFETY EVALisATIO!1
SUMMARY
FOR UCl3 92-007 (SE 97-0008)
TITLEn-Centerior Management Reorganization CilA11GE :
The proposed changes to the USAR consist of c.ne consolidation of cettain corporate upper management responsibilities and the creation of a new position '
within the corporate organization.
REASoll J OR CHAtlGE: -
With the retirement-of Centerior Energy's current Chairman and Chief Executive Officer, the Centerior Energy Board of Directors appoir.ted the current President and Chief Operating officer to succeed the Chairman upon his retirement. The-President and_ Chiel-Operating officer position will be eliminated.
Along with this change the Centerior Energy Board of Directors also created a new- position of benior Vice President-Legal. Human and _ Corporate Af f airs.
SAFETY EVALUATIOll SUlt1ARY t The proposed changes to USAR Section 13.1 and 17.2 have no effect on any structures, systems, and componenta or their associated safety functions. The proposed changes are administrative in nature and do not affect ths operation of any plant system.
The technical qualifications necessary to operate the DBt1PS continue to be
~
E provided by the Toledo Edison nuclear organization. There continues to ce f -
established and well-defined lines of authority, responsibility, and c anunication f rom the highest management levels through intermediate levels to and including all onsite operating organization pasitions involved with activities affecting the rafety of the plant. Therefore. A11SI N18.1-1971 remains satisfied.
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SAFET: EVALUAT10!! SUIDIARY FOR UC11 92-023 (SE 4't-0017)
, TITLE:
Charu'y vi9it or anil vehicle control from ownet Cont rol Area to Pr ot ect ed At en in USAR Section 13.7 GilAllGE i in Sortion 13.7 Industtial Setutity, the uutding *0wnet Cont. tolled Aren* 19 being, replaced by " Protected Area".
EEAGoll _FOR CllAllGE:
This change is in accordance with the requireinent s of 10 CFR 73.55. There is no requirement for control of visitot s and vehicles on the owner Controlled Area. Visitor and Vehicle conttols begin at the Prot ected Area boundary by Federal Regulation.
S AFETY EVALUATIO!! SUl11tARY:
There ate no tequirements or qualifications for visitors for the Owner Conttolled Area that are mandated by Federni Law. The name is true for vehirles in the Owner Constolled atea an opposed to the Protetted Atea.
The proposed change will have no adverse affects on the safety functions of systems, components. or attuctures, either directly or indirectly. This change in administ rative and only involves the changing, of Owner Cont rolled Area to I' t o t e r t e d A t'e ft .
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I i SAFETY EVALUAT1011 SUlitiARY FOR UC11 91-02S (SE 90-0116)
TITI.E r Eemoval of Shift Technical Advisor (STA) Log tilAt1GE :
UCll 92-025 was writ ten to remove from the USAR the referente to tht Shift Terhnical Advisor (STA) log. The STA log is used to documer.t evaluations of oper ating event s, incident assessments, and off-normal events. It is present ly maintained by the Shift finna ge r .
REA!uAl FOR CllAtlGE:
Refetence to the STA log was otiginally placed in the USAR t o capture the results of activities now performed in Performance Engineeting. Since Shift lianagers do not perf orm activities involving t ransient assessment and induntry event teview, it is inappt opriat e to maintain such a log.
SAFETY EVALUATIOli SUlit4ARY:
o certain evaluations, assessments, and descriptions of operating events are pr esent ly tecorded in the Unit Log. The Unit tog is maintained by the Shift Supetvisot and documents all that le requited by USAR Section 13.6.1.4. The proposed activity does not change any physical aspects of the plant. Deletion of the reference to the STA log is intended t o provide flexibility and ,
eliminate redundant logkeeping.
Deleting the ref erence to the STA Log doer not involve a change t o Technical Specif'. cation 6,10. All required log critries will continue to be made in accordance with DB-OP-00005, operator Logs and Reading Sheets. The recording -
of important plant data tollowing a teactot trip is coveted by Attachment 9 to DibOP-06910. Review of other events and incidents is addressed by Ell-DP-01310, Transient Anseasment l'rogr am.
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SAFETY EVALUATIO!! SU!DtARY l
FOK UCil 92 034 (FE 92-0024) ,
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,T I TLE :
480V Auxiliary Outdoor Distribution System CllAl1GE :
Update USAR Section 8.3.1.1.5 reflecting the existing 480V auxiliary outdoor dist ribution syst em' for outdoor et teet lighting and other miscellaneous loads.
RF.AS011 FOP. CilA!1GE:
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1 Two outdout non-essential distribution centern originally supplied power for 1 verious loads. Temporary modifications88-559 and 88-612 disconnected'one of I these distribution-centerre, BY2, f rom its nortnal and alternate feeds. The tenson BY2 was disconnected was that it was no longer-required for
- miscellaneous-loads or street lighting'.
SAFETY EVALUAT1011
SUMMARY
The function of the 480V auxiliary power non-essential distrabution system is not affected by UCl1 92 034 since the miscellaneous loads and outdoor lighting loadi, on BY2 are no=1onger required, and were removed. UCt1 92 034 reflects the disconnection of BY2 previously completed by.TH 88-612 and 88 559.
Since there were no longer any l'oads connected to BY2, the function of the 480V non-essent.lal ' auxiliary system remains tinchanged. UCf3 92-034 has no effect on safety or the 480V non-essencial auxiliary function important to safe plant operation.
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SAFETY EVALUAT1011 SUliliARY FOR 0C11 92 036 (SE 92-0030 R01)
TITLE:
Change in Section Title frem Radiological Control to Esdiatien Protection CHA!1GE :
Replace reference to Radiological Control Section with Radiation Protection throughout the USAP and redefine responsibilities of U.e 11anager . Radit. tion Prote. tion.
REASOll FOR CilAf1GE:
To align Davis.Besse with the Perry Huclear Power Plant and initiate compliance with' draft. American flational Str.ndardo Institute standard hHSI/A11S-3.1, the Radiological Control Section proposes retitling the section as the Radiation Protection Protection-Section.
TAFETY EVALUATIO11 SUlillARY:
. These changes describe the proposed organizational structure. All present administrative and programmatic responsibilities are incorporated in the submitted changes. The proposed changes consist of title changes and exchange
- of certain responsibilities within the organization. There continues to be catablished and defined lines of authority, responsibility and communication.
Thete are no Effects on safuty due to the proposed organitational change. l a
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SAFETY EVALUATI0li SUl&!ARY FOR I Ucli 92-051 (SE 92-0022) I
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i TITIE:
Receipt Inspection Process i
CilANGE:
{
Revise USAR Section 17.2.7.3 Acceptance of Materials and Items.
REASON FOR CHAllGE:
The change proposed by USAR Change Notice 92-031 more clearly reflects the existing intent for the conduct of receipt inspection including documentation ;
and proceaning of non-cenforming items. ;
SAFETY EVALU/ TION - SUlt!ARYi-USAR Section 17.2.7.3,. Acceptance of Haterials and Items addresses the requirements for acceptance and un:onditional release of items for use in quality related applications at Davis-Besse. An item is considered as ;
non-acceptable until sufficient quality documentation has been provided. The inst sentence of Section 17.2.7.3 states that unacce,, table materials or items discovered during the receipt inspection orocess are documented and processed ,
I as non-conforming material. There has never been an intent to document
- material as non-conforming for documentation that is forthcaming or in the process of being evaluated. The-receipt inspection process wpecifies that-material cannot_be unconditionally teleased for use until all aspects of the innpectica-have been satisfactorily completed, ineDading review of L
documentation. The proper procedure is to identify any non-conformances at the '
completion of the inspections tequired.
_W tile the_specifled combination.cf required inspections __are proceeding, material is maintaineo in a hold status.
t The proposed change is an adrainistrative, programmatic c'enge and .loes not I l directly or indirettly affect the.sciety function (s) of any struiture, system, or componeat. Only acceptable materials thnt satisfy the technical and quality requirements of material or items will be statused as acceptable and unconditionally released.
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SAITTY EVALUATID11 SUlitiARY TOR 110 01 - 0 000 $ , Revisien 2 (SE 02-0010)
T11kri Kevinion 2 to Db.Ol'-00005, Opetator L e p t. an;l Reading Sheets L }l A!K.E :
Revit:lon 2 eliminatet the I;e a c t o r Operatot 1.og and moves the logging t equi t ernent.c pr eviottoly in t he Rent tot Operatot Log to the Unit Log.
11ASn11 FO R CllAtK;Es lili- O P - 0 0 0 0 5 Revibjon 2 Operatot Logs atid Reading Sheets, wa s: wtitten to eliminate redundant logging that is ottutting in the Unit Log and the Reactor Opetaton Log.
SAFETY EVAL (IAT1011 SUll!1AHY Set t ion 13.0.1.2 of the USAR ntates *The log contains inioimat ion concerning thonges in tote t eact ivit y" . This information will now be logged in the Unit Log The USAR 5ection 13.6.1.2 alto delineat es tequirements fot logging alarms pet t nining t o cot e condit ione with an explanation, abnormal condit ions of i e a c t. o t opetation due to auxilisty equipment and notation of radioact Ive want e relennes. 1hase items will be entered in the Unit Log.
The proposed activity doen not change any physical aspects of t..e plant and will eliminate redundant log keeping. The proposed activit y has no ef f ect on the nnlet y of t h' plant . Deleting the React at Operat01 Leg does not involve a c h s tmo to fechnical Specification 6.10.
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