Text

1993 Annual Rept of Facility Changes,Tests & Experiments
	ML18059B126
Person / Time
Site:	Palisades
Issue date:	12/31/1993
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
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ML18059B125	List: ML18059B123; ML18059B126;
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ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 1993 ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS July 8, 1994

~~Rso~00140 940727 R

OCK 05000255 PDR

ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 1993 ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS July 8, 1994 175 Pages

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER-DOCUMENTS Table of.Contents FACILITY CHANGES FC-864 Rev 3.........

DRY STORAGE OF SPENT FUEL FC-934 RELOAD 0 CYCLE 11 1

10 FC-940 /

Rev 1.....

22 DIESEL GENERATOR BREAKERS-AND PROTECTIVE TRIP LOGIC UPGRADE -

AUTOMATIC TRANSFER FROM PARALLEL TO UNIT FC-940 * /

Rev 1................. *.........

25 DIESEL GENERATOR BREAKERS AND PROTECTIVE TRIP LOGIC UPGRADE -

UPGRADE GENERATOR OVER CURRENT LOGIC FC-940 /

Rev I 26

DIESEL GENERATOR BREAKERS AND PROTECTIVE TRIP LOGIC UPGRADE -

REPLACE SOLENOID OPERATED D/G CIRCUIT BREAKERS 152~107 (152-213}

FC-944 ASDV'S BACKUP NITROGEN SUPPLY 29 FC-946 30 BORONOMETER REMOVAL FC-947 32 OFFICE FACILITIES EXPANSION FC-949 33 CONTAINMENT SUMP pH CONTROL -

FC-950~01......... -.............

37 TEMPORARY ALTERNATE SPENT FUEL POOL COOLING FC-950-02... *.......................

42 MAINTENANCE OF SERVICE WATER (SW) AND SPENT FUEL (SFPl SYSTEM VALVES

. FC-951

~..*.....

~...............

44 CONTAINMENT AIR COOLER (CAC) COOLING COIL REPLACEMENT i

. CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents FUNCTIONALLY EQUIVALENT SUBSTITUTION FES-93-015 47 REPLACE CK-CA440 FES-93-031 47 RPS AUXILIARY TRIP UNIT ZENER DIODE REPLACEMENT FES-93-087 48 REPLACEMENT OF PRINTING DEMAND METERS FES-93-90.......

49 REPLACE MV-CA698 FES-93-091 49 REPLACE MV-CA903 FES-93-162 50 STEAM GENERATOR E-SOA AND E-SOB TUBE PLUGGING (UP TO 10 TUBES PER STEAM FES-93-204 51 INSTALLATION OF R0-2122 FLOW ORIFICE IN FI-2122 INLET FLANGE FES-93-285 51 REROUTING PIPING FOR RADWASTE VENT MONITORING PUMP P-1809 FES-93-286 REROUTING PIPING FOR EAST ENGINEERED SAFEGUARDS ROOM VENT MONITORING PUMP P-1810 FES-93-'287 REROUTING PIPING FOR WEST ENGINEERED SAFEGUARDS ROOM VENT MONITORING SAMPLE PUMP P-1811 SPECIFICATION CHANGES SC-91-167...........

CONDENSER TUBE PLUGGING SC-92-050....................

FEED PUMP RECIRCULATION VALVE POSITIONER MODIFICATION 53 54 56 57 ii

CONSUMERS POWER COMPANY PALISADES NUCLEA~ PLANT FACILITY CHANGES, SPECIFICATION tHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents

. Page

SC-92-124.*......**........**.*..* **..

.. ' 58 PROVIDE PERMAN.ENT 480 VAC POWER FEED TO THE STEAM GENERATOR (SG)

TEMPORARY SC-92-127 *.*....*..... '..... ;. *......

58 REMOVAL OF AUXILIARY FEEDWATER RELAYS R/0727 AND R/0749.

SC-92-128. _.. : -..... *.... ~.......

60 REPLACE VARIOUS PRIMARY SYSTEM TRANSMITTERS SC-92-169. *...... *.....................

62 -

REPLACEMENT OF FUEL HANDLING AREA RADIATION MONITORING CHANNELS RE/RA/RIA-2316 AND 2317

SC-92-171.**.... *... ;.*....... -.

HOGGING AIR EJECTOR CONTROLLER REPLACEMENT SC-92-177.

~............. -*....

ATMOSPHERIC STEAM DUMPS' DRIP LEG STEAM TRAP SYSTEMS MATERIAL.

SPECIFICATION CHANGE SC-92-191... *.. ; **......... *..*....

REPAIR HANDRAIL IN CONTAINMENT AT 607' ELEVATION SC-92-199 *...*.......*.....

MODIFICATION VC-10/VC-ll CONTROL CIRCUITS SC-92-220.. *.... *.. *. * *........

ADDITION OF A SUCTION LINE ACCUMULATOR TO C-17 SC~92-223.. *..

~.. *....

REPLACING MV-CRN127 AND 181 SC-,92-224. *.. *.......*. *. *..

E-6A AND B VENT REPLACEMENT SC-92-227. * *.. * * * * *. *

REPLACEMENT OF MV-,CVC2070 64 65 66 67

68.

68 69 70 SC-93-006....................

70 COMPONENT COOLING WATER HEAT EXCHANGERS. E-54A AND E-548. VENT

AND DRAIN VALVES REPLACEMENT SC-93-007...............................

. 70 REDUCED SET POINT ON TS-1824. TS-1825 AND TS-1826 FROM 104.F TO 10o*F iii

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents SC-93-019.................

UPGRADE OF C-42 SAMPLE' LINE FILTERS SC-93-026... *.................

FUEL BUILDING RUNWAY CRANE RAIL MODIFICATION 71 72 SC-93-027.. ;............................

73 RECONNECT EXISTING CHLORINATION SYSTEM TO ABANDONED CHLORINATION PIPING SC-93-032....-..........................

74 INSTALL A WELDING OUTLET AND LIGHTING FOR THE SPENT FUEL POOL (SFP) TOOL CRIB SC-93-038.......-....................

PERMANENT DISCONNECTION OF THE ALARM PORTION OF PIA-1066 SC-93-044...........

REPLACEMENT OF TIC-0821 SC-93-050......... *......................

CHANGE MOTOR PINION/WORM SHAFT GEAR SET OAR AND REPLACE SPRING PACK IN M0-3062. M0-3064, M0-3066 AND M0-3068 74 75 76 SC-93-054 ~......................-......

77 SERVICE WATER TEMPERATURE CONTROL ON CONTAINMENT AIR COOLERS SC-93-087..... *. *...........*.....

78 WELD MODIFICATION OF TE-0101 AND TE-0102 NOZZLES TEMPORARY MODIFICATIONS TM-92:..os8. *.... *.. *.......... *........ *.

80 TEMPORARY FIRE WATER SUPPLY TO WESTINGHOUSE TRAILERS ON TURBINE DECK TM-93-003 *. *......... *........

BLANK FLANGE INSTALLED IN PIPE LINE MB-1-3 11 TM-93-004.....................

INSTALLING THERMOCOUPLES ON CABLES IN TRAY TK-032 AND INSTALLING FIBERGLASS INSULATION UNDER THE COVER; AND INSTALLING THERMOCOUPLES ON CABLES IN ADJACENT TRAY TK-030 80 81 iv

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents TM-93-009. * * * ** * * * * * * * * * * * * *. * * * * * *. * * *.

82 TEMPORARY CONSTRUCTION POWER FROM V-49A OR V-49B OUTLETS IN CONTAINMENT 649' ELEVATION TM-93-010 * * * * * -. * * *. ;. * * * * * * *. * *. * * * * *

PROVIDE TEMPORARY POWER TO 52-206 FROM 52-1109 DURING BUS lD OUTAGE 83 TM-93-013 * * * * * * * *. * * * * * * * * * * * *. *. * * * * *

83 CONTAINMENT TEMPORARY ELECTRICAL POWER FROM TRANSFORMER EX-15A TM-93--014 * * *. * * * * * * * * * * * * * * * * *...........

84 TURBINE BUILDING TEMPORARY ELECTRICAL POWER TM-93-017 * * *. * *. * *. * * * * * * * *. *. * *. * * * *. *.

85 RESTORATION OF MATRIX LIGHTS AND DISABLING PIP DEVIATION ALARM FOR CONTROL ROD 8 TM-93-019 * * *

.~ * * * *. * * ; *. * * * * * *.. *. *.

INSTALL IMPROVED PLUG IN CONTAINMENT AIR COOLER VHX-2 TM-93-023 *****.****..**..*

TEMPORARY MODIFICATION TO FI-1073 TM-93-026 * * * * *. * *

REROUTE E~20B DRAIN TM-93-029 * * * * * * * * * * * * * *

INSTALL ENCAPSULATED TUBE PLUGS 86 87 88 88 TM-93-035 * * * * * * * * * * * * * * * * * * * * * * * *. * * * * *

89 LINE THE BOTTOM OF CABLE TRAY TK-030 AND TK-032 (13' EACH TRAY, 26' TOTAL) USING MARINITE INSULATING BOARD TM-93-052 * * * * * * * * * -. * * * * * * * * * * * * *. * * * *

90 TIE-JN TEMPORARY AIR COMPRESSOR TO MV-CA200 FOR SUPPLYING PLANT AIR TM-93-053

~ * * * * * * * * * * * * * *.- * * * * * *.. * * * *

91 INSTALL TUBE PLUGS TO ISOLATE A 1 3/4 11 NON-CLEANABLE U-BEND TM-93-055 * * * * * * * * * * * * * * * * * * * * * * * * *. * * *

92 INSTALL AND REMOVE SWAGELOK TUBING PLUGS ON FJ-0881 TM-93-057.........................

93 INSTALLATION OF LEAK STOP DEVICE ON FLANGE WELD PIN HOLE LEAK v

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents TM-93-059 93 INSTALL FILTER ON DRAIN LINE FROM MV-SFP505, REACTOR SIDE TILT PIT DRAIN TM-93-061.... *....... ~.........

.. INSTALLATION OF LEAK STOP DEVICE ON VALVE BODY (MV-SW136) TO

- PIPE DOWNSTREAM WELD PIN HOLE LEAK

1M-93~064_. *.....................

SERVICE WATER SUPPLY TO FLUSH CONDENSER TUBES TM-93~070......... -..... -.. *. :....

SERVICE WATER LEAK NON-CODE REPAIR HB-23-4 11 TM-93-071.....................

SERVICE WATER LEAK NON-CODE REPAIR HP-23-16 11 TM-93-074................... :....

REMOVE BROKEN INSTRUMENT TUBING FOR PX-0653 AND CAP TM-93--079..... *.....

BLOCK CLOSED CV-947 TM-93-080..... _.....

SEE TM-93-086 TM-93-086......

0 BLOCK CLOSED CV-0948 94 95

~ 95 97 98 99 99 99 '

TM-93-088....

i REMOVAL OF CHP RELAY 5Pl

......... 100 TM-93-089.... -.... _.. -............

101 OPEN LINKS TO 20/AST TM-93-090........

~..............

102 INSTALL JUMPER ACROSS CONTACT 3 OF SIS-4 RELAY TM-93-091........ *..

102 BLOCK CLOSED CV-947 TM-93-092..................... *....

.... 103 BLANK FLANGE INSTALLED ON OUTLET FLANGE OF MV-CHM750

OOOoooooOUoooooooooooOOOOOUOOOOOOOOOOOHOOoOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOHOOOOOOoOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOOooooooooooooooooOOooooooooooooooooooooooooroooooooooooooooooooooooooooooooooooooooooooooooooooooooooo**ooooOoooooooo vi

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents

TM-93-093................. *..............

104 ADDITION OF VIBRATION MONITORING ACCELEROMETERS TO THE REACTOR HEAD TM-93-096 * * * * * * * * * * * * *. *. * * * * * * * * * * * * * * *

105 TEMPERATURE MONITORING OF THE, REACTOR HEAD INSTRUMENT FLANGES TM-93-107 **.**********

REPLACE BREAKERWITH SPARE TM-93-109 ** ~ * *.** * *.********.****..*** :.

INSTALL SPARE BREAKER 252-SPARE4 IN CUBICLE 152-204 106 108 TM-93-145.***********.- ** *, *******

110

. SERVICE WATER SUPPLY TO FLASH CONDENSER TUBES OTHER DOCUMENTS C-136L Ill SPECIFICATION FOR THE SUPPLY OF VSC-24 SYSTEM CASKS EA-A-NL-91-169-01. **

~ ****** ~.

114 FUEL HANDLING ACCIDENT ANALYSIS EA-D-PAL-93-207--01.**** ; * * * * * * * * * * * * * * * * * * * * *

il5.

LOCA CONTAINMENT RESPONSE ANALYSIS WITH REDUCED LPSI FLOW USING THE CONTEMPT El-28 CODE EA-ELEC-LDTAB-004 * * * * * * * * * * * * * * * * * * * * *. * * *...

117 UPDATE OF THE PALISADES CLASS lE STATION BATTERIES LOAD PROFILES AND RESERVE CAPACITY MARGINS (125V DC)

EA-GCP-93-01

~ * *.

117 REVIEW.OF CURRENT SPENT FUEL POOL CRITICALITY ANALYSIS AND

.. BOUNDING CONDITIONS EA-RDR-93-07 119 DETERMINATION OF THE MOST REACTIVE ASSEMBLIES CURRENTLY IN REGION I I SPENT FUEL RACKS EMF-93-086(P) ****** : ******

PALISADES LOSS OF LOAD ANALYSIS 121 SE93.:.. I031...............................

122 ACCOUNTABILITY AND RADIOLOGICAL CONSEQUENCES OF EOC 10 FAILED FUEL

- - - - .............................................. ~**.. ************........................................................................................................................ ******....................................... *****....

vii

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents FSAR EDITORIAL CHANGES 125 FSAR CH 2.............................

125 FSAR CHANGE TO UPDATE INFORMATION ON SOUTH HAVEN AIRPORT AND RAILROAD LINES FSAR CH 2, 11...........

FINAL SAFETY ANALYSIS REPORT FSAR CH 3...........

SUSTAINER SOURCE REMOVAL 126 127.

FSAR CH 3.........

..................... 127 SIEMENS CODE CHANGES FSAR CH 3. * *. *.......... *... *. *.. *... *..

128 FSAR CHANGE REQUEST - ADD PELLET-CLAD INTERACTION (PCI) AS A FUEL FAILURE MECHANISM FSAR CH 3.....................

129 FUEL RECONSTITUTION EVALUATION REQUIREMENTS FSAR CH 3 & 6.............................

130 FSAR CHANGE REQUEST - TYPOGRAPHICAL ERROR AND SHUTDOWN COOLING SYSTEM SAMPLING DESCRIPTION FSAR CH 4...................

FSAR CHANGE REQUEST - REMOVES REFERENCES FSAR CH 5 *..............

o 131 TEMPORARY LOADING ON PERMANENT PLANT STRUCTURES OR EQUIPMENT FSAR CH 5.......... *....................

133 FSAR CHANGE REQUEST - CHANGE TABLE 5.2-3 CONCERNING STEAM LINE CODE CLASS FSAR. CH 5.....................

SEISMIC REQUIREMENTS FOR CERTAIN COMPONENTS FSAR CH 5..............

FSAR CHANGE REQUEST TO CHAPTER 5 134 134 FSAR CH 6.................

135 FSAR CHANGE REQUEST - ADD

SUMMARY

OF GENERIC LETTER 90-05 TO THE FSAR viii

CONSUMERS POWER COMPANY

. PALISADES NUCLEAR PLANT FACILITY CHANGES, SPECIFICATION CHANGES, TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table of Contents FSAR CH 6 AND 9 FSAR CHANGE REQUEST FOR SECTION 6.10.2 AND 9.8.2 FSAR CH 7. * * ;. *......... *... -..

FSAR CHANGES TO CHAPTER 7 ON CONTROL RODS FSAR CH 7 *AND 9................

CLARIFY DESCRIPTION OF FOGG OPERATION FSAR CH 8....................

FSAR CHANGE REQUEST FOR SECTION 8.4.1.2 136 137 137

... 138 FSAR CH 8....................

139 VOLTAGE PROTECTION AND LOAD SHEDDING SYSTEMS FSAR CH 8............... *.......

139 EMERGENCY DIESEL GENERATORS (EOG) DAY TANK FSAR CH 8. **..*...........

140 FSAR CHANGE REQUEST CHAPTER 8 FSAR. CH 9.... * *.... *.......

FSAR CHANGE REQUEST FOR SECTION 9.3.3.2 142 FSAR CH 9. :................ *..............

143 FSAR CHANGE REQUEST FOR SECTION 9.3.2.1. TABLES 9-5, 9-7, AND 9-20 FSAR CH 9.. *........ *.. *. *. *. *....*.

143 FSAR CHANGE REQUEST REGARDING CCW HEAT EXCHANGERS FSAR CH 9.................._........

144 FSAR CHANGE REGARDING STARTING OF FIRE PUMPS FROM CONTROL ROOM F SAR CH 9. *

~....*........ *. ;...... ;.....*.

14 5

FSAR CHANGE REQUEST CONCERNING DESCRIPTION OF. SPENT FUEL POOL CRANE FSAR CH 10 146

. CHANGE FSAR TABLE 10-9 °COOLING TOWER PUMPS" TO.CORRECT TO AS BUILT CONDITIONS FSAR CH 11

14 7 RADIOACTIVE MATERIAL STORAGE BUILDINGS ix

CONSUME~S POWER.COMPANY

.PALISADES NU~LEAR PLANT FACILITY tHANGES, SPECIFICATION CHANGES,.

TEMPORARY MODIFICATIONS AND OTHER DOCUMENTS Table 6f Contents*

FSAR CH 11 *..........

STACK MONITORING SYSTEM M-34

. SPECIFICATION FOR HORIZONTAL CENTRIFUGAL PUMPS

... 152 153

. M-60A.... *..*.......

~.

~...............

. 155 SPECIFICATION FOR CONTAINMENT AIR COOLERS REPLACEMENT COOLING.

COILS MP* 2. 7. 1 CPAL-.1..... ;........

~.............

156 WESTINGHOUSE FIELD SERVICE PROCEDURE 11 FUEL BUNDLE ALIGNMENT PIN.

GAGING AND STRAIGHTENING STANDING ORDER 62..... *......... *....

.... 160 TECHNICAL SPECIFICATION INTERPRETATIONS/GUIDANCE 10CFR72.48 SAFETY REVIEWS.

fC..:864 163 10CFR12.48 REVIEW - DRY STORAGE OF SPENT NUCLEAR FUEL FC-LID.............. **......

171 10CFR12.48 REVIEW - AUTOMATED WELDING PROCEDURE FOR MSB C-136A ************.*.****.******* -~ * **

172.

10CFR12.48 REVIEW - FABRICATION SPECIFICATION FOR THE MULTI-

ASSEMBLY SEALED BASKET SM-LID **

~............. * *.*..**..

.173.*

10CFR12.48 REVIEW - MANUAL.WELDING PROCEDURE FOR MSB

x

FACILITY CHANGES

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facilit~ Changes; Specification Changes, Temporary Modifications and Other Documents FC-864 Rev 3 SE93-0504 EXPLANATORY NOTE for FC-864 Safety Eva 1 uat ion*

The following 1 OCFR50.59 Safety Evaluation was prepared prior to NRC issuance of the Certificate of Compliance for the VSC-24 spent fuel storage cask. *This Safety Evaluation concluded that at that time an Unreviewed.Safety Question would exist if fuel storage in the VSC-24 system were implemented before NRC approval of the Certificate. In accordance with Palisades' procedures, this Safety Evaluation was.

sufficient *to permit cask and storage pad construction activities to proceed, but NRC approval (i.e., issuance of the Certificate) was required before any spent fuel could be loaded into a cask. The NRC subsequently issued Certificate of Compliance No. 1007, effective May 7, 1993, for the VSC-24 Ventilated Storage Cask. Pursuant to 10CFR72.210, when this Certificate was issued, Palisades gained NRC authorization as

a General Licensee to begin using the VSC-24 system for fuel storage. Fuel storage in

the VSC-24 casks was no longer an Unreviewed Safety Question..

DRY STORAGE OF SPENT FUEL The document being reviewed is the Facility Change Package that documents the approvals to add a dry fuel storage system to the Palisades Facility. Since Palisades*

began operation in the early 1970's all sp*ent nuclear fuel discharged from the reactor core has been stored on site in the spent fuel pool. Storage of spent fuel on-site will continue until. at least the year 2000 when the United States Government repositories.

are scheduled to be operational and will accept spent fuel. To provide adequate storage capacity on site, and to maintain the full core off-load capacities in the fuel pool; an alternate storage system will be needed. The alternate storage system selected will use dry storage concrete casks, which will be placed on a concrete foundation within the Palisades protected area. The Independent Spent Fuel Storage Installation (ISFSI) has been designed to store up to 600 assemblies in 25 storage casks, 24 assemblies per cask. The dry storage system will consist of five major components:

.1)

A multi-assembly basket (MSB) which will hold the. spent fuel - Safety Related.

2)

A transfer cask (MTC) which will move the f uel/MSB from the fuel pool to the ventilated concrete cask - Sat ety Related.

3)

The ventilated concrete cask (VCC) which will provide storage for the fuel/MSB -

Safety Related.

4)

A heavy haul transfer trailer (HHT) which will move the loaded VCC from the track alley to the storage pad - Non-Safety Related;.

5)

A track alley load distribution system (LOS), which is made up of steel members bridging the track alley slab. The movement of the VCC down track alley is allowed only if the VCC is on top of the LDS - Safety Related.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changesi Specification Changes, Temporary Modifications and Other Documents The general sequence of events for moving spent fuel from the pool to the final dry storage location begins Vlt'.ith the loading of the MSB in the fuel pool. The 11 x 11 spent fuel rack will be removed from the_ northeast corner of the fuel pool and replaced Y\\'ith an impact limiting pad. The impact limiting _pad will provide protection to the fuel pool liner and structure in the unlikely event of a cask drop during th_e fuel loading sequence. Another impact limiting pad will also be placed in the cask washdown area.

The MSB will be filled with borated water and, along with the transfer cask (MTC), will be placed in the fuel pool. 24 fuel assemblies that have decayed to < 1 kW (each assembly) will be loaded into the MSB using the spent fuel handling machine. The support plate and shield lid will be placed on the MSB while in the pool. The MSB and MTC will be transferred from the pool to the cask washdown area. While remaining in the washdown pit area the shield lid will be welded in place and the structural lid will be installed and welded on the MSB, when the MSB will then be drained, vented, and vacuum dried, followed by a helium flush and fill. The heliul]l fill and vent connections will then be sealed. At this point the spent fuel is totally encapsulated within the MSB.

The track alley structure was reviewed and the slab was found inadequate to support the loads associated with the ISFSI. Therefore the track alley load distribution system (LDS) must be installed prior to the VCC being moved to the hatch opening. The purpose of the LDS is to transfer the weight of the VCC and other heavy loads directly to supporting concrete walls beneath the track alley fioor. With this LDS design the track alley floor only supports the ISFSI loads where a concrete wall exists directly below. Next the MSB and MTC will be moved to the VCC that has been placed in the track alley using the HHT and the VAT. The VCC will be moved outside the Auxiliary Building with the heavy haul trailer (HHT), and there it will be set on the LOS. Using a ventilated cask air transporter (VAT) under the VCC, the VCC will slide on top of the LDS until it is directly under the hatch opening from EL 649. The MTC is placed on top of the open VCC. Sliding doors on the bottom of the MTC slide open by hydraulic cylinders, allowing the MSB to be lowered by the overhead crane (L-3) vertically into the VCC. Once the transfer cask is removed, the VCC shield ring and weather cover is attached on top of the VCC. At this point the loaded MSB/VCC is moved out of track alley using the VAT on the LDS and onto the HHT located outside the track alley area.

The HHT then moves the VSC to the storage pad. A ventilated cask air transporter (VAT) is then used to lift the VCC from the HHT and move the cask to its final storage location. This system has been designed and will be provided by Pacific Sierra Nuclear (PSN) as described in their Safety Analysis Report (SAR) Rev 0, "Ventilated Storage Cask System for Irradiated Fuel," approved by the NRC. Operation of the Palisades ISFSI will be according to 1 OCFR72 K and the Certificate of Compliance for the PSN VSC-24 Dry Cask Storage System.

Safety Analysis Summary FSAR Sections 14.11, "Postulated Cask Drop Accidents," and 14.19, "Fuel Handling Incident," are the only accidents considered generally applicable to this evaluation.

Because fuel handling is performed in the usual manner (by Operating Procedure FHS0-17), the fuel handling incident while in the pool does not require an additional evaluation. Loading the MSB requires the same equipment and same activities as CONSUMERS POWER COMPANY -

PAL~SADES NUCLEAR PLANT Facility Changes, SpecifiCation Changes, Temporary Modi fi cat i ans and Other Documen.ts moving fuel from one spent fuel storage rack to another. However, the cask drop accident is significantly different from the activity being addressed in the FSAR..

Section 14.11 discusses the drop of a 25-ton cask while the dry fuel process uses a

. tra.nsfer cask weighing approximately 96 tons. Since all lifting devices, including the transfer cask and connection points such as cask trunnions and MTC lifting yoke, shall be designed and tested according to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and ANSI 14.6; "Special lifting devices for shipping containers weighing 10,000 pounds or more for nuclear material," (Ref EA-FC-864-30) and since the fuel pool crane (L-3) has rated lifting capacity of 100 tons, lifting the 96-ton transfer cask (MTC) over the fuel pool does not increase the probability of an accident above that already addressed iri plant licensing bases: (Reference FSAR Sections 14.. 11 and 14~ 19). In addition the fuel pool crane (L-3) will be load tested at more than 96 tons before any movements over the fuel pool as part of the FC-864 pre-operational test. *

All lifting devices used over the fuel pool will be designed and tested according to

NUREG-0612 and, where applicable, ANSl-14.6. This applies to all devices used to install the impact pad and MSB lids. Other measures taken to assure safe operation include the installation of an impact limiting pad in the cask loading area that will protect. the fuel pool liner and the pool structure from damage in the unlikely event the trarisf er cask (MTC) drops into the pool. Pacific Sierra Nuclear (PSN) analysis (Reference EA-FC-864-09) CPC-109.002.1 "Design Calculations: MSB Transfer Cask Drop Analysis and Impact Limiter Design" which has been reviewed by qualified CPCo pers'onnel has shown that a drop of the loaded MTC will not damage the Palisades fuel pool liner or structure. The entry and exit from the fuel pool will be from the washdown pit directly east of the cask loading area of the fuel pool. This activity will be strictly controlled by the dry fuel storage loading procedure (FHS-M-32).

The track alley structure was reviewed and the slab was found inadequate to support the loads assoCiated with ISFSI. Sierra.Nuclear Corp (SNC) was se.lected to design a steel bridging system (LOS) to span above the track alley slab and carry the ISFSI casks from the roadway to the hatch opening in the Auxiliary Building. Design Input for normal and seismic load combinations was provided by CPCo to ensure the. design's

. compliance with the Palisades FSAR. Design Output is in the form of two analyses. A calculation addressing the normal loads (dead weight and cask loading), which for the most part governs the structure's design, is by SNC, Reference SNC Analysis EA-FC-864-019, "Structural Analy~is of Palisades Track Alley Bridge for VSC-24 Loading Operations." A calculation addressing the seismic loads (dead weight and seismic loading), which governs anchorage design, is by CPCo, Reference CPCo Analysis EA-FC-864-024, "Seismic Analysis of Palisades Track Alley Load Distribution.

System" (LOS). These structural checks conclude the movement of the ISFSI casks using the LOS is acceptable, therefore the probability of an accident due to the failure of the supporting structure will not* increase.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The consequences of accidents are not increased, for the fuel handling accident, because loading fuel into the MSB by_ the spent fuel handling machine is no different from moving fuel from one spent fuel rack to another. Therefore,. the accident consequences are not increased from those described in FSAR Section 14.19.* The consequences of a cask drop in the fuel pool and areas around the fuel pool for a.*

loaded cask has been analyzed to show no adverse effects will be possible. Reference EA-A-NL-91-1 69-02. This EA, "Radiological consequences of a cask drop in the spent.*

fuel pool with the 96-ton transfer cask for dry fuel storage" was written to evaluate the worst case scenario for a cask drop in the spent fuel pool. This engineering evaluation uses the Standard Review Plan an9 Regulatory Guide 1.25 as analysis guidelines.

Many elements.of NUREG 0612, Section 5.1 are contained in the Standard Review Plan. Those elements not in the Standard Review Plan are addressed by EA-864-09, "Design Calculation MSB Transfer Cask Prop Analysis and Impact Limiter Design,"

Permanent Maintenance Procedure EHS-M-23, "Overhead Crane Inspection,"

MSM-M-13, EA-A-NL-91-169-02 and EA-FC-864-30.

No other consequences of a malfunction of equipment important to safety will _increase*

since all other possible accidents have been evaluated and show no other safety related equipment will be affected by a drop of a load.ed cask. Reference analysis completed -

by PSN, "Design Calculation: M_SB Transfer Cask Drop Analysis and Impact Limiter

Design" CPC-0109.002.001 (Reference EA-FC-864-09). As part of the fuel loading process, two lids must be placed onto the loaded MSB while in.the fuel pool. These two li,ds must be moved over the 24 fuel assemblies already in the MSB. The lifting devices are designed and tested according to NUREG-0612 and handling is controlled by the Fuel Loading Procedure (FHS-M-32). However, if a lid were to drop onto the loaded MSB, the radiological consequences are bound by the analysis (EA-A-NL-91-169-02) which shows the release rates for 73 fuel assemblies in spent.

fuel Rack #10 are below 10CFR100 limits as discussed above. With the addition of the -

track alley bridge the failure of these structures is not a* credible event. Reference EA-FC-864-09, EA-FC-864-19 and EA-FC-864-24 for LOS structural analysis. The drop of the MTC/MSB lid (structural shield lid, etcJ onto the loaded MTC/MSB is bounded by EA-A-NL-91-169-02 as it would be a lesser accident. The heaviest load to be moved by the fuel pool crane will be the MTC/MSB when loaded with spent fuel, water, and the support plate/shield lid as it is being removed from the fuel pool (loaded weight is 96 tons). At the location in track alley where.the MSB/MTC will be loaded onto the VCC the LOS Will be in place sitting above the track alley slab and, no safety related equipment necessary for safe shutdown of the Palisades Plant exists at the elevation immediately below. In addition, Chapter 14 of the FSAR has been reviewed, specifically Section 14.24; and the consequences of these analyses have not changed.

Since a drop of the. MSB/MTC into the track alley will not cause the MSB to open (Reference EA:-FC-864-11 "Design Calculation:.Evaluation of the MSB for Drop Loads for a Hypothetical.Drop on the LOS. in the Track Alley"), th~ radiological consequences are not different from those previously evaluated in FSAR Section 14.24. Per EA-FC-864-09 and EA-FC-864-24 the LOS will assure that the track alley floor will not

fail if an MTC/MSB onto the VCC occurs. If the MSB were dropped prior to placement into* the VCC, or the MSB were dropped into the VCC, an analysis has been completed CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents by PSN which shows the MSB will not fail under these conditions. Reference PSN Calculation CPC-109.002.012 (Reference EA-FC-864-11 ). Once the MSB is placed inside the VCC, the VCC has been analyzed for tip over, seismic, tornado, wind loading, missiles, etc. (Reference SAR Rev 0, Sections H and 12, "Accident

Analyses.")

Accidents of a different type than any previously evaluated in the FSAR will be created.

Because of the unique design of the Independent Spent Fuel Storage Installation (ISFSI), a number of potential accidents and malfunctions different from any previously

. evaluated are now possible. Those identified within the PSN SAR include the.

following:

l)

Unusual environmental conditions such as extreme heat may reduce the ability of ventilated storage cask (VSC) to remove heat from the fuel. Extreme cold might affect the cask materials. An analysis has shown that continuous temperatures up to 1 00 degrees F with average solar load and as low as -40 degrees F without solar load will have no detrimental effect on the system operation. *Reference Section 11. 1. 1 of the PSN SAR. Reference EA*FC-864-02 documents the Palisades site specific parameters enveloped by the dry storage system limits addressed in the NRC approved SAR Rev 0.

2)

An.analysis has been performed to show that 1 /2 of the air inlets may be blocked without preventing proper cooling* operation of the VSC. Additional analysis shows that complete inlet blockage will be acce.ptable without detrimental effect on the.

safe operation of the system as long as the total blockage (all four inlets blocked) has not been in place more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If total blockage is in place more than

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the concrete cask must be taken out of service. Reference Secti'ons 11.1.2 and 12.2.3 of the PSN SAR Rev 0. A daily surveillance will be incorporated to inspectthe inlets to assure they remain open. Reference Section 11 ~1.2 of the PSN SAR and Design Calculations EA-FC-864-05, "Air Flow

. Calculation for Loaded (24Kw) VSC,". EA-FC-864-43, "VCC Temperature Monitoring at ISFSI Pad." *

3)

An evaluation has been performed assuming an alignment. interference occurs during the lowering of the MSB from the MTC into the VCC. The corrective actions and consequences associated with the freeing of the MS1;3 have been evaluated within the SAR Rev 0 Section 11. 1.3. The MSB, VCC and MTC are sufficiently strong to support the loads developed during this occurrence so no over stressed conditions will exist in these components.

4)

The SAR has evaluated the radiological consequences of an inadvertent loading of a-contaminated MSB and its subsequent storage in the ISFSI. Reference Section

11. 1.4 of the SAR states no adverse effect will take place. The fuel loa~ing procedure (FHS-M-32) will also provide controls to minimize the potential of this type occurrence from happening.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents

5)

An evaluation has been completed which addresses total failure of the MSB and all its contained fuel. (Reference Section 11.2.1 of the SAR states release rates are below limits set by 10CFR72.106.). *

6)

An evaluation has been completed to show the VSC tip over incident during transfer due to failure of the roadway between the track alley and the storage pad will not adversely affect the integrity of the MSB. Reference SAR Rev 0 Section 11.2.6.2 and PSN Evaluation CPC-109.002.5_ "Existing Plant Road Evaluation *

(Reference EA-FC-864-016)." In addition, a load test will include moving the loaded trailer over the roadway between the track alley and the storage pad as part of the pre-operational test procedure for the dry storage system to verify integrity of the roadway (Reference test procedure T-FC-864-01). Also, as part of the

equipment preparation procedure (FHS-M-33), a load test of the roadway will. be

performed before moving a loaded cask over the road if an equal load has not be~n moved over the road for one year or more. Section 1 i.2.6, "Earthquake Event," of SAR Rev 0 shows that a vertical ground displacement of approximately 5.6 feet would be required to move the center of gravity over the corner of the. cask so that th.e cask would topple. It should be noted that the trailer bed and the LDS on which the VCC sits is approximately 26" high so even a postulated trailer. failure will not overturn the VCC. If the roadway were to fail during cask transporting, those plant structures that may be affected are all non-safety related and will not in any circumstances affect safety operation of the plant. Structures that may be affected are the support building, temporary office trailers and the construction office building.

7)

The VSC has been evaluated for tornado loading and tornado generated missiles~.

with no adverse effects on the system operation. Reference Section 11.2.4 of the SAR. Reference EA-FC-864-02 documents the Palisades site specific parameters enveloped by the dry storage system li_mits addressed in the NRC approved SAR

- Rev 0.

8). The VSC has been designed to withstand flooding conditions, without adversely affecting safe operation of the system. Reference Section 11.2.5 of the SAR. The VSC is the only component of the dry storage system that may be exposed to flooding conditions. The final location of the ISFSI will minimize the potential for flood because the Palisades *1sFSI is located above any flood plain. Reference.

EA-FC-864-02 documents the Palisades site specific parameters enveloped by the dry storage system limits addressed in the NRC approved SAR Rev 0.

9)

The VSC has been designed conservatively with respect to the Palisades design basis earthquake, without adversely affecting the safe operation of the system.

Reference Section 11.2.6 of the SAR. The PSN SAR Design Basis Earthquake (0.25g horizontal and 0. 17 vertical accelerations) encapsulates the design basis earthquake at the Palisades site. Reference EA-FC-864-02 documents the Palisades site specific parameters enveloped by the dry storage system limits addressed in the NRC approved SAR Rev 0. '

CONSUMERS POWER COMPANY - PALISADES. NUCLEAR PLANT Fa~ility Changes, Specification Changes, Temporary Modifications and Other Documents

10) The MSB has been analyzed for internal pressurization caused by failure of all the contained fuel. The analysis shows no adverse effect on the safe operation of the system. Reference Section 11.2. 7 of the SAR.

11) An analysis as required by the NRC has been completed assuming fresh unburned fuel assemblies are loaded into the MSB: The analysis shows that no adverse effects will be generated under these circumstances. Reference Section 11.2.8 of

the Topical Report. Reference EA-FC-864-02 documents the Palisades site specific

parameters enveloped by the dry storage system generic design (PSN SAR Rev 0) which has been approved by the NRC.

l2) The MSB was designed and analyzed for a lifetime of 50 years. Note that the general license for the storage of spent fuel in each cask fabricated under a *

. Certificate of Compliance terminates 20 years after the date that the particular cask is first used by the gen'eral licensee to store spent fuel, unless the. cask's.

_Certificate of Compliance is renewed, in which* case the general license terminates 20 years after the cask Certificate of Compliance renewal date. (Reference 1 OCFR72.212(a)(3)). However, the exterior of the MSB is coated so that no corrosion is expected.

Malfunctions of a* different type_than any previously evaluated in the FSAR will be created. Because of the unique design of the Independent Spent Fuel Storage Installation (ISFSI), a number of potential malfunctions different from a-ny previously evaluated are now possible. See above.

The margin of safety defined by Plant Licensing Bases is not reduced by the addition of casks for storage of spent fuel because the addition of storage casks does not violate the acceptance limits for plant operations. The actual cask system to be installed and used will be under the control of an approved facility change by 1 OCFR50.59, 1 OCFR72 Subpart K and AP 9.03, when all acceptance limits unique to the cask system will be met during initial and future operation. Administrative controls are in place to assure the appropriate evaluations are documented before completing the heavy loads activities, thus assuring the margin of safety is not compromised. In addition, design criteria for environmental conditions and natural phenomena are considered and meet the requirements established with ANSI 5 7. 9-1984, _..Design Criteria for an Independent Spent Fuel Storage Installation (dry storage type). II Reference SAR Rev 0 Section 2.2.

To comply with Technical Specifications Section 3.2.1.2e and f which states no heavy loads shall be moved in the fuel pool area unless an evaluation to meet Section 5.-1 of NUREG-0612 has been completed. D~sign Calculation (EA-FC-864-09) completed by Pacific Sierra Nuclear, shows that a drop of the. loaded MSB/MTC will not adversely affect the fuel pool structure as long as the impact pad is installed in the fuel pool and washdown pit. Installation of these impact pads are controlled by the pre-operational test procedure T-FC-869-01, and normal system procedure FHS-M-33. This evaluation has been completed to meet Section 5.1 of NUREG-0612. In addition, precautions CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Ch~nges, Specification Changes, Temporary Modifications and Other Documents have been taken to assure the potential for a load drop in the fuel pool area-is.

extremely small as defined by generic letter 85-11. Actions to assure the potential for load drops is extremely small include load testing the crane to equivalent loads; designed, tested, and inspected all lift rigs to the requirements of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" and ANSI 14.6, "Special Lifting*

Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials" where applicable; all heavy load movements within the fuel pool area will be controlled by procedure (FHS-M-23) which defines specific heavy load paths and will not allow movement of any heavy loads over spent fuel stored in storage racks; only trained crane operators and signalmen will perform the heavy load move, and finally, the overhead crane will be inspected following existing procedure MSM-'M-13, "Overhead Crane Mechanical Inspection," before starting the fuel loading process.

GENERAL COMMENT

S Other safety issues that will be addressed in the operating system procedure (FHS-M-32) shall be administrative controls for maintaining proper boron levels in the spent fuel pool as described in the SAR, and compatibility of all equipment materials that may enter the fuel pool.

Selection of the appropriate fuel assemblies for placement into the dry storage system will be completed by Reactor Engineering and documented within the fuel loading procedure. The assemblies selected will be compared against fuel characteristic.s..

Specified within the approved SAR to assure the assemblies selected are bounded by the approved cask design and the Certificate of Compliance for the VSC-24 system.

Selection of the proper fuel assemblies will assure operation of the dry cask system within design limits as described in the PSN SAR Rev 0 (which includes criticality control, ALARA concerns short and long term, and expected dose rates on the MTC during the loading process). All operational. procedures used to control activities associated with the dry fuel process shall be reviewed and approved using the appropriate administrative procedures (Ref AP 10.41). All safety related procurement shall include a documented Quality Assurance program implementing the applicable requirements of N45.2. All crane activities shall be controlled by qualified operators and signal men. A pre-operational test procedure (T-FC-864-01) which verifies all aspects of the fuel 1.oading sequence will be completed bet ore declaring the system operable. Health Physics coverage shall follow the appropriate plant administrative procedures (Reference AP 7.00 and 7.02). All site construction activities will be completed outside areas containing safety related equipment. A supplemental steel structure has been designed to bridge the track alley floor so the structure can adequately support the heavy casks and associated equipment (Reference EA-FC-864-019). Security requirements necessary to maintain and operate an ISFSI (Reference 10 CFR 72.212) will be completed under a separate Facility Change (FC-925) which will be tested and declared operable (per AP 9.03) before placing spent fl,lel at the ISFSI. I I

I

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT

Faci1ity Changes, Specification Changes, Temporary Modifications and Other Documents Since the.concrete casks are highly resistant*to the effects of fire and explosion, a fire needs to burn for days before much of the. concrete wall thickness would be affected.

And, since *a fire or explosive event is highly unlikely, no fire or explosion protection is required. (Reference SAR Section 2.3.6). Because of brittle fracture concerns relating to the potential damage caused by dropping the MTC or MSB during transport with.

extremely cold ambient temperatures, limitations will be imposed as part of the normal loading procedure (FHS-M-32) which will prohibit moving the MTC or MSB if the ambient temperature approaches these limits. The* heavy haul trailer (HHT) and its prime mover, which will move the VCC from the track alley area to the storage pad and the ventilated cask air transporter' (VAT) and its prime mover, which will move the VCC from the track alley via the LOS to the HHT and then again from the HHT to the long term storage location on the storage pad, are all non-safety related since the VCC can withstand any credible accident during transport to the storage pad (Reference SAR Section 1.1 ). The Fuel Loading Procedure will provide limitations on the speed at which the HHT and VAT will be allowed to move and the roadway between the track alley and the storage pad is flat; grade changes are not a concern. The automatic welding system (AWS) will be used to seal weld the shield and structural lids onto the MSB using a qualified process controlled by the approved loading procedure (FHS-M-32).

This is also true for the draining, vacuum drying, and helium filling processes used on the MSB. A part of the pre-operational testing of all of the above activities will be performed on an MSB mock-up of identical materials as the actual MSB. These testing

activities will be successfully completed before loading the first MSB with spent fuel.

All instrumentation calibrations will be performed according to the existing M&TE program at Palisades or, in the case of the welding equipment, according to the qualified welding procedure. Revisions to the Palisades procedures that are affected by the dry fuel loading process will be completed before the start of loading the first MSB.

Revisions to the Palisades FSAR describing the dry fuel storage system will be initiated bet ore the closeout of FC-864. All dry fuel storage system equipment periodic maintenance activities will be established before closeout of FC-864 and included into the existing maintenance program at Palisades. A permanent maintenance procedure will also be written and approved according to AP 10.41 to provide for preparation of all dry fuel storage system equipment before the initiation of fuel loading. This procedure will be initiated before closeout of the FC'-864 package. (Reference Permanent Maintenance Procedure.FHS-M-33). In addition, the unloading of the MSB will be demonstrated as part of the pre-operational test procedure (T-FC-864-01 )_and-a Permanent Maintenance Procedure (FHS-M-34) will be developed before the close out of FC-864 which will provide instructions for off loading the MSB in the event a VSC/MSB must be taken out of service. Over the time to complete FC-864 the vendor has changed names from Pacific Sierra Nuclear Associates (PSN) to Sierra Nuclear.

Corporation (SNC). *_The expertise/analysis and design work, etc associated with PSN or SNC has been completed by the same individuals under the s_ame QA program.

Therefore, the company names may be interchangeable.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents FC-934 SE93-1058

-R/1 SE93-1025 R/1 SE93-0159 RELOAD 0 CYCLE 11 SE R/2 SE R/1 The Cycle 11 reload will include 60 new assemblies; and 52 M, 68 N (standard), 8 N (shield) and 16 Modified I assemblies currently in Cycle 10.

Reload 0 continues to l.JSe the design changes made in Cycle 1 0 to improve debris resistance. Cycle 11 is a low leakage design, which will minimize fluence on the critical reactor vessel axial welds. A comprehensive description and evaluation are included in Siemens Power Report EMF-92-177 Revision 2, "Palisades Cycle 11 Safety Analysis Report."

Safety Analysis Summary The Cycle 11 reload involves the replacement of 60 assemblies with fresh fuel, and the replacement of 16 "I" series assemblies with 16 "L" series assemblies. Fresh fuel is required for continued reactqr operation. The replacement of fuel does not increase the probability of an accident previously evaluated in the FSAR.

Siemens Nuclear Power Corporation has issued a report, EMF-92-17 8 Revision 1 "Disposition of Standard Review Plan Chapter 15 Events for Palisades Cycle 11." This report lists the accident consequences bounded by existing analysis as well as those events requiring additional analysis. Per Siemens letter (HGS:312:93), incorporation of the "L" assemblies in the reload is bounded by the current report. The cycle 11 core design required to change to the Technical Specifications which increases the radial peaking factor limits for Cycle 11 (Reload 0 fuel assemblies). The increased radial peaking limits for Cycle 11 caused the predicted minimum DNBR to decrease and peak linear heat rate to increase for anticipated operational occurrences. The MDNBR is predicted to Jemain above the ANSFP correlation limit and the peak LHR is predicted to remain below the fuel centerline melt criteria for all AOO events. Therefore, the consequences of all AOO events are within the specified fuel design limits.

All but four postulated accidents remain bounded by the previous analyses. The effect of increased radial peaking limits on MDNBR and peak LHR was assessed for the reactor coolant pump rotor seizure and single rod withdrawal events. The effect of increased radial peaking limits on radiological consequences was assessed for the fuel handling and spent fuel cask drop accidents.

MDNBR, for the reactor coolant pump rotor seizure and single rod withdrawal events, is predicted to remain above the AFNP correlation limit and the peak LHR is predicted to remain below the fuel centerline melt criteria. Thus, the consequences of these events are within the specified acceptable fuel design limits.

. CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes~

Temporary Modifications and Other Documents The predicted radiological consequences for the fuel handling and spent fuel cask drop accidents are less than those predicted by the previous analyses of record. Though higher peaking factors are allowed, the use of dose conversion factors from ICRP 30, which are consistent with the latest revision to 1 OCFR20, results in lovver predicted consequences.

Therefore the consequences of *a11 events remairi less than the acceptance criteria and operation of the facility by the revised Technical Specifications would not result in a.

significant increase in the consequences of an accident previously evaluated.

The mechanical design of the "O" fuel is basically the same as the previous cycles, including the debris resistant features. There is no* impact ori reactor vessel internals, active fuel region, or primary systems components because of this modification.* The probability of malfunction of equipment important to safety is not increased.

SAN-8 Upoer Tie Plate Change Fuel assembly SAN-S, which lodged in the UGS when the UGS was lifted, has had its upper tie plate replaced with a modified upper tie plate. The modified tie plate, designated SAN-SR, has chrome plated alignment pin hole inside diameters, an increased diameter undercut section at greater depths, and a slightly wider entrance chamfer. A 0.22 inch length throat is provided for minimum hot engagement.

The undercut should accommodate bent alignment pins at reduced interference forces.

The chrome plating sho.uld reduce engagement forces. The slightly wider entrance chamfer should help proper engagement of a bent pin. These changes should help reduce the change of assembly SAN-S's lift during a UGS lift. The chrome plating thickness provided.in the throat of 0.001 to 0.0025 inches is equal to or greater than that used on guide,tube wear sleeves in C.E. 14_X 14 plants.

"L" Assembly Usage After the failure of a fuel rod in the 1-24 assembly in Cycle 10, and subseque~t fuel inspections, it was concluded that the 1-Haf nium assemblies would need to be replaced for Cycle 11.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specificat~on Changes, Temporary Modifications and Other Documents Root Cause Summary of 1-24 Cycle 10 Failure In determining the root cause of this failure, the following. possibilities were considered:

1.

Damaged during fuel moves during previous cycle(s) 2; Damaged during EOC 9 Ultrasonic Test Inspection

3.

Damaged during fuel moves this refueling.

4; Fuel Failure due to loose spacer grid. *

5.

Fuel failure due to increased PCS flow

6.

Fuel failure due to core barrel vibration

7.

Fuel failure due to manufacturing failure

8.

Fuel failure due to shroud/fuel assembly interface problem Possibilities 1, 2, 3, 4 have been eliminated by inspection results and review of*

logbooks and records. Possibilities 5, 6, 7 are considered unlikely. However,.

possibilities 4, 5~ and 6 may be factors. Possibility 8 is considered the most likely cause of the 1-024 rod failure. Though actual failure cause may not be definitely determined, modifications to the "I" replacements were made to address possibilities 4, 5, 6, and 8.

Loading oversized stainles.s rods in the corners of the replacement assemblies addresses possibility 4, by tightening potentially loose spacer corner cells. Loading stainless rods into the replacement assemblies addresses.possibilities 5, 6, and 8 by placing sacrificial non fuel rods in the vulnerable corner locations of the replacement assemblies.

Use of stainless steel rods will not stop the fretting. of grid spacers as observed on l'-24 assembly. The stainless steel rod placement design is a remedial corrective action.to cover use of the L-bundles in the core shroud corner locations for one for one cycle during cycle 11. For Cycle 11, any fretting of the spacers at the core shroud corners will potentially only affect adjacent stainless steel rods. To date, fuel and reactor internal inspection results have not shown a significantly degrading condition that would not be mitigated by the I-Hafnium replacements.. The stainless steel rod placement design as a stainless steel rod barrier. from guide bar to guide bar for the

~ssembly corner that faces the core shroud-corner.

.CONSUMERS POWER COMPANY

~ PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The selection criteria that were used -for the I-Hafnium replacement fuel assemblies are:

a.

Burnup < 37,500, MWD/MTU @ EOC 11 for J and K assemblies.

b.

Previously not ori one of the eight octant symmetric corner shroud positions that assembly 1-024 was in when it failed.

c.

Consider bundle bow to avoid largely bowed assemblies.

d.

Consider the advantages the "L" series bundles have in spacer plate stamping.

orientation and leak free status.

e.

Flu-ence rate values should not exceed Cycle 9 values.

The "L" assembly modifications are as follows:

. a.

Eight (8) stainless steel rods are replacing fueled rods in all 16 L assemblies in the

assembly corner that will be at the core shroud corner during Cycle 11 operation..

Thus, fuel rod failures *similar 'to that which occurred in assembly 1-024 during Cycle 10 will be avoided. Also, should spacer fretting occur similarly to what happened to 1-024, the corner stainless steel rod will remain in place captured by spacers 1 through 5 and 10.

b.

For the other three assembly corners, one stainless steel rod of 0.437 inch outer

.diameter (00) has replaced the one 0.417 irich fueled rod in each corner. This modification was deemed prudent to address fuel rod fretting that was seen.

during inspections of four and five times burned fuel assemblies. An additional

stainless steel rod in each of the three c-orners, for* a total of two stainless steel rods in each corner, was used to replace fueled rods to provide additional conservatism. The two stainless steel rods in these corners act to place a symmetric force on the adjacent fuel rods through the lantern spring. The stainless steel rods with their larger OD tighten up all four rods in each corner since they share a common lantern spring. This is supported by data taken on withdrawal forces of surrounding rods both before and after the stainless steel rod replacements in these corners (see item c below). Additionally, the stainless steel rod will not fret during operation like a zircaloy clad fuel rod. Therefore, any subsequent loosening of the four rods in any one corner due to an initial fretting of the very corner rod will be avoided. The greater stiffness of the stainless steel

rods-and their projected minimal wear will reduce any vibration response amplitude and thereby reduce any mechanically or hydraulically transmitted motion to adjacent fuel rods.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facil itY Changes, Speci fi cation Changes, Temporary Modifications and Other Documents

c.

The effect on the fueled rods that share a lantern spring with two larger diameter stainless steel rods has been evaluated by using design documents, testing on archive spacers, as well as withdrawal force data collected during the recent L-bundle reconstitution effort. The primary effect is to displace the lantern spring

toward the neighboring f u*eled rods providing an estimated spring force increase of 0.6 lbs. The resulting increased spring forces are well below the maximum beginning of life design spring force.

Several fuel assembly position changes have been made to the original planned Cycle 11 core design to keep peaking factors and reactor vessel fast neutron fluence within limits.

Thermal Hydraulic Design Fourteen oversized (0.437-inch O.D. for a fuel rod) stainless steel rods will replace fuel

. rods in the sixteen replacement L assemblies. The primary hydraulic impact of these rods will be to reduce the total flow in these L assemblies by reducing the flow area..

Because the resulting reduction in the total flow area of the core is insignificant, the flow through all the subchannels in the core without stainless steel rods.on their boundaries will be essentially unaffected. The flow in subchannels that contain one or more of the stainless steel rods will be less than that of the "intact" flow subchannel.

The two rods sizes will cause a change in the internal flow distribution within each of the-L assemblies but will not affect the inter-assembly cross flows between the L assemblies and the rest of the core. The flow distribution within the L assemblies will develop in the region just downstream of the entrance to the assembly and will be consistent with the two different rod sizes and flow subchannels. There will be no significant intra-assembly cross flows* within the L assemblies because of _the two different rod sizes and this distribution will remain nearly unchanged axially in the L.

assemblies.

. Concerning DNB margins, the flow area affected by the replacement of fuel rods by stainless steel rods represents a very small fraction of the full flow area of the core and the impact will be a_lmost undetectable. Small though it may be, the core-wide hydraulic impact of the stainless steel rods will be positive; The reduction in flow area

. at the periphery of the core will force a minute increase in flow in the interior where the limiting assemblies reside. This increase in flow increases the core pressure drop with a minute increase in the calculated MDNBR. The DNB margin of the L assemblies themselves is not a concern because the power in the fuel rods in these assemblies is such that the L assemblies cannot become DNB limiting assemblies and the subchannel flows of the "intact" subchannels are slightly greater than those with the stainless steel rods.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Standard design fuel assemblies (Bimetallic spacers) have a lower pre'ssure drop for a given flow rate due to their smaller, less restrictive spacer grid design than the pressure drop expected fo*r an assembly with a High Thermal Performance (HTP) design spacer.

A concern has been raised that with an HTP assembly installed adjacently to a standard assembly, flow will tend to be channeled away from the HTP assembly and flow

. crosswise to the standard assembly. This assembly crossflow could result in increased rod fretting and axial loading beyond the design capability of the assembly. The following paragraphs provide a qualitative discussion of the impact on localized flow.

conditions with an HTP design fuel assembly installed adjacently to a standard design fuel assembly in location B-19. SPC has determined that the flow resistance is 5%

higher at each spacer grid in the HTP assembly design. Other flow losses along with the spans between the spacers and through the upper and lower tie plates are the same as the standard design. Consequently; there will only be a localized flow differential in

.areas immediately adjacent to spacer grids.

The B-1 9 core location is in a core shroud corner with the other two sides facing a control rod gap. If there are two HTP assemblies adjacent to a standard assembly in core location B-19, *it is possible that the flow will redistribute in the vicinity of the spacers. However, with the wide Water gap associated with the control rod gap, the flow will preferentially redistribute into the gap (not the standard assembly) and there will be reduced crossflow into the standard assembly.

Crossflow induced fretting damage to fuel rods related to the design of the fuel assembly is usually worse at the inlet or the outlet of.the fuel assembly. The 'wear observed on fuel assemblies at Palisades is not close to either the inlet' or the outlet. It is therefore concluded that the fuel rod fretting caused by cross flows related to fuel design is not indicated and, consequently, it is not considered a root cause. The root cause for the wear phenomenon appears to be a local condition outside fuel design considerations.

The impact of the increased. core flow after the steam generator replacement on the flow-induced vibration and spacer fretting was evaluated. With the increased core,

flow, the projected end of life restraint is greater than the projected vibration force.

Siemens' correspondence (ENH:93:029) provides detailed discussion on the restraint force design and flow-induced vibration.

Reconstitution Considerations Siemens prepared a generic report, ANF-90-0082, entitled "Application of ANF Design Methodology, for Fuel Assembly Reconstitution," which summarized the application of NRC approved design methods to justify insertion of irradiated fuel assemblies, which have been "reconstituted" by replacing fuel rods with inert rods, into a* reactor. This document is generic in that it describes and justifies the criteria required to qualify such a reconstituted assembly for continued irradiation. The reconstitution performed on the L assemblies for Cy~le 11 is not completely consistent with that described in the generic report.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT.

Facility Changes, Specification Changes, Temporary Modifications and Other Documents Specifically, the generic report identifies replacement *rods as inert rods, filled with zircaloy or stainless steel, or water rods. These two types of rods were specified in the generic report as Siemens typically uses this type of replacement rod in the repair of fuel assemblies with failed fuel rods. Fuel assemblies that have been reconstituted according to the generic report can be used in any core location. The analytical methodology described in the generic report called for an evaluation of the impact of the reconstitution in the following areas.

Neutronics Power Distributions -local peaking, axial shapes, etc..

Kinetic Parameters Control Rod Reactivity Core Monitoring Coefficients Thermal Hydraulics Hydraulic Compatibility Thermal Performance (DNBR)

Fuel Centerline Temperatures Rod Box LOCA Rod Ejection Accident Mechanical Design Differential Rod Growth and Assembly Growth Assembly Liftoff Linear Heat Generation Limits The Palisades Reload L assemblies are being reconstituted use fourteen s_olid stainless steel rods that are approximately twenty mils larger in diameter than a standard fuel rod (0.437-inches O.D. versus 0.417-inches O.D.). The inconsistency with this reconstitution* and the generic report is that the generic report and the reconstituted L assemblies have a difference in the number of adjacent inert rods per subchannel.

However, the generic report identifies the evaluations required to support the use of a reconstituted fuel assembly (see above). Safety implications of the reconstitution of the Reload L assemblies have been reevaluated in a safety evaluation that is specific for proposed operation of Cycle 11 and a revised Safety Analysis Report (SAR) was prepared. The SAR includes explicit modeling of the fourteen replacement stainless steel rods inserted into the sixteen L assemblies.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The principal impacts of the insertion of the stainless steel rods in the L assemblies are the changes that occur in the local power and flow distributions. These are discussed in Thermal Hydraulic Design, above. The reconstituted L assemblies were also explicitly treated in the generation of the core monitoring coefficients.. The impact that these reconstituted Reload L assemblies will have on the monitoring of assembly power and power peaking is therefore accounted for by these coefficients~

The use of the fourteen stainless steel rods per "L" assembly for Palisades does not degrade the safety margins at Palisades because of the planned location of the

. reconstituted assembly in the core. The inert rods are in the corners of very low

( s... 20%) power assemblies located next to the core baffle. DNB is *not a concern in these very low power locations. Flow distribution is also not a concern because the inert rods are located far from DNB limiting fuel assemblies.

The mechanical impact will tend to be channeled away from the HTP assembly and flow crosswise to the standard assembly. This assembly crossf low could result in

.increased rod fretting and axial loading beyond the design capability of the assembly.

The following paragraphs provide a qualitative discussion of the impact on localized.

flow conditions with an HTP design fuel assembly installed adjacent to a standard design fuel assembly in location B-19. SPC has found that the flow resistance is 5%

higher at each spacer grid in the HTP assembly design. Other flow losses along with the spans between the spacers and through the upper and lower tie plates are the same as the standard design. Consequently, there will only be a localized flow differential in areas immediately adjacent to spacer grids.

The B-19 core location is in a core shroud corner with the other two sides facing a

control rod gap. If there are two HTP assemblies adjacent to a standard assembly in core location B-19, it is possible that the flow will redistribute near the spacers.

However, with the wide water gap associated with the control rod gap, the flow will preferentially redistribute into the gap (not the standard assembly) and there will be reduced crossflow into the standard assembly.

Crossflow induced *fretting damage to fuel rods related to the design of the fuel assembly is usually worse at the inlet or the outlet of the fuei assembly. The wear observed on fuel assemblies at Palisades is not close to either the inlet or the outlet. It is therefore concluded that the fuel rod fretting caused by cross flows related to fuel design is not indicated and, consequently, it is not considered a root cause. The root cause for the wear phenomenon appears to be a local condition outside fuel design considerations.

.. The impact of the increased core flow after the steam generator replacement on the flow-induced vibration and spacer fretting was evaluated. With the increased core flow, the projected end of life restraint is greater than the projected vibration force.

Siemens' correspondence (ENH:93:029) provides detailed discussion on the restraint force design and flow-induced vibration.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Reconstitution Considerations Siemens prepared a generic report, ANF-90-0082, entitled "Application of ANF Design Methodology, for Fuel Assembly Reconstitution," which summarized the application of NRC approved design methods to justify insertion of irradiated fuel assemblies, which have been "reconstituted" by replacing fuel rods.with inert rods, into a reactor. This document is generic in that it describes and justifies the criteria required to qualify such a reconstituted assembly for continued irradiation. The reconstitution performed on the L ~ssemblies for Cycle 11 is not completely consistent with that described in the generic report.

Specifically, the generic report identifies replacement. rods as inert rods, filled wi~h zircaloy or stainless steel, or water rods. These two types of rods were specified in the generic repo_rt as Siemens typically uses this type of replacement rod in the repair of fuel assemblies with tailed fuel rods. Fuel assemblies that have been reconstituted according to the generic report can be used in any core location. The analytical methodology described in the generic report, called for an evaluation of the impact of

the reconstitution in the following areas.

Neutronics Power Distributions - local peaking, a_xial shapes, etc.

Kinetic Parameters Control Rod Reactivity Core Monitoring CoettiC:ients Thermal Hydraulics Hydraulic Compatibility Thermal Performance (DNBR)

Fuel Centerline Temperatures

Rod Box LOCA Rod Ejection Accident Mechanical Design Differential Rod Growth and Assembly Growth Assembly Liftoff Linear Heat Generation Limits The Palisades Reload L assemblies are being reconstituted use fourteen solid stainless steel rods that are approximately twenty mils larger iri diameter than a standard fuel rod (0.437-inches O.D. versus 0.417-inches O.D.). The inconsistency with this reconstitution and the generic report is that the generic report and the reconstituted L assemblies have a difference in the number of adjacent inert rods per subchannel. *.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents However, the generic report identifies the evaluations required to support the use of a reconstituted fuel assembly (see above). The safety implications of the reconstitution of the Reload L assemblies have been reevaluated in a safety evaluation that is specific for proposed operation of Cycle 11 and a revised Safety Analysis Report (SAR) was prepared. The SAR includes explicit modeling of the fourteen replacement stainless steel rods inserted into the sixteen L assemblies.*

The principal impacts of the insertion of the stainless steel rods in the L assemblies are the changes that occur in the local power and flow distributions. These are discussed

in Thermal Hydraulic Design, above. The reconstituted L assemblies were also explicitly treated in the generation of the core monitoring coefficients. The impact that these reconstituted Reload L assemblies will have on the monitoring of assembly power and power peaking is therefore accounted for by these coefficients.

The use of the*fourteen stainless steel rods per "L" assembly for Pal.isades does not degrade the safety margins at Palisades because of the planned location.of the reconstituted assembly in the core. The inert rods are in the corners of very low *

(""" 20%) power assemblies located next to the core baffle. DNB is not a concern in these very low power locations. Flow distribution is also not a concern because the inert rods are located far from DNB limiting fuel assemblies.

The mechanical impact of inserting larger and stiffer stainless steel rods in the fuel assembly has also been addressed for Cycle 11 operation. Stainless steel rods are

. about 0.2 lbs heavier than the fuel rods they replace due to their larger dia.meter. This increased weight has a positive effect on assembly holddown. The presence of oversized stainless steel rods tends to make the L assemblies more rigid and less prone to vibration. The potential for fretting of neighboring fuel rods by the stainless steel rods should thus be lower than in an L assembly without stainless steel rods.

CPCo has experience with the use of these p*articular rods, as they are being taken from a selection in the "H" reload neutron shield assemblies used during Cycle 8. The use of the "H" assemblies in Cycle 8 was addressed as part of the SPC reload methodology similar to the analytical methodology described in the generic report.. In addition, stainless steel partial shield assemblies used in Cycle 10 (which will receive their second burnup in Cyde 11) have been addressed under the SPC reload methodology in a similar manner. However, it is noted that the outside diameter of the stainless steel rods used in the "N" partial shield assemblies is the same as the fuel rod.

As a point of further clarification, the following discussion is presented to address the general subject of Generic Letter 90-02.and its supplement.

1 OCFR50.59 allows licensees to make changes to the plant without prior NRC approval if the changes: a) do not involve an unreviewed safety question and, b) do not involve a Technical Specification (TS) change.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents

Since the Palisades TS do not specify the number of fuel rods in a fuel assembly, replacement of fuel rods with non-fuel rods does not conflict with the Palisades TS.

Therefore,*such replacements may be completed without an unreviewed safety question.

The NRC has issued a Generic Letter, GL 90-02, which discusses replacement of fuel rods with inert rods or with open water channels. GL 90-02 provides guidance to licensees whose existing TS do not allow such replacement. This GL suggests changes that may be proposed to TS that would amend an explicit fuel assembly description; like that in the former Standard Technical Specifications (STS), to allow such replacement of fuel rods if the associated analyses have been completed using NRC approved methodology. GL 90-02 is similar, in intent, to GL 88-16 that suggests proposing TS changes that relocate specific core operating limits from the TS to a Core Operating Limits Report and add the TS requirement that these limits be calculated according to NRC approved methodology.

In contrast to the Palisades TS, the former STS contained an explicit description of a fuel assembly in Section 5.3. (See the former CE STS, NUREG 0212)

Since this description stated the number of fuel rods in a fuel assembly; fuel rods could not be replaced with non-fuel rods without an amendment to the TS and the associated NRC review. GL 9-02 provides relaxed TS wording which would allow replacement of fuel.

rods with certain non-fuel rods or with open water channels without prior NRC approval "if justified by cycle-~pecific reload analyses using NRC-approved methodology." *

The NRC issued Supplement 1 to GL 9-02 on July 31, 1992. The stated purpose of that supplement is to "clarify the limitations on the application of currently NRC-approved analytical methods and to _withdraw and replace the model technical specifications (TS) which are recommended by GL 90-02... "

The supplement states, in part, "licensees are not required by this supplement, or by

GL 90-02, to change their TS." Since Palisades has not chosen to submit a TS change request based on GL 90-02 or the supplement, those sections of the supplement discussing TS changes that emulate the m*odel TS of GL 90-02 and its supplement, and the interpretations of limits therein, do not apply to Palisades.

"L" Assembly Bow As part of the Cycle 11.core design, the Batch L assemblies selected were evaluated for assembly bow. All sixteen assemblies will be oriented with their burn up gradient away from the shroud to minimize the possibility of contact between these assemblies and the shroud. Eight of these assemblies will be pla.ced next to control blades and.

their burnup gradients will be toward the control blades. Since burnup gradient

_correlates positively to fluence gradients that correlate positively to fuel assembly bow, burnup gradients have been evaluated to determine acceptable assembly bow.

-CONSUMERS_POWER COMPANY - PALISADES NUCLEAR PLANT

Facility Ch~nges, Specification Changes, Temporary Modification£ and Other Documents Expected burnup gradients at the end of Cycle 11 have been calculated for the L assemblies, which will be used on the periphery. The maximum projected gradient calculated was approximately 6,400 MWD/MTU. This maximum was for an assembly

,which was not next to a control blade. The maximum projected burnup gradient toward a control blade is about 4,600 MWD/MTu.*

During operation of Palisades, it has been quite common to have burnup gradients more than 9,500 MWD/MTU not oriented toward a control blade and 6,000 MWD/MTU toward a control blade. In some cases, the burnup gradients have exceeded 7,000 MWD/MTU toward a control blade. Palisades has not*experien_ced any difficulties related to the operation of the control blades or to contact between the fuel assemblies and the control blades because of such gradients. Given the projected burnup gradients for the L assemblies, past operating history for Siemens Power Corporation fuel at Pe1lisades bounds Cycle 11. The assembly bow for the L assemblies in Cycle 11, based on the conservative nature of the burnup gradients in the L assemblies, is also expected to be bounded by past operating history for Siemens fuel at Palisades.

The consequences of any malfunction to equipment important to safety are bounded by the analyses described in the report referenced in Part 2.

Relocating and replacing fuel assemblies does not increase the possibility of an accident*

of a different type than _evaluated in the FSAR. The function and interfaces of the core remain the same.

The change to Technical Specification Table 3.23-2 increases the assembly and total radial peaking factor limits for Cycle 11 (Reload 0 fuel assemblies). This change is in core neutronics parameters due to changes in the fuel design and fuel management scheme. No changes to plant hardware (other than the new fuel and the 16 modified "L" bundles) are involved. There-are no associated changes in plant systems, operating procedures, or in instrument alarm or trip settings. Therefore, operation of the facility following the revised Technical Specifications would not create the possibility of a new or different kind of accident from any previously evaluated.

Refueling does not introduce the possibility of a malfunction different from any previously evaluated in the FSAR. The mechanical design of the ~ore is unchanged.

The loading of Core 11 does not affect reactor system hardware or operating procedure.

The margin of safety as defined by Plant Licensing bases will not be reduced by loading Core 11. Operation of Core 11 as intended was dependent upon successful completion

.. of the acc\\dent analysis and approval of the Technical Specification change submittals.

This approval was received June 16, 1993 in Amendment #156.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents FC-940 I Rev 1 Modification 3 SE93-0172 DIESEL GENERATOR BREAKERS AND PROTECTIVE TRIP LOGIC UPGRADE - AUTOMATIC TRANSFER FROM PARALLEL TO UNIT All Emergency Diesel Generator systems operate in one of two modes, Isochronous

("Unit") or Droop ("Parallel").

By design the default mode of the control system is Isochronous or "Unit," which is the desired mode of operation to supply the bus under emergency conditions. The Droop mode ("Parallel") is only used for periodic testing of the EOG. A brief discussion of both methods of operation is provided to help the reader in understanding the changes proposed by this modification. It is important to understand that the mode of operation selected is not electrically interlocked with the AC source breaker operating controls but determined by the AC source breaker position. Voltage regulators are used to control the flow of reactive power.

When the current lags on the voltage waveform, it transmits less power then when it is in phase. The apparent power is found by multiplying voltage by current, but the real power is less than this by the power-factor percentage. It is convenient to divide the apparent power into two vector components - real power and reactive power. Both components are important in power generation and must be metered and controlled separately. The power factor is a way of representing the extent to which alternating current drawn by the plant is out of phase with the voltage. It is expressed as the ratio (or percentage) of real power (watts) to apparent power (volts x amperes). Most industrial users draw a lagging current and have a power factor somewhere between 70 and 90 percent. The concept of power factor becomes less useful when in plant generation is involved because it is difficult to tell whether the power factor is leading or lagging. That is why VARS are typically monitored. Reactive current is that component of total current that is 90 degrees out of phase with the voltage. Reactive power is expressed in volts-amperes reactive (VARS). The power output of a generator

is controlled by varying the torque applied to its shaft by the engine. The VARS output is controlled by varying the field excitation of the generator. This is done automatically by the voltage regulator. A potential transformer produces a signal proportional to the voltage, and the regulator adjusts the excitation to a level at which the generator would produce the same voltage if it were operating in isolation. This excitation level is then trimmed to control the flow of VARS from the generator.

DROOP (PARALLEL) MODE Droop is a decrease in speed (frequency) or voltage proportional to the applied load.

That is as the load increases, the speed (frequency) or voltage decreases. Droop is expressed as a percent that the speed or voltage drops below no load speed or voltage when the system is fully loaded. When the generator is connected to the utility bus, the utility bus determines frequency. Speed Droop allows the operator to use the raise/lower speed control to vary the load since the speed cannot change.When

°CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents operating in parallel an additional circuit is required in the voltage regulating system to allow the paralleled generator to share reactive load and reduce circulating reactive currents between the paralleled AC sources. An auxiliary contact on the AC source breaker used to connect the isolated bus to the utility bus is typically used to toggle the additional circuit into the loop only when breaker position requires the parallel mode of operation. As shown by Figure B when the generator is supplying the bus while in the Droop mode the bus fre.quency will equalize to a lower steady state value following a load transiE;mt. While Figure B shows that the start point is 60 HZ before the transient, the real starting point will be whatever the.raise/lower potentiometer was last set at while controlling the load. In conclusion, when operating in the droop mode the load on the EDG is determined by the reference speed setting of the droop governor.

ISOCHRONOUS MODE (UNIT) MODE

Isochronous means repeating at a single rate or having a fixed frequency or voltage.

For the sake of these discussions it means the EDG will maintain an output of a constant 60 cycles (HZ) and a constant 2400V regardless of the load it is supplying, up to the full load capabilities of the engine-generator set.

PROPOSED MODIFICATION For Palisades Plant the default mode of the electronic control system is isochronous or "Unit", which is the desired mode _of operation to supply the bus under emergency conditions.. The existing system has a switch to allow the operator direct control of the operating mode of the controller system. The"Unit/Parallel" Switch (DSR) is used to place the generator in either "Unit" or "Parallel" mode of operation. In "Parallel" the DSR actuates the Droop Relay (DR) which in turn provides the signals to the governor control system and the voltage regulator control.in "Parallel" op_eration. In the Unit mode _of operation (Isochronous) the diesel generator is the sole source of AC power to the vital bus and the Droop Relay is de-energized.

IEEE 387-1984 and Reg. Guide 1.9 requires an automatic actuation of the diesel generator controls when emergency conditions warrant it. This feature does not currently exist in the control system, the operator is procedurally required to manually position the selector switch*to the proper position following an event. The standard i~dustry practice for automatic transfer of the droop/isochronous mode is to use the AC source breaker position. Palisades is equipped with a fast transfer system to provide alternate AC sources before requiring the diesel generators to function, thereby requiring a combination of AC breaker positions to ensure correct mode selection.

Since the interlocks are breaker position interlocks they are present whether the breaker closure is by manual or automatic means. The Automatic transfer from Parallel to Unit will be achieved by using the breaker position inputs from the two AC feeder breakers in series with the EDG breaker position to ensure the electronic speed control is in the parallel mode when the* EDG is operated with another AC source (See Figure 3-1).

Currently while operating in either mode there is no direct indication of either power factor or KVARS in the control room to help the operator in accurately assessing the CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents reactive load on the EOG. The "Unit/Parallel" switch will be disconnected from the control circuit and removed from the control panel. Flat head screws will be installed in the vacated mounting holes on the control panel. This modification will also require new cables to be routed from the switchgear rooms to the Diesel Engine Control Panels. Currently there a~e both local and remote status lights associated With the Unit/ Parallel transfer switch. At the Conceptual Design Review Meeting the decision was *made to maintain the Unit/Parallel indicating lights.

Note: According to DBD 5.06 section 3.3.2 both the Droop Relay (DR) and the Droop Switch Remote (DSR) are non-class 1 E but are directly connected to class 1 E control circuits. This modification resolves the non 1 E issue of the DSR, in that it will be removed from the control panel. The issue of using a non-1 E Droop Relay is not addressed by this modification.

Safety Analysis Summary Fr9m the above detailed description of the proposed modification, the "Unit/Parallel" switch (DSR) used to place the generator in either "Unit" or "Parallel" mode of

operation will be. eliminated, and replaced with automatic actuation/transfer from Parallel mode to Unit mode of operation. This will enhance the system design and increase reliability by eliminating any human error associated with having the operator direct control of the operating mode of the controller system. Presently, the operator is procedurally required to manually position the sele.ctor switch to the proper position following an event. In addition this modification will align operations with IEEE 387-1974 and Reg. Guide 1,9 requiring an automatic actuation of the Unit (Isochronous) when emergency conditions warrant it. Currently, DBD 5.06 Section 3.3.2 states the Droop Switch Remote (DSR) is non-class 1 E directly connected to class 1 E control circuits. This modification resolves the non 1 E issue of the (DSR) by eliminating it from the system.* From the above analysis it can be seen that this modification. wlll not increase the probability of occurrence of an accident or the consequences of an ac.cident which was previously analyzed in the FSAR. Nor does it create the possibility of an accident not analyzed in the FSAR.

Based on the* above evaluation this modification is considered an enhancement increasing system reliability. The probability of malfunctions of equipment important to safety is not increased nor is the possibility of a malfunction of a different type other than preciously analyzed in the FSAR been cre~ted. The consequences of a

malfunction remain unchanged.

Based on the above analysis and after thorough review of the plants' Tech Specs and plant licensing bases this modification will not reduce the margin of safety as defined or implied by the said bases and thus an unreviewed safety question does not exist.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents*

FC-940 I Rev 1 Modification No 5 SF93-0J Z3 DIESEL GENERATOR BREAKERS AND PROTECTIVE TRIP LOGIC UPGRADE - UPGRADE GENERATOR OVER CURRENT LOGIC

The present 1 out of 2 Diesel Generator (DG) over current logic provides adequate DG protection, but does not provide coincident logic as outlined in IEEE Std. 387-1984 and

. Reg. Guide 1.9.

Presently the over current trip circuit is an on-line trip function. Three inverse time over

current relays are connected on the generator side of the EG 2400 V breaker to protect the EG. circuit from long-term overloads. Each phase of the generator is monitored for current. Over current on the "Z" phase ~elay results only in an over current alarm at 100% of the generator nameplate_ loading. Over current on the "X" or "Y" phase relays (1 out of 2 logic) at 110% of the generator nameplate rating, results in actuating the same 186 lockout trip relay as the generator differential trip feature, resulting in both _an engine and breaker trip. The present 1 out of 2 logic, although adequate for diesel generator protection, does not provide coincidence logic as outlined in IEEE 387-1984 and Reg Guide 1.9 for diesel generator operation during an accident condition.

This modification will revise the existing 1 out of 2 over current trip logic to become a coincidence trip logic. The new trip logic will provide an engine/bre_aker trip when any two of the three phases sense over current. The necessary contact cont iguration required to achieve the coincidenc.e trip logic is presently not available. It will be

obtained by installing three new Class 1 E auxiliary relays for the coincidence logic and a new over current relay in phase "Z" for the overload alarm. No changes to the Main*

Control Room annunciator logic will be required. The auxiliary relays together with the new over current relay required to implement the design modification will be specified in the detailed design phase of the modification.

Presently, DBD-5.02, Section 3.3.4.2 indicates that the existing "Z" phase over current relay is set to provide an alarm signal at 100% of the generator nameplate loading.

This modification will keep the alarm on the "Z" phase at 100%. The breaker/engine over current trip signal will remain at 110% of the generator nameplate rating,* based on 3125 KVA at 80% power factor. This will require new tap and time dial settings for the existing phase "Z" over current relay. The new phase "Z" settings should match the existing settings for phases X and Y. The new auxiliary relays and over current alarm relay will be mounted in the upper compartment above the diesel generator breaker cubicle.

The over current relay will be semi-flush mounted like the existing over current relays. A seismic analysis in the form of an EA will be performed for all new equipment mounted in the EOG bre_aker upper compartment.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, ~pecification Changes, Temporary Modifications and Other Documents Safety Analysis Summary This modification will provide the coincident logic outlined in IEEE 387 and comply with Reg. Guide 1.9. The new trip logic will provide an engine/breaker trip when any 2 of the 3 phases sense over current. The DG overload alarm function provided from the Z phase current will be retained. The new trip logic will reduce the possibility of the over current relays causing a nuisance trip of the DG, enhancing the total system design and increasing the reliability oL the electrical supply system. Therefore, this modification will not increase the probability of occurrence of an accident or the consequences of an accident previously evaluated in the FSAR.

Based on the above item description and evaluation this modification is considered an enhancement increasing system reliability reducing the probability of malfunctions of*

equipment important to safety.

As described above, the new trip logic scheme keeps.the over current trip function for the DGs and enhances it by reducing the probability of nuisance tripping. Since the protective function (over current trip) is enhanced and its set point unchanged, this modification does not increase the consequences of a malfunction of equipment.

important to safety. By the same reasoning, no accident of a different type than any previously evaluated in the FSAR is created. Again by the same reasoning, no malfunction of a different type than any previously evaluated in the FSAR is created.

Based on the above discussion, this modification does not reduce the margin of safety as defined or implied by the plant licensing bases and thus it can be co*ncluded that an.

unreviewed safety question does not exist.

FC-940 I Rev 1 Modification No 10 SE93-0743 DIESEL GENERATOR BREAKERS AND PROTECTIVE TRIP LOGIC UPGRADE - REPLACE SOLENOID OPERATED D/G CIRCUIT BREAKERS 152-107 (152-213)

. Presently, due to problems with reliability, maintainability and vendor support, the existing solenoid-operated emergency generator circuit breakers 152-107 ( 152-213)

. will be replaced with new stored energy circuit breakers (Siemans Allis Type MA-

. 250C 1). It should be noted that installation and design of these breakers are based on the fact that the existing breakers (in the warehouse) have been qualified and proper

. documentation to that fact is provided by Consumers Power Company or an outside vendor and retained in the Detailed Design Package for FC-940. A review was performed by ABB lmpell on the vendor's test report to assess the applicability of the report for the installation of the Siemens Allis Stored Energy Breakers that have been purchased by Consumers Power Com.pany (CPCo). Based on the review, it was concluded that the testing on the similar breakers does not satisfactorily meet the requirements of IEEE 344-75 and CPCo Technical Specification C-175 (0). Presently, CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specificatiori Changes, Temporary Modifications and Other.Documents

. the Palisades Plant is trying to qu_alify the Medium Voltage Switchgear Breakers through

the Seismic Qua I if ication Utility Group (SQUG) Program with the Generic Implementation Plan (GIP) criteria. Final determination will be made during the detail design phase of the modification and the NRC commission will be notified in writing.

stating that the SOUG program will pr.oceed. In addition, this modification will also

. install one indicating light (white) on each associated switchgear and two indicating lights (white) on the DIG Control Panel in the Main Control Room to monitor the status of the charging springs on the new breakers (See Fig. 10-1 and 10-2). The status indicating lights will be purchased 0 for safety related use. It is anticipated that the existing fuse holders in the switchgear will be reused to accommodate the new fuse sizes for the charging spring motor and indicating lights. The relative weights of the new indicating lights used at the station will not compromise the seismic mounting of the breaker. The surge currents associated with the solenoid breaker coil and the stored energy charging motor of the replacement breaker evaluated against the existing surge currents proved to be less than that of the existing breaker and therefore will have a minimal impact on the existing Class 1 E 125V DC Load System. Presently, (if available spares are not identified) new control cables (2) will be required from their associated switchgear to the new indicating lights in the Main Control Room. All cable routing shall be per Palisades Circuit and Raceway Schedules during the Detailed Design Phase. In addition, all secondary installations shall be seismically mounted per IEEE-344 for mounting interaction but not fqr function. The breakers shall be seismically installed per IEEE-344 for mounting interaction as well as for function and a seismic analysis performed in a Supporting Engineering Analysis.

All panel modifications will be covered in installation procedures provi"ded during the detailed design.

In a conversation with Siemans Allis on 5/27 /92, it was confirmed that the new stored energy breakers type MA-250C 1 (presently in stock in the Palisades Warehouse) are.

interchangeable with the old solenoid operated breakers type MA-2508. They stated these breakers were a "Like-for-Like" replacement and would not require any physical switchgear alterations to accept the new breakers.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The associated anti-pump circuitry is designed to prevent cycling of the circuit breaker between the closed and tripped (open) positions when both automatic closure and automatic trip signals exist concurrently. The anti-pump circuitry prevents.repeated attempts to close the breaker under valid trip (fault) conditions. In the Palisades evaluations for IN 88-75 and Supplemen_t 1, the acceptability of the Pal.isades E/G Breaker Control Design was based on the fact that these breakers are solenoid-operated, and therefore are not subject to the logic/timing problems arising from spring-charging delays experience in stored energy breakers. Because of replacing these breakers with Stored Energy Breakers the basis for acceptability of the existing E/G Breaker control design will no longer be valid. The new design will include the concerri of these information notices and will be used.as design* input in the detailed design phase of the modification. Now it is anticipated that the new breakers will not be subject to logic/timing problems rising from spring-charging delays as the charging cycle is minimal.

Safety Analysis Summary By replacing the old outdated solenoid operated E/G Circuit Breakers 152-107 (152-213)_ with new stored energy circuit breakers this will eliminate the existing problems of reliability, maintainability and vendor support thus, enhancing the overall system design. In addition, status indicating lights will be installed to monitor the status of the charging springs on.the new breakers increasing operator system awareness. This modification is a "Like-for-Like" replacement and will not require any physical switchgear alterations to accept the new breakers.

Seismic Analysis will pe performed for the status indicating lights so as not to* compromise the seismic mounting of the breakers or switchgear. The surge currents associated with the solenoid breaker coil and charging motor of the replacement breakers proved to b~ less than that of the existing breaker. Therefore, it will have a nominal impact on the existing Class 1 E 125VDC Load System. Overall this modification is considered an enhancement transparent to operations. Therefore, this modification will not increase the probability of occurrence of an accident or the consequences of an accident which was previously analyzed in the FSAR. Nor does it create the possibility of an accident not analyzed in the FSAR.

Based on the above evaluation this modification is considered an enhancement increasing system reliability/maintainability reducing the chances of a malfunction. The probability of a malfunction of equipment important to safety is not increased nor is the possibility of a malfunction of a different type other than previously analyzed in the FSAR been created.

Based on the above discussion this modification does not reduce the margin of safety

as defined or implied by the plant licensing bases and thus an unreviewed safety question does not exist.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility. Changes, Specification Changes~

Temporary Modifications and Other Documents FC-944 SE93-0095

.ASDV'S BACKUP NITROGEN SUPPLY

_ This modification will add a line connecting the bulk nitrogen system to the ASDV's (Atmospheric Steam Dump Valves) instrument air supply. This will allow the ASDV's to be operated during a station blackout (SBO) since the compressed air system will be inoperable. Palisades must withstand a SBO for four hours. Regulatory Guide 1.155 provides guidance in the SBO issue.

The bulk nitrogen tank provides nitrogen at approximately 250 psig. This pressure will be reduced to the required pressure for the ASDV's via the new pressure control valve.

Operation's rounds sheets state that minimum liquid nitrogen height in the bulk nitrogen tanks is approximately 2.5 gallons of liquid nitrogen, an amount insignificant when compared to the 717 gallons available. The liquid nitrogen supply is therefore more than adequate.

The backup nitrogen will simply provide a means of operating the ASDV's during an SBO and is not necessary for the normal operation. Check valves will be placed in the system to prevent the nitrogen from interfering with the normal operation of the air supply. The modification will be installed using an approved design and approved procedures.

Safety Analysis Summary Per review of the FSAR Chapter 14 Analyses as shown in the safety review, connecting the ASDV air supply to the bulk nitrogen system could not initiate an accident either directly or indirectly. Therefore the modification could not increase accident probability. An approved design and approved procedures are being used to install the modification.

Per review of the FSAR Chapter 14 Analyses as shown in the safety review, connecting the ASDV air supply to the bulk nitrogen system could not increase the consequences of a previously evaluated accident. The Design uses check valves so that one system does. not interfere with the other's function.

Tying the air line and nitrogen line together could not increase the consequences of a malfunction of equipment important to safety. The proposed modification will not change, degrade, or prevent actions for which credit is taken or implied in the FSAR.

The modification will simply provide a means of operating the ASDV's during a SBO.

Using an analyzed and approved design for the modification will ensure that the possibility of an accident of a different type than any previously evaluated in the FSAR will not be created.

- CO~SUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and-Other Documents This modification will simply enhance the operation of the ASDV's dui"ing station blackout, and will not affect any systems normal operation. Using an analyzed and approved design for the modification will ensure that the possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created.

The system is designed to perform the same function as before the modification. The added benefit being nitrogen supply to the ASDV's during a SBO. The change will have no direct or indirect effect on the fission product boundaries or on accident consequences.

FC-946 -

SE93-0297 BORONOMETER REMOVAL The boronometer (AE-203) is no longer necessary for plant operation and has been retired in place since 1988 (Reference SC-87-384). This modification is to remove the boronometer and associated piping, electrical cable and instruments from the Auxiliary Building.

The boronometer contains an internal radiation source that contributes to dose when entering the 602 pipeways due to its proximity to the ladder leading to 602' elevation.

Removing the nonfunctional boron_ometer will eliminate several potential.leak paths

-from the Chemical and Volume Control_ System's (CVCS) letdown side. The common piping to the fail fuel -monitor (RE-202) will be rerouted out of the way, to improve accessibility to 602 pipeway access ladder and to alleviate tripping hazards over the 1 /2" pipe in the middle of the room.

This activity will involve rerouting the 1 /2" pipe by the wall at the same elevation.

Stainless steel pipe will be used to replace the existing pipe, according to piping class sheet M-260 Sh HC. Some existing valves will.be replaced or reused, based on the level of existing contamination inside the valve body. The connections to the existing pipe will be flanged to simplify the removal of the pipe section to flush contaminants.

Chemistry department will determine the neutron absorption *in the coolant by sampling and analysis of primary coolant water per Tech Spec requirements Table 4.2. 1 Item 2.

Safety Analysis Summary Boronometer AE-203 has been out of service since 1988, other reliable means were established to measure the boron concentration in the primary coolant water. Those sampling techniques are proven to be reliable, dependable and meet the Tech Spec requirement Table 4.2.1 for min frequency of sampling. Therefore, removing an abandoned piece of equipment out of the Auxiliary Building will not increase the probability of an accident.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents For the same reasons described above, since the pipe replacement is equivalent to the existing, and the installation is controlled by Palisades approved. procedures and code requirements, the consequences of an accident previously evaluated in the FSAR will not be increased. Refer to FSAR Section 14.23 and DBD 1.04 Table 4.2, "Radiological Consequence of Failure of Small Lines Carrying Primary Coolant Out.side Containment."

The existing boronometer AE-203 hasn't functioned since 1988; therefore', the

probability of malfunction of equipment important to safety will not increase. On the contrary, removing this piece of equipment from its location will eliminate several potential leak paths from the eves system and" will reduce the general dose level in Room 121 in the Aux;iliary Building.

Based on the same ~easoning as number 3, the consequences of malfunction of this equipment are eliminated by removing it from the plant. Therefore, a malfunction of equipment important to safety will not be increased.*

No different type of accident is possible since the safety related piping to be rerouted is equivalent to existing pipe. Replacing the pipe and valves will be controlled by plant approved procedures. Construction practices will be followed.

No different type of maifunction is possible for the same reasons determined above.

Plant Licensing Basis is being met without having the boron6meter in operation.

Therefore, this abandoned piece of equipment will not affect the margins of safety.

previously defined. Removing this from the Auxiliary Building will reduce the dose contribution in the general area. FSAR, DBD-1.04, and operating procedure, applicable sections will be revised to reflect the change.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents FC-947 SE93-0666 OFFICE FACILITIES EXPANSION A 33,000 square foot building addition will be made to the west side of the existing Service Building. This addition will consist of offices, conference rooms, a lunch room and associate uses. The third floor and north end of the second floor of the existing Service Building will also be improved under this FC package. Minor changes will be made to the first Jloor of the existing building. In addition an emergency exit will be constructed for the offices on the second floor of the Turbine Building. The Service Building addition will be sprinkled and will have fire alarms. The building will be serviced by the 1 O" fire water main located approximately 1 O' west of the existing Service Building. This main dead ends after serving the PMC& T Building, fire hydrant 9 and lO and the existing Service Building. The only Fire Hydrant that FSAR Section 9-,

6. 7.1.4 requires to be operable near this modification is #3 which is located near the

southwest corner of the existing Service Building. Its operability will not be affected by this modification. Temporary fire water service will be provided during construction to provide protection for the areas served by the existing fire main.

Office buildings were not evaluated as a concern for the safety of the plant in the FSAR, unless they served a purpose other than office space. None of the systems that will be affected by this modification are discussed in the Technical Specifications. The sole function of this facility change is to provide office space and related functions.

The facility change will not affect the FSAR other than chang'es to the fire protection system on the P&IDs.

Safety Analysis Summary The modification will not be altering the function of systems or equipment related to the safety of the plant. *Therefore, the probability or consequences of an accident previously evaluated in the FSAR will not be increased.

This modification will not affect any equipment important to safety of the plant.

Therefore, the probability or consequences of any safety related equipment malfunctioning will not be increased and there will be no increase in consequences of a malfunction of safety related equipment.

The only change to the fire protection system is the addition of another sprinkler system to the fire protection system. This addition falls within the design flow of the fire protection system. Thus the margin of safety as defined by Plant Licensing Bases will not be reduced.

- 32

CONSUMERS POWER COMPANY - PALISADES NUCLEAR -PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents FC-949 SE93-0629 CONTAINMENT SUMP pH CONTROL

. This modification requires the installation of Trisodium Phosphate (TSP) baskets in the containment which will ensure that a pH of 7.0 to 8.0 is achieved at the initiation of Recirculation Actuation Signal (RAS). The baskets containing TSP dodecahydrat'e will be placed on the floor or raised slightly above the floor (590') of the containment building. This system is a passive form of pH control for post LOCA containment spray and core cooling water and requires no operator action. The current NaOH system will be retired in place.

Safety Analysis Summary Installation of TSP baskets will not increase the probability of an accident as previously evaluated in the FSAR. The current revision of the Standard Review Plan Section 6.5.2 requires that containment sump pH be above 7.0 at the initiation of Recirculation Actuation Signal (RAS) to ensure iodine is retained in solution. Also, the MHA analysis EA-A-NL-92-012-03 assumes a pH of 7.0 or greater at RAS. Additionally Palisad_es hydrogen generation analysis EA-GFP-92-02 requires a containment sump pH below 8;0. The current system controls containment sump pH by manual addition of sodium hydroxide (NaOH) from tank T-103. T.his could result in up to an eight-hour time delay after RAS before a containment sump pH of 7.0 to 8.0 is reached.

High temperatures and low pH, which would be present after Loss of Coolant Accident (LOCA), tend to promote Stress.Corrosion Cracking (SCC) which could lead to

the failure of necessary safety systems or components. Installation of TSP baskets with the required quantity of TSP as calculated in EA-FC-949-01 will ensure a containment sump pH of 7.0 to 8.0 before RAS and help inhibit sec and reduce the probability of an accident. Since the TSP system is a passive system that performs its design function only after an accident, it will not affect normal operation of the plant.

Hence, the TSP baskets will not increase the probability of an accident.. *

The TSP baskets will be installed in the containment building at a floor elevation of 590'. The baskets will be raised about 6".from the ground to avoid loss of TSP due to any spillage or leakage. TSP will dissolve in water because of any accident that would cause flooding inside the containment and the water with dissolved phosphate will be drained into the containment sump through the five 16" and one 24" drain lines. There are three cases discussed in Chapter 14 of the FSAR that could cause flooding inside containment. The effects of TSP on the three cases are discussed below. For all accidents in the FSAR other than those discussed below, the TSP baskets will not affect the consequences since it is a passive system installed on the containment floor elevation.*

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes~ Specification Changes, Temporary Modifications and Other Documents (a)

Case I. Loss of Coolant Accident (LOCAl.

A large break LOCA activates the Safety Injection System. Borated water is pumped from the SIRW tank to the prim.ary coolant system to provide core cooling. Borated water from the SIRW tank is also discharged through two heat exchanger to a dual set of spray headers and spray nozzles in the containment to limit the containment building pressure rise and reduce the potential for release of airborne radioactivity. The borated water will collect on the containment floor and _

will dissolve the TSP, as it drains into the containment sump. The containment sump pH level will be maintained between 7.0 and 8.0 (Ref. to EA-FC-949-01) before the Recirculation Actuation Signal (RAS). This also meets the

-req-uirements set forth in the Standard Review Plan Section 6.5.2. -

The TSP baskets will not affect containment spray system or containment air cooler operation, and will therefore not affect containment pressure following a -

LoCA. The TSP, once dissolved will prevent iodine re-evolution from the sump solution by maintaining a pH greater than 7.0 as assumed in the MHA analysis EA-A-NL-92-012-03. Thus, TSP will not increase the radiological consequences of a LOCA. The TSP baskets are designed to maintain the sump solution pH less _

than 8.0 following recirculation. This is within the pH level used in the Hydrogen

generation analysis EA-GFP-92-02. Thus, the TSP baskets will not increase hydrogen generation following a LOCA. Hence, the consequences of a Loss of Coolant Accident previously evaluated in the FSARwill not i_ncrease because of the modification.

In case of a small break LOCA the recirculation phase may not occur at all or may set in few hours after LOCA. If recirculation phase does set in the effect of TSP_

would be the same as described above.

The TSP basket frames will be fabricated out of carbon steel and painted with zinc primer for corrosion resistance. The plant hydrogen generation analysis and FSAR Table 14.22-4 must be revised to account for the additional zinc in contain-ment.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Faciltty Changes, Specification Changes, Temporary ~odifications and Other Documents (b)

Case II. Main Steam Line Break inside containment (MSLB)

Following a postulated Main Stream Line Break inside the containment, the contents of the ruptured steam generator will be released to the containment. To limit the containment building pressure and the temperature. rise the Containment Spray System will be activated and borated water will be sprayed iriside the con-tainment. The borated water from the containment spray and the water from the main steam line break will dissolve the TSP and drain into the containment sump niaking the sump water caustic. Per FSAR Section 6.1.2.1 "there is no recircula-tion phase following a main steam line break" hence the caustic water will not be.

recirculated into the Primary Coolant System or Engineered Safeguards System.

The TSP baskets will not affect containment pressure following a MSLB.

Therefore, the consequences of main steam line break will not be *increased because of this modification.

(c)

Case Ill. Control Rod Ejection In case of a Control Rod Ejection a Loss of Coolant Accident is induced (Refer FSAR Section 14.16.3.1), the situation discussed in Case I above would result.

As discussed above, the TSP baskets will not increase the consequences of a LOCA if induced by control rod ejection.

The probability of malfunctions of equipment important to safety will not be increased because.of this modification. TSP is a passive form of containment sump pH control and requires no pumps, piping, heaters, etc. as employed by the current sodium hydroxide system. Retirement of T-103 and the corresponding installation of TSP baskets will decrease the number of components required to operate following a LOCA.

In case of a LOCA or induced LOCA (Control Rod Ejection) TSP will control the sump pH before RAS to an approximately neutral level between 7.0 and 8.0. At Recirculation Actuation Signal water will be circulated inside the containment atmosphere. Since, the safety equipment inside the containment is designed to operate at a pH near 7.0, the probability of malfunction of equipment important to safety will not be increased in case of a LOCA.

After a LOCA, the Primary Cooling System and the Shutdown Cooling System will be exposed to high temperature borated water. Prolonged exposure to core cooling water combined with stresses imposed on the components can cause SCC. High temperature and low pH, which would be present after a LOCA, tend to promote SCC. By controlling the containment sump pH within 7.0 and 8.0 TSP will help inhibit SCC. The fuel cladding will be exposed to the TSP dissolved in the borated water during recirculation. This solution will have a non-aggressive pH that will be closer to neutral than could be expected with the pH control achieved by the existing sodium hydroxide system.

CONSUMERS POWER COMPANY.- PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,

.Temporary Modifications and Other Documents Trisodium Phosphate wi[I remain dissolved in solution within the temperature range of 77°F and 283°F. Therefore, TSP will not precipitate out when cooled down and will not block the sump drain screens. Addition of TSP baskets will not increase the maximum. flood level as determined by EA-GCP-91-04 because any contribution from the volume of the eql1ipment would be small compared to the tota*I volume injected.

Following a main steam line break inside the containment, the sump pH will be basic.

Since, there is no recirculation phase in case of a MSLB caustic water will not be recirculated through the Primary Coolant System or Engineered Safeguard System.

Containment sump isolation valves CV-3030 and CV-3029 are made of stainless steel and are resistant to caustic solution. Hence, these valves will not fail and the.

modification will not increase the probability of malfunction of equipment important to safety.

The consequences of a malfunction of equipment important to safety will not be increased because of this modification. As mentioned above, post LOCA, TSP is a passive form of controlling containment sump pH and requires no pumps, piping, and electrical components. The TSP will be monitored per a Technical Specification surveillance test, which will ensure that the containment sump pH post LOCA and at.

RAS is between 7.0 and 8.0.

The possibility of an accident not previously evaluated in FSAR has not been created.

The potential for a new type of event that would cause a degradation of one or more fission product barriers and result in a radiological risk to the general public* more than 10CFR100 limits has not been created. The modification will ensure a containment sump pH of 7.0 or above at RAS and will increase the integrity of the Iodine Removal System. Installation of the TSP baskets will not create any new or abnormal operating conditions.

The possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be c_reated. As previously discussed, the installation of TSP baskets will not result in the creation of any malfunction of equipment important to safety not previously evaluated in the FSAR. The TSP baskets are a passive system. Retiring T-103 decreases the number of components required to operate following a OBA.

. Installation of TSP baskets will help inhibit SCC and this modification will play a direct role in mitigating the consequences of an accident.

The margin of safety as defined by the Plant Licensing Bases will not be reduced by this modification.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents FC-950-01 SE93-1229 TEMPORARY ALTERNATE SPENT FUEL POOL COOLING Cooling of the Spent Fuel Pool by a temporary system before the 1995 Refueling Outage (Refout) and cooling of essential loads during the 1995 Refout will be required while the Service Water and Component Cooling Water systems are out of service for valve replacements. Two systems of cooling will be provided and coordinated with operating procedures to ensure proper operation.

Prior to Refueling Outage Prior to the refueling outage*while the *heat.load from the spent fuel is at its lowest

level, one temporary system will consist of a movable heat exchanger located. in the pool with service water used as the cooling medium. The heat exchanger will be floating in the pool on redundant pontoons and rigged to the side. Hoses/piping will be used for supply and discharge. The discharge piping will be routed to the service water discharge side of diesel generators K6A and K68 to provide radiation monitoring of the alternate SFP heat exchanger discharge via Radiation Monitor RE-0833. Mixing of the SFP water will be assured through natural circulation and the use of skimmers. The system is expected to be placed in service anywhere from 60-0 days before the r-efueling outage for approximately two-weeks. During operation of the temporary system, activities in the Spent Fuel Pool heat exchanger room will include heat exc'1anger eddy current testing, valve maintenance, and piping modifications to the suction of the Spent Fuel Pool pumps. The testing and modification of the Spent Fuel Pool system will be performed by Facility Change FC-950-02. Due to the requirem~nts of Technical Specifications Section 3.21.2.b, the floating heat exchanger will be removed from the pool after the work in th~ Spent Fuel Pool heat exchanger room is completed and before the beginning of the core off load.

During Refueling Outage with Full Core Off L.oad Another system will be configured to cool the Spent Fuel Pool during the full core off

1oad. The system will comprise a service water pump, the spent fuel pool heat

. exchangers, diesel generators K6A and K6B service water supply (Critical Headers 'A' and 'B'), service water return lines to Lake Michigan, and temporary piping and fittings to make connections. Flow to all essential loads will be accomplished with appropriately sized temporary connections and piping: SFP heat exchangers will be tested before the outage to insure reliability. In the unlikely event of a leak to the service water system, normal monitoring is maintained by routing the discharge through the diesel generator service water return lines before the cooling tower pump

structure and Radiation Monitor RE-0833(see M-208, Sheet 1 A).

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT*.

Facility Changes, Spetification Changes, Temporary Modifications and Other Documents This alternate configuration for spent fuel pool cooling will be placed in service early in the refueling outage. The Spent Fuel Cooling system will be returned to its normal configuration after maintenance and repair work on the Service Water and Component Cooling Water systems are completed per FC-950-02.

All general design criteria for the temporary systems, as specified in the Design Input.

Checklist, will be met.* The temporary systems will be non-seismic and non-0.

Engineering Analyses will be provided to evaluate the routing and quality of components to be used to prevent a potential failure in a seismic event. Should an event occur, the pre-outage system will be floating in the pool and will not be seismically sensitive and there will be 2 - 4 days for required compensatory measures to be completed before the fuel pool temperature limits being exceeded. For the temporary outage system, the Shutdown Cooling System will be used as the backup or other compensatory measures evaluated.

Safety Analysis Summary Prior to Refueling Outage The only accident in Chapter 14 of the FSAR related to this Facility Change is a fuel handling accident. The floating heat exchanger will be movable while in the pool. In this manner, the heat exchanger can be placed away from any locations that could obstruct any required fuel movements before the outage. FSAR Chapter 14 also addresses the possibility of dilution of the boron concentration in the spent fuel pool. The alternate system components will be designed and tested to assure that leakage into the pool or out of the pool is prevented. Fuel will not be moved through the transfer canal or be located in the tilting mechanism during the use of the pre-outage alternate SFP cooling system. Thus, the placement and use of the alternate SFP heat exchanger in the SFP will not increase or decrease the probability of a fuel bundle being dropped during fuel movement activities before the outage. Therefore, the probability of an accident previously evaluated in the FSAR will not be increased for the temporary Spent Fuel Pool Cooling system before the outage.

Failure of the hose or heat exchanger during the 60 days before the refueling outage will not increase the consequences of an accident previously evaluated in the FSAR.

The heat load in the pool is at its lowest levels for all modes of operation, therefore, there is more time to respond to a loss of cooling medium or hose failure. Should this

system fail, the alternate system can be isolated and remain isolated for 2-4 days depending on the pool temperature at the time of isolation. The consequences of boron dilution in the SFP by the failure of the alternate heat exchanger or the supply and discharge hoses is bounded by the FSAR Section.14. 19. 1 evaluation that states that the design of the storage racks is such that the fuel will' be maintained subcritical even with zero boron concentration in the spent fuel pool storage water~ Fuel will not be in the transfer canal or tilt machine while the temporary cooling system is in place because these locations do not meet the subcriticality requirements for the zero boron concentration event.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,

. Temporary Modific~tions and Other Documents Before.taking the Spent Fuel Pool.Cooling system out of service, the alternate system will be tested and assured of proper operation. Testing of the alternate components will verify the alternate system can meet the Service Water system pressure requirements and will not increase the probability of leakage into the Spent Fuel Pool.

Alternate system components exterior to the pool will be evaluated to include protection which may be required to prevent inadvertent damage from *local traffic, etc.

in the temporary routing area. The heat exchanger will be redundant and capable of being moved around the pool. Additionally, should failure of the temporary hose/piping..

occur, the heat exchanger can be isolated, if required, and remain isolated for 2-4 days depending on the pool temperature at the time. of isolation while compensatory actions are taken. The hoses/piping, heat exchanger, and all temporary connections will be tested before use to prevent any possible leakage either into the pool or other safety-

related areas. The potential for flooding caused by the unlikely failure of the hose/piping will be evaluated to determine the required design features necessary to prevent the likelihood of flooding plant areas.

The activities associated with the temporary cooling system will not change, degrade or prevent any actions as stated in the FSAR or Technical Specification from being accomplished. Additionally, assumptions made in the evaluation of consequences will not change and bound the possible malfunctions associated with the alternate system.*

Also, should this system fail, the alternate system can be isolated for 2-4 days depending on pool temperature at time of isolation. The consequences of a possible flooding incident caused by the temporary system are bounded by previous flooding analyses described in the FSAR. Therefore, the consequences of a malfunction of equipment important to s~fety will *not increase.

Construction activities to Spent Fuel Pool, Service Water, and Component Cooling Water systems will occur while each system is out of service. The potential for boron dilution in the SFP by the alternate system is bounded by the FSAR Section 14. 19. 1 evaluation (and supported by EA-GCP-93-01) which states that the design of the storage racks is such that the fuel will be maintained subcritical even with zero boron

concentration in the spent fuel pool storage water. Also, the FSAR already evaluates the flooding and loss of spent fuel pool cooling accidents. Therefore, the possibility of an accident of a different type than any previously evaluated in the FSAR will not be created.

While construction activities are performed, the respective systems will be out of service. Additionally, safety-related systems will function if needed and, should the alternate Spent Fuel Pool Cooling system fail, the system can be isolated and remain isolated for 2-4 days depending on *pool temperature at the time of isolation.

Therefore, the possibility of a malfunction of a different type than any previously evaluated in the FSAR will. not be created.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,.

Temporary.Modifications and Other Documents Before the refueling outage the temporary heat exchanger will be installed and preliminarily expected to remove 4 million BTU/hr of heat. The design of the heat exchanger will be capable of removing 6 million BTU/hr of heat. Should the alternate.

Spent Fuel Pool Cooling system fail the alternate system can be isolated and remain isolated for 2-4 days depending on the pool temperature at the time of isolation.

Engineering analyses will be provided to verify the SFP heat load before the outage and the time required to initiate compensatory actions on failure of the alternate system.

Design of the alternate heat exchanger will require that the weight of the portion of the.

system placed in the pool (heat exchanger, pontoons, hoses/piping) will be greater than 1300 pounds. Per Technical Specifications Section 3.21.2.a states that "Heavy loads shall not be moved over fuel stored in the main pool zone." However, Technical

Specifications Section 3.21.2.b provides that "Heavy loads shall not be l"(loved over.

areas of the main pool zone that do not contain fuel unless the fuel stored in the main

. pool zone has decayed a minimum of 30 days when the charcoal filter is operating, or the fuel stored in the main pool zone has decayed a minimum of 90 days when the charcoal filter is not operating." The floating heat exchanger will be located in area's of_

the pool that do not contain spent fuel. This will be achievable once spent fuel has been removed via the dry cask storage process. A condition for use of the floating heat exchanger will be that boundary conditions for Technical Specifications Section 3.21.2.b (no fuel in the transfer canal or tilt machine) will be incorporated in the requirements for use of the pre-outage temporary system.

During Refueling Outage with Full Core Off Load FSAR Section 9.1.3.3 discusses the possibility of a service water discharge line rupture. The *service water requirements during the use of the alternate cooling system will be much less than the analyzed 16,386 gpm required during normal operation. No*

other changes will be made to the service water discharge piping. Due to the reduced.

discharge flows required during the use of this alternate cooling system, the possibility of the previously analyzed discharge line rupture event.is also reduced. Service water pumps P-7 A and P-7C will remain operable until the full core off load is completed. In this way, Tech Spec requirements (two pumps operable) will be maintained while fuel is in the* vessel. The integrity of the spent fuel pool heat exchangers will be verified before use of the alternate system. In this way, the probability of the boron dilution accident discussed in Section 14. 1 9 of the FSAR is not increased. Also, this system*

will not have components located in or around the spent fuel pool that could possibly increase the probability of a fuel handling accident. There were no other FSAR Chapter 14 accidents identified which were applicable to the alternate system to be placed in service after the refueling outage has begun.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specificaticin Changes,*

Temporary Modifications and Other Documents.

The Service Water and Component Cooling Water systems will be out of service during valve repair/replacement activities. To ensure that the heat exchanger can *perform its function before connecting the temporary service water system to the heat exchanger, the heat exchanger will be eddy current and leak tested per FC-950-02. The alternate

. cooling hose/piping and components will be.designed and tested to verify that the

temporary system meets the pressure requirements of the service '1\\(ater system.. This will ensure high reliability of the heat exchanger as well as the temporary system. The alternate cooling systems will not affect the severity of any FSAR Chapter 14 accidents

. (e.g., dilution event) and will have no adverse effects on radiological barriers relied on in an accident. Therefore, the consequences of an accident previously evaluated in the FSAR will not increase.

Maintenance to the Spent Fuel Pool system and design of the temporary system piping and components will ensure high reliability of the Spent Fuel Pool heat exchanger as well as the reconfigured system. Testing of the alternate components will verify the alternate system can meet the Service Water system pressure* requirements and will not increase the probability of leakage into the Spent Fuel Pool. Alternate system components will be evaluated to include any protection that may be required to prevent inadvertent damage from local traffic, etc. Jn the 'temporary routing area. However,

failure of the Spent Fuel Pool Cooling system or temporary piping during a refueling outage would require the Shutdown Cooling system to be aligned to cool the spent fuel pool. This will be achievable with the spent fuel pool gate removed and the core fully off loaded. However, alternate compensatory actions will be required and evaluc~ted for the scheduling and installation of particular service water system valves to be replaced during the use of this alternate system (i.e. CV-0823 and CV-0826). Also, the.

alternate system will be 'evaluated to determine the required design features necessary to prevent increasing the likelihood of flooding plant areas.*

A requirement of this facility change arid the Technical Specifications is that two Service Water. pumps are operable when fuel is in the vessel. This configuration will.

be maintained until all fuel assemblies have been off loaded at which time the alternate cooling lines will be connected to the manifold in the Spent Fuel Pool heat exchanger room. Redundant lines run via different paths will be connected to the manifold. Also, a backup supply of cooling water can be provided through a connection to the fire protection system. Thus, there will be redundancy, independence and a backup supply provided to mitigate the consequences of a malfunction of equipment important to safety.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Until the last fuel assembly is removed from the vessel, two Service Water pumps will

be operable anq will be available to provide required cooling. When the last fuel assembly is placed in. the Spent Fuel Pool, the point of emphasis for accidents will be the cooling of the Spent Fuel Pool. This situation is analyzed in Section 9.4.3. l of the

. FSAR. This section indicates that, in the case with no Spent Fuel Pool pumps operating, the time to reach boiling (212 ° FHn the pool is 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months . The conditions for this analysis (0 SFP pumps operating) bound the possible accident scenarios for this*

facility change in that the possible failure of the temporary system cannot create an accident more severe than the accident involving no SFP pumps operating.

Any possible malfunctions are bounded by the conditions analyzed in Section 9.4.3.1 of the FSAR (0 SFP pur:nps oper-ating). No other credible malfunctions exist which co*uld create conditions different (more severe) than those previously analyzed.

The design of the existing heat exchanger allows 30.8 million BTU/hr of heat to be removed. The actual removal need during the fuel core off load after 17 days of decay

. is approximately 24 million BTU/hr of heat. The existing system will be -used with service water temporarily connected to provide the cooling medium, while service water and component cooling water system valves are being repiaced. The temporary system has the capability of removing the original design required heat load. Using the service wate.r, which is at a lower temperature than CCW, allows the system to be operated with less water flow to remove the same amount of heat. Engineering analys.es. will be provided to verify the SFP heat load with the full core off load and the r~quired service *water flow to the SFP heat exchangers. Therefore, the margin of

safety as defined by the Plant Licensing Bases will not be reduced.

FC-950-02 SE93-0800 MAINTENANCE OF SERVICE WATER (SW) AND SPENT FUEL CSFP) SYSTEM VALVES This facility change involves the replacement of valves in the SW and SFP systems.

Valves and valve operators that will be replaced will be of a similar design to the existing valves. Changes will be made to the suction piping of pumps P-51 A and P-51 B and the Spent Fuel Pool heat exchangers. An alternate Spent Fuel Cooling system will be required and installed per FC-950-01 for th~s work.

Safety Analysis Summary The valves being replaced act in mitigating accidents identified in the FSAR (i.e.,

providing cooling medium for the containment air coolers). Their failure would not.in.

itself cause an accident previously evaluated in the FSAR so the probability of such an accident is not increased by their replacement.

CONSUMERS POWER COMPANY

~ PALISADES ~UCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents All the replacement valves will meet or exceed the operating requirements (flow, pressure,.temperature, leakage, failure position, etc.) of the existing valves. Thus, the basic design requirements of the valves, to allow the required flow, stop flow when required, and maintain the system pressure boundaries are not being changed by the replacement of the valves identified in Table 1 with similar valve designs. The

identified valves will provide their design function such that the consequences of an accident previously evaluated in the FSAR will not be increased.

Rupture of the common service water discharge line as described in Section 9.1.3.3 of

the FSAR is not affected since the required action for this event, isolation of either of the main supply lines, is still available after installation of the new valves in positions CV-844, 845, 846, and 857. Changes to the SFP suction piping _will not alter the

reasoning provided in FSAR Section 9.4.3.1 describing a potential failure of the inlet spent fuel pool piping system.

All the valve positions, as identified in Table 1, required Seismic Category I, CPCo Design Class 1, and ASME Ill, Section 3 components. The replacement valves, valve.

operators, and piping will be designed to meet or exceed these requirements. Failure

. positions for the air-operated valves (AOVs) will be maintained and system leakage, pressure, and temperature requirements will be met or exceeded for the new components.

The changes to the suction piping* of pumps P-51 A and P-51 B will include the installation of basket or cup type strainers between the existing flanges. The purpose of these strainers will be to remove large objects which could potentially damage the

pumps. Existing strainers in these locations were removed per Work Order 24706996 and were not replaced. The purpose of removing these strainers was to eliminate hot spots in the suction piping. This resulted in foreign material entering pumps P-51 A and P-51 B as documented in D-PAL-88-107. The new strainers will have a mesh size large enough to prevent plugging and hot spots created by the accumulation of small material but small enough to prevent large objects from damaging the pumps. Thus, the proba.bility of the failure of these components is equal to or less than the failure probabilities for the existing equipment.

The possible malfunctions of these components include:

Failure to maintain pressure boundaries and, Failure of the valves to open or close when required.

As discussed above, the replacement components will be required to meet or exceed the design requirements of FSAR Table 5.2-3,. the system pressure/temperature requirements, and will be designed to satisfy the designed failure conditions. Thus, the consequences of a malt unction of this equipment are not changed from those previously evaluated.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modi fi cations *and Other Documents The replacement components will have the same design functions as the existing components (Le., to maintain the pressure boundaries and allow or stop flow when required). There are no new flow paths or failure positions crated by this facility.

change* such that an accident of a different type could be created.

As discusseq above, the only credible malfunctions possible are the failure to maintain pressure boundaries and failure of the valves to open or close when required. By requiring the replacement components tb meet or exceed the system design requirements, the probability of the credible malfunctions is not increased and the possibility of different malfunctions is not created.

The margins* of safety for the systems involved (SW and SFP) are based on providing redundancies to the system flow paths as identified in the "Margin of Safety" FSAR section.for each system. This facility change provides a 1 for 1 replacement of the.

existing valves with similarly designed valves with the same design function (normal

. and failure positions). Therefore, system redundancies are maintained and the margins of safety are not reduced.

. FC-951 SE93-0847 CONTAINMENT AIR COO_LER (CAC) COOLING COIL REPLACEMENT The proposed modification will replace the cooling coil units on three of the four Containment Air Coolers. The coolers have been experiencing leaks, primarily in the manifolds and headers. Work history has documented over 30 such leaks. Non-code repairs have been allowed by the NRC until such time as replacement of the coils becomes feasible. The existing coo[ers consisting of inlet and outlet headers, * *

. manifolds, and coils will be replaced with new units. The replacement units will be designed to maintain the existing heat removal capacity under postulated accident conditions. The design will match existing pressure drops and flows in order to maintain compatibility with interfacing equipment. The new coolers will meet CPCo Design Class 1 requirements..

Safety Analysis Summary

The proposed change involves the replacement of safety related cooling coils within the three (VHX-1, -2, and -3) safety related Containment Air Coolers. A review of all design basis accidents listed in Chapter 14 of the Plant's FSAR found that none of the scenarios were initiated by the failure of a component of coolers on the Containment Air Coolers as units. *.I

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes~

Temporary Modifications and Other Documents Per a review of FSAR Chapter 14, it was found.that the Containment Air Coolers are used to mitigate the consequences of two previously analyzed accidents, Main Steam Line Break (MSLB) and Loss of Coolant Accident (LOCA). In both scenarios, the

. coolers' function is to limit the peak pressure and temperature attained in the containment building following initiation of either event. The proposed replacement cooling coils are to be made with equally qualified components which do not alter the function of the coolers. Since the proposed design capacity of the cooling coils with respect to accident conditions will not be changed, the probability of the containment being challenged is not changed, hence, the consequences of these postulated accident scenarios are not increased.

The heat *removal capability of the coolers is also used as an input to the analysis of the fuel. under. LOCA conditions. In this case, the maximum expected heat removal will be used to ensure that specified fuel design limits are maintained.

This modification will be installed in accordance with code, regulatory and qualification requirements or, be reconciled with the original design and construction of the affected piping system. In addition, FSAR Table 5.2-3 indicates the Containment Air Cooling System and the Service Water System supply being CPCo Design Class 1. As such, a reverification of the affected stress packages with appropriate interfaces will be performed as nece*s~;ary. Thus, the probability of equipment failure is not increased from that of the original installation.

The proposed modification does not change the function of the cooling coils. Also, the modification will be installed consistent with or reconciled with the original requirements. In addition, redundancies with the remaining coolers and containment spray pumps ensure cooling of the containment environment following either of the two applicable events. In conclusion, the effectiveness of the barriers in limiting the consequences of any previously analyzed malfunction would remain unchanged.

The component's function in the system has not changed and, therefore, its operational characteristics could not result in any new system transientS. The proposed replacement coils do not introdu.ce any new or different failure mechanisms which haven't been previously evaluated or bounded. It is concluded that no new accidents will be. introduced by this modification.

The replacement coils will remain subject to all typical failure modes considered in the original design and will be installed consistent with or reconciled with the original requirements of the cooling coil~.

CONSUMERS POWER COMPANY.. PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Ot'her Documents Design characteristics of equipment which serves to protect the margin of safety.

possess certain design criteria.. These criteria include diversity, separation, redundancy, and independence. The proposed modification does not change the function or alter the inherent design characteristics mentioned above of the Containment Air Coolers or any of their components. Since the new coolers will be designed for a service water temperature of 85°F, the actual margin may be increased. Therefore, the margin of safety in the Plant's Licensing Bases has not been reduced.

FUNCTIONALLY EQUIVALENT e

SUBSTITUTIONS

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT.

Facility Changes, Specification Cha*nges, Temporary Modifications and Other Documents FES-93-015 SE93-0103 REPLACE CK-CA440 Replace CK-CA440 with stainless steel identical valve se_ri~s 8C4-10 NUPRO.

Safety Analysis Summary

The stainless steel check valve will no*t be susceptible to wear like the original brass model. This change will improve the safety and reliability. This modification will not increase the probability or consequences of an accident previously evaluated by the FSAR. The probability and consequences of equipment important to safety will not be increased. Neither the possipility of an accident nor malfunction of a different type.*

than any previously evaluated in the FSAR will be created. The margin of safety of Licensing Bases will not be reduced.

FES-93-031 SE93-0977 RPS AUXILIARY TRIP UNIT ZENER DIODE REPLACEMENT The Auxiliary Trip units are experiencing false telltale light actuations. This is due to plant generated noise causing transistor 02 or 04 to "turn on II firing ci 1 or 03 turning trip or pretrip telltale lights on. Increasing the voltage value of zener diodes VR2 and VR 1 will help mitigate this problem. The Auxiliary Trip Units will experience no change to their function.

Safety Analysis Summary

No accident identified in Chapter 14 of the FSAR will be affected by this modification.

The affect on the RPS is transparent. The _Auxiliary Trip Unit Circuit Boards will operate

. exactly as before. The. small change in the voltage applied to transistors 02 and 04 will not adversely affect the operation of the rest of the circuit. Work Order (WO)

2430014 7 documents this and shows how the changes should mitigate the telltale illumination problem. The RPS Auxiliary Trip units are Class 1-E (safety related). The

new zener diodes are procured to meet all applicable design, material, and construction standards for a Class 1-E system.

No system.design limits will change due to this modification. Only a small decrease in current supplied by* the *power supply will be experienced. This will not adversely affect.

the power supply.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents No accidents previously evaluated in Chapter 14 of the FSAR are affected by this modification. Consequently the consequences of an accident previously evaluated cannot be increased. Doses received by the general public and doses received-by plant personnel as a direct result of an accident are, therefore, not of concern as a result of thi$ modification. No assumptions made or the methodology used in evaluating the consequences of an accident described in the FSAR are altered.

This modification does not change, degrade or prevent actions for which credit is taken or implied in the FSAR description of an accident or malfunction. No assumptions made in the evaluation of the consequences of an accident or malfunction will be altered. The system will operate exactly as before modification, so a direct role in mitigating the consequences of an accident or malfum;:tion is not a concern.

No new type of accident.can be postulated due to the zener change out. All failure modes are identical implying that all existing accidents are considered.

The replacement will use components and procedures compatible with existing.

Therefore, the function of the system in no way changes. No new loads are present due to change so no new types of equipment malfunctions can occur. No new failure modes can occur.

This modification has no affect on Instrument Channels, Reactor Trip, Reactor Protective System, Logic Degree of Redundance, and Engineered Safety Features System Logic described in Section 1-2 of Technical Specification. The Limiting Safety System Settings, Section 2.3 of Technical Specification, are not affected by this modification. Instrumentation Operating Requirements for RPS are outlined,in Table.

3.17.1 are not affected. No acceptance limits will change due to the malfunction. The change ha-s no direct or indirect effect on fission product boundaries or accident consequences due to the likeness of the system.

FES-93-087 SE93-0742 REPLACEMENT OF PRINTING DEMAND METERS Replace GE type PD-57F printing demand meters with new digital meters.

Safety Analysis Summary The probability or consequence of a previously evaluated accident will not increase since the meter cannot create an accident. The meter will be seismically mounted within the C-04 panel using available space vacated by current meters.

The probability or consequence of a malfunction t<? equipment important to safety will not increase. This equipment does not interact wit_h any equipment important to safety.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The possibility of an unevaluated accident or malfunction will not increase since this meter will be seismically mounted in the space vacated by the older meter and does not interact with any equipment important to safety.

The margin of safety as currently defined will not be reduced. This meter is 0-listed for seismically mounted equipment only and does not interact with any safety related equipment.

FES-93-90 SE93-0358 REPLACE MV-CA698 Replace MV-CA698 (CV-0701 A/S equalization valve).

Safety Analysis Summary MV-CA698, a 1 /2" gate valve is being replaced by a 3/8" ball valve. This type of ball valve is* used widely throughout the plant with success. This replacement will improve the reliability of the air supply equalization line isolation for the MFW Reg valve CV-0701. The valve (MV-CA698) is normally in a closed position. The replacement will

. not affect plant or system operation. The replacement will not increase the probability, consequences, or possibility.. of accidents either previously or not already analyzed by the FSAR. Nor, will the probability, or consequences of malfunctions already analyzed, or new malfunctions not analyzed be increased. There is no effect on the margin of safety.

enceFES-93-091 SE93-0357 REPLACE MV-CA903 Replace MV-CA903 (CV-0703 A/S equalization line isolation).

Safety Analysis Summary A 3/8" needle valve is being replaced by 3/8" ball valve with more reliable shut off characteristics. The replacement will not increase the probability or consequences of

accident both analyzed or a different type than those in the FSAR. Nor, will the probability or consequences of previously analyzed or new malfunctions not analyzed be increased. There is no effect on margin of safety.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Change~, Specification Cha~ges,.

Temporary Modifications and Other Documents FES-93-162 SE93-0824

. STEAM GENERATOR E-SOA AND E-SOB TUBE PLUGGING CUP TO 10 TUBES PER STEAM GENERATOR)

Remove up to 10 tubes per steam generator from services based on Tech Spec and procedure acceptance limits. The tubes to be_plugged exceed the plugging criteria specified in the 1993 Steam Generator Eddy Current Test Outage Plan. Plugs used will be Westinghouse E-690 "Mechanical Plugs." Plugs are qualified and designed for use in Palisades steam generators.

Safety Analysis Summary Removal of degraded tubes from service preserve the low probability of an accident, specifically a steam generator tube rupture event.

Plugging tubes cannot affect any FSAR accidents or the consequences of an accident except to decrease its probability by removing defective tubes from service.

The probability of a malfunction to the steam generator, which is important to safety will be decreased by plugging unacceptable or defective tubes.

Consequences of a malfunction of equipment impo_rtant to safety will not be increased.

This change merely removes defective tubes from service and has no-effect on any steam generator failure.

. Plugging only affects the steam generator tube rupture accident and removal of.

defective tubes prevents that from happening in those tubes.

Plugging defective tubes prevents the tube leakage. No malfunctions of a different

'type are created.

Plugging de.fective tubes prevents the defects in those tubes from falling below acceptable limits for this design. The margin of safety is preserved by this activity.

NOTE: The number of tubes to be plugged will not exceed 15% of the total number of tubes which is the value assumed in plant accident analyses.

-------------------------------: ---------~-~- --

CONSUMERS POWER COMPANY - PALISADES NUCLEAR P.LANT -

Facility Changes, Specification Changes, Temporary Modifications and Other Documents FES-93-204 SE93-0959 INSTALLATION OF R0-2122 FLOW ORIFICE IN FI-2122 INLET FLANGE R0-2122 is an orifice sized to restrict the flow to flow indicator Fl-2122 (Fuel Pool Recirc Booster Pump Flow). The flow indicator is currently not working. Adding the R0-2122 will only increase operations* flexibility to measure the flow by restoring proper operation of Fl-2122.

Safety Analysis Summary The restricting orifice is of adequate d,esign to be placed in the 1" flange. (Reference EA-FES-93-201-01). The orifice will not fail.

Existing Fl-21 22 is reading incorrectly. The failure of the orifice to restrict flow would _

not change the current status of the indicator.

The restricting orifice has a very low probability of 'malfunctioning, Addition of the restricting orifice will not introduce the occurrence of an accident or malfunction different than those evaluated in the FSAR.

. The margin of safety as defined by the licensing basis will not be reduced as a result of the installation of the restricting orifice in the 1" flange on the inlet side of Fl-2122.

FES-93-285 SE93-1324 REROUTING PIPING FOR RADWASTE VENT MONITORING. PUMP P-1809 This Analysis is written to evaluate/document the acceptability of rerouting piping for the Radwaste Vent Sample Pump, P-1809. Rerouting of the piping will include the following:

1)

Relocating valves MV-VA 113 and MV-VA115 such that they will no longer interfere with the sampling skid removal for maintenance.

2)

Replacing globe valves MV-VA 113 and MV-VA 115 with ball valves.

3)

Replacing the existing piping and tubing between the new locations for MV-VA 113 and MV-VA 115 and the skid with hose suitable for this application.

4)

Installing unions on the suction and discharge sides of P-.1809 to ease removal of the pump during maintenance.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specificatirin Changes,*

Temporary Modifications and Other Documents Safety Analysis Summary MV-VA 113 and -VA 115 are normally open valves. The purpose of these valves is to provide isolation between the Radwaste Ventilation System and the RE-1809 sampling skid for normal maintenance. The safety function of the piping and valves is to provide flow to the sampler. Per FSAR Section 7.4. 5.3, "the radiation monitor provides a ventilation shutdown signal for the area in the event of spillage. The area is maintained under negative pressure to prevent radioactive leakage out of the building with the supply fan and damper shut off while one of the exhaust fans is not shut off. Alarms in the control room warn the operator that one or more rad waste area ventilation fans have tripped. Per FSAR Section 9.8.2.4.16.f,."ln the event of a spillage of radioactive material in the radwaste area, the radiation monitor at the filter plenum senses the activity and stops the supply fan, closes the radwaste area supply Damper P0-1809, and stops t_he selected exhaust fan; however, a low flow alarm will override the high radiation signal and keep the standby exhaust fan running. The duct to access control remains open and is isolated from the radwaste area by Damper P0-1809." The identified piping and valves assure flow to the sampler in case of spillage in the radwaste area. They do not contribute to the possibility of the spillage or provide any active fuhctiori to alleviate the spill. Therefore, their replacement will not increase the probability of the spill occurring.

The new piping and valves will have no affect on the consequences of a spill in the radwaste area as the replacement components will provide the required flow (one complete sample change through the system every 10 seconds per Specification M-217) as the new configuration requirements (pressure retention and flow) are equal to or exceed the existing configuration requirements.

Operability and leak testing of the piping and sampler skid shall be performed prior to returning the sampler to service. This will provide assurance that the new configuration does not leak and will provide the required flow to the sampler. Since the valves and piping provide only a passive function (not required to be operated and only

p.rovide flow); an increase in* the probability or consequences of a malfunction of this

equipment is not credible.

The new valves and piping are a one-for-one replacement for the existing configuration and do not introduce any new or different operating parameters fro_m the existing configuration.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT

Facility Changes, Sp~cification Changes, Temporary Modifications and Other Documents FES.;.93-286 SE93-1325 REROUTING PIPING FOR EAST ENGINEERED SAFEGUARDS ROOM VENT MONITORING PUMP P-1810 This Analysis is written to evaluate/document the acceptability of rerouting piping for the East Safe.guards Room Vent Sample Pump, P-1810. Rerouting of the piping will include the following:

1)

Relocating yalves MV-VA 120 and MV-VA 124 such that they will no longer interfere with the sampling skid removal for maintenance.

2)

Replacing globe valves MV-VA 120 and MV-VA 124 with ball valves.

3)

Replacing the existing piping and tubing between the new locations for MV-

. VA 120 and MV-VA 124 and the skid with hose suitable for this application.

4)

Installing unions on the suction and discharge sides of P-1810 to ease removal of the pump during maintenance.

Safety Analysis Summary MV-VA 120 and -VA 124 are normally open valves. The purpose of these valves is to provide isolation between the Engineered Safeguards Room Ventilation System and the RE-1810 sampling skid.for normal maintenance. The safety function of the piping and valves is to provide flow to the sampler. Per FSAR Section 7.4~5.2, "One radiation monitor for each engineered safeguards pump room to provide a room isolation signal upon high radioactivity levels in the applicable room. The automatic isolation allows maintenance of acceptable dose levels at the site boundaries (Chapter 11 ). " Per FSAR Section 9.8.2.4.16.g, "In the event of significant airborne contamination in the engineered safeguards rooms, the supply and exhaust dampers of those rooms are closed on a signal from the individual radiation monitor for each e_xhaust duct. II The.

identified piping and valves assure flow to the sampler in case* of a high airborne indication in this area. They do not contribute to the possibility of the event or provide any active function to alleviate the event. Therefore, their replacement will not increase the probability of the event occurring.

The new piping and valves will have no affect on the consequences of a high airborne event in the engineered safeguards area as the replacement components will provide the required flow (one complete sample change through the system every 10 seconds per Specification M-217) as the new configuration requirements (pressure retention and

. flow) are equal to or exceed the existing configuration requirements.

CONSUMERS POWER COMPANY

~ PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Do~uments Operability and leak testing of the piping and sampler skid shall be performed prior to returning the sampler to service. This will provide assurance that the new configuration does not leak and will provide the required flow to the sampler. Since the valves and piping provide only a passive function (not required to be operated and only provide flow), an increase in the probability of a malfunction of this equipment is not credible.

The new valves and piping are a one-for-one replacement for the existing configuration and do not introduce any new or different operating parameters from the existing

configuration.

FES-93-287 SE93-1326 REROUT.ING PIPING FOR WEST ENGINEERED SAFEGUARDS ROOM VENT MONITORING SAMPLE PUMP P-1811 This Analysis is written to evaluate/document the acceptability of rerouting piping for the West Safeguards Room Vent Sample Pump, P-1810. Rerouting of the piping will include the following:

1)

Relocating valves MV-VA 121 and MV-VA 125 such that they will no longer interfere with the sampling skid removal for maintenance.

2)

Replacing globe valves MV-VA 121 and MV-VA125 with ball valves.

3)

Replacing the existing piping and tubing between the new locations for MV-VA 121 and MV-VA 125 and the skid with hose suitable for this application.

4)

Installing unions on the suction and discharge sides of P-1810 to ease removal of the pump during maintenance. -

~

CONSUMERS POWER COMPANY - PALISADES'NUCLEAR PLANT Facility Changes, Specification Changes; Temporary Mod*i fi cations and. Other Documents Safety Analysis Summary.

MV-VA 121 and -VA 125 are normally open valves. The purpose of these valves is to provide isolation between the Engineered Safeguards Room Ventilation System and the RE-181.0 sampling skid for normal maintenance; The safety function of the piping and valves is to provide flow to the sampler. Per FSAR Section 7.4.5_.2, "One radiation monitor for each engineered safeguards pump room to provide a room isolation signal upon high radioactivity levels in the applicable room. The automatic isolation allows maintenance of acceptable dose levels at the site boundaries (Chapter 11). " Per FSAR Section 9.8.2.4.16.g, "In the event of significant airborne contamination in the engineered safeguards rooms, the supply and exhaust dampers of those room.s are closed on a signal from the individual radiation monitor for each exhaust duct." The

identified piping and valves assure flow to the sampler in case of a high airborne indication in this area. They do not contribute to the possibility of the event or provide any active function to alleviate the event. Therefore, their replacem*ent will not

increase the probability of the event occurring.

The new piping and valves will have no affect on the consequences of a high airborne eve11t in the engineered safeguards area as the replacement components will provide

the required flow (one complete sample change through the system every 1 0 seconds per Specification M-217) as the new configuration requirements (presswe retention and

flow) are equal to or exceed the existing configuration requirements.

Operability and leak testing of the piping and sampler skid shall be performed prior to returning the sampler to service. This will provide assurance that the new

co.nfiguration does not leak and will provide the required flow to the sampler. Since the valves and piping provide only a passive function (not required to be operated and only provide flow), an increase in the probability or consequences of a malfunction of this equipment is not credible.

The new valves and piping are a one-for-one replacement for the existing configuration and do not introduce any new or different operating parameters from the existing configuration.

SPECIF/CATION. CHANGES

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Docum1~nts SC-91-167 SE93-0988 SE R/1

sE91-1306 CONDENSER TUBE PLUGGING Install tube plugs for leaking or damaged tubes. The condenser is over designed, therefore, removal of a small number of tubes from service will not have any effect on the condenser's performance.

Safety Analysis Summary Review of the accidents analyzed in Chapter 14 of the FSAR indicate that no analyzed accidents will be affected by this change. The main condenser is not identified in any analyzed accident scenario. Therefore, the plugging of condenser tubes does not affect the probability or consequences of an accident previously analyzed in the FSAR.

The probability of malfunctions of equipment important to safety will not be increased.

The main condenser is non-Q and nonsafety related. The plugging of condenser tubes does not have any effect on the reliability nor does it affect any equipment important to safety. The plugging will prevent circulating water from entering the feedwater and detrimentally affecting the steam generators.

The consequence of a malfunction of equipment important to safety will not be increased. The main condenser does not play any role in mitigating the consequences of an accident er malfunction. The plugging of condenser tubes doe;; not affect the function of the condenser or the overall performance of the condenser since it was designed to have tubes plugged over its life.

The probability of an accident_ of a different type than previously evaluated in the FSAR will not be created. The plugging of the main condenser tubes will not cause a degradation to one or more fission product barriers or result in radiological risk to the

general public in excess of the 10CFR100 limits. The main condenser is not involved in any accident scenario and the plugging of condenser tubes will not change the function or the overall performance of the condenser.

The possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created. The main condenser.does not perform a safety function..

The plugging of the main condenser tubes will prevent leakage of circulating water into the condensate and feedwater. Therefore, the change will not result in any failures which could directly challenge another safety system or introduce any new types of failures.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes~

Temporary Modifications and Other Documents The margin of safety as_ defined by Plant Liqmsing Bases will not be reduced. The main condenser does not have any direct or indirect effect_ on fission product boundaries or accident consequences. The plugging of condenser tubes will prevent the leakage of circulating water into the condensate and feedwater. Therefore, the margin of safety is not reduced.

sc~92-oso SE93-1049 SE R/l SE93-0924 FEED PUMP RECIRCULATION VALVE POSITIONER MODIFICATION Spe_cification changes for feed pump recirculation valve positioner modification. This

-. SC adds positioner (POC-071 O/POC-0711) on the main feed pump recirculation valves (CV-071 O/CV-0711) for improved recirc control.

This modification is in support of eliminating main feedwater control system transients caused by feed pump recirculation valve opening and closure. This will be accomplished by installing a positioner on the recircul_ation valves CV-0710 and CV-0711. The positioner will ramp open and close the valve per a "programmed" curve of suction flow versus required recirculation flow.. As the feed pump suction flow increases, the recirculation valves will close.

Addition of these positioners will allow better feedwater control and pump protection without impacting operations as far as functional changes are concerned.

Safety Analysis Summary The probability/consequences of an _accident or malfunction previously evaluated -in the*

FSAR are not increased. This modification is for decreasing transients in the main feed system by providing better control over the recirculation portion of this system. This system is located in the turbine building and is not required by plant Tech Specs.

This modification does not create the possibiiity of accident/malfunction of different type because it will decrease transients in the main feedwater through better -

recirculation control and thus protect the feed pumps.

The margin of safety as defined by Plant Licensing bases is not changed.

-CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Te~porary Modifications and Other Doc~ments SC-92-124 SE93-0482 PROVIDE PERMANENT 480 VAC POWER FEED TO THE STEAM GENERATOR CSGl TEMPORARY STORAGE BUILDING The SG temporary storage building will be used permanently by the plant. This Specification Change (SC) will provide a permanent 480 Vac power feed to the building.

from the primary side of the parking lot feeder transformer No X 1 08 which is currently fed from the MCCB200 at the Auxiliary Building via Breaker No 52-20006.

Safety Analysis Summary

The modification is nonsaf ety related.

_The probability of an accident will not be increased since it does not affect safety related components.

The consequences of an accident will not be increased since _the change will not have an effect on components that mitigate an accident.

The probability or consequences of a malfunction of equipment important to safety will not be increased since the change does not affect safety related equipment. MCC -

B200 was evaluated and it was determined that it has enough capacity to support the_

load at the SG Temporary Storage Building.

A different type of accident or malfunction will not be created since the change affects only nonsafety related equipment.

The margin of safety as defined by the Plant Licensing Bases will not be reduced since the change is on nonsaf ety related equipment that does not affect safety related components.

SC-92-127 SE93-0128.

REMOVAL OF AUXILIARY FEEDWATER RELAYS R/0727 AND R/0749 The relays being removed by this modification (R/0-727 and R/0749) presently perform the transfer function when control of the Auxiliary Feedwater System is desired from the Auxiliary Safe Shutdown Panel (C 150). These relays are energized by actuating hand switches (HS-1012A/HS-O 102B) in the C 150 panel. These hand switches can directly perform this transfer function. This will allow for the elimination of the relays.

Since they are located in a harsh environment, qualification will no longer need to be maintained.

- 58 _-

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modi.fi cations and Other Documents Safety Analysis Summary The probability of an accident previously evaluated in the FSAR will not be increased.

This modification will be eliminating the EEO transfer switches from the component cooling water room as their qualified life is nearing its end and suitable replacements can not be obtained. By eliminating unnecessary equipment from the instrumentation loop, the probability of equipment failure is reduced and the probability of an accident is not increased.

The consequences of an accident previously evaluated in the FSAR will not be increased. This modification will eliminate the redundant relays from the instrumentation loop and their function will be completed by another existing *

instrument. The same function will still be r~quired from the remaining equipment in order to mitigate an accident. However, reducing the number of instruments in, the

loop reduces the chance for failures iri the equipment which helps ensure the equipment will be able to perform its intended function. Therefore, this modification will not detrimentally change the consequences of an accident previously evaluated in the FSAR.

The probability of malfunctions of equipment important to safety will not be increased.*

Removal of the relays will reduce the amount of equipment in the safety scheme. The malfunction probability of the relay will be eliminated. The hand switch will still be performing the same function and its probability for malfunction will not be changed.

Therefore, the probability of malfunctions of equipment important to safety will not be increased.

The consequences of a malfunction of equipment important to safety will not be increased. No new equipment will be added to the instrument loops. The hand switch will still be required to operate in the same manner as before the modification. If it were to malfunction, the consequences would be the same as previous to the modification. The failure of the hand switch would still inhibit the transfer of control of the Auxiliary Feedwater System to the C150 panel. Therefore, this modification does not effect the consequences of a malfunction of equipment important to safety.

This modification does not result in the system operating in a different manner than it does presently. Therefore, an accident of a different type than ariy previously evaluated in the FSAR is not a possibility.

The equipment in the auxiliary feedwater instrumentation loops is the same as previous*

to the modification with the exception of the elimination of the relays (R/0727 and R/0749) from the loop. The equipment will function in the same manner as before the modification and therefore the possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created.

. CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Chang~s, Temporary Modifications and Other Documents Since no new equipment will b.. e added to the instrument loops and the cable being used will be purchased for the application the margin of safety as defined by Plant Licensing Bases will not be reduced.

SC-92-128 SE93-0285 REPLACE VARIOUS PRIMARY SYSTEM TRANSMITTERS Specification Change will upgrade fifteen "primary" system transmitters and their instrument loops. It will remove voltage limiting diodes from eight "primary" system loops and it will provide a power supply for FT-021 OA as it is not being upgraded. It will also delete the existing multiple instrument power supply P/S-0202.

The following transmitters will be replaced:

EQUIPMENT ID DESCRIPTION LT-0116 PT-0131A PT-0131B PT-0132A PT-0132B PT-0133A PT-0133B PT-0141A PT-0141B PT-0142A PT-0143A PT-01438 Pt-0205 LT-0205 FT-0202 Quench Tank Level PCP P-50A Vapor Seal Pressure

PCP P-50B Vapor Seal Pressure

PCP P-50A Middle Seal Pressure PCP P-50B Middle Seal Pressure

.. PCP P-50A Upper Seal Pressure

PCP P-50B Upper Seal* Pressure

PCP P-50C Vapor Seal Pressure PCP P~50D Vapor S~al Pressure PCP P-50C _Middle Seal Pressure.

PCP P-50C Upper Seal Pressure PCP P-500 Upper Seal Pressure Volume Control Tank Pressure Volume Control Tank Level Letdown* Flow The following transmitters will have zener diodes removed:

EQUIPMENT ID DESCRIPTION PT-0142B PT-0116 PT-0202 PT-0206 PT-0208*

PT-0215 FT-0306 FT-0315 PCP P-500 Middle Seal Pressure Quench Tank Level Letdown Pressure Boric Acid Discharge Pressure Boric Acid Disch_arge Pressure PCS Bleed Off Header Pressure Shutdown Cooling Flow SI Tanks Leak Off CONSUMERS ~OWER COMPANY - PALISADES NUCLEAR PLANT Facility.Changes, Specification Changes, Temporary Modifications and Other Documents The following equipment will be deleted from instrument loops:

EQUIPMENT ID DESCRIPTION PA-0132A PA-.01328 PA-0133A

. PA-01338 PA-0142A PA-01428 PA-0143A PA-01438 FY-0202 P/S-0202 Middle Seal Pressure Alarm Middle Seal Pressure Alarm Upper Seal Pressure Alarm Upper Seal Pressure Alarm Middle Seal Pressure Alarm Middle Seal Pressure Alarm Upper Seal Pressure Alarm Upper Seal Pressure Alarm Letdown Flow Square Root Extractor.

Multiple Instrument Power supply

P/S-0202 will be replaced by a new power supply for FT-021 OA.

Safety Analysis Summary The probability of an.accident previously evaluated in the FSAR will not be increased.

Thi~ modification only upgrades aging and obsolete equipment with new, more reliable equipment. The equipment's function will not be altered. In some cases, instrument loops will be simplified by combining the operation of two pieces of equipment into one. This will reduce the possible equipment failures in the loop without changing

actual function.

Tha consequences of an accident previously evaluated in the FSAR will not be increased. The equipment will function in the same manner as before the modification.

Some of the instrumentation loops will be simplified slightly, but the operation of the loops will remain the s~me. Any foriction they performed to mitigate an accident will remain the same.

The replacement equipment has the possibrlity of malfunctions similar to the existing equipment; However, the probability or consequences of these malfunctions are not increased by the modification,. as the modification results in new, more reliable equipment in the loop and in some cases-a more simplified scheme.

The possibility of an accident of a different type than previously evaluated in the FSAR will not be created. The instrument loops will be performing the same function as previous to the modification. These replacements will not require the instruments to perform differently thereby possibly creating an accident of a different type than previously evaluated in the FSAR.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Speci fi cat ion Changes, Temporary Modifications and Other Documents Since the equipment is being upgraded to newer and more reliable equipment, and since it is performing the same basic function in the same basic manner the possibility of ari accident or malfunction or the consequences of an accident or malfunction is not increased. Therefore, the margin of safety as defined by the plant licensing basis will not be reduced.

SC-92-169 SE93-0473 REPLACEMENT OF FUEL HANDLING AREA RADIATION MONITORING CHANNELS RE/RA/RIA:...2316 AND 2317 RE-2316 and RE-2317 are being changed from Victoreen Model 847-1 analog detectors/preamps to Victoreen Model 977-1 digital detectors/preamps. RIA-2316 and RIA-2317 are being changed from Victoreen model 846-1 readouts to Victoreen. Model 946A readouts. The Fuel Handling area radiation monitors are more commonly known as the Refueling containment isolation radiation monitors. The RIA's are used to close containment isolation valves that are not required for engineered safeguards, following

. an unexpected change in radiation levels during refuelfng operations. The RIA's are.

only used during refueling and each isolates one channel of containment isolation valves.

Safety Analysis Summary There are two previously evaluated accidents in the FSAR which occur during refueling or fuel handling. The accident analyses are found in sections 14.3.1 and 14.19.

Section 14.3. 1 deals with "Boron Dilution During Refueling" and section 14. 19 deals with a "Fuel Handling Incident." The isolation of containment valves (which are not required for engineered safeguards) is not an initiator of either of these accidents; this change, therefore, will not increase the probability of their occurring.

The consequences of an accident previously evaluated in the FSAR will not be increased. The replacement of RIA-2316/2317 with digital readout will not effect the out come of either of the fore mentioned accidents. The containment isolation valves are not used in the "Boron Dilution During Refueling" accident. The containment

. isolation valves are mentioned in the*"Fuel Handling Incident" accident analysis. The analysis states "If the assembly were dropped in the containment building, the release would trigger the area alarms resulting in. an automatic isolation of the reactor building ventilation system, resulting in near zero release to the environment if the equipment hatch is closed. This analysis is referring to the Fuel Handling Area radiation monitors, but the automatic containment isolation may be irrelevant; the* Plant'.s equipment hatch is open during refueling operations. In this condition the Spent Fuel Rod Area ventilation charcoal filters will limit off-site releases.* Therefore, the failure of the new digital radiation monitors will have no effect.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The replacement of RIA-2316/2317 will. not increase the probability of a malfunction to the containment isolation system. The new radiation monitoring system will be connected to the containment isolation system using existing wiring and in the same electrical configuration as the old radiation monitoring system. It will also use the existing failure logic and fail safe conditions, one-out-of-two logic for channel isolation and valve closure on high radiation alarm condition or monitor power failure. Also

,based on Plant operating experience; the digital systems are more reliable than the analog systems.

The consequences of a malfunction of equipment important to safety will not be increased. As stated above the refueling isolation radiation monitors are used to isolate non-essential containment valves only. Therefore, will not affect the safe shutdown of the Plant.

The possibility of an accident of a different type than any previously evaluated in the FSAR will not be created. A malfunction of RIA-2316/2317 could leave the containment isolation valves in either the open or closed position. Neither one of these valve positions would initiate an accident or cause it to propagate. The containment isolation valves being closed is the fail safe and most conse~vative mode. The valves being opened or falling to close is accounted for in the Plants design. FSAR sections 9.8.1 states that the containment building and fuel handling area are designed for containment of radioactive particles. The exhaust air from these areas is ducted to high-efficiency filters to minimum activity levels for the stack discharge and to maintain containment of radioactive particles in these areas of possible contamination. The fuel handling area also has a charcoal filter in parallel with the high-efficiency filter which is placed in operation during fuel handling operations.

The possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created. From a radiation monitoring system point of view, the failure of a digital monitor is not different than the failure of an analog monitor. The digital and analog monitors both rely on contact closure to cause the containment isolation valves to close. The Victoreen 900 series radiation monitoring systems have been in service at the Plant since 1988 and there hasn't been any evidence of a common mode failure in any of the numerous locations were they are installed. We do not have the radiation monitoring systems' software V&V, but based on our past operating experience we don't think it is needed. As stated above, the safety significance of these radiation monitoring systems is so small that a software V&V program is not required.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT.

Facility Changes, Specification Changes, Temporary Modifi cat fons and Other Documents The change to a digital radiation monitoring system will not impact the margin of safety of the Plant. The new digital radiation monitoring system will not be connected to any engineering safeguards valves. If the radiation monitors failed to close the containment isolation valves; the malfunction is evaluated in FSAR section 14.19.3.3.. "For the worst case analysis the entire source term, as described on Table 14. 19-1, is assumed released via the Plant stack with no credit for plate-out or cleanup. Table 14.19-2 presents this worst case results. As can be seen, the worst case dose is within the guidelines set by 10 CFR 100. In the case of the actual incident however, the dose rate is expected to be significantly less." Therefore, just like the old analog radiation monitoring system it will not have an impact on the margin of safety of the Plant.

SC-92-171 SE93-1391 HOGGING AIR EJECTOR CONTROLLER REPLACEMENT Specification Change SC-92-171 will upgrade the existing controls for the Hogging Air Ejector. The changes consist of replacing the old controller, PIC-0632 and simplifying the circuit by eliminating extraneous components such as the position switch on PCV-0632 and the solenoid valve on CV-0634.

Safety Analysis Summary The Hogging Air Ejector is not considered in FSAR Chapter 14, "Safety Analysis."

Therefore, the probability or consequences of an accident previously evaluated cannot be increased.

There are no requirements for the Hogging Air Ejector in the Palisades Technical Specifications. It is not safety related, and not subject to the Palisades QA program.

No safety related structures, systems, or components are affected by the control circuit changes. However, the Palisades Appendix R Post Fire Safe Shutdown Analysis does refer to use of the hogger for emergency steaming operations. These are accomplished under various emergency operating and off normal procedures, The changes to the controls do not impact the design functions. They are to be made only to improve reliability and restore automatic operation to the system, as originally designed.

Therefore, the probability or consequences of malt.unctions of equipment important to sat ety cannot be increased.

The changes are insignificant to operation of the plant. The normal operation of the hogger is unaffected, and the ability to perform emergency steaming operations is preserved. Therefore, the possibility of an accident or malfunction of a different type is not created and no safety margins are reduced.

. CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents SC-92-177 SE93-0321 ATMOSPHERIC STEAM DUMPS' DRIP LEG STEAM TRAP SYSTEMS MATERIAL SPECIFICATION CHANGE Replace the Atmospheric Steam Dumps' drip leg steam trap systems with a chrome-moly material. Existing systems are carbon steel. Erosion corrosion is the root cause f<;>r the numerous leaks we have experienced. Chrome-Moly is a corrosion resistant material.

Part of SC-92-177 replaces the existing Armstrong inverted bucket type steam trap with a Bestobell steam trap. The Bestobell steam trap has an internal wye strainer as part of its design. Therefore, the existing steam trap inlet wye strainers (YS-0789, 0790; 0791, & 0792) will be eliminated.

Safety Analysis Summary

. SC-92-177 replaces the Atmospheric Steam Dumps (ADVs) drip leg steam trap systems' piping and components with a Chrome-Moly material. The material specification change will have no effect on the previously evaluated accidents.

Th~ existing steam trap inlet wye strainer will be replaced by the steam traps internal

wye strainer. This wye strainer replacement is functionally and diagrammatically equivalent to the existing. Therefore, this change will have no effect on the above previously evaluated accidents.

The ADVs drip leg steam trap inlet piping is safety related, ASME XI boundary. The bypass valves' safety related function is to close and be pressure retaining. The inlet valves' and piping's safety related function. is retaining pressure. The steam traps' safety related function is to pass fluid and retain pressure.

The replacement piping and valves are identical in fit, form, and func:tion. The material specification change upgrades the systems to a corrosion resistant material, which will minimize future leaks. Therefore, the probability or consequences of a malfunction of equipment important to safety will not be increased.

The replacement ADVs' drip leg steam trap systems are functionally equivalent to the original systems. The replacement steam trap and material specification change is considered a system upgrade and more reliable systems will result. Therefore; the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR will not be created.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes~

Temporary Modifications and Other Documents SC-92-191

.SE93-0469 REPAIR HANDRAIL IN CONTAINMENT AT 607' ELEVATION The handrails provide safety for personnel around the craneway opening on 607'

elevation in containment. Currently the amount of carbon steel in containment is 2,210 square feet, and the amount of organic zinc coating in containment is 4,393 square feet. Addition of 2 anchor plates and 10 inches of handrail will increase the a*mount of zinc coating by 10.24 square feet, and this will be an inorganic zinc coating and not an organic zinc coating.

The handrails provide safety to personnel that are near the craneway opening on 607' in containment. During a plant shutdown or accident condition the failure of this has -

no safety related functions.

Safety Analysis Summary The increase in the amount of carbon steel in containment will not increase the probability of an accident, nor will the increase in zinc coating, as evaluated in the FSAR. This is described in the FSAR Section 14.18.1.2.1, Table 14.17.1-2, and Table 14.118.1-1.

The consequences of an accident will not be increased due to the extra handrail and zinc coating being added. The carbon steel handrail and 2 anchor plates will increase the heat sink. The zinc coating being added will increase the amount of hydrogen.*

produced during a locC!I event, by an insignificant amount.

The additions of carbon steel and zinc coating will not increase the probability or consequences of a malfunction of equip-ment.

No new accidents or malfunctions of a different type will be created by increasing the amount of carbon steel and zinc coating. The amount of carbon steel and zinc coating is a very small amount and is insignificant.

The margin of safety will not be affected by the increased heat sink from the carbon steel or the zinc coating. This will have insignificant effects and neither of these will affect Plant Licensing Bases.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents SC-92-199 SE93-1170 MODIFICATION VC-10/VC-ll CONTROL.CIRCUITS Control room HVAC condensing unit vc-1o and VC-11 co~trol circuits will be modified.

An existing relay will be replaced with a time delay relay to provide a timeout period for blocking the compressor low suction pressure cutout during compressor start-up operation. _

Safety Analysis Summary VC-1 O/VC-11 can not directly nor indirectly initiate any of the FSAR Chapter 14 a*ccidents and thus the probability of such an event is not changed.

C-1 O/VC-11 are designed to provide the cooling capacity for the control room.

SC-92-199. will only impact the operation of VC-10 and VC-11. SC-92-1 99 will allow VC-1 O/YC-11 to operate during specific conditions when they would presently trip.

From a control room temperature perspective, it is conservative for VC-10 or VC-11 to be in operation. Therefore, the consequences of an accident will not increase following SC-92~ 199 because VC-1 O/VC-11 wiil be more reliable for maintaining control room temperatures suitable for continued operation of safety related equipment.

SC-92-199 is designed to reduce the number of inadvertent trips of VC-1 O/VC-11 which inherently increases their reliability during normal plant operation. The. new design meets the environmental requirements of seismic and mild environment qualification. EA-SC-92-199-01 provides the technical justification for the Increase.in reliability. The increase in reliability justifies that no increase in the probability of malfunctions of VC-10/VC-11, nor any safety rel~ted equipment which VC-1 0/VC-11 protect, will occur:

VC-1 O/VC-11 do not have a direct effect on failure of a fission product boundary nor are they designed to impact cont~ol-room dose levels. VC-1 O/VC-11 will be vulnerable to combined loss of freon and oil temperature trip malfunctions for approximately 120 seconds during start-up, but the loss of freon alone will cause failure of the condensing unit. The action items for inoperability of one or both condensing units are covered under Standing Order 54 section 3.14 and thus no new consequences are introduced. -

SC-92-199 only involves changes to VC-10/VC-11 control circuits. No failures.

associated with VC-1 O/VC-11 would cause degradation of a fission product boundary and thus none would.be considered accidents.

Coincident loss of freon and compressor oil temperature trip malfunctions could cause long term failure of VC-1 O/VC-11 following this modification, but failure of a compressor is not a new failure mode and the loss of freon alone will disable the condensing unit.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Chartges, Specifitation Changes; Temporary Modifications and Other Documents SC-92-199 will increase the reliability of VC-1 /VC-11 at the expense of reducing redundant compressor protection against loss of freon for approximately 120 seconds during start-up.* But a loss of freon malfunction itself will cause failure of the condensing unit, so this protection is pointless when associated with risk. Since this scenario is the only identified malfunction the modification will impact and the overall reliability of VC-1 O/VC-11 should be increased, the margin of safety is not reduced.

SC-92-220 SE93-0090 ADDITION OF A SUCTION LINE ACCUMULATOR TO C-17 The current design of the EC-42 turbine sample panel Sample Cooling System allows liquid freon to enter the suction of the sample chiller compressor C-17. This liquid freon dilutes the compressor's oil causing it to trip off on low oil pressure.. The addition of a suction line accumulator would eliminate this problem.

Safety Analysis Summary*

.This cooling system is not safety related and will have no effect on. any accident evaluated in the FSAR..

This. Specification Change (SC) will enhance our ability to monitor secondary plant chemistry, therefore reducing the probability and consequences of equipment malfunction.

This Sample Cooler System has no effect on other plant systems, therefore it will not create a new type of accident or malfunction not evaluated in the FSAR.

This system is not safety related and will have no effect on any plant safety margin.

SC-92-223 SE93-0066 REPLACING MV-CRN127 AND 181 Replacement of MV-178 and 181, P-75A _and B suction valves. The present valves are Aloyco gate valves and this SC will replace them with Worcester ball valves.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Fac~lity Changes, Specification Changes, Temporary Modifications and Other Documents Safety Analysis Summary The valves being replaced are in a non-0, non-safety system and do not _figure in any previously evaluated accident, therefore, do no increase the probability or consequences of an accident or malfunction.

Since the valves meet all design requirements of the original valves they can not create an accident of a different type than previously evaluated.

Since the valves are equivalent to original no new malfunction mode is introduced.

The valves have no affect on the margin of safety as defined by Plant Licensing Bases.

SC-92-224 SE93-0102 E-6A AND B VENT REPLACEMENT Feedwater heaters E-6A and B vent headers will be changed from carbon steel to stainless steel and wall thicknesses will be increased. The angle of the vent lines into the main header will also be changed.

Safety Analysis Summary There is no reference in Chapter 14 of the FSAR to the feedwater header operating vents. Therefore, the probability of an evaluated accident will not be increased.

Since the feedwater header vents are not inputs to any FSAR chapter 14 accident they can not effect the consequences of a previously analyzed accident.

The vent header does not connect to or run in the vicinity of any safetY equipment.

Therefore failure of the line will have no effect on equipment important to safety.

The proposed change only changes the vent piping to stainless steel. The modification

meets requirements of pipe and weld class GC-17 and ANSI/ ASME B31. 1, 1 986.

Changing the line will reduce the normal mode of failure, erosion/corrosion, which we have seen. Therefore the possibility of a new type of accident will not be created.

No analysis beyond those performed for ANSl/ASME B31.1, 1986 have been required on the vent lines. Therefore, by meeting the same standards for the modification, no new malfunction modes will be created.

The proposed change does not affect the margin of safety as defined by Plant Licensing Bases. Feedwater heater operating vents are not referenced in the FSAR or Technical Specification. They will not be required during or after a LOCA or HELB, therefore, the proposed change does not involve a reduction to the margin of safety.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Chang~s, Temporary Modifications and Other Documents SC-92-227 SE93-0509 REPLACEMENT OF MV-CVC2070 Replace manual globe valve with manual ball valve. Minor changes.are allowed for connecting reach rod to new valve.

Safety Analysis Summary*

Ball valve will provide better isolation, better flow characteristics and easier access for maintenance. This valve is normally closed and prevents unidentified* PCS leakage from.

filter F-54A. This valve is also used to drain and flush filter F-54A. The only change to the FSAR is a drawing change to show a ball valve versus a globe valve. The affected drawing is Figure 9-25 Sheet 1. No unreviewed safety question exists.

SC-93-006 SE93-0243 COMPONENT COOLING WATER HEAT EXCHANGERS. E-54A*AND E-548. VENT AND DRAIN VALVES REPLACEMENT The "Piping and Instrumentation Diagram, Service Water System" M-208 Sh1A, FSAR Figure No 9-1 Sh 1 A, will be changed to reflect the replacement of MV-SW529, 531, 533, 535 (Globe valves), and MV-SW530, 532, 534, 536 (Gate valves) with ball valves as necessary. Seismic capabilities of Hx are not affected per EA-SC-93-006-01.

Safety Analysis Summary Safety as defined by Plant Licensing Bases will not be reduced by replacing these valves with ball valves. The ball va,ves will function in the same manner as the existing gate and globe valve vent and drain applications.

SC-93-007 SE93-0125 REDUCED SET POINT ON TS-1824. TS-1825 AND TS-°1826 FROM 104°F TO l00°F These temperature switches provide an alarm to the control room that the switchgear

and cable spreading area are in high ambient temperatures. Currently the set point 104°F is set at the design temperature value for these rooms and equipment.

Reducing the set point 4°F provides added time margin to correct these high temperatures.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Safety Analysis Summary The reduction in set point will not increase the probability of an accident as evaluated in the FSAR. This is described in FSAR Section 9.8.2.21 and 9.8.5.2.6. The new set

point of 100°F is conservative with respect to providing increased safety margin. It allows an increase in operator time to respond to the necessary _corrective measures as described in ARP 8 Annunciator #67. The 104°F current value.is the upper design temperature rating for electrical equipment in these rooms. The plant needs to manage the ambient air temperatures in these areas in order to maintain equipment qualifications.

The consequences of an accident will not be increased, these temperature switches actuate an alarm only, and it is passive with respect to equipment operation.

These temperature switches provide no safety related function, their failure or malfunction is not significant to plant shutdown or accident conditions. The reduction in set point will not increase the probability or consequences of a malfunction of equipment.

No new accidents or malfunctions of a different type will be created by reducing the temperature switch set point. The reduction*is conservative and passive.

The margin of safety will be increased by reducing the set point. Thus change will not affect the Plant License Bases.

-SC-93-019 SE93-0667 UPGRADE OF C-42 SAMPLE LINE FILTERS The EC-42 sample inlet filters will be upgraded from 2 sq in to 80 sq in effective filter area. This will increase service life about 40 times (40 weeks). The filters will also be_ -

moved downstream of the Samplex rough coolers to eliminate the burn hazard _

associated with filter element change out.

Safety Analysis Summary There is no accident evaluated in the FSAR for the EC-42 turbine plant sample panel.

A malfunction of these filters will have no effect on safety related equipment.

These filters have no effect on any plant systems other than the sample panel itself, and therefore, will not cause an accident or malfunction involving plant safety equipment.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Neither the filters* nor the panel are safety related and will have no effect on the plants safety margins.

~\\

c~j SC-93-026 SE93-0459 FUEL BUILDING RUNWAY CRANE RAIL MODIFICATION Isolation joint in crane rail between Auxiliary Building and Auxiliary Building addition to be eliminated. This will reduce maintenance requirements of crane and provide safer more reliable operation.

Safety Analysis Summary EA-SC-93-026-01 demonstrates that in accordance with FSAR Section 9.11.4.3 that the fuel building crane will not be dislodged and fall on any equipment or structures

situated below during an Operating Basis Earthquake (OBE) or Safe Shutdown Earthquake (SSE) seismic event, even though the crane rail will in all likelihood fail at the interface between the Auxiliary Building and the Auxiliary Building addition.

EA-SC-93-026-01 also demonstrates that the loads transmitted from one building to.

another during a seismic event will be insignifica.nt because of the crane rail failure. As the two buildings remain isolated at.all other locations, they will still act independently of each other. The seismic analysis of either has not been invalidated. Requirements of FSAR 5. 7.3.3 are* still met.

Splicing the crane rail in the Auxiliary Building to the crane rail in the Auxiliary Building addition has no significant impact on operation of the crane, other than reducing maintenance requirements.- Operating criteria, particularly lift capacity has not been changed.

The crane has_no safety related function. It is safety related only in that it must not be

. dislodged and fall down on any safety related equipment/structures situated below it.

EA-SC-93-026-01 demonstrates that crane will still satisfy this criteria, even with the implementation of this modification.

Therefore probability or consequences of accidents or malfunctions are unaffected. No new type of accidents or malfunctions are created, and margins of safety are not reduced.. __ _J

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents SC-93-027 SE93-0577 RECONNECT EXISTING CHLORINATION SYSTEM TO ABANDONED CHLORINATION PIPING Reconnect existing chlorination piping to abandoned chlorination piping to piping for the.

purpose of chlorinating the intake structure for zebra mussel population control.

. The original chlorination system injection points were upstream of the traveling screens.

By injecting upstream of the traveling screen the entire intake structure was **

chlorinated. The cooling tower modification abandoned this portion of piping and moved the injection points to the service water bay and cooling tower lines. This Specification Change (SC) reconnects the abandoned piping in order to control the.

zebra mussel infestation of the intake structure.

Safety Analysis Summary**

The FSAR does not address any accident associated with the Chlorination System. Nor does the FSAR address chlorination in any accident scenario to any system that interfaces with the Chlorination System.

Due to chlorination of the intake structure for the control of zebra mussels, the probability or consequences of malfunction to equipment important to safety (i.e.;

service water pumps, fire pumps) will not be increased.

This SC reinstates the original chlorination design of injecting upstream of the traveling

screens. The existing chlorination injection points will also remain as is. Injection of hypochlorite is not a safety related function. Any failure of the PVC piping below the floor level would be caught by the traveling screens, Therefore, the probability of an accident of a different type than previously evaluated will not be created.

The control of zebra mussel population is important to assure proper operation of equipment using water from the intake structure. This SC is being written for the sole purpose of controlling the zebra mussel population of the intake structure. Therefore, the possibility of malfunction of a different type than any previously evaluated will not be created.

No Tech Spec requirements apply to the Chlorination System. lherefore, the margin of safety as defined by Plant Licensing will not.be reduced.

CONSUMERS POWER COMPANY *- PALISADES NUCLEAR PLANT Facility thanges, Specification Changes, Temporary Modifications and Other Documents SC-93-032 SE93-0682.

SE93~1387 SE Rev 1 INSTALL A WELDING OUTLET AND LIGHTING FOR THE SPENT FUEL POOL {SFPl TOOL CRIB This modification involves replacing the existing nonsafety related circuit for the ultrasonic cleaners (no longer used) with a welding outlet circuit-and an SFP tool crib lighting circuit. All breakers arid wiring will be upgraded as necessary.

Safety Analysis Summary Neither the probability nor the consequences of an accident or malfunction previously evaluated in the FSAR will be increased because the welding outlet and lighting for the SFP tool crib will be fed from P-53 which is fed from 1 E MCC-81; breaker 52-8121.

These loads only replace the existing load shown as "Ultrasonic Cleaners Power Panel P53" on FSAR Figure 8-10. All breakers and wiring sizes will be sized properly for the loads. All supports will be seismically analyzed.

Neither the possibility of an accident nor a malfunctio~ of a different type than any previously evaluated in the FSAR will be created because the change only involves*

replacing the existing non safety related circuit shown as "Ultrasonic Cleaners Power Panel P53" on FSAR Figure 8-10 with a welding outlet and lighting circuits for the SFP tool crib.

The_ margin of safety as defined by Plant Licensing Bases will not be reduced because the change will not cause any equipment to operate outside the design capability.

sc~93~03s

. SE93-1038 PERMANENT DISCONNECTION OF THE ALARM PORTION OF PIA-1066 PIA-1066 alarm function is being disabled because it is a distraction to the control operator. The alarm is in continuously because it is seeing system pressure due to valve leakage. Only the alarm portion is being disabled not the indication portion. PIA-1066 does not perform a safety function.

Safety Analysis Summary The probability or consequences of an accident previously evaluated in the FSAR will not be increased by this change. The alarm is used to alert the operator of a potential off-normal condition.

CONSUMERS POWER COMPANY ~ PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The probability or consequences of malfunctions of equipment important to safety will not be increased by this change. The alarm is used to alert the operator of a potential off-normal condition. However, the pressure indication is still present. The existing alarm* condition provides no indication of an off-normal condition.

The possibility of an accident of a different type than any previously evaluated in the FSAR is not created by this change. The possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created by this change. This alarm is used only to.alert the operator of a potential off-normal condition.

This change will not reduce the margin of safety as defined by Plant Licensing Bases.

No margin of safety is associated with this alarm.

SC-93-044 SE93-0844 REPLACEMENT OF TIC-0821 SC-93-044 is replacing component cooling water (CCW) temperature controller TIC-0821. This controller provides nonsafety related automatic temperature control of the CCW between 65°.F and 90°F by regulation of service water through CV~0821 and CV-0822 to compensate for short term fluctuations in service water temperature and operational heat loads. Gross adjustment is also required due to seasonal temperature variations in the service water and is accomplished by adjustment of the hand-operated butterfly valves in the heat exchanger service water outlet headers. The original temperature controller had a capillary and bulb inserted into a thermowell to monitor CCW temperature at a common point in the outlet of the two CCW heat exchangers.

The replacement temperature controller will have an RTD inserted into the existing thermowell to monitor CCW temperature. Both controllers provide a 3-15 psi pneumatic output to control CV-0821 and CV-0822. Control valves CV-0821 and CV-0822 are also designed to go closed to conserve service water upon a recirculation actuation signal RAS which occurs when the SIRW tank level switches sense that the tank level is low. This design function will not be changed. Controller mounting is also being modified to.satisfy Civil/Structural Analysis EA-SC-93-044-02.

. Safety Analysis Summary Temperature controller TIC-0821 only provides automatic temperature control of the CCW for short term fluctuations of service water temperature and operating heat loads.

Temperature controller TIC-0821 is not discussed as a safety related instrument used to mitigate any accident. The temperature control valves, CV-0821 and CV-0822, controlled by TIC-0821 are designed to close upon a RAS to conserve service water during an accident and this function will not be changed. Therefore, the probability or consequences of an accident previously evaluated in the FSAR will not be increased.

CONSUMERS POWER COMPANY*. - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,

.Temporary Modifications and.Other Documents The replacement controller is modern equipment designed to be more reliable, will

perform the same function,* and the design of CV-0821 and CV-0822 closing upon a RAS is not being changed. The type of temperature sensing element is being changed

but will use the existing thermowell. Since there are no significant changes, the consequences of a malfunction of equipment important*to safety will not be increased.

Temperature controller TIC-0821 does not perform a safety function. It merely.

. provides automatic temperature control of the CCW for short term fluctuations in

. service water temperature and operational heat loads.. The replacement controller will perform the same function. The two extreme failure modes of TIC-0821 would cause temperature control valves CV-0821 and CV-0822 to fully open or fully close eliminating the automatic temperature control. These two extreme failure modes* are no different than the existing controller. The temperature sensing element is also being replaced but its failure is bounded by the two extreme failure. modes of the con.troller.

CCW temperature is indicated and a*ny abnormal, high or low, temperature would be annunciated in the control room by separate instruments. The function of CV-0821 and CV-0822 closing upon receipt of a RAS, to conserve service water during an accident, is not being changed. Since TIC-0821 does not perfor_m a sat ety function, the replacement controller will perform the same function as the existing controller, and the extreme failure modes of the controller will not be changed, the possibility of an

.. accident or malfunction of a different type than any previously evaluated in the FSAR will not be created.

Temperature controller TIC-0821 does not perform a safety function and the replacement temperature controller will be performing the same function as the existing obsolete~ controller, therefore the margin of safety as defined by Plant Licensing Bases will not be reduced.

SC-93-050 SE93-0848

.CHANGE MOTOR PINION/WORM SHAFT GEAR SET OAR AND REPLACE SPRING PACK IN M0-3062. M0-3064. M0-3066 AND M0-3068 During MOV program preliminary calculations" it was determined that the minimum

.required thrust/torque to close was nearly equal to maximum actuator torque. The motor to actuator gear ratio (OAR) was changed to increase actuator torque. Also, the spring pack was replaced. The hew model spring pack increased the band available for torque switch setting.

Safety Analysis Summary Modification meets m*anufacturers standards. Although OAR changes valve stroke, time falls within Tech Spec and Safety Analysis limiting stroke times.

The component could not initiate an accident or malfunction or increase consequences beyond that already addressed in the FSAR.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes; Temporary Modifications and Other Documents OAR and spring packs are approved vendor parts. New OAR and spring improves actuator torque and torque switch set ability.

No new accident or malfunction possibilities are introduced. New OAR torque

*developed has been shown to be below for valve actuator limitations.

SC-93-054.

SE93-0843 SERVICE WATER TEMPERATURE CONTROL ON CONTAINMENT AIR COOLERS The containment air cooler temperature control valves are 4" bypasses around the 8".

.. high capacity outlet valves. The temperature controller function is to maintain a containment air cooler outlet temperature of 60°F (set point of controllers). Due to instrumentation problems, these temperature controllers have not maintained a consistent containment air temperature. The verification of control function cannot be accomplished during plant operation since all controls are located inside containment.

An evaluation of the temperature controls performed during the 93 REFOUT showed the controls to be only *marginally adequate. Removing the controllers and operating the temperature control valves full open at all times will assure adequate containment air cooling. An isolation valve is maintained for manual control of any of the three temperature control valves.

. Safety Analysis Summary A review of all Chapter 14 FSAR events showed that credit is not taken for the containment air cooler temperature control valves CV-0838, CV-0863 and/or cv~0872 or assoCiated instrumentation. The main service water air cooler high capacity outlet valves are not affected in any way through this modification. The containment air cooler function to maintain containment air temperature to within desired limits in accident and/or non-accident conditions has been maintained. Therefore, this modification does not affect probability or consequences of any previously evaluated accidents.

The temperature control valves are to be positioned fully open on VHX-1, 2 and 3.

VHX-4 is failed closed and is not affected through this modification. The failure of any of the three temperature control valves to remain open will cause the containm~nt air *

temperature to rise. Upon detection, the high capacity valve on any of the 4 coolers can be manually opened from the control room to increase service water flow thus lowering containment air temperature. In addition, the accident mode for the service water flow to VHX-1, 2 and 3 is to have the inlet and outlet valves full open for maximum cooling capability. The bypass valves being open during an accident will not reduce the cooling capability of VHX-1, 2 and/or 3 since the water will take the path of least resistance* (8" valve). This operational configuration is really no different than the existing configuration with temperature controllers holding the bypass valve open to some position. The probability or consequences of malfunctions will not be increased.

. CONSUMERS POWER COMPANY - PALiSADES NUCLEAR PLANT.

Facility Changes, Specification Changes, Temporary Modifications and Other Documents Per the discussions above, this modification will not create the possibility of an*

accident or malfunctiori of a diff~.rent type.

Operating the VHX-1, 2 and/or 3 containment air coolers with the temperature control vahies fully open does not violate any licensing basis for the plant, thus the margin of safety is not reduced.

SC-93-087 SE93-1217.

SE92-1244 SE R/l WELD MODIFICATION OF TE-0101 AND TE-0102 NOZZLES The modification will allow a very small amount of primary coolant into contact with a very limited area of the pressurizer low alloy carbon steel pressure vessel shell.. This primary coolant will not circulate through the carbon steel area, but will be frapped in the location between the inner and outer surfaces of the pressurizer. EA-SC-93-087-01 evaluates the corrosion potential of the above condition in section 4.4.d and determines that, per an ABB/Combustion Engineering letter~ the m. aximum corrosion rate expected for the pressurizer shell material is 3 mils per year. This ABB/CE letter references an

industry study of low alloy steel's corrosion rate in the presence of borated primary

coolant and the test conditions are extremely conservative for the conditions this modification creates. The biggest difference is that the test coupon was in contact with a much larger sample of coolant than the pressurizer shell will be. The only.

coolant in contact with the pressurizer shell will be the coolant that penetrates the nozzles due to existing cracks in TE-0101 and TE-0102. Since the modification will move the pressure boundary.to the outer surface of the pressurizer, the space interior to the outer weld will become pressurized with primary coolant. However, once this space fills, there will be little or no exchange of coolant with the bulk of the pressurizer.

. Since there will be little exchange of coolant there will not be any refreshing of the solution to cause additional corrosion of the carbon steel beyond the very small amount that will occur on initial contact. Whatever small amount of corrosion that does take place will be insignificant in relation to the thickness of the pressurizer shell at.either nozzle location (maximum highly conservative estimate of 3 mils/year versus a 4 inch minimum thickness pressure vessel Wall). Since th.e corrosion is insignificant (and in fact as the above EA also points out in paragraph 4.4.d that industry experience with leaks has shown no significant corrosion or erosion of carbon* steel in contact with borated primary coolant from pressurizer heater leaks that existed for most of a fuel cycle) this modification does not constitute an Unreviewed Safety Ouestior:'.

Safety Analysis Summary This modification does not increase the probability of an accident previously or consequences evaluated in the FSAR. The pressure boundary is reestablished per the ASME code with negligible effects on the pressurizing system.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents This modification does not increase the probability or consequences of malfunctions of equipment important to safety. The modification does not adversely affect the pressurizer.

This modification does not create the possibility of an accident or malfunction of a different type that any previously evaluated in the FSAR.- The original design bases of

the pressurizer are maintained, thus the likelihood of new accidents is not created..

This modification does not reduce the margin of safety as defined by Plant Licensing Bases. As discussed above, even though the carbon steel pressurizer shell will now be exposed to primary co'olant, and will thus* experience some c*orrosion, that corrosion will be so slight as to be insignificant as compared to the design basis of the pressurizer shell. Thus the margin of safety is not reduced.

TEMPORARY MOD/FICA TIONS

CONSUMERS POWER COMPANY PALISADES.NUCLEAR PLANT Facility'Chahges, Specification Changes, Temporary Modifications and Other. Documents TM-92-058 SE93-0710 TEMPORARY FIRE WATER SUPPLY TO WESTINGHOUSE TRAILERS ON TURBINE DECK A tee will be installed to allow Fire Protection System to supply water to installed sprinklers in temporary trailers on the turbine deck. It will supply static pressure and only flow water in the case of a fire actuating the sprinklers.

Safety Analysis Summary The function of the Fire Protectfon System water supply is not considered in any.

previously evaluated accident. Therefore it can not increase the probability or consequence of any accidents.

While the Fire Protection System is important to safety of the plant equipment the Temporary Modification (TM) will not effect the level of protection. In the event of an actuation of the trailer sprinklers or in the event of a hose rupture we could not exceed

fire pump capabilities. We also have the ability to isolate the temporary installation if a failure develops.

The possibility of a new accident or malfunction not previously evaluated in the FSAR will not be created. A failure of the hose or actuation of the sprinkler will have no significant effect on the Fire System. We will have the ability to close MV-FPS521 restoring the Fire System to its design configuration.

Margin of sat ety as defined by Plant Licensing Bases will not be affected by this modification.

TM-93-003 SE93-.0107 SE R/1

sE93-0083 BLANK FLANGE INSTALLED IN PIPE LINE MB-1-3"

A elbow in pipe line MB-1-3", just upstream of CV-1301 and CV-1302, will be removed and a blank flange installed in the upstream side of the opening. The Temporary Modification (TM) will allow chlorination qf the Service Water System, while ensuring no sodium hypochlorite is able to enter the cooling tower circulating water piping, during maintenance on that system.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Safety Analysis Summary The probability or consequences -of an accident previously evaluated will not be increased because the.affected non-critical service water line is a small (1 % " diameter

- at FIS-1338), low pressure (75 psig design pressure) line. It is very unlikely that the blank flange would fail.. The TM does not affect the capability of the non-critical service water header to isolate.

The probability or consequences of malfunctions of equipment important to safety will not be increased because the TM does not directly affect any safety related equipment.

Even blank flange failure would not cause any safety related failure because the.flow

would not be great enough to cause service water pump damage. The TM.does not change, degrade on prevent any safety related actions from occurring.

The possibility of a different type of accident or malfunction will not be created because; the TM only affects noncritical service water. The blank flange removes no safety related equipment from service and the blank flange could only fail in a manner similar to other piping component failures.

The margin of safety will not be reduced because the TM is in noncritical service. water piping which will be isolated during an accident.

TM-93-004

_SE_9_3_-0_0_9_1 ____________________________________________________ ~

INSTALLING THERMOCOUPLES ON CABLES IN TRAY TK-032 AND INSTALLING FIBERGLASS INSULATION UNDER THE COVER; AND INSTALLING THERMOCOUPLES ON CABLES IN ADJACENT TRAY TK-030 Two issues to be evaluated are placing thermocouples on cable jackets and placing fiberglass insulation in the cable tray.

Placing the thermocouples on the cable will cause no change to the cable or cable tray, they remain in their normal operating configuration.

Placing the insulation in the tray, beneath the trays' cover, will have an expected*

minimal impact on the cables ability to dissipate heat, making this Temporary Modification (TM) a test not described in the FSAR. During a previous similar test the average hot spot temperature was recorded as 47.6°C on the.cable insulation. Add 1 °C temperature increase to account for the 6. T between the cable conductor and insulation skin, and we have a conductor temperature of 48.6°C. This is for a conductor design temperature of 90°C maximum.

i L

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes) Specification Changes, Temporary Modifications and Other. Documents Safety Analysis Summarx

. The plant operates with cables sized in accordance with IPCEA or NEC ampacity values, where the ampacity listed is the amperage that will produce the design conductor temperature*. As described above the cables were operating much below their design temperature. Because of the large margin a small change in the convection cooling characteristics will not *cause the cable temperatures to exceed their design temperature. This is also shown by test data where temperatures were monitored on cable jackets with a cable tray cover and fiberglass insulation added at the cover ends.

The thermocouples remained as during the open top tests. Cable temperatures increasep less than + 5°C.

To ensure that temperatures remain below the ca_ble design temperature, the TM requires that after installing the fiberglass insulation, the datalogger will be monitored.

until thermal equilibrium is reached, defined as no temperature increase for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Should equilibrium not occur below 85°C instructions say to remove the fiberglass insulation, discontinue the test and allow temperatures to return to normal. It should be noted that temperatures discussed here are those recorded on the jacket of the three (3) most heavily loaded circuits in the cable tray.

TM-93-009 SE93-0133 TEMPORARY CONSTRUCTION POWER FROM V-49A OR V-498 OUTLETS IN CONTAINMENT 649' ELEVATION Additional power is needed in containment at 649' elevation. Temporary Modification*

(TM) adds 480 Volts 3 phase temporary construction power from outlets normally used for CRDM fans V-49A and 498.

Safety Analysis Summary

C_RDM fans V-49A and V-498 are not needed or used during plant outages. This Temporary Modification uses outlets (480 Volt, 30), which normally supply V-49A and V-498, to supply additional construction power in containment at 649' elevation.

CRDM vent fans are not reql1ired for safe shutdown or for plant shutdown activities.

This temporary modification will have no effect on any postulated accidents or equipment malfunctions as described or different than described in the FSAR. This temporary modification will be removed prior to start-up.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-010 SE93-0134 PROVIDE TEMPORARY POWER TO 52-206 FROM 52-1109 DURING BUS ID OUTAGE This Temporary Modification (TM) will provide temporary power to vital security loads (52-206) from LC-11 breaker 52-1109. A TM is required during Bus 1 D outage.

During Bus 1 D outage, MCC 2 will be de-energized. This TM will supply power to 52-206 to allow continued operation of security loads.

Safety Analysis Summary The probability of an evaluated FSAR accident will not be increased as the temporary modification follows IEEE-384 criteria for isolation. Utilization of existing Class 1 E Breaker 52-1109 from Load Center 11 ensures that electrical isolation criteria for power circuits as defined in IEEE-384 paragraph 7.1 are met. The addition of fused disconnects provides protection for the cables between the disconnect and MCC #2.

Review of the FSAR for loss of off-site power and related Diesel Generator (D/G) failures found the worst case electrical accident to be loss of off-site power and loss of one Emergency Diesel Generator. This Temporary Modification (TM) will not increase the probability of the r13maining Bus or DG being lost (FSAR 14. 14-2). The consequences of the evaluated accident is therefore not increased.

The TM will not increase the probability or consequences of malt unctions of equipment important to safety since IEEE-384 compliance ensures this. Breaker 52-1109 is Class 1 E and serves as the isolation device between load center 11 and loads fed from CC #2 Breaker 52-206.

Compliance with IEEE standards and proper breaker setting and fuse setting ensures the possibility of a different type of accident or malfunction will not be created.

This TM will be installed and removed while the plant remains in cold shutdown.

Breaker 52-1109 and the disconnect fusses will protect Load Center 11 and Bus 1 C.

No Technical Specification margin of safety is reduced.

TM-93-013 SE93-0376 CONTAINMENT TEMPORARY ELECTRICAL POWER FROM TRANSFORMER EX-ISA Provide temporary containment construction power for outage work activities from pressurizer heater transformer.

CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT Facility Changes, Specification Changesi Temporary Modifications and Other Documents Safety Analysis Summary This Temporary Modification (TM) will not increase the probability or consequences of accidents previOl:'Sly evaluated in the FSAR since the load being placed on Bus 1 E does.

not exceed any equipment ratings. This temporary modification will be installed and removed while the plant is in cold shutdown.

This modification will not increase the probability or consequences of malfunctions of

. equipment important to safety. Pressurizer heater XFMR EX-15 is nonsafety related and is not used in cold shutdown. Breaker 152-305 will function as before and protective devices will be set to ensure proper coordination for the temporary circuit

. configuration.

This modification does not increase the.possibility of a different accident or malfunction. The operation of Breaker 152-305 will not be changed. Only the load fed

from the breaker will be changed. Bus and XFMR loading and protective device settings have seen evaluated by Engineering.

TM-93-014 SE93-0168 TURBINE BUILDING TEMPORARY ELECTRICAL POWER Provides temporary construction power for outage work activities from a heater drain pump breaker:

Safety Analysis Summary.

This Temporary Modification (TM) will not increase the probability or consequences of.

accidents previously evaluated in the FSAR since the load being placed on Bus 1 E does.

not exceed any equipment ratings. This modification will be installed and removed while t.he Plant is in cold shutdown.

This modification will not increase the probability or consequences of malfunctions of equipment important to safety. Heater drain pump P-1 OA is nonsafety related. Breaker 152-307 will function as before and protective devices will be set to ensure proper coordination for the added load.

This modification does not create the possibility of a dif.ferent accident or malfunction.

The operation of the breaker and associated equipment will not be changed. Only the load fed from the breaker will be changed.

This TM will not reduce any licensing basis* margin of safety. Bus 1 E availability will not change due to the modification. Since Bus 1 E is not required when PCS is below 325 °, *this TM is not a safety concern.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, **

Temporary Modifications and Other Documents TM-93-017 SE93-0166 RESTORATION OF MATRIX LIGHTS AND DISABLING PIP DEVIATION ALARM FOR CONTROL ROD 8 This Temporary Modification_ (TM) removes control rod 8 inputs to the Primary Information Processor display, jumpers PIP control rod 7 inputs to the PIP control rod 8 display, and jumpers across auxiliary relay contacts for a PIP permissive to the Rod Position Display Matrix.

The TM eliminates the PIP deviation alarm caused by the PIP error for control rod 8.

This will allow Tech Spec testing of control rods to be performed by eliminating the

standing PIP deviation alarm caused by control rod 8 (bad servo). An attribute of the Tech Spec test is to check that a deviation is produced by moving a rod *within a rod group. Verification of _an operable deviation alarm is not a Tech Spec requirement.

This modification allows monitoring of rod position with placement of jumpers across the auxiliary relay contacts. The Rod Position matrix Display contains four lights for each rod; Red for fully withdrawn (requires PIP and position switch contact logic), Blue for withdrawn (requires PIP contact logic), White for inserted (requires PIP and position switch contact logic,), and Green for fully inserted (requires PIP and position switch contact logic). Since the PIP position sensing servo for control roc:l 8 is working*

improperly, jumpering the PIP.auxiliary relay contacts w_iil allow the Red and Green lights on the Rod Position Matrix Display to properly operate via the position switches.

PIP control rod position indication for control rod 8 will remain inoperable and be identified as such by Operations via a Caution Tag. SPI indication will be relied upon for position indication of control rod 8.

Safety Analysis Summary Probability or consequences cannot be increased by this change._ No accident can be initiated by failures of primary indication or loss of deviation alarm. SPI remains available so Tech Spec requirements for indication of rod position remain satisfied.

Even with loss of PIP rod position indication and the PIP-generated deviation alarm, SPI remains available for monitoring rod position. Technical Specification only require a type of position indication and has no requirement for PIP-generated deviation alarm.

Tech Spec-requirements continue to be met. Since the changes being made have no impact on SPI, probability of its failure is unchanged. Equipment which is affected by

the TM is not required for continued plant operation. Therefore, probability of malfunction of that equipment required by Tech Specs remains unchanged.,,

.,,..,....,,y.....

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modificattons and Other Documents Since rod position monitoring is still performed by normal installed equipment in a mode.

permitted by Tech Specs, and the ability to move or trip control rods cannot be affected, the consequences of any malfunction of position indication cannot change.

Since rod position indication for rod 8 *remains available and since ability to move and trip the control rod is unaffected no potential for a new type of accident can be created.

Since all normal equipment is being used for position indication and all Tech Spec requirements will be sa.tisfied, no new conditions or design features are created which could fail in a different mode than those already considered. Since the Red (UEL) and Green (LEL) light indications to the operator will be normal, there is no potential for misleading information to cause inappropriate operator action.

Since no accident types or consequences can be affected by the TM, no safety or design margins can be affected. All equipment used by operators for rod position and deviation monitoring is normally installed equipment and all Tech Spec requirements for equipment availability remain satisfied.

TM-93-019

sE93-0189 INSTALL IMPROVED PLUG IN CONTAINMENT AIR COOLER VHX~2 Expandable tube plugs which can be encapsulated will be installed in a leaking 3" cleanable return bend to seal a pin hole leak from the service water pressure source.

The return bend is part of a cooling coil in VHX-2. This TM removes a portion of

leaking coil from service; it does not repair the coil. The coil will be repaired in the next refueling outage.*

Safety Analysis Summary Plugging a leak in a cooling coil return bend will not increase the likelihood of an accident. The plugging is a simple interference fit process that will only affect VHX-2.

The only system affected is the Service Water System, for which ther.e are not evaluated accidents.

Plugging one return bend will not increase accident consequences because of the minute impact created. The total loss of cooling capacity for one serpentine coil subsection being isolated would be equal to 0.09% of the CAC's total surface area.

Also mitigating the loss is the excess margin of service water flow ( 114 gpm per data for Special Test T-216) which.would lessen the effect on one coil subsection loss.

Service water flow margin is equal to 2. 1 % of required flow. Special Test T-318, showed air flow to the cooler at 38,250 cfm which is 27.5% above the minimum accident flow.of 30,000 cfm. Therefore the total imp_act of losing 0.09% of the cooling area is negligible compared to the total margin of the cooling system.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The probability of malfunctions of equipment important to safety will not be increased.

Tube plugging is a proven repair technology that has been used many times at Palisades without causing equipment malfunction. These expandable plugs, which are encapsulated in the cooling coil reduce the probability of a plug failure resulting in a significant service water leak. The worst case leak due to a plug failure would merely be the original pin hole leak.

Consequences of equipment malfunction will not be increased because the tube plugs can not cause a higher level of equipment malfunction, i.e., more than one tube subsection can not fail if the plugs fail.

No accident or malfunction of a different type than previously evaluated would be created since the only affected equipment would still be able to perform it's intended safety function.

No. margin of safety reduction would result since the margin present in service water flow and air flow will easily make up for the minute loss of CAC surface.

TM-93-023 SE93-0230 TEMPORARY MODIFICATION TO Fl-1073 Place a blank flange on inlet to Fl-1073. This will allow*restoration of Primary Makeup Water (PMW) to the Chemical and Volume Control System (CVCS) while work is being performed on MV-CRW193 (T-61 inlet).

Safety Analysis Summary Since the blank flange will replace the function of normally closed valves which would be closed in all analyzed accidents, it does not increase the probability of a previously evaluated accident happening.

Failure of the blank flange could result in a loss of PMW makeup to the CVCS. Loss of PMW will not increase the consequences of any evaluated accident.

The blank flange will not effect any equipment important to safety. Therefore will not increase the probability or consequences of a malfunction of equipment required for safety.

Installation of the TM will not cause a possibility of any accident. The blank flange can not increase the possibility of an accident or malfunction different from any previously evaluated.

The margin of safety as defined in the Plant Licensing Bases will not be reduced by this TM.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specificatio~ Changes, Temporary Modifications and Other Documents TM-93-026 SE93-0275 REROUTE E-208 DRAIN The condensate drain line on E-208 gland seal condenser will be cut and rerouted to a turbine building floor qrain. This is non-safety balance-of-piant equipment.

Safety Analysis Summary Loss of E-208 can only cause a loss of gland sealing vacuum which would only cause steam loss to atmosphere and does not affect any equipment important to safety.

TM-93-029 SE93-0350 -

INSTALL ENCAPSULATED TUBE PLUGS Encapsulated expandable tube plugs will be installed in a leaking 3" cleanable return.

bend to seal a pin hole leak from the Service Water (SW) pressure source. The return bend is part of a cooling coil in VHX-2. This Temporary Modification (TM) merely removes a portion of the leaking coil from service; it does not repair the coil. The coil will be repaired in the next refueling outage..

Safety Analysis Summary Plugging a leaking return bend will not increase the likelihood of an accident. The plugging is a simple interference fit process that will only affect VHX-2. The only system affected is the Service Water System for which there are no evaluated

. accidents in the FSAR.

Plugging one return bend in addition to the one plugged under TM-92-106, will not *

. increase accident consequences because of the minute impact created. The total loss of cooling capacity for one serpentine coil subsection being isolated would be equal to 0.09% of CAC's total surface area. Also mitigating the loss is the excess margin of service flow ( 114 gpm per data from Special Test T-216) which lessen the effect of one coil's subsection loss. SW flow margin is equal to 2. 1 % of required flow. Special Test T-318, showed air flow to the cooler at 38,250 cfm, which is 27.5% above the minimum accident flow of 30,000 cfm. Therefore, the total impact of losing 0.18% ( 2 sections isolated) of the cooling area is negligib.le compared to the total margin of the system.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The probability of malfunctions of equipment important to safety will not increased.

Tube plugging is a proven repair technology that has been used many times at Palisades without causing equipment malfunction. These expandable plugs are encapsulated in the cooling coil to reduce the probability of a plug failure resulting in a significant service water leak. The worst case leak due to a plug failure would merely be the original pin hole leak.

Consequence of equipment malfunction will not be increased because the tube plugs can not cause a higher level of equipment malfunction, i.e., more than one tube subsection can not fail if the plugs fail.

No accident or malfunction of a different type than previously evaluated would be created since the only affected equipment would still be able to perform it's intended safety function.

No margin of safety reduction would result since the margin present in service water flow and air flow will easily makeup for the minute loss of CAC surface area.

TM-93-035 SE93-0427 LINE THE BOTTOM OF CABLE TRAY TK-030 AND TK-032 (13' EACH TRAY. 26' TOTAL)

USING MARINITE INSULATING BOARD Three issues to be evaluated are:

Decreased Convective Cooling of Cables - Placing the insulating board beneath the cable tray is expected to have a minimal impact on the cables ability to dissipate heat, causing an insignificant temperature increase.

The change, caused by this Temporary Modification (TM), in convection cooling characteristics is similar to TM-93-004 (worked in concert with this TM), the identical circuits and tray areas were instrumented to monitor temperatures in open - top TK-030 versus covered TK-032. The temperatures in covered tray TK-032, where the circulatory motion of air is disrupted by the cover, was from + 2.2°C to a high of

+ 5.6°C above the equivalent measurement in tray TK-030, the open top tray where no disruption of the circulatory motion of air occurs. Also, the worst temperature recorded was 30.8°C recorded on the cable insulation surface. Add 1 °C (an industry standard for power cable) temperature increase to account for the l1 temperature between the conductor and insulation surface, and we have a conductor temperature of 31.8°C. This for a cable design temperature of 90°C conductor temperature.

Adding Potential Fire Load - The insulating board used is 54 % " square feet of marinite.

Marinite is made. from calcium silicate.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes; Specification Changes, Temporary ModificatiOns and Other Documents

. Marinite is approved for plant use. It's main use is fire stop repair. Per UL Guide CERZ, the material has a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months minimum rating.

Use of marinite meets the intent of Branc~ Technical Position (BTP) APCSB 9.5-1.

The Effect of Adding an Additional 104 lbs to the Tray Hangers and-Supports.

The 104' static load added by the TM is expected to have no impact on the hanger system. The hanger system has a design limit of 100% fill (see EA-TM-93:..035). The worst case tray fill is 26.3%. The weight difference between 100% of cable fill and 26.3 % of cable fill far exceeds the additional 104#.

Seismic survivability of a cable tray and associated hanger system, is dependent upon being welldesigned for static load. As determined by_ EA-TM-93-035 a large margin.

exists between static load of the cable and marinite board and design limit of the cable fill.

Safety Analysis Summary The plant operates with power cables sized in accordance with ICEA or NEC ampacity.

values, where the listed ampacity is the amperage required to produce the cable design conductor' temperature. Because the test recorded temperatures provide a large margin when.compared with the cable design temperatures (30.8°C + 1 °C = 31.8°C versus 90°C) the decreased convection cooling created by this temporary modification is not expected to produce cable tef!1peratures of 90°C, thereby having no significant effect on the circuits routed in the cable* trays.

TM-93-052.

SE93-0660 TIE-IN TEMPORARY AIR COMPRESSOR TO MV-CA200 FOR SUPPLYING PLANT AIR Tie-in temporary air compressors to MV-CA200 to supply air to plant.

Safety Analysis Summary This Temporary Modification (TM) wi.11 not increase the probability or consequences of an accident previously evaluated in the FSAR. Failure of instrument air is evaluated in the FSAR and would not initiate a FSAR accident.

The probability of malfunctions of equipment important to safety will not be increased..

The plant air compressors and Feedwater Purity (FWP) air compressors will remain.

available in standby mode. Therefore, the probability of malfunctions will not increase..

The consequences of a loss of instrument air at power operations (not applicable to this (TM) and Shutdown Cooling (SOC) operation are already discussed in FSAR 9.5.3.3..

I I

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Also CV-3025 and 3055 have 'handjacks' to allow positioning on loss of air supply.

Consequences of malfunctions are not increased.

The possibility of an accident of a different type than any previously evaluated will not be created. Loss of instrument air is analyzed/discussed in FSAR for differing plant conditions.

Loss of instrument air while on SOC is addressed in FSAR 9.5.3.3. Therefore, this TM will not create the possibility of a malfunction of a different type than any previously evaluated in the FSAR.

This TM will not reduce the margin of safety as defined by the Plant Licensing Basis.

TM-93-053 S.E93-647 INSTALL TUBE PLUGS TO ISOLATE A 1 3/4" NON-CLEANABLE U-BEND Encapsulated expandable tube plugs will be installed in cleanable u-bends in order to

. isolate a pin hole leak in a non-cleanable u-bend. The non-cleanable return bend is a

part of a cooling coil in VHX-2. This TM merely removes a portion of the leaking coil from service; it does not repair the coil. The coil will be repaired in the next refueling outage..

Safety Analysis Summary Plugging a leaking return bend will not increase the likelihood of an.accident. The plugging is a simple interference fit process that will only affect VHX-2. The only system affected is the Service Water System, for which there are no evaluated accidents in the FSAR.

Plugging one return bend in addition to the two already plugged will not increase acCident consequences because of the minute impact created. The total loss of cooling capacity for one serpentine coil subsection being isolated would be equal to 0.09

percent of the CAC's total surface area. Also mitigating the loss is the excess margin of service water flow ( 114 gpm per data from special test T-216) which would lessen the effect on one coil's subsection.loss. Service water flow margin is equal to 2.1 % of required flow. Special Test T-318 showed air flow to the cooler at 38,250 cfm, which is 27.5% above minimum accident flow of 30,000 cfm. Therefore, the total impact of losing 0.27% (3 sections isolated) of the cooling area is negligible compared to the

total margin of the system.

~~---

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT FaciJitY Changes, Specification Changes, Temporary*Modifications and Other Documents The probability of malfunctions of equipment important to safety will not be increased.

Tube plugging is a proven repair technology that has been used many times at Palisades without causing an equipment malfunction. These expandable plugs are encapsulated in the cooling coil, reducing the probability of a plug failure resulting in a significa_nt service water leak. The worst case leak due to a plug failure would merely be the original pin hole leak.

Consequences of equipment malfunction will not be increased because the tube plugs can not cause a higher level of equipr:nent malfunction, i.e., more than one tube subsection can not fail if the plugs fail.

No accident or malfunction of a different type than previously evaluated would be created since the only affected equipment would still be able to perform its safety function.

No margin of safety reduction would result since the margin of safety present in service water flow and air flow will easily make up for the minute reduction of CAC surfaces area.

TM-93-055 SE93-0678 INSTALL AND REMOVE SWAGELOK TUBING PLUGS ON FI-0881

swagelok tubing plugs will be installed on the tubing down stream of the Fl's root

. valves (MV-SW635 and MV-SW636) at the existing swagelok fittings. Tubing plugs will isolate service water from a leaking swagelok fitting..

Safety Analysis Summary.

The probability of an accident previously evaluated in the FSAR will not be increased because isolating service water from a non-0 flow indicator will not affect the service water system negatively. There are no previously evaluated accidents caused by the Service Water System. The probability of consequences of a malfunction *of equipment important to safety will not be increased because the Temporary Modification (TM) will only affect a non safety related component. The instrument has no safety function..

The TM will not create the possibility of a different type accident or malfunction

. because the TM isolates a section of instrument tubing that's already isoiated, i.e., the root valves have been closed. The instrument has no safety function.

The TM will not affect the margin of safety as defined by the Licensing Basis because.

the ability of the SWS to mitigate the effects of an accident will not be changed. No SW flow to any safety related component is affected by this TM.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-057 SE93-0709 INSTALLATION OF LEAK STOP DEVICE ON FLANGE WELD PIN HOLE LEAK A temporary leak stop device will be used to seal a pin hole leak in a flange to pipe weld. The flange is the HP-23-16" flange downstream of CV-0826. The code does not allow components having through wall failures to be considered operable.

Safety Analysis Summary The probability of a previously evaluated accident being increased does not exist. No evaluated accidents directly result from a SWS failure.

The consequences of a previously. evaluated accident will not be increased because the only failure that could occur is a resumption of the existing service water leak. The leak is so small it can not affect any other interrelated system.

The probability of malfunctions to equipment important to safety will not be increased because the leak stop device is passive with respect to the SWS and its failure could only result in a resumption of a very small servic:e water leak.

The probability of a different type of accident or malfunction than previously evaluated will not be created because the TM does not affect any required service water flows. *

. TM failure would only result in the very small service water leak going to the floor.

The margin of safety as defined by the Plant Licensing Bases will not be reduced because the TM does not result in any service water flow reduction.

NOTE: Per Generic Letter 90-05, the plant is allowed to use a temporary no.ncode repair to limit leakage from flawed piping. The SWS is a Class 3 low energy piping system.

TM-93-059 SE93-0719 INSTALL FILTER ON DRAIN LINE FROM MV-SFP505, REACTOR SIDE TILT PIT DRAIN A hot spot filter will be connected via rubber hose to the outlet of MV-SFP505. This will allow water to be filtered, to remove suspended radioactive particles, prior to going into floor drain. system. This will help maintain cleanliness of the sump and limit hot

spot buildup later. -------- ----------------- -------

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Safety Analysis Summary No accident previously evaluated can be affected by the addition of a filter unit on the drain line to the floor drain.

No equipment important-to safety will.be affected by the addition of the filter unit.

No accident or malfunction of a different type is created by the addition of this filter unit.

The margin of safety as defined by Plant Licensing Bases is not affected.

TM-93-061 SE93-0734 INSTALLATION OF LEAK STOP DEVICE ON VALVE BODY CMV-SW136) TO PIPE DOWNSTREAM WELD PIN HOLE LEAK A temporary leak stop device will be used to seal a pinhole leak in a valve to pipe weld.

The valve is MV-SW1_36. Code does not allow components with through wall failures to be considered operable.

Safety Analysis Summary Probability or consequences of previously evaluated accident being increased does not exist. No evaluated accidents directly result from a SWS-failure. The only failure that could occur is the resumption of existing leak. Leak is so small it will not affect inter-related system.

Probability or consequences of malfunctions to equipment important to safety will not increase. Leak stop device is passive with respect to SWS. It's failure could only result in resumption of small leak.

Possibility of a different type of accident or malfunction than previously evaluated will not be created since TM affects no other required SW flows. TM failure would only result in resumption of pinhole leak.

Margin of safety defined. by Plant Licensing Bases will not be reduced because TM does not result in any SW flow reductions.

NOTE: Per Generic Letter 90-05, plant is allowed to use temporary noncode repair to.

limit leakage from flawed piping. SWS is Class 3 low energy piping system.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-064 SE93-0763 SERVICE WATER SUPPLY TO FLUSH CONDENSER TUBES A temporary water supply will be connected to P-44 suction line drain (MV-SW 564) to be used for flushing main condenser tubes during cleaning. This connection is non-critical SW and flow rates will be less than 1 00 gpm.

Service water system pipe break is already evaluated iri FSAR and is bounded by a main header rupture. In this case we are limited by the small size of the line (1 ") and the fact we are only going to affect non-critical components during shutdown conditions. Non-critical header can be isolated by closing its isolation valve preventing any effects on critical loads.

Safety Analysis Summary No equipment important to safety are connected to non-critical service water. Non-critical can be isolated from critical at any time during shutdown conditions with effecting any equipment important to safety.

No new accident or malfunction can be causes by this TM. Service water leaks are previously analyzed an we will not cause any other accident or malfunction with this TM.

Margin of safety will not be affected by this TM. Flow rates are very low compared to system capabilities and the tie in is to the non-critical header.

TM-93-070 SE93-1002 SERVICE WATER LEAK NON-CODE REPAIR HB-23-4 11 A pinhole leak has developed in the toe of the butt-weld between the HB-23A" service water piping and valve MV-SW136. The location is downstream of CCW heat exchanger E-548.

A code repair cannot be completed during plant operation or during a normal refueling outage because the leak location cannot be isolated without removing the entire service water system from service. A full-core off load is required to perform a code repair.

1 OCFR50.55a(g) (5 and 6) state that if the licensee determines that conformance with a certain code requirement is impractical, the Commission may grant relief. With respect to 1 OCFR50.55a(g) (5 and 6), Generic Letter 90.05 is being used as guidance for the temporary non-code repair, and the NRC has approved the non-code repair. The repair will consist of welding a 5" schedule 80 pipe (sleeve) over the existing HB-23-4" line.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents Safety Analysis Summary Per review of the FSAR Chapter 14 analyses, welding a sleeve over the existing HP-23-4" pipe could not initiate an accident either directly or indirectly. Therefore the.

modification could not increase accident probability. A NRC approved design and analysis are being used. Per 1 OCFR50.55a(g) a temporary non-code repair is acceptable with NRC concurrence.

Per review of the FSAR Chapter 14 analyses, welding a sleeve over the existing HP-23-4" pipe could not increase the consequences of a previously evaluated accident.

The item does not change, degrade, or prevent any actions described in the FSAR. The modification will actually repair a pipe that is leaking into the CCW room, thus*

improving an* adverse situation. The repair will help ensure the system can perform its intended design function. A NRC approved design and analysis are being used.

The

modification will have no affect on the severity of an accident or mitigation of its -

consequences, therefore consequences are not increased.

The sleeve repair is designed to meet or exceed the strength of existing piping components per analysis. Therefore the sleeve will not fail during normal or accident conditiqns. The modification will help ensure the system can perform its intended design function. Other anticipated events will not cause the SSC to malfunction; the

~epair simply consists of welding a pipe over an existing pipe. Per reference "a",

systems engineering will perform monthly walk-downs to inspect the leak locations.

If the sleeve were to fail, the consequences would be no different than if the pipe were to fail. In fact, the modification is being completed to ensure that the leaking pipe does not fail.

A riew accident which could cause degradation of one or more of the fission product barriers could not occur. The modification is simply a pipe repair. *The modification will be installed with approved welding procedures. An approved design and analysis are being used along with NRC concurrence per 1 OCFR50:55a(g).

The repair could not cause a malfunction of a different type. The failure type would be a leak, in other words the same malfunction as if the pipe were to fail. The modification will be installed with approved welding procedures. An analysis has been completed showing that the sleeve meets or exceeds the strength of existing components in the line to ensure that the possibility of a different type of malfunction will not occur.

Fission product barriers and radiation doses received by the general public are unaffected. Analysis ensures the temporary repair will perform its intended function.

Per the engineering analysis the repair does not violate FSAR Chapter 5 pipe stress limits.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes; Specification Changes, Temporary Modifications and Other Documents TM-93-071 SE93-1004 SERVICE WATER LEAK NON-CODE REPAIR HP~23-16 11 A pinh9le leak has developed at toe of weld between HB~23-16" service water piping and-the slip-on flange just downstream of CV-0826.

A code ~epair cannot be completed during plant operation or during a normal refueling outage because the leak location cannot be isolated without removing the entire service water system from service. A full core-off load is required to perform a code repair. -

1 OCFR50.55a(g) (5 and 6) state that if the licensee determines that conformance with a certain code requirement is impractical, the Commission may grant relief. With respect to 1 OCFR50.55a(g) (5 and 6), Generic Letter 90.05 is being used as guidance for the temporary non-code repair~ The repair will consist of welding a 18... schedule 80 pipe (sleeve) over the existing HB-23-16" line.

Safety Analysis Summary Per review of the FSAR Chapter 14 analyses, welding a sleeve over the existing HP 16" pipe could not initiate ari accident either directly or indirectly. Therefore the modification could not increase accident probability. A NRC approved design and analysis are being used. Per 1 OCFR50.55a(g) a temporary non-code repair is

acceptable with NRC concurrence.

Per review of the FSAR Chapter 14 analyses, welding sleeve over the existing HP 16" pipe could not increase the consequences of a previously evaluated accident. The item does not change, degrade, or prevent any actions described in the FSAR. The modification will actually repair a pipe that is leaking in the CCW room, thus improving an adverse situation. The repair will help ensure the system can perform its intended design function. A NRC approved design and analysis are being used. The modification will have no affect on the severity of an accident or mitigation of its consequences, therefore consequences are not _increased.

The sleeve repair is designed to meet or exceed the strength of existing piping components per analysis. Therefore the sleeve will not fail during normal or accident conditions. The modification will help ensure the system can perform its intended

design function. Other anticipated events will n*ot cause the SSC to malfunction; the repair simply consists of welding a pipe over an existing pipe. Per reference "a",

systems engineering will perform monthly walk-downs to inspect the leak locations.

If the sleeve were to fail, the consequences would be no different than if the pipe were to fail. In fact, the modification is being completed to ensure that the leaking pipe does not fail...*I

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specific~tion Chahges, Temporary Modifications and Other Documents A new accident which could cause degradation of one or more of the fission product barriers could not occur. The modification is simply a pipe repair.* The modification will be installed with approved welding procedures. An approved design and analysis are being used along with NRC concurrence per 1 OCFR50.55a(g).

The repair could not cause a malfunction of a different type. The failure type would be a leak, in other words the same malfunction as if the pipe were to fail. The.

modification will be installed with approved welding procedures; An analysis has been completed showing that the sleeve meets or exceeds the strength of existing components in the line to ensure that the possibility of a different type of malfunction will* not occur.

. Fission product barriers and radiation doses received by the general public are unaffected. Analysis ensures the temporary repair will perform its intended function.

Per the engineering analysis the repair does not violate FSAR Chapter 5 pipe* stress limits.

TM-93-074

SE93-0832 REMOVE BROKEN INSTRUMENT TUBING FOR PX-0653 AND CAP A piece of sensing tubing between the extraction steam line to E-1 A and PX-0653 has broken off. This line is located inside the condenser and is beneath and to the east of the LPA turbine. The exposed ends will be capped.

. Safety Analysis Summary Capping and abandoning the sensing lines for PX-0653 will not increase the probability or consequences of an accident previously evaluated in FSAR. This Temporary Modification (TM) will essentially have no impact on plant operation in any mode.

The probability or consequences of malfunctions.of equipment important to safety will not be increased. The extraction steam pressure nor its indication will not affect safe shutdown or operation of the plant.

This TM will not create an accident or malfunction of a different type than previously evaluated in FSAR. The only possible impact this TM might have is to slightly improve the efficiency of E-1 A as compared to the condition it is now in with the broken tubing and the tubing uncapped.

This TM will not impact the margin of safety as defined by the Plant Licensing Bases.

There are rio radiological or environmental consequences in any operating mode while in this condition.

- 98

-coNSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Chan~es, Temporary Modifications and Other Documents TM-93*079 SE93-0888 BLOCK CLOSED CV-947 Safeguard pump cooling train supply inlet valve CV-947 will be blocked closed to allow work to _be completed on a flow switch downstream of the control valve. The valve is being used as a tagging point to provide worker protection preventing the valve from going open if instrument air is -lost.

Safety Analysis Summary The probability or consequences of an accident or malfunction as described in the FSAR will not increase since.the plant is in cold shutdown. The purpose of the safeguard pump cooling system is to provide seal/bearing cooling during elevated temperatures caused by RAS.

Blocking the valve closed during cold shutdown will not cause a different type of accident or malfunction not described in the FSAR to being created. Seal/bearing cooling is not needed during plant shutdown.

The margin of safety as defined in the Plant Licensing Bases will not be reduced.

TM-93-080 -

SE93-0887

'SEE TM-93-086 TM-93-086 SE93-0975-BLOCK CLOSED CV-0948 Safeguard pumps cooling train supply inl_et valve, CV-0948, will be blocked closed to

_ allow work to be completed on FS-0958. The valve is being used as a tagging point to provide worker protection. The mechanical block will prevent the valve from going open if instrument air is lost.

Safety Analysis Summary--

The probability or consequences of an accident or malfunction as described in the FSAR will not be increased because the plant is in cold shutdown. The purpose of the pump CCW cooling system is to provide pump cooling during -elevated temperatures caused by pumping RAS water and during SDC.

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT F~cility Changes, Specification Changes, Temporary Modifications and Other Documents

. Blocking the valve closed during cold shutdown will not cause a different type of

accident or malfunction not described in the FSAR from being created. Pump cooling is.

not needed during plant shutdown (cold shutdown).

The margin of safety as defined in the Plant Licensing Bases will not be reduced because.the pumps can be run without CCW cooling at PCS cold shutdown temperatures, if needed.

TM-93-088 SE93-0993 REMOVAL OF CHP RELAY 5Pl

. Remove (electrically) CHP relay 5P-1 from scheme 503.. Removal is temporary to allow for electrical testing of 5P-1 relay while maintaining. operation of 5R-1, 3, 5, 7 relays.

Safety Analysis Summary The probability and consequences of previously evaluated accidents (in FSAR) will not be increased. The plant is in cold shutdown with the PCS depressurized. All right channel containment high pressure relays will remain operable. All left and. right channel containment high radiation relays will be operable. NOTE: Left channel CHR relays will be inoperable during TM installation/removal as fuses will be removed in Circuit 503.

The probability and consequences of malfunctions of equipment important to safety will not be increased. Per E-208 Sheet 2 the CHR relays perform the identical functions as the CHP relays. As both CHP and CHR relays are available/operable on the right channel, redundancy is maintained. This TM will* be installed and removed while the plant remains in cold shutdown with the PCS depressurized. All CHR relays (left and right channel) will be operable for refueling operations.

The possibility of an accident or malfunction* of a different type than previously evaluated within the FSAR will not be created. This TM only affects left channel CHP relays 5P-1, 5P-3, 5P-5, 5P-7.. Left and right channel CHR relays will remain operational to isolate containment (via associated valves) should a CHR event occur.

As the plant will remain in cold shutdown with the PCS depressurized, inoperability of the left channel CHP relays does not reduce the margin of safety. All CHR relays will be operational for containment isolation.

100 -

. CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT -

- Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-089 SE93-"1034 OPEN LINKS TO 20/AST

This Temporary Modification (TM) will be in place whil'e the plant is in cold s~utdown.

Whether the turbine is latched or tripped has no affect on secondary or primary side parameters such as flows, pressures, temperatures, etc., when in a cold shutdown condition. Therefore defeating the turbine solenoid trip can have no affect on plant safety. The TM will be removed before leaving cold shutdown and the turbine trfp solenoid will be restored to operability.

Safety Analysis Summary The 20/AST turbine trip solenoid activation could indirectly initiate an accident by tripping the turbine, challenging the reactor trip system and safeguards.. However, the challenges would only happen above 15 % power. This temporary modification will.

only be installed during cold shutdown. By defeating the 20/AST trip solenoid per the

TM, it can not trip the turbine and therefore can not indirectly initiate an accident therefore, the implementation of this temporary modification does not increase the probability of an accident..

Those accidents which are affected by a turbine trip are all evaluated at power. This TM will be in place only during cold shutdown. Therefore the consequences described in the accident evaluations will not be changed by the implementation of this TM.

Rendering the 20/AST turbine trip solenoid inoperative per the TM has no affect on any safety related equipment whatsoever. Therefore, the 20/AST operational state can not increase the probability of safety related equipment failure, nor can it increase the consequences of safety equipment malfunctions.

The capability to trip the turbine from a remote source such as MSIV trip or reactor trip during cold shutdown does not create an accident or malfunction not evaluated in the FS,AR: Whether the turbine is tripped or latched m_akes no difference during cold shutdown because there is no steam flow.

This TM will be in place only during cold shutdown, at which time there is no secondary steam flow. Whether the turbine can be remotely tripped by the operator; reactor trip, or MSIV closure has no bearing on licensing bases because no changes in secondary or primary operating parameters occur whether the turbine is tripped or latched. Therefore, safety margins are not affected.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-090 SE93-0995 INSTALL JUMPER ACROSS CONTACT 3 OF SIS-4 RELAY Contact 3 is a normally closed contact, but voltmeter measurements indicate contact is open. Jumper will temporarily restore circuit until relay is replaced.

Safety Analysis Summary As per E-209 Sheet 1, aside from testing, SIS can only be actuated via CHP. Low pressurizer pressure actuation of SIS is disabled per ESS-E-24. As the-plant is in cold shutdown and the PCS is depressurized, CHP will not occur. As such; the probability of an accident (and consequences of) previously evaluated in the FSAR will not. be increased. The probability and consequences of safety equipment malfunction will not increase.

This Temporary Modification (TM) jumpers SIS-4 contact 3. In effect, this TM will maintain CV-2136 (boric acid storage tank T-53B recirc inlet) open. If an SIS signal should somehow be generated, the PCS is already depressurized and at shutdown boric acid concentration. Recirc of boric acid would be of no consequence. AdditionaJly, boric acid pump P-56A (from T-53A) remains capable of supplying boric acid..

Review of sections listed on the safety review form found no.Plarit Licensing Bases margin of safety reduced.

TM-93-091 SE93-1010 BLOCK CLOSED CV-947

- Safeguard pump cooling train supply inlet valve CV-94 7 will be blocked closed to allow work to be completed on a flow switch down stream of the control valve. The valve is being used as a tagging point to provide worker protection preventing the valve from going open if instrument air is lost.

The probability or consequences of an accident or malfunction as described in the

FSAR will not increase since the plant is in cold shut down. *The purpose of the.

. safeguard pump cooling system is to provide seal/bearing cooling during elevated.

temperatures caused by RAS.

Blocking the valve closed during cold shutdown will not cause a different type of accident or malfunction not described in the FSAR from being created.

Since seal/bearing cooling is nof needed during plant shutdown, the margin of safety as defined in the. Plant Licensing Bases will not be reduced.

- 102 -

CONSUMERS POWER COMPANY

~ PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-092 SE93-1042 BLANK FLANGE INSTALLED ON OUTLET FLANGE OF MV-CHM750 A blank flange will be installed on the outlet flange to MV-CHM750. This blank flange will allow isolation of T-18A for repairs.

Safety Analysis Summary Review of the accidents analyzed in Chapter 14 of the FSAR indicate that no analyzed accidents will be affected by this change. The chlorination system is not identified in any analyzed accident scenario. Therefore installing a blank flange will not affect the

probability or consequences of an accident previously analyzed in the FSAR.

The probability of malfunctions of equipment important to safety will not be increased.

The chlorination system is not safety related. The installation of a blank flange does not have any effect oh the reliability or the nonaccident design basis events of the chlorination system nor does it affect any equipment important to safety. The installation of a blank flange will allow for continued operation of the chlorination system while also allowing for repairs to be made to T-18A.

The consequences of a malfunction of equipment important to safety will not be increased. The chlorination system does not play any role in mitigating the consequences of an accident or malfunction; The installation of a blank flange won't affect the function of the chlorination system since it will use T-188 as the source for sodium hypochlorite.

The probability of an accident of a different type than previously evaluated in the FSAR will not be created. The installation of a blank flange will not cause a degradation to one or more fission product barriers or result in radiological risk to the general public in excess of the 10CFR100 limits. The chlorination system is not involved in any accident scenario and the installation of flange on the outlet flange of MV-CHM750 will not aft ect the operation of the chlorination system.

The possibility of a malfunction of a dif.ferent type than any previously evaluated in the FSAR will not be created. The chlorination system does not perform any safety function. The installation of a blank flange will allow for continued operation of the

chlorination system while at the same time allow for repairs on T-18A. This change will not result in any other failures' which could challenge another safety system or.

introduce any new types of failures.

The margin of safety as defined by Plant Licensing Bases will not be reduced. The chlorination system does not have any direct or indirect effect on fission product boundaries or accident consequences.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-093.

SE93-1063 ADDITION OF VlBRATION MONITORING ACCELEROMETERS TO THE REACTOR HEAD This Temporary Modification (TM) will be adding a passive vibration monitoring system temporarily to the reactor head flange for the detection of vibration (noise) in the reactor. The system will include the addition of three accelerometer~ to the reactor head near the flange. The accelerometers are approximately two inches high and.

weigh about one ounce. They are made of inconel and are rated for high temperature (750 degrees F). They will be mounted with a magnet (strength about forty pounds).

The cable leaving the head will be high temperature coax cable routed with other reactor instrumentation cables. Once the route is far enough from the head, regular coax cable will be used. The cable will pass through penetrations EZ-0242 or EZ-0241 and a signal conditioning junction.box will be mounted seismic 11/1 in the north penetration room. Monitoring equipment will then be hooked up to this junction. box when required.

Safety Analysis Summary The probability or consequences of an accident previously evaluated in the FSAR will not be increased. The vibration monitoring system is a passive, low voltage system which will only be operational periodically. The additional weight on the reactor head is negligible and the magnetic mounting is strong enough to ensure the accelerometer will not become dislodged from its mounting location. The material, inconel, is compatible with the reactor head and containment systems and it will have no affect on existing systems.

The probability or consequences of malfunctions of equipment important to safety will not be increased. The accelerometers will be mounted on the reactor head flange, however, the effect on the head will be negligible. Similarly, the cable will be routed with other instrumentation loops, but the signal is low voltage and will not affect the operation of these circuits. The signal conditioning junction box will be seismically mounted (II/I) on the north penetration room wall.

The possibility of an accident or a malfunction of a different type than any previously evaluated in the FSAR Will not be created. As discussed above, the new system will have no affect on any existing equipment or systems; it will be passive and only in service periodically.

The margin of safety as defined by the plant licensing bases will not be reduced. As supported above, the new equipment to be installed temporarily will not affect the

operation of safety related equipment or systems.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents*

TM-93~096 SE93-1219 TEMPERATURE MONITORING OF THE REACTOR HEAD INSTRUMENT FLANGES This Temporary Modification will involve the installation of twelve thermocouples to the reactor head instrumentation flanges/incores. Three thermocouples will be -installed on each of the following flanges.: 2, 4, 6, 8. One thermocouple will be installed on the incore stalk above the instrument flange. One will be installed directly on the connector between the incore and its extension cable and the other thermocouple will be installed on the extension cable. The thermocouples are the surface measurement type and will be installed with metal wire ties to securely mount them in position. The

.cable will b.e installed only briefly in cable trays and will be routed on the floor to the open hatch which drops down to the 590 elevation where it will be routed to the air room. There a recorder will be temporarily mounted to take data on the environment which the incore cables are subjected to at the reactor head flanges. This information will be used to validate the engineering analysis previously performed to calculate the normal operating temperature which the.cabl13s and connectors in this location have to be qualified to..

Safety Analysis Summary

The probability of an acci_dent previously evaluated in the FSAR will not be increased.

The temperature monitoring system is a low voltage system and all the components are small and nonintrusive. The _additional weight on the reactor head flanges is negligible

and the mounting is strong enough to ensure the thermocouples will not become dislodged from their mounting locations. The material is compatible with the reactor

. head and containment systems and it will have no affect on existing systems. The recorder will be mounted in the air room and will be restrained so as not to become dislodged and damage other equipment.

The consequences of :an accident previously evaluated in the FSAR will not be increased. As discussed above, the tiew temperature monitoring system is compatible with the reactor and containment systems and its low voltage signal will have no affect on existing instrumentation loops. The installation. of the cable and the recorder shall also be such that they will not contribute to other equipment failures.

The probability or consequences of malfunctions of equipment important to safety will not be increased. The thermocouples will be mounted on the reactor head instrumentation flanges, however, the effect on the head will be negligible. Similarly, the cable will be routed through containment, but the signal is low voltage and will not affect the operation of other circuits. Also, the cable will be secured so it cannot mechanically affect other equipment. The recorder will be temporarily mounted in the air.room on an existing work bench.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,.

Temporary Modifications and* Other Documents The possjbility of an accident or a malfunction of a different type than any previously

evaluated in the FSAR will not be created. As discussed above, the new system will have no affect on any existing equipment or systems.

The new equipment to be installed temporarily I will not affect the operation of safety related equipment or systems, therefore the margin of safety as defined by the Plant Licensing Bases will not be reduced.

TM~93-107 SE93-1356 REPLACE BREAKER WITH SPARE Within. Temporary Modification TM-93-'107 breaker 152-113 (HPSI P-66B) will be replaced with spare solenoid breaker 252-SPARE4. This breaker replacement is required to aid in the evaluation of D-PAL-93-235. D-PAL-93-235 was initiated due to the discovery of. a blown closing coil fuse in the 152-113 breaker's closing c9il circuit.

. Within this TM, breaker 152-113 will be removed from its cubicle. The breaker's internal wiring and auxiliary switch contact configuration will.be compared to the vendor print. Breaker 252-SPARE4 has been previously verified to match the vendor print. If breaker 152-113 and breaker 252-SPARE4 are determined to have equivalent internal wiring and auxiliary switch configurations, breaker 252-SPARE4 will be deemed acceptable for installation in cubicle 152-113. Breaker 152-113 will then be inspected to determine if a physical problem exists with the breaker which may have caused the "closing coil" fuse to blown.

Nameplate data for breaker 152-113 (walk-down data recorded 4-1-92) lists its closing current as 73 amps. Nameplate data for breaker 25.2-SPARE4 lists its closing current as 96 amps (walk-down data recorded 8-4..:93). To ensure the increase in listed closing current will not cause actuation of the installed FRN-12 closing coil fuses, FUZ/A.1113-2, FRN-R-15 fuses will be installed in their place. The larger fuses will be installed only while the breaker 252-SPARE4 remains in cubicle 152-113. The coordination. curve for the FRN-12 fuses has been reviewed and shows that the fuses should actuate in approximately 3 seconds when carrying 75 amps. The coordination curve for the FRN-R-1 5 fuses shows that the fuses should actuate in approximately 3 seconds when carrying 100 amps. Thus, increasing the fuse *size from FRN-12 to FRN-R-15 ensures that approximately the same time margin for fuse actuation will exist

. between the normal breaker/fuse combination and the TM installed breaker/fuse combination. The coordination curve of the FRN-R-15 fuses has also been compared against the coordination curve of the DC control circuit feeder breaker (GOULD EH2B 100). **installation of the FRN-R-15 maintains proper fuse/breaker coordination such that the fuse will actuate before the breaker.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT.

F~cilitY Changes, Specif~cation Changes, Temporary Modifications and Other Documents Safety Analysis Summary The probability and consequences of an accident previously evaluated in the FSAR will not be increased. Breaker 252-SPARE4 contains the same style operator as breaker 152-153. As noted above, the internal wiring and auxiliary switch configuration of*

breaker 152-113 will be verified to match that of breaker 252-SPARE4 prior to*

installation of the spare breaker. In the case that the two breakers do not have the same internal wiring or auxiliary switch configuration, breaker 152-SPARE4 will not be installed until the difference is incorporated and documented.into the temporary modification. Larger (FRN-R-15)-fuses will be installed in place of the existing FRN-12 fuses for FUZ/A 1113-2, to ensure that the increased closing current ~equired by breaker 252-SPARE4 will not cause fuse actuation. The FRN-R-15 fuses will maintain fuse/breaker coordination and allow approximately the same actuation time margin (with 252-SPARE4) as the FRN-12 fuses (with 152-113).

The probability and consequences of malfunctions of equipment important to safety will not be increased. Both breaker 152-113 and 252-SPARE4 have identically functioning solenoid operators. The internal wiring and auxiliary switch configurations of the two

. breakers will be verified to ensure that they are the same prior to installation of the

252~SPARE 4 breaker into cubicle 152-113. Increasing the closing coil fuse size, FUZ/A 1113-2, to FRN-R-15 will ensure sufficient time margin for the breaker to operate (close) V¥hile maintaining proper fuse/breaker coordination.

The possibility of an accident or malfunction of a different type than previously evaluated in the FSAR will not occur. Both breaker 152-113 and 252-SPARE4 have identically functioning solenoid operators. Increasing.the closing coil fuse size to operate (close) while maintaining proper"fuse/breaker coordination. Once the TM is in place, Operations will test cycle breaker 252-SPARE 4 in cubicle 152-113 to verify operability.

The margin of safety as defined by Plant Licensing Bases will not be reduced. This TM does not defeat any safety systems or substitute a less capable breaker for use in*

cubicle 152-113. With the exception of the increased closing coil fuse size, the TM essentially substitutes a function ally equivalent breaker for the one normally* installed in cubicle 152-113.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents

TM-93-109 SE93-1372 INSTALL SPARE BREAKER 252-SPARE4 IN CUBICLE 152-204 Within Temporary Modification TM-93-109 solenoid breaker 152-204 (P-7 A breaker) will be replaced with spare solenoid breaker 252-SPARE4. This breaker replacement is required to allow the re installation. of the 152-204 breaker's Push to Trip" rod found

on the 152-204 cubicle's floor. Inspection of the ~Push to Trip" rod and associated parts initially indicates that a cotter pin failed in the trip rod assembly causing the trip rod and associated parts (spring, washer, cotter pin) to fall to the cubicle floor. With the exception of a small cotter pin fragment, all pieces of the trip rod assembly have been recovered. Operations has successfully cycled breaker 152-204 (12-6-93) in the CONNECT position to verify past and present operability. The breaker was not disturbed prior to the operability testing.

The "Push to Trip" rod directly acts on the breaker's trip latch and functions to allow the manual opening of the. breaker in the event of a loss of DC control power. Manual

opening of the breaker can still be performed by partly.depressing the breaker foot lever

. (pedal) which also actuates directly on the breaker's trip latch. Lack of the "Push to Trip" rod has no impact on the electrical functioning of the breaker.

Within this TM, breaker 152-204 will be removed from its cubicle. The breaker's internal wiring and auxiliary switch contact configuration will be compared to vendor print. Breaker 252-SPARE4 has been* previously verified to match vendor print. If breaker 152-204 and breaker 252-SPARE4 are determined to have equivalent internal wiring and auxiliary switch configurations, breaker 252-SPARE4 will be deemed acceptable for installation in cubicle 152-204. Breaker 152-204 will then be inspected for its general condition and have preventative maintenance performed.

Nameplate data for break~r 152-204 is not available as the breaker is presently racked in. Nameplate data (from walk-downs) for other solenoid feeder breakers ( 152-103 P-78, 152-207 P-66A) list breaker closing current as 73 amps. Nameplate data for breaker 252-SPARE4 lists its closing current as 96 amps (walk-down data r.ecorded

8-4-93). To ensure the increase in listed closing current will not cause actuation of the installed FRN-12 closing coil fuses, FUZ/A 1204-2, FRN-R-15 fuses will be installed* in their place. The larger fuses will be installed only while breaker 252-SPARE4 remains in cubicle 152-204. The coordination curve for the FRN~ 12 fuses has been reviewed and shows that the fuses should actuate in approximately 3 seconds when carrying 75

. amps. Th.e coordination curve for the FRN-R-15 fuses shows that the fuses should actuate in approximately 3 seconds when carrying 100 amps. Thus, increasing the fuse size from FRN-12 to FRN-R-1 5 ensures that approximately the same time margin for fuse actuation will exist between the normal breaker/fuse combination and the TM installed breaker/fuse combination. The coordination curve of the FRN-R-15 fuses has

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents also been compared against the coordination curve of the DC control circuit feeder break~r 72-403 (Gould HE3A 100). Installation of the FRN-R-15 fuses maintains proper fuse/breaker coordination such that the fuses will actuate before the breaker. Wiring within the 252-SPARE4 breaker's closing coil circuit and from the load side of the FUZ/A1204-2 fuse block is #14 AWG SIS which is rated at 35 amps (Table 310-17, 1993 NEC). Increasing the closing coil fuse size to FRN-R-15 maintains protection of the wiring.

Safety Analysis Summary The probability and consequences of an accident previously evaluated in the FSAR will not be increased. Breaker 252-SPARE4 contains the same style operator as breaker 152-204. As noted above, the internal wiring and auxiliary switch configuration of breaker 152-'204 will be verified to match that of breaker 252-SPARE4 prior to installation of the spare breaker. In the case that the two breakers do not have the same internal wiring or auxiliary switch configuration, breaker 252-SPARE4 will not be installed until the difference fs inc_orporated and documented into the temporary modification. Larger (FRN-R-15) fuses Will be installed in place of the existing FRN-12 fuses for FUZ/A 1204-2, to ensure that the increased closing current required by breaker 252-SPARE4 will not cause fuse actuation. The FRN-R-15 fuses will maintain ftise/breaker coordination and allow approximately the same actuation time margin (with 252-SPARE4) as* the FRN-12 fuses (with 152-204).

The probability and consequences of malfunctions of equipment important to safety will not be increased. Both breaker 152-204 and 252-SPARE4 have identically functioning solenoid operators. *The internal wiring and auxiliary switch configurations of the two breakers will be verified to ensure that they are the same prior to installation of the 252-SPARE4 breaker into cubicle 152-204. Increasing the closing coil fuse size, FUZ/A 1204-2, to FRN-R-15 will ensure sufficient time margin for the breaker to operate (close) while maintaining proper fuse/breaker coordination.

The possibility of an accident or malfunction of a different type than previously evaluated in the FSAR will not occur. Both breaker 152-204 and 252-SPARE4 have identically functioning solenoid operators. Increasing the closing coil fuse size to FRN-R-15 will ensure sufficient time margin for the breaker to operate (close) while maintaining proper fuse/breaker coordination. Once the TM is in place, Operations will test cycle breaker 252-SPARE4 in cubicle 152-204 to verify operability.

The margin of safety as defined by Plant Licensing Bases will not be reduced. This TM does not defeat any safety systems or substitute a less capable breaker for use in cubicle 152-204. With the exception of the increased closing coil fuse size, the TM essentially substitutes a functionally equivalent breaker for the one normally installed in cubicle 152-204.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents TM-93-145 SE93-0763 SERVICE WATER SUPPLY TO FLASH CONDENSER TUBES

~temporary water supply will be connected to P-44 suction line drain (MV-SW564) to be used for flushing main condenser tubes during cleaning. This connection is noncritical service water and flow rates will be less than 100 gpm.

Safety Analysis Summary Probability or consequences of accidents or malfunctions are not increased. Service Water System pipe break is already evaluated in FSAR and is bounded by a main header rupture. In this case we are limited by the small size of the line (1 ") and the fact we are only going to affect noncritical components during shutdown conditions.

Noncritical header can be isolated by closing its isolation valve preventing any effects*

on critical loads.

No equipment important to safety are connected to noncritical service water.

Noncritical can be isolated from critical at any time during shutdown conditions without affecting any equipment important to safety.

No new accident can be caused by this Temporary Modification (TM). Service water leaks are prev.iously analyzed and we will not cause any other accident or malfunction

. with this TM.

Margin of safety will not be affected by this TM. Flow rates are very low compared to system capabilities and the tie in is to the noncritical header.

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SUMMARY

OF GENERIC LETTER 90-05 TO THE FSAR This FSAR change adds a description of Generic Letter 90-05 which contains guidance to be considered by the NRC when evaluating relief requests submitted by licensees for Class 3 systems.* Since certain Generic Letters can be considered to be part of ttie plant's licensing basis, those which are relevant should be described in the FSAR.

This change also adds a subsequent memorandum issued by the NRC which clarified GL 90-05 and added that stop-gap measures may be used to stop leaks in Class 3 systems when certain criteria are met. This information is also considered to be part of the plant's licensing basis and the plant has taken advantage of this in mitigating certai'n leaks in the.service water system.

Safety Analysis Summary This FSAR change does not increase the probability or conseq_uences of an accident

. previously evaluated in the FSAR. This change is merely administrative and reflects NRC policy regarding nori-code repairs in Class 3 systems. By itself, this change has no adverse affect on the integrity of piping because non-.code repairs cannot be made without NRC approval. In order to obtain NRC approval, the plant must justify the non-code repair in terms of piping integrity.

This FSAR change does not increase the probability or consequences of malfunctions of equipment important to safety.. This administrative change merely reflects NRC policy and would not~ by itself, allow the pl~nt to perform a non-code repair.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents This change does not create the possibility of a different type of accident or malfunction than any previously evaluated in the FSAR. This is just an administrative change being made since Generic Letters are considered to be part of the plant's licensing basis.

The margin of safety as defined in Plant Licensing Bases will not be reduced. Non-code repairs would have to be approved by the NRC. Approval of the repair would be based on maintaining system integrity, so margin of safety is not compromised.

FSAR CH 6 AND 9 SE93-0042 FSAR CHANGE REQUEST FOR SECTION 6.10.2 AND 9.8.2 During CRHVAC normal operation, the control room envelope pressure is controlled by DPIC 1659 and 1660 as discussed in FSAR Section 9.8.2.1. The DPIC's modulate dampers to control outside air intake to maintain pressure. Air intake is through normal air intake louvers.

During emergency operation, the normal air intake modulating dampers are closed and a fixed flow of outside air is drawn in through the emergency air intake located some 95 meters from the containment. The DPIC's are not operative during the emergency mode.* FSAR Section 9.8.1.4a prpvides the proper design bases for pressurization during th!3 emergency mode of CRHVAC operation*. The statement in FSAR 6.10.2 refers t9 Section 9.8 but implies that during normal and emergency modes of operation, the dP is "greater than all adjacent areas" - when, in fact, for normal operation, the Dp is measured for only one adjacent area..

Safety Analysis Summary The CRHVAC system cannot create an accident of the type analyzed in FSAR Chapter 14. The CRHVAC is a support system for heat removal from the control room and fo control post accident radiation doses to control room operators. Therefore, neither the probability nor the consequences of an accident previously evaluated in the FSAR will be increased.

The FSAR changes do not result in any equipment changes. The FSAR changes are being made to provide consistency in the FSAR. Therefore, neither the consequences nor the possibility of malfunction of equipment important to safety will be increased.

Since the CRHVAC does not initiate accidents, the possibility of an accident different than those previously evaluated in the FSAR will not be created.

No hardware changes or operational changes are involved, therefore the possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Chariges, Specification Changes,*.

Temporary Modifications and Other Documents Since the changes are made to provide consistency and no hardware or op*erational

- changes are being made, the margin of safety as defined by the Plant Licensing Bases will not be reduced.

. FSAR CH 7 SE93-0964*

FSAR CHANGES TO CHAPTER 7 ON CONTROL RODS Parts of Section 7. 5. 3. 1 still describe the controlling functions for the control rods as still having an automatic function. The text also* describes the sequencing alarms as being generated by both computers and panel lights. Only the SPI alarms the sequence of rods.

Safety Analysis Summary The control of rods by automatic circuits, and the alarming method for showing out of sequence, are not necessary and do not contribute to analyzed accidents in the FSAR.

Equipment important to safety is not affected by this change. This change affects FSAR description only. The ability of the control rod.~ to respond to the RPS actions and operation input is unaffected. *

' The possibility of an accident of a different type is not created because the change does not change the ability of control rods to respond to plant conditions as designed.

The basic scram and manual controls remain unaffected.

The possibility of a malfunction of a different type is not created because the out of sequence alarm still works to alert the operator of incorrect rod withdrawal and the disconnection of automatic control makes permanent an administrative control to prevent. the use of automatic control. No present plant operation is affected.

The margin of safety* in Plant Licensing Bases is not affected. Alarm functions and control functions remain unchanged. Rods out of sequence are still alarmed and the rods will still scram. The change only affects one part of FSAR, Section 7.5.

FSAR CH 7 AND 9 SE93-0750 CLARIFY DESCRIPTION OF FOGG OPERATION This clarifies the description in the FSAR regarding use of the FOGG valve~ in a main steam line break accident. The FSAR says that the Emergency Operating Procedures direct the operators to use the FOGG valves to isolate the low pressure steam generator after a MSLB. This is incorrect. The EOPs actually tell the operators to isolate a steam generator by closing the associated flow control valves. According to

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifi cat fons and Other Documents Operations, the EOPs _have never taken credit for using the FOGG-valves to isolate a steam generator. In fact, of all the EOPs, only EOP 7.0, Loss of Feedwater Flow, uses the FOGG valves and that is to throttle auxiliary feedwater flow if the flow control valves are not available.

-Safety Analysis Summary The probability or consequences of an accident previously evaluated in the FSAR will.

not be increased. This change has no effect on the plant's ability to respond to an accident. It merely reflects measures already being taken in the EOPs to mitigate a MSLB. The EOPs have been reviewed under 1 OCFR50.59.

The probability or consequences of malfunctions of equipment important to safety will not be increased. This change merely reflects actions taken iri the EOPs.

The possibility -of an accident or malfunction of a different type than any previously

_evaluated in the FSAR will not be created.

The margin of safety as defined by Plant Licensing Bases will not be reduced. This FSAR change has no effect on any margins of safety.

FSAR CH 8 SE93-0081 FSAR CHANGE REQUEST FOR SECTION 8.4.1.2 Currently FSAR Section 8.4.1.2 states "Operation of these switches/slide links is governed by Off Normal Procedure 25.2." ONP 20, not ONP 25.2, now governs the above described operation.

Safety Analysis Summary This revision to delete FSAR Section 8.4.1.2's reference to ONP 25.2 does not increase the probability or consequences of an a,ccident previously evaluated in the FSAR or malfunction of equipment important to safety being increased.

In addition, as this is only a change to make the FSAR reference the proper ONP in Section 8.4. 1.2, possibilities of an accident or malfunction different than previously -*

evaluated in the FSAR will not be created. The margin of safety will not be reduced by this FSAR revision.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, ~pecification Changes, Temporary Modifications and Other Documents FSAR CH 8 SE93-b762 VOLTAGE PROTECTION AND LOAD SHEDDING SYSTEMS Correct paragraph_ which describes testing to accurately describe the testing that is performed.

Safety Analysis Summary Neither the probability nor the consequences of an accident previously identified in the FSAR ar~ increased because there is no change to the facility or its operation.

The probability of malfunction of equipment important to safety is not increased. The implied testing which is being -deleted in the FSAR change ("... and load sequencing, once the diesel generator is back on line... ") is redundant to the testing described in the first sentence of the paragraph. The testing described in the first sentence is a complete test of the_ load sequencing feature. Testing this feature again after simulating a loss of the diesel generator would not test any different components that -

had not already been tested.

The consequences of a malfunction of equipment important to safety is not increased_ -*

because no -change to the facility is being made.

The possibility of an accident or malfunction of a different type than any previously evaluated in* the FSAR is not created because there is no change to the facility or its operation.

_The margin of safety as defined by the Plant Licensing Bases is not reduced because there is no change to the facility or its operation.

FSAR-CH8

_ SE93-1148 EMERGENCY DIESEL GENERATORS (EDGl DAY TANK Provide current Emergency Diesel Generators (EOG) day tank run time in FSAR and move FSAR Section 8.7.3.2 information to 8.4.1.

Safety Analysis Summary The EOG fuel oil system is not associated with any FSAR Chapter accident initiation.

Changes to this system wili not increase the probability of any Chapter 14 accident.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The significant Chapter 14 accident for which consequences may be affected are Main Steam Line Break (MSLB) and Loss Of Coolant Accidents (LOCA). Decreasing the EOG run time on the day tank will not increase the consequences of an MSLB or LOCA. D-PAL-93-181 F determined current plant procedures are adequate in support of maintaining fuel oil to the EOG following an accident.

Decreasing the EOG run time -on the day tank will not increase the probability of a malfunction. D-PAL-93-181 F determined current plant procedure are adequate in maintaining fuel oil to the EOG following an accident.

D-PAL-93-1 81 F identified diverse methods of replenishing the EOG day tank fuel oil supply (i.e., re power P-10 or fuel truck). Therefore, malfunction of equipment does not have increased consequences Decreased EOG day tank run time does not create an accident or malfunction of a different type than previously evaluated in the FSAR. D-PAL-93-181 F determined plant procedures are adequate to maintain-fuel oil to the, EOG.

The margin of safety has not been decreased. The licensing basis for the EOG day tanks is to provide capacity such that a method of transferring a continuous supply of fuel oil to the day tanks can be implemented prior to the EOG starving for fuel. D-PAL-93-181 F determined plant procedures maintain fuel oil to the day tanks within the required time to prevent the EOG from starving; FSAR CH 8 SE93-1316 FSAR CHANGE REQUEST CHAPTER 8 Revision of discussion about protecting power cables with circuit breakers per the National Electric Code.

Safety Analysis Summary It is conceivable that power system protective action could be involved in the electrical

.disturbance of-FSAR events 14.7, 14.12, 14.13, 14.15, and 14.17. However, the probability of a previously evaluated accident is not increased, because:

The electrical system was designed to "minimize the effects of any electrical fault and to maximize the availability of on-site and off-site power sources" (FSAR 8.1.1) This philosophical basis for design subordinates interruption of service to availability. In general the National Electrical Code subordinates availability to interruption of service. Therefore, losses of power are less probable.

The National Electrical Code has limited applicability, to nuclear plant design. In fact, the NEC exempts power facilities.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The consequence of a previously evaluated accident are not increased, because:.

The electrical system was designed to "minimize the effects of any electrical fault and to maximize the availability of on-site and off-site power sources." (FSAR 8.1.1 l This philosophy, which favors availability of power, assures accident mitigation capability beyond that which wou.ld be provided. i,f all power cables..

were protected with circuit breakers per the NEC.

As an example, the thermal overload protection of certain essential moto.r operated. valves is disabled, because this is considered more important to nuclear safety than cable or equipment protection.

Power. circuit protecti'on criteria affect all plant power systems. Therefore, this item is applicable to equipment requiring electrical power to manage Chapter 14 events and other events.

As "passive" devices, electrical protective equipment cannot directly cause safety related equipment malfunctions. However, interruption of electrical power to safety related equipment could be consider.ed a real function of that equipment. The probability of this type of ma_lfunction of equipment important to safety is not increased because:

The.electrical system was designed to "minimize the effects of any electrical fault

. and to maximize the availability of on-site and off-site power sources." This is not generally the objective of the National Electrical Code, which is strongly biased in favor of interrupting service.

The above qµoted electrical design philosophy makes safety related equipment "malfunctions" (loss of power) less probable than if the National Electrical Code (which is of limited applicability) had been followed.

The consequences of a malfunction of equipment important to safety are not increased because:

As "passive" devices, electrical protective equipment cannot increase the

consequence of safety related equipment malfunctions beyond interruption of service.

The electrical system was designed to "minimize the effects of any electrical fault and to maximize the availability of on-site and off-site power sources. (FSAR 8.1.1) This criteria, which focuses on availability, tends to decrease the consequence of equipment malfunctions when coupled to National Electrical Code criteria (NEC criteria have a strong focus on interruption of service).

The possibility of an accident of a* different type than any previously evaluated in the FSAR is not created because:

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The electrical system design criteria to "minimize the effects of any electrical fault and to maximize the availability' of on-site and off-site power sources" (FSAR 8.1.1) concerns the supply and availability of electrical power The supervision and protection of electric power circuits per this criteria (as opposed to the National Electrical Code) can cause power passes (already designed and accounted for). However, the plant is designed to absorb power losses without adverse effects to the plant.

The possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created because:

Electrical power system protection criteria address availabili~y, interruption of service, faults, overloads, equipment protection, etc.

These issues are all addressed elsewhere in the FSAR. Removing the erroneous citation of the NEC in Section 8.5.3.2 of the FSAR does not affect the prior consideration of these issues - including the primary related "malfunction" the loss of power.

The margin of safety as defined by the Plant Licensing Bases is not reduced because:

Power cable protection is an issue of a supervisory nature which is not directly related to the margin of safety for the plant.

The electrical system criteria intended to "min.imize the effects of any electrical fault and to maximize the availability of on-site and off-site power sources" (FSAR 8.1.1) are more supportive of supporting the plant margin of safety (indirectly) than would the erroneously - cited NEC. This is true because the NEC, which has limited applicability to a nuclear power plant, is generally biased toward a complete interruption of service.

FSAR CH 9 SE93-0006 FSAR CHANGE REQUEST FOR SECTION 9.3.3.2 The FSAR change request will modify FSAR Section 9.3.3;2 5 to show the "as built" condition; that is, provisions are provided for ISi, operability and performance testing.

This change was not included when test connections were installed.

Safety Analysis Summary Providing test connections does not change the operation or function of the system, therefore the probability or consequences of an accident previously evaluated in the FSAR will not be increased.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents ISi, operability and performance testing enhances the reliability of equipment, therefore the probability or consequences of a malfunction of equipment important to safety will not be increased.

Since the test provisions do not change the function or operation of the system, the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR will not be created.

Incorporation of test provisions for ISi, operability and performance testing do not change the function or operation of the system. The provisions, together with the testing, enhances reliability of the system. Therefore, the margin of safety as defined by Plant Licensing Bases will not be reduced.

FSAR CH 9 SE93-0146 FSAR CHANGE REQUEST FOR SECTION 9.3.2.1. TABLES 9-5, 9-7, AND 9-20

FSAR change request making miscellaneous changes to FSAR Section 9.3.2.1 and FSAR Tables 9-5, 9-7 and 9-20.

Safety Analysis Summary The changes for this FSAR change request are all either typo's, corrections to maintain consistency, or are supported (as identified in the description) by calculations, drawings or other documentation. The changes do not affect either probability or consequences of accidents or malfunctions evaluated by the FSAR nor do they create the opportunity for a different accident or malfunction than those evaluated in the FSAR. There also is no affect on the margin of safety as defined by Plant Licensing Bases.

FSAR CH 9 SE93-0222 FSAR CHANGE REQUEST REGARDING CCW HEAT EXCHANGERS This FSAR change request makes two changes.. One, a statement is added that specifies that two component cooling water heat exchangers be in service anytime the plant is above cold shutdown operation and that damage would occur due to excessive flow rates if only one heat exchanger was in service and a second CCW pump were started due to a sat ety injection signal. Two, the design capacity (shell side) of a CCW heat exchanger is added to Table 9-6. This heat exchanger capacity is taken from the Design Basis Document for the CCW system. The reference in the DBD from which this value was taken added to the references of Chapter 9 of the FSAR.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents This FSAR change is the result of a corrective action to E-PAL-93-001 *which* requested that the FSAR description of the CCW system be revised to reflect th.e design.

characteristics of the system.

SafetV Analysis Summary This change does not increase the probability of an accident previously evaluated in the FSAR. Adding the design limitations of the CCW heat exchangers to the FSAR has no eHect on the probability of any FSAR accidents.

This change does not increase the consequences of an accident previously evaluated in the FSAR. Adding design limitations of the CCW heat exchangers has no effect on any radiological barriers.

The probability or consequences. of malfunctions of equipment important to safety will not be increased.. This change could. only decrease the likelihood of failure of a CCW

heat exchanger due to excessive flow since it makes clear that two heat exchangers must be in service to prevent this type of malfunction. No radiological barriers are affected by this change.

. The possibility of an accident of a different type that any evaluated previously evaluated in the FSAR is not created. This change merely identifies the operating

limitations of the CCW heat exchangers.

The possibility of a malfunction of a different type than any previously evaluated in the FSAR is not created. No new malfunctions :are created by this change.

The margin of safety as defined by Plant Licensing Bases are not reduced. Requiring that both CCW heat exchangers be in service during all plant operation above cold shutdown assures that there is sufficient capacity to remove heat from the CCW system under all operating conditions. This is consistent with guidance provided in SOP 16 (CCW System).

FSAR CH 9 SE93-0373 FSAR CHANGE REGARDING STARTING OF FIRE PUMPS FROM CONTROL ROOM This FSAR change merely clarifies the as-built condition of the fire pumps. The fire pumps are capable of being starfed from the local control panel or the control room.

The current FSAR revision ( 14) implies that the fire pumps may only be started from the local control panel.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other.Documents Safety Analysis Summary The probability or consequences of an accident previously evaluated in the FSAR are not increased. This FSAR change merely reflects the actual design of the fire pump start circuitry and has no effect on any of the postulated FSAR accidents.

The probability or consequences of a malfl.inction of equipment important to safety will not be increased. This FSAR change merely reflects actual plant design and could not adversely affect safety equipment.

The possibility of an accident or malfunction of a different type than previously evaluated in the FSAR could not be created. This FSAR change has no effect on plant equipment; it merely reflects plant design. As such, it could not create a new accident or malfunction.

The margin of safety as defined by Plant Licensing Bases is not reduced. Existence of the capability to start the fire pumps from the control room has no eHect on any safety margins.

FSAR CH 9 SE93-1108 FSAR CHANGE REQUEST CONCERNING DESCRIPTION OF SPENT FUEL POOL CRANE This FSAR Change Request corrects inaccurate information in FSAR Table 9-9 Sheet 1 regarding its description of the trolley rail for the spent fuel pool crane. According to the vendor drawing for the crane (M-61 Sheet 45), the FSAR description actually pertains to the runway *rail and also contains a typographical error (i.e., "and" should be "lbs"). The trolley *rail should be desc~ibed as" 175 lbs USS" per the vendor drawing (USS is United States Steel).

This FSAR Change Request also corrects in Section 9.11.4.3.8 the number of mechanical brakes on the spent fuel pool crane main hoist. The hoist was originally supplied with one brake but a second brake was added later.

Safety Analysis Summary This change request has no effect on the probability or consequences of an accident described in the FSAR. The change corrects the description of the spent fuel pool crane rails and the main hoist brakes, and the crane has no impact on any of the accidents described in the FSAR.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents This FSAR change has no effect on the probability or consequences of malfunctions *Of equipment important to safety. This is merely a change to the FSAR description of the crane and has no effect on the crane itself. Correcting the description of the crane main hoist brakes and the rails, in accordance with the crane vendor drawing, could not affect safety equipment.

The possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR will not be created. This is just an administrative change to the FSAR.

The margin of safety as defined by Plant Licensing Bases will not be reduced.

Correcting the description of the fuel pool crane rails and the number of main hoist brakes in the FSAR does not adversely impact the margin of safety.

FSAR CH 10 SE93-0817 CHANGE FSAR TABLE 10-9 "COOLING TOWER PUMPS" TO CORRECT TO AS BUILT CONDITIONS Change FSARtable "Design Head ft" from "106" to "90" and change "Pump Speed,

. r/min" from "257" to "253" per vendor information. Currently FSAR incorrectly lists the design head and pump speed for the cooling tower pumps.

Safety Analysis Summary Review of the accidents analyzed in Chapter 14 of the FSAR indicate that no analyzed accidents will be affected by this change. The cooling tower pumps are not identified in any analyzed accident scenario. The change to FSAR Table 10-9 updates the table to the actual design conditions as verified by the vendor. This change does not involve

. a modification to the cooling tower pumps.

The probability or consequences of malfunctions of equipment important to safety will

not be increased. The cooling tower pumps are non-0, nonsafety related and therefore are not designated as equipment important to safety. The proposed change to the FSAR does not have any effect on the reliability or the non-accident design basis events of the cooling tower pumps. The cooling tower pumps design condition change does not affect any equipment important to safety. The change just identifies the correct design conditions of the pumps.

The possibility of an accident or malfunction of a different type than previously evaluated in the FSAR will not be created. The cooling tower pumps design parameter change will not cause a degradation to one or more fission product barriers or result in radiological risk to the general public in excess of the 10CFR100 limits. The cooling tower pumps are not involved in any accident scenario and the proposed design parameter change just clarifies the actual design conditions of.the pumps.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The margin -of safety as defined by Plant Licensing Bases will not be reduced. The cooling tower pumps do. not have any direct or indirect effect on fission product*

boundaries or accident consequences. The proposed change just identifies the corrE3ct design paramete.rs for the pumps. Therefore, the margin of safety is *not reduced.

FSAR CH 11 SE93-0448 RADIOACTIVE MATERIAL STORAGE BUILDINGS

. Miscellaneous tools and equipment become radioactively contaminated during maintenance activities. Not all of these tools and equipment can be stored inside the

radiologically controlled area of the plant because of lack of storage space. Therefore, radioactive material storage areas outside of the main plant are needed.

These radioactive materials are not wastes. This material has a purpose and is being stored until ne~ded for that purpose.

Currently, two buildings are used to store radioactive materials until they are needed.

The North Storage Building is located north of the plant inside the protected area fence.

It is an engineered steel building with a concrete floor. The South Storage Building is located south of the p_lant just outside of the protected area fence but inside a locked cyclone fence. It is. an engineered steel building.with a blacktop floor. Both buildings are locked at all 'times except when storing or removing equipment. Th~ Mechanical Maintenance Department is responsible for inventory, housekeeping and accessibility of work groups into the storage buildings. The Radiation Services Department oversees the movement of radioactive material to and from these buildings.

Potentially, a third building will be needed to store radioactive material. Because of the weight of much of the dry fuel storage equipment it will be necessary to store this

equipment in the Blue Building south of the parking lot. This building has a special floor.

built to withstand this extra weight. None of the other storage buildings can take the weight. The dry fuel storage equipment is not contaminated at this time but through the course of normal work it is expected that this equipment will become contaminated. The Blue Building is an engine*ered steel building with a concrete floor located outside the protected area fence but within the owner controlled area. The*

building will be locked when it is used as a radioactive material storage area except when storing or removing equipment. NECO is responsible for inventory, housekeeping*

and accessibility of work groups into this building. The Radiation Services Department oversees the movement of radioactive material to and from this building.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,*

Temporary Modifications and Other Documents Design Basis No regulatory guidance exist solely for the purpose* of radioactive material storage buildings. Therefore, References A, B, and C will be used as guidan~e to evaluate radioactive material storage building. Professional judgement will be used in areas

where these references do not apply or are overly conservative in controls as they.

would apply to radioactive material.

References A.

NRC Information Notice No. 90-09, "Extended Interim Storage of Low Level Radioactive Waste by Fuel Cycle and Materials Licensees" (2/90).

B.

NRC Generic Letter 81-38, "Storage of Low Level Radioactive Wastes at Power Reactor Sites" ( 11 /81).

C.

NRC Generic Letter 85-14, "Commercial Storage at Power Reactor Sites of Low Level Radioactive Waste Not Generated by the Utility" (8/85).

D.

10 CFR 100 - Reactor Site Criteria.

E.

IE Circular No. 80-18, 10 CFR 50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems" (8/90).

F.

10 CFR 50 - Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criterion 60, 63, and 64 and Appendix I, Design Bases Efflu~nts.

G.

10 CF.R 50.59-Changes, Tests and Experiments.

H.

10 CFR 20 - Standards for Protection Against Radiation.

I.

EA-E-PAL-91-030*01 - Effluent Dose Evaluation from Radwaste storage Building.

J.

EA-E-PAL-91030*02 - Direct Dose Evaluation to Site Boundary from Wastes

Stored in the Radwaste Buildings.

Analysis A.

Bounding Event There are no limits for radioactive material storage in the existing Technical Specifications. Evaluation of accident and direct dose calculations for radwaste stored fn East Radwaste Building and South Radwaste Building in EA~E-PAL 030*01 and EA-E-PAL-91-030*02 support not having a limit for radioactive

material storage buildings.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents These analyses can be used to very conservatively estimate the accident and direct dose consequences of the radioactive material storage buildings for the following.

reasons.. The Curie content of radioactive material packages are generally much less than radwaste packages since the material is not compacted or concentrated.

The radioactive material storage buildings, like the radwaste buildings, will not physically support more than 150 DAW boxes, which was the calculational basis for these engineering analyses.

Radiological Consequences A.

Gaseous B.

Accident releases from radioactive*material storage buildings are not considered credible because of ~he type of material. This material is externally and/or internally contaminated with varied amounts of contamination. The only scenario that could release this radioactive material in a gaseous form would be a fire.

. However, most of this material would not readily burn. Accident case 1 and 3 of EA-E-PAL-91-030*01 burned the contents of un-boxed dry active waste (DAW) and the entire contents of the South Radwaste Building (150 DAW boxes and 150 55 gallons drums of asphalt solidified concentrates). The accident *cases were less

than 10% of 10 CFR 100 limits and direct dose to the site boundary was less than i mR/yr as required by Generic Letter 81-38.

Liquid There are no liquid effluent consequences because all material in the buildings must meet dry radioactive material requirements. All materials packaged for.

storage are inspected for the presence of freestanding liquid prior.to packaging and transfer to the storage area.

C.

Direct Dose The direct dose consequences from storing radioactive material in these buildings are minimal. EA-E-PAL-91-030*02 evaluated the direct dose rates from 150 DAW boxes and 150 asphalt solidified concentrate drums and found that the direct dose at the site boundary was significantly less than 1 mR/yr. This evaluation conservatively estimates the direct dose from radioactive material stored in these

. buildings for the previously stated reasons.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification -Changes, Temporary Modifications and Other Documents Design Requirements A.

Container Sele-ction Items to be placed in containers for storage should be wiped down or wrapped.

Packages shall be sturdy enough so as to allow handling without breaching the package. Heavy or bulky packages should permit handling with a forklift or overhead crane. Containers made of wood should be made of fire retardant lumber. Container surfaces should be painted, lined or otherwise finished to prevent them from becoming contaminated and permit easy decontamination.

Shielding should be provided internal or external to the package when necessary to reduce the radiation levels. Posting requirements of 10 CFR 20 shall be maintained.

B.

Surveillance Permanent long term radioactive material storage areas are surveyed for radiation and contamination on a monthly basis.

C.

Transportation and Location The North Storage Building is located within the protected area fence. The South Storage Building is located just outside of the protected area fence and is surrounded by a locked cyclone fence. The Blue Building. is located south of the parking lot. Buildings are locked as well as subject to periodic security patrols.

Transport routes used to reach the storage buildings are all within the owner controlled area.

D.

Control of Releases of Radioactive Materials to the Environment ( 10 CFR 50.

Appendix A. Criterion* 60)

The normal activities that take place in the radioactive material storage buildings do not produce any gaseous or liquid effluents. Even under the worst accident scenario, burning the entire contents of the building, less than 10% of the 10 CFR 100 limit was exceeded and the dose to the site boundary was less than 1 mR/yr.

Therefore, no controls for the release of gaseous or liquid effluents are necessary.

E.

Monitoring Equipment (10 CFR 50. Appendix A. Criterion No processing activities take place in the North or South Storage Buildings.

Materials are either stored or removed. Materials are surveyed upon packaging and the storage buildings are surveyed on a monthly basis. Therefore, no area radiation monitoring or airborne radiation monitoring equipment is necessary.

150 -

CONSUMERS POWER COMPA~Y - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents F.

Because of the size and weight of much of the dry fuel storage equipment it may at times be necessary to work on this equipment in the Blue Building. When and if it

. becomes necessary to perform work on contaminated equipment in this building the necessary controls will be evaluated and implemented to monitor for radioactive releases and contain any contamination or effluent to the extent practical.

Effluent Monitoring (10 CFR 50 Appendix A. Criterion 64)

There will be no effluent monitoring from the North South Storage Buildings. No processing is allowed in these buildings which would create a radioactive gaseous or liquid* effluent pathway.

Effluent monitoring will be performed for the Blue Building in accordance with the above section on monitori.ng equipment.

G.

Flooding Flooding is not considered a threat to the North Storage Building or the Blue Building. The South Storage building has been subject to minor flooding and heavy rain. For this reason, materials stored in this building are to be raised off of the floor whenever possible.

H.

Tornado The buildings were not designed or built to withstand the impact of a tornado.

Since the structures are engineered steel buildings, they are expected to withstand some of the im*pact and remain partially intact. The roof may be lost and the material may be released through that pathway. The loss of this material presents more of a spread of contamination problem than an effluent problem. However, accident case 3 of EA-E-PAL-91-030*01 released all of the material of 150 DAW boxes and 150 concentrate drums and did not exceed 10% of 10 CFR 100 values.

I.

ALARA Materials will be stored with lower dose rate reading materials near the walls to minimize dose rates outside of the radioactive materia! storage areas and to provide shielding for higher dose rate materials stored in the center of the buildings.

"'. 151 -

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT

~acility Changes, Specification Changes, Temporary Modifications and Other Documents Safety Analysis Questions Radioactive Material handling accidents are not evaluated in the present FSAR.

There is no equipment important to safety in the radioactive material storage buildings.

Accidents involving storage of radioactive material are contained in the attachments and meet acceptance criteria in design basis.

There is no margin of safety defined for radioactive material storage.

The design bases of described abov.e have been met. Therefore, radioactive material storage buildings are not an unreviewed safety question.

.. FSAR CH 11 SE93-1132 STACK MONITORING SYSTEM The Backup Stack Gas Monitors; RIA-2318 and RIA-2319, that have been serving as the backup to the Radioactive Gaseous Effluent Monitoring System (RGEMS) installed in

1983, are obsolete and will no longer be used as a backup to the RGEMS. Grab samples will be used as the backup when RGEMS is not operational. This is consistent with current practice as RR~9G had required grab samples during calibration of RIA-2318, if RIA-2326 in the RGEMS system was also inoperable, and radioactive effluent release via the stack was to continue during the period of operability.

Safety Analysis Summary The backup stack gas monitors monitor gaseous effluents leaving the plant throughout the stack discharge system as a backup to the primary RGEMS to detect abnormal gaseous releases (which may indicate a_n accident). The monitors could not initiate an accident~ either directly or indirectly, therefore, the probability of a previously evaluated accident could not increase.

Effluent monitoring and sampling systems indicate. increases in the levels of.radiation and radioactive materiai in plant effluents, but do not effect the radiation dose criteria used for the design basis accident, therefore the consequences of any previously evaluated accident could not increase.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT

Facility Chan~es, Specification Changes, Temporary Modifications and Other Documents The RGEMS was added in parallel to the original system so discontinuing use of the.

original system will not degrade the reliability of RGEMS, or effect its survivability or ability to function during nonaccident events.

Since grab samples will be used as the backup for RGEMS, as.is currently the case when both the primary and backup systems are inoperable, the. consequences of a malfunction of the RGEMS could not increase.*

Discontinuing the use of the backup stack gas monitors does not require any changes to*

plant operations or to the function performed by any system, structure, or component that could cause a degradation of one or more fission product boundaries.* Therefore there is no possibility of an accident not already addressed in the FSAR.

Discontinuing the use of the backup stack gas monitoring system does not effect the operation of the any plant equipment, and could not create the potential for a new type of equipment malfunction or failure.

Since installation of RGEMS extended the monitored range of stack gas effluents, discontinuing use of the old monitoring system will not reduce the margin of safety as defined by Plant Licensing Bases.

M-34 SE93-1011.

SPECIFICATION FOR HORIZONTAL CENTRIFUGAL PUMPS In August 1969, the Bechtel design group requested a field change to the shield cooling pumps to increase their capacity based on revisions to piping. and the addition of the reactor cavity floor cooling coils. In response to their request, the Bechtel field representative contacted a local pump vendor requesting.guidance on changes to P-77 A/B to provide the additional required capacity. The vendor responded recommending that the pumps be run at 3 500 rpm with a 5-1 /4 inch impeller and a 7-1 /2 Hp motor from the original 1750 rpm, 6-3/4 inch impeller and 3 Hp motor.

These changes were intended to rerate the pumps from 125 gpm at 38 feet head to* 180 gpm at 79 feet head. In January 1970, the Bechtel field representative reported to the design group that the field change as recommended by the vendor had been made.

Evidently, none of the supporting system design bases or vendor inform.ation was updated since* the current FSAR and vendor files still reflect the original design of 125 gpm at 38 feet head and associated pumpspecifications. This field change rep~esents a change to the facility as described in the FSAR even though it has been installed since the plant started operating ih 1971.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Teinporary Modifications and Other Documents Safety Analysis Summary The probability of an accident previously evaluated in the FSAR will not be increased since the shield cooling system is not required after any DBA. Therefore, the increased flow through the system will not increase the probability of any previously evaluated accident.

The.consequences of an accident previously evaluated in the FSAR will not be increased again because the shield cooling system is not required after any DBA. Therefore, the increased system flow will not cause additional dose to the general public or Plant personnel as the result of any accident.

The probability of malfunctions of equipment important to safety will not be increased.

The field change made to the shield cooling pumps during original construction increased its capacity but did not make them any more susceptible to failure during accident or nonaccident events than the original lower flow design. This system does not perform any accident mitigation functions and the field change actually improves the system's ability to perform its function during normal operating conditions by providing additional heat removal capability through the increased system flow.

The consequences of a malfunction of equipment important to safety_ will not be increased. The field change does not increase the pump's susceptibility to any malfunction and the system is not relied upon to mitigate the effects of any accident or malfunction. Therefore, the dose received by the general public or Plant personnel will not be increased by this field change.

The possibility of an accident of a different type than any previously evaluated in the FSAR will not be created. This system is not required after any DBA so the slightly increased capacity provided by this field change will not create any new type of accident which differs from those already evaluated in the FSAR. -

The possibility of a malfunction of a different type than any previously evaluated in the FSAR will not be created. This field change slightly increased the capacity of the shield cooling pumps and associated heat removal capability of the system but did not provide a revised configuration which would introduce any n*ew types of failures.

- 154 -

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modif~cations and Other Documents The margin of safety as defined by the _Plant Licensing Bases is actually increased by this item. The FSAR states that the shield cooling coils are capable of removing 180,000 Btu/h based on the original 125 gpm shield cooling pump design. With the flow capacity of the pumps increased to 180 gpm, the shield *cooling coils are now capable of removing additional heat load. The FSAR also states that the maximum heat load on the system is only 120,000 Btu/h.* Therefore, the increase in the heat removal capability of the shi~ld cooling system will increase the already appreciable capacity margin.

Additional calculational work is being performed by the Reactor Safety and Analysis group to determine the specific effect that this field change has on the design basis of the shield cooling $YStem. It is sufficient to state here.that the increased shield cooling pump capacity does not represent an Unreviewed Safety Question.

M-60A

SE93-1295 SPECIFICATION-FOR CONTAINMENT AIR COOLERS REPLACEMENT COOLING COILS The proposed specification was developed for replacement cooling coil units for three (VHX-1, -2, and 3) of the four Containment Air Coolers.* The existing coolers have been experiencing leaks, primarily in the manifolds and headers. Work history has documented over 30 such leaks. Non-code repairs have been allowed by the NRC until such time as replacement of the coils' becomes feasible. The existing coolers consisting of inlet and outlet headers, manifolds~ and coils will be replaced with new units. The replacement units will be designed to maintain the existing heat removal capacity under postulated accident conditions. Pressure drops and flows will be designed to maintain compatibility with interfacing equipment. The new coolers will meet CPCo Design Class 1 requirements.

Safety Analysis Summary A review of all design basis accidents listed in Chapter 14 of the Plant's FSAR found that none of the scenarios were initiated by the failure of a component of coolers or the Containment Air Coolers as units.

Per a review of FSAR Chapter 14, it was found that the Coritainment Air Coolers are used to mitigate the consequence of two previously analyzed accidents, Main Steam Line Break (MSLB) and Loss of Coolant Accident (LOCA). In both scenarios, the coolers' function is to limit the peak pressure and temperature attained in the Containment Building following initiation of either event. The pr_oposed replacement cooling coils are to be made with equally qualified components which do not alter the function of the coolers. Since the proposed design capacity of the cooling coils with respect to accident conditions will not be changed, the probability of the containment being challenged is not changed, hence, the consequences of these postulated accident scenarios are not increased. The coolers must also function as a containment pressure boundary. and so, the post-accident environmental conditions will be specified.

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CONSUMERS POWER COMPANY.:. PALISADES NUCLEAR PLANT

Facility Changes, Specification Changes, Temporary Modifications and Other Documents This specification has been developed in accordance with code, regulatory and qualification requirements more stringent than the original design requirements. In addition, FSAR Table 5.2-3 indicates that the Containment Air Cooling System is CPCo Design Class 1. Thus, the probability of equipment failure is not increased from that of the original equipment.

The proposed spetification does not change the function of the cooling coils. Also, the specification is more.stringent than the original specification. In addition, redundancy with the remaining coolers and containment spray pumps ensure cooling of the containment environment following either of the applicable events. In conclu_sion, the effectiveness of the barriers in limiting the consequences of any previously analyzed malfunction would remain unchanged.

The component's function in the system has not changed and, th~refore, its operational characteristics could not result in any new system transients.. The proposed specification for the replacement coils does not introduce. any new or different failure mechanisms which have not been previously evaluated or bounded. It is concluded that no new acCidents wlll be introduced by.this modification.

The replacement coils will remain subject to all typical failure modes consi.dered in the original design and will be installed consistent with or reconciled with the original requirements of the cooling coils.

Equipment which serves to protect the margin of safety typicaily possesses certain design characteristics. These criteria include diversity, separation, redundancy,and

independence. The proposed specification does not change the function or alter the inherent design characteristics mentioned above of the Containment Air Coolers or any, of their components. Since the new coolers will be designed on the basis of a service water temperature of 85°F, the actual margin may be increased. Therefore, the margin of safety in the Plant's Licensing Bases has not been reduced.

MP 2.7.1 CPAL-1 SE93-0906 WESTI'NGHOUSE FIELD SERVICE PROCEDURE "FUEL BUNDLE ALIGNMENT PIN GAGING AND STRAIGHTENING This procedure governs the activities associated with the ga-ging and straightening of selected fuel bundle alignment pins on the bottom of the Upper Guide Structure. In addition, as an option, the procedure contains steps to gage holes in the Core Support

Plate where fuel bundle lower tie plate alignment pins engage. The procedure includes guidance and administrative controls to monitor and document the required work activities. Most of the activities will require the Upper.Guide Structure to be lifted from

_its storage location and rotated while suspended. The procedure also provides the option to install UGS support stands as necessary to assist with inspection and straightening operations. Caution statements have been placed in the procedure and

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents work order to ensure the straightening tool and support stands are not set on reactor cavity liner welds or on cavity leak chases. The actual work activities will be performed by Westinghouse Electric Corp. in accordance with Palisades Administrative Procedure and Consumers Power approved Quality Program Plans. These inspections are being performed as part of the root cause evaluation for E-PAL-93-019 which documents fuel bundle SAN-8 being suspended from the Upper Guide Structure when the UGS was lifted the second time from the reactor vessel as part of 1993 refueling activities.

The Upper Guide Structure (UGS) assembly consists of a flanged grid structure, 45 control rod shrouds, a fuel alignment plate and fuel alignment pins. The UGS aligns and supports the upper end of the fuel bundles, maintains the control rod channel spacing, prevents fuel bundles from being lifted out of position during a severe accident condition and protects the control rods from the effect of coolant cross flow in the upper plenum.

The UGS is handled as one unit during installation and refueling.

The fuel bundle alignment plate is designed to align the upper ends of fuel bundles and to support and align the lower ends of the control rod shrouds. Two alignment pins are attached to the fuel bundle alignment plate at each core location to align the fuel bundle.

This procedure is intended to inspect the alignment pins to ensure they are still correctly oriented. This inspection/gaging will be accomplished with a functional gage containing two holes representative of the top of a fuel assembly to verify straightness and relative position of the two alignment pins corresponding to one core location. If the functional gage fits properly (the pins pass the functional gaging test), further inspection of those pins may not be required. Should the pin functional gage not fit properly, two different types of gages will be used to characterize the problems associated with the pins. The primary gage will use two Linear Differential Transformer (LVDT) displacement measuring probes to measure pin straightness. As a contingency in the event of malfunction of the electronic gage, a mechanical gage consisting of a bushing with a circle of axially sliding pins can provide a profile of the relation of the alignment pin. to the underside of _the UGS. These pin angles measurement tools may also be used to characterize pins which are successfully tested with the functionality gage.

Pin straightening criteria is contained in the procedure. Engineering justification for this criteria is documented in Westinghouse Letter Report no. MED-PCE-11845, "Investigation of Fuel Bundle Alignment Pins for Palisades Nuclear Power Station." This report provides laboratory test results of pin straightening to ensure the acceptability of the pin straightening operations, if it is required. Westinghouse has performed alignment

. pins gaging and straightening at other nuclear plants, specifically Byron and Indian Point 3.

If align_ment pin straightening is required, the pin will be straightened using a cup die engaged on the pin from underneath. The die will be pressed over the pin into contact with the underside of the UGS, restoring ttie alignment pin to perpendicular relation to the UGS. Water-filled hydraulic jacks will be used to press the die over the pin.

Quantity of deionized water used in the jacks (less than 5 gallons) will be administratively limited to ensure a negligible effect on reactor cavity boron

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CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT.

Facility Changes, Specification Changes, Temporary Modifications and Other Documents concentration_ should the equipment leak.* The reactor cavity floor will resist the downward force of the hydraulic jacks. An evaluation of the effects of the load on the re~ctor cavity floor has been performed and found to be acceptable.

Procedure and Work order cautions require the straightening tool placement on the reactor cavity floor avoid stainless steel liner welds and liner leakage monitoring trenches.

If required, inspection of the pin holes in the Core Support Plate will be performed using an underwater video camera. Following visual inspection, the pin holes will be checked with a ball type gage as a "Go-No Go" inspection device. The gage, with the diameter of the fuel assembly lower tie plate alignment pin, will be mounted to a handling. pole such that the ball can be inserted manually into each of the four pin holes at core location selected to gage. The use of a ball type gage will reduced to probability of the gage sticking in the core support plate hole.

Safety Analysis Summary The probability of an accident previously evaluate in the FSAR will not be increased by activities performed during this procedure. The activity will be authorized through the.

work order process. Permanent Maintenance Procedure FHS-M-24 "Movement of Heavy.*

Loads in the Containment Area" will be used when UGS lifting is required to support the*

inspections and/or repairs. Use of FHS-M-24 ensures administrative controls are in place to limit the potential of a heavy load drop.. Permanent Maintenance Procedure MSM-M-.

47 "Foreign Material Exclusion in the Reactor Cavity and Spent.Fuel Pool Areas" will also be used to control the work activities around the reactor cavity. Westinghouse Letter Report no MED-PCE-11845 documents the engineering evaluation of the straightening to ensure that any repair activities will not impact the load carrying capacity of the alignment pin or fuel bundle alignment plate as a result of bending and straightening operations. Therefore, the integrity of the reactor vessel internals will be maintained and be capable of performing its intended design function once the Upper Guide Structure is reinstalled.

The fuel assembly guide pins do not contribute to the postulation of any accident in the FSAR. Potential loading on the pins during normal operations would be flow-induced vibration loads acting on the reactor internals. However, because the weight of the a fuel assembly is greater than the primary coolant system flow uplift force, there would be no relative motion between the fuel assembly a.nd the guide pins. Therefore, straightening of pins will have no effect on the probability of any currently postulated accidents.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,*

Temporary Modifications and Other Documents The consequences of previously evaluated *accidents will not be increased due to the performance of the activities contained in this procedure. The procedure in conjunction with the associated work order provides guidance and administrative controls to ensure the work activities will have a limited impact on refueling outage activities. The function of the guide pins during postulated accidents is to provide lateral restraint to the fuel assemblies and to maintain the geometry of the core. If required, the use of the

. straightening tool will return the UGS alignment pins to as close to original design as possible thereby ensuring previously assumed accident results are unchanged.

The probability or consequences of malfunctions of equipment important to* safety will not be increased by the gaging and straightening of the fuel alignment pins or the gaging of the core support plate pin holes. The normal operational function of the equipment will not affected by the inspection of the alignment pins or core support plate pins holes or straightening of the alignment pins. Furthermore, under postulated accident conditions the function of the guide pins will be maintained~ Westinghouse letter report no. MED-PCE-11845 provides engineering justification of the straightening operations to ensure that any repair activities will not affect reliability and survivability of the guide

pins.

The possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR will not be created. The guide pin and core support plate inspections will have a negligible impact on the associated equipment. Westinghouse*

Letter Report MED-PCE-11.845 provides engineering justification that a straightened pin will ber able to accomplish its design function.

The margin of safety as defined by Plant Licensing Bases will not be reduced by the activities contained *in this procedure. The margin of safety of the guide pins is defined by the structural criteria in the design codes. The inspections will have no impact on the structural integrity of the guide pins or core support plate pin holes. Westinghouse Letter Report MED-PCE-11845 provides engineering justification of the pin straightening operation will ensure the guide pins meet or exceed the design codes..

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents STANDING ORDER 62 SE93-0054 TECHNICAL SPECIFICATION INTERPRETATIONS/GUIDANCE Revision prevents use of the LCO of Technical Specification (TS) 3.3.2.d for the CCW heat exchangers during power operation and requires both CCW heat exchangers to be operable with PCS temperature > 325 °F; clarifies HPSI pump operability requirements of Technical Specification 3.3.2g.. Defines alternate method of monitoring reactor vessel inventory during RVLMS inoperability (presently found in SOP 1, Revision 24, "Primary Coolant System," Section 7.4). Deletes fire protection items removed from Technical Specifications and relocated in FSAR Section 9.6. 7 and FPIP-4. Reflects relocation of incore alarm compensatory measures from SOP 35, "Neutron Monitoring System," to SOP 34, "Data Logger System," and deletes RETS items removed from Technical Specification by Amendment 154 and relocated to the ODCM.

D~PAL-89-032 determined that if one ( 1) CCW heat exchanger was removed from service during normal operation as permitted by Technical Specification 3.3.2.d, excessive flow could result through the remaining CCW heat exchanger should an SIAS occur and cause a second CCW pump to start. To prevent this scenario, a plant requirement was added to SOP 16 that both CCW heat exchangers be in service with PCS temperature > 325°F, and failing this, Technical Specification 3.03 was applicable.

Taken literally~ TS 3.3.2.g (3) could preclude rendering a HPSI pump inoperable during power operation, even though such action is specifically permitted by 3.3.2.c. In the Basis of TS 3.3, the intention of the HPSI pump operability requirements of Section 3.3.2.g is mitigation of a LOCA under L TOP conditions; thus its inclusion in Section

.3;3.2 (applicable during power operation) is inappropriate. The alternate method of monitoring reactor vessel inventory in the event of RVLMS inoperability was inappropriately located in SOP 1 (Rev 22) during system installation 1988; it is relocated to SO 62.

Deletions of FPS and RETS items aligns SO 62 with Tech Specs through Amendment 154.

Safety Analysis Summary The probability of an accident previously evaluated in the FSAR is not increased by requiring 2 CCW heat exchanges to be operable above 325°F; this condition is assumed both by the accident analysis and by the design analysis of FSAR 9.3.3.

The consequences of an accident previously evaluated by the FSAR are not increased, as this is the normal mode of operation of the system.

The probability of malfunctions of equipment are not increased, as the proposed change will prevent reliance on one (1) heat exchanger for CCW cooling.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Fa~iiity Changes, Specification Changes, Temporary Modifications and Other Documents

. The consequences of malfunctions of equipment might be considered to be increased if one took the view that the inability to implement the LCO of TS 3.3.2d to.mitigate the effects of the malfunction or perform necessary repair is the loss of an opportunity to avoid a plant transient. Ho_wever; applying TS 3.03 is permissible that operation under

those circumstances.

The possibility of an accident or malfunction of a different type than previously evaluated is not created, as the change is designed to prevent failure of an in-service CCW heat exchanger due to excess flow during an accident. Two in-service CCW heat exchangers

is the normal mode of operation of the system.

The margin of safety defined by licensing basis is not reduced. The change is more restrictive than TS 3 ;3 and complies with the margin of safety discussed by FSAR 9.3.3.1.

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72.48 SAFETY REVIEWS

CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT F~cility Changes, Specification Changes, Temporary Modifications and Other Documents.

FC-864 SE93-0587 10CFR72.48 REVIEW - DRY.STORAGE OF SPENT NUCLEAR FUEL Since Palisad.es began operation in the early 1970's all spent nuclear fuel discharged from the reactor core has been stored on site in the spent fuel pool. Storage of spent fuel on-site will continue until at,least the year 2000 when the Department of Energy will accept spent fuel from Palisades.. To provide adequate storage capacity on site, arid to maintain full core off-loa*d capacities in the fuel pool, an alte.rnate storage system is needed. The alternate storage system is the Sierra Nuclear VSC-24 System selected for use at Palisades. This system uses concr~te storage casks, placed on a concrete foundation within the Palisades protected area. The Independent Spent Fuel Storage Installation (ISFSI) has been designed to store up to -600 assemblies in 25 Ventilated Storage Casks (VSC's), 24 assemblies per cask.

The VSC-24 System will consist of seven major components:

1)

A Multi-Assembly Sealed Basket (MSB), which will hold the spent fuel assemblies. *

(Safety Related)

2)

A Multi-Assembly Sealed Basket Transfer Cask (MTC), which will provide shielding and transportation of the MSB during the fuel loading. (Safety Related)

3)

A Ventilated Concrete Cask (VCC), which will provide storage for the MSB.

(Safety Related) *

4)

A Heavy Haul Transfer Trailer (HHT), which will move the Ventilated Storage Cask (VSC) from the Track Alley to the ISFSI pad. (Non-Sat ety Related)

5)

A Track Alley Load Distribution System (LOS), which is made up of steel members bridging the Track Alley Slab. The movement of the VCC down Track Alley is allowed only if the VCC is on top of the LOS. (Safety _Related)

6)

A Vacuum Drying Skid (VOS), which will be used to drain and vacuum the loaded MSB. (Non-Safety Related)

7)

A Ventilated Air Transporter (VA~), which will be used to lift the VSC., (Non-Safety Related)

PROCEDURES

Moving the spent fuel from the spent fuel pool to the ISFSI will be as follows:.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification.Changes, Temporary Modifications and Other Documents

1)

Prepare the VSC-24 System for loading using (FHS-M-33), which prepares the MTC, MSB, VCC, VAT, VDS, Spent Fuel Pool Impact Limiting Pad (ILP), Decon Pit ILP, HHT, rigging, and other system components required for the loading procedure.

FHS-M-33 will:

_Move the Spent Fuel Pool ILP to'the spent fuel pool, Move the MTC and the Decon Pit ILP to the Washdown Pit, Move the VCC with the MSB to Track Alley, Transfer the MSB into the MTC in the Washdown Pit, Set up the Automated Welding System (AWS) and the VDS on the pool floor and Load Test the roadway, LDS and L-3 Crane if necessary._

2)

The-Loading Procedure (FHS~M-32) Wlll:_

Transfer the MTC I MSB to the spent fuel pool, Load the MSB with 24 fuel assemblies, Transfer the MTC I MSB to the Washdown Pit, Decontaminate the MTC I MSB, Weld and vacuum dry the MSB, Transfer the MTC I MSB to the VCC in Track Alley, Transfer the MSB from the MTC to the VCC, and Transport the VSC to the ISFSI.

Sierra Nuclear Corporation (SNC) has prepared Letter SNC-A TL-93-092 which provides a comparisons of approved cask design, specific CPCo cask design, and specific NRC documents. SNC has established that the VCC, MSB, and MTC are within the certified cask design as described in the C of C including design change notices (DCNs) and nonconformance reports (NCRs).

RX-277 WATER ABSORPTION

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CONSUMERS POWER.COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents

During Design Review closure an issue* with the RX-277 absorption of water during the pre-op test and potential release as steam during the drying process was identified. This issue is addressed as an evaluation of the "po.ssibility *of a malfunction of a* different type than any previously evaluated in the VSC SAR". The SAR and SERs do not address the issue..

The VSC-17 Report had a statement on Page 2-2 indicating that the RX-277 shielding off gassed during the heat up in the form of water vapor and hydrogen as temperature increased.

VSC-17 had 8 penetrations through the lid to instruments in the fuel region. This provided a direct interaction between the RX-277 and the fuel area. In addition, the RX-277 was not baked as the material was not thought to off gas water at that time.

The VSC-24 RX-277 material was baked at 350°F for 24. hours to resolve the concern.

The additional concern with the Number 1 shield lid being soaked for approximately 26 days was discussed to determine if this would be detrimental. The restricted flow path in the VSC-24 system to the RX-277 cavity will prevent the absorption of much water but even if it did, most of the moisture wpuld be driven out during vacuum drying.

Because of the stainless steel design, the water would not affect anything in the RX-277 cavity.

The pressure resulting from the partial pressure of the air and partial pressure of the water vapor in the RX~277 /Swagelok fitting cavity was conservatively estimated to be 33 psi. The effects of this pressure on the lid cover weld and the shell surrounding the RX-277 were found to be well within the allowables.

Consequently, the presence of water in the RX-277 /Swagelok cavity is considered to be within.. the design basis.

VENTILATED AIR TRANSPORTER Section 1.1 of the SAR and Section 3.1 of the SER identify the method for moving the concrete casks on and off the heavy haul trailer (HHT) using a "hydraulic roller skid towed or pushed by a tractor." Per Palisades Procedure FHS-M.:.32, the movement of the concrete casks will be performed using a Ventilated Air Transporter (VAT).

EA-FC-864 "VCC Jacking" evaluates the different load distribution between the hydraulic rollers as referenced and the four jacks which will be used for lifting the casks to place air pads underneath.

The probability of an accident previously evaluated in the SAR will not be increased by the use of the VATs. The VATs perform the same function as the hydraulic rollers in that they are capable of safely lifting the loaded VSC's. The SAR identifies that an engineering equivalent may be used to move the concrete casks.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents The VA Ts will be used to transport the MTC into Track Alley underneath the equipment hatch.

. The VATs will be used*to transport the VCC I MSB into Track Alley The VATs will be used to move the VSC from track alley onto the HHT and will also be used to move the VSC's onto the ISFSI.

The VATs are safer than the hydraulic rollers because the cask movement can be quickly stopped by releasing the air from the air pads.

The consequences of an accident previously evaluated in the VSC-24 System SAR will

. not be increased due to the use of the VATs. The Certificate of Compliance requires

that the loaded VSC not be handled at a height of greater than 1 8 inches. The VSC will be handled at a height of approximately 2" with the VATs. The height is not enough to cause damage to the VSC upon sudden loss of air to the air pads..

The probability of malfunction of the VSCs or associated equipment important to safety will not be increased by the use of the VATs to transport the VSCs. The VSC functions

. independently of the equipment used to transport the loaded casks. The VA Ts have no input into the safe function of the VSC.

The consequences of a malfunction of the VSC-24 System equipment important td safety will not be increased due to the use of the VATs~ The VATs perform no safety related function. The VAT's sole function is to transport the Storage cask arid other safety related equipment but is not in of itself classified as a safety related component as defined by the NRC in the Certificate of Compliance.

The possibility of an accident of a different type than any previously evaluated in the VSC-24 System SAR will not be created by the use of the VATs to move VSC-24 Safety related components. The safety related components (i.e. the MTC, VCC and MSB) will be moved by VA Ts prior to performing their safety related function(ref. FHS".'M-33). The only time the VATs move a safety related component performing a safety related function is when the. loaded VSC is being moved on the storage pad and on the LOS in the Track Alley. The possibility of an accident occurring during the transportation down the Track Alley to the HHT and also during movement of the VSC on the storage pad is bounded by the accident scenarios that the VSC is designed for as described in Section 2 of the SAR.

The margin of safety as defined by the VSC Licensing Bases will not be reduced by the use of the VATs to move the VSC-24 System components. The design safety margins

_tor the loaded VSC are bounding for postulated failures of the air pads used to transport the fuel storage cask.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Change~, Specification Ch~ngesj Temporary Modifications and Other. Documents There will be no increase in occupational exposure for transporting the loaded VSCs with the VATs.. The exposure to workers will be limited to that which has been evaluated and approved in the SAR Rev.. 0 and site specific calculations(Ref. EA-FC-864-07).

The VATs will not affect the environmental impact of the ISFSI.

WATER REMOVAL FROM ANNULUS SAR Section 10.1.3.2 and SER Section 8.1 address.the use of water in the MSB-MTC annulus during MSB closure, drying, and inerting operations.

Per FHS-M-32, the water in the annulus of the MSB-MTC will be allowed to drain after removal of the MSB-MTC from the pool to the decontamination pit. Justification for this change is provided in that:

1) the resulting increase in dose rates (see Table 4 of EA-FC-864-40) while work is performed in the decontamination pit is not a significant hazard to personnel and will not increase off-site dose to the public.

2) the work area around the MSB-MTC will be restricted to essential personnel and continually monitored by Health Physics personnel when the MSB is closed, dried, and inerted.

The accidents and malfunctions of equipment previously evaluated in the VSC SAR arid SERs deal with inadvertent criticality of the fuel stored in the MSB and drops of the VSC-24 system equipment. The draining of water from the annulus between the MSB-MTC does not provide for criticality control in that the fuel to be stored in the MSB has been analyzed such that criticality will not occur in the normal condition (dry)_

without the MTC. Remov_al of water during the lift and transfer to the decontamination pit will decrease the consequences of a potential drop by decreasing the weight of the MSB-MTC. The water in the annulus does not represent an active portion of the equipment and its removal will have rio affect on the probability or consequences of a malfunction of equipment important to safety. The possibility of an accident of a different type from those previously evaluated in the VSC *sAR is not increased in that the only effect of the water drainage will be the slight increase in radiation as identified above, a decrease in the weight of the MSB-MTC, and potentially, a decrease in the cooling medium for the MSB while the MSB is being vacuum dried. The worst case heating situation is when the MSB is vacuum dried which has been analyzed for thermal integrity. The process of removing the water from the MSB-MTC annulus takes place entirely within the fuel pool/decontamination pit area and will not create a significant unreviewed environmental impact.*

CASK TEMPERATURE MONITORING

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents SAR Section 1.2.2 identifies that the VSC system operation is totally passive and no active systems are required. The VCC air outlet temperature monitoring system will not be passive as required by the NRC in the C of C.

The probability of an accident previously evaluated in the VSC-24 System SAR will not be increased by the addition of the temperature monitoring system to the VSCs. The temperature monitoring device consists of an RTD fastened to the air outlet screen of the storage *cask..There will be four RTDs, one for each outlet air vent. The RTDs will not inhibit the safety characteristics *of the storage cask. EAs have be.en written to

support the installation requirements for the design change. EA-FC-864-27 "VCC 1 & 2 Heat Flow" has been revised to show that the addition of the temperature monitoring RTD's to the air outlets will not inhibit the air flow. EA-FC-864-43 documents the operational characteristics of the temperature monitoring system on the VSCs.

EA-FC-864-27 calculates the safe temperature operating band for the first two casks placed into service. T-338 will be used to verify the cask outlet temperatures are acceptable. The outlet temperatures for both casks will be measured until each cask tias reached equilibrium. After the cask temperature reaches equilibrium the cask

temperature will henceforth be measured by DW0-1.

The consequences of an accident previously evaluated in the VSC-24 System SAR will not be increased due to the use of the RTDs for thermal performance monitoring of the*

VSC's. The temperature monitoring devices verify the outlet temperatures of each storage cask on the ISFSI. The verific~tion that the RT Os are functioning properly is by daily terT'lperature measurement by DW0-1.

The probability of malfunction of the VSCs or associated equipment important to safety will not be increased by the RTDs. The RTDs and associated equipment will not have any affect on the VSC equipment. The RTDs will be placed on the cask so as_ not to inhibit designed function.

The consequences of a malfunction of the VSC-24 System equipment important to safety will not be increased due to the RTDs. Cask outlet temperature will be monitored on a daily basis.

The possibility of an accident of a different type than any previously evaluated in the VSC-24 System SAR will not be created by the use of the temperatu*re monitoring system. The design does not affect the cask accident scenarios in the SAR Section 11.

The possibility of an malfunction of a different type.than any previously evaluated in the VSC-24 System SAR will not be created by the temperature monitoring system. The design does not affect the cask to cause it to malfuncti.on..

The margin of safety as defined by the VSC Licensing Bases will not be reduced as a result of the temperature monitoring devices placed onto the VSCs. The temperature monitoring will increase the margin of safety of the VSC-.24 casks as it will provide for daily monitoring. of cask outlet temperatures.

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CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents There will be no significant increase in occupational exposure as a result of a remote temperature monitoring system. The system will reduce the amount of time that personnel are required to spend at the ISFSI to take required temperatures.

The temperature monitoring system will not affect the environment.

WELDING SAR Section 3.4.. 2 describes the positive closure system that is composed of multi-pass welds at five locations.

The shield lid seal weld will be made using a single-J partial penetration joint design.

. The joint design used by CPCo will use a 45 degree groove angle as opposed to the* 20 degree angle. This minor change in joint designs was made to increase the weldability and is supported by the actual shield lid edge preparation which shows a single-bevel joint preparation with a groove.angle of 45 degrees. The 45 degree groove angle increases the weldability of the joint by allowing greater access to the root of the joint.

In addition, the 45 degree groove angle will decrease the chance of weld hot cracks by increasing the width to depth ratio. The final joint will still be a single-J partial penetration joint.

FHS-M-32 will address welding and helium sniff requirements by procedures, FC-Lid, SM-Lid and ES-DC-1. Helium sniff checks will be performed on t_he MSB shield lid seal weld, MSB structural lid weld and MSB shield lid to structural lid port hole fillet weld, and final cover plate weld. The helium sniff procedure is ES-DC-1.

Dye penetrant inspections will be performed on the MSB shield lid root weld, MSB shield iid seal weld, structural lid root weld, structural lid seal weld, and three fillet ~elds.

The Swagelok valve on the water drain line In the MSB shield lid was replaced with a NPT hose fitting to allow faster draining which reduces radiation dose by reducing time for draining and reduces the potential of icing during initial vacuum pump down by removing the pressure loss associated with the Swagelok valve. This line will be closed.

with a helium tight fitting after the draining and initial pump down to remove the water vapor. The Swagelok was left on the helium back fill line to allow a helium tight closure when the connecting line is removed after helium back filling. Neither fitting is safety related as they are not part 9f the pressure retaining boundary; they are both covered and welded over with the two valve cover plates. Thus, this line does not play a role in increasing probabilities or mitigating consequences of the accidents previously evaluated in the VSC SAR. This line and valve do not interface with any safety related components and, since the line function is now passive, the probability and consequences of malfunctions of equipment important to safety is, if anything, decreased by this change. The use of the NPT fitting can not credibly introduce any new accidents or malfunctions of equipment different than those previously evaluated. The

decrease in draining time actually will increase the margin of safety in that the 4 7 hour8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months requirement for drainage of the MSB will more easily accomplished by this change;

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes; Temporary Modifications and Other Documents

. Radiation exposure should be reduced by reducing the time required for draining the

MSB. Since the drainage of the MSB takes place entirely within the fuel pool/

decontamination pit, there can be no credible unreviewed environmental impact.

'FHS-M-32 will address welding and helium sniff requirements by procedures, FC-Lid,

. SM-Lid and ES.,.DC-1. Helium sniff checks will be performed on the MSB shield lid seal weld, MSB structural lid weld and MSB shie.ld lid to s~ructural lid port hole fillet weld, and final cover plate weld. The helium sniff procedure is ES-DC-1.

. OTHER ISSUES Table 2.3-1 of the SAR identifies that demineralized water would be used in the MSB and MSB-MTC annular gap. The actual process will use filtered borated water. This meets the requirements of SAR Section 2;3.2.1 which allows the use of "demineralized borated water".

SAR Section 9.1.3 identifies that the first VSC placed in service using the VSC-24 System will be loaded with fuel producing 24 kw and temperature and radiation measurements will be taken in the loading area prior to moving the VSC to the storage.

pad. FC-864 will not be moving fuel producing 24 kw of energy in the *first VSC.

SER Section 12.0 identifies that the MSB shall be loaded with 24 spent fuel assemblies, constituting a heat source of approximately 24 kw and then. loaded into the VCC to test.

heat transfer by measurement of air temperatures.

C of C, Section 1. 1. 7 allows for the storage of spent fuel with a heat load less than 24

kw provided that a calculation of the temperature *difference between the inlet and outlet *

. temperatures is performed using the same methodology and inputs in the SER and SA~,

with the lesser load as the only exception. This calculation is documented in EA-FC-864-05.

Table 12.1-1 of the SAR identifies "General Areas where Controls and Limits are Necessary". Additional procedures and changes to those identified.(i.e. daily vs. weekly inspections of the air inlets and outlets) are required. The daily surveillance requirement is based upon C of C requirements identified in Section 1.3.1 which take precedence over the SAR requirements..

Procedure FHS-M-32 provides for the loading of the MSB while it is located in the fuel pool. This procedure, the shield lid is lowered into the MSB in the pool.. As part of the design review process noted, the possibility existed of dropping the shield lid.* By letter dated 2/9/93, it was determined that this accident is bounded by existing analysis WNEP 8626, "Design Report of Region Two Spent Fuel Storage Racks," Revision 2, 1987 and that no _further analyses are needed.

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CONSUMERS POWER COMPANY - PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents FC-LID SE93-1375 10CFR12.48 REVIEW - AUTOMATED WELDING PROCEDURE FOR MSB Page 5-4 of the SER, dated April 28, 1993 lists the exemptions to the ASME Section Ill code requirements for the MSB closure welds. During the review of the SER it was identified that the NRC exemption did not address the ASME code requirements for post

weld heat treatment for the MSB shield lid to shell weld and the three valve access c*over welds. An exception to the code can be taken due to the Charpy V-notch impact testing performed for the welds. The impact testing showed that the lowest energy absorbed for any of the tests was 35 ft-lbs, which was a specimen removed form the unaffected base metal. The lowest energy absorbed by a specimen from either the weld or base metal heat affected zone was 43ft-lbs.

The lack of a 200°F preheat for the shield lid to shell weld and the three valve access cover welds is not a problem due to the following:

1.

A 516 Gr. 70 material is not significantly hardenable

2.

The three welds are multi-pass welds.

3.

The impact properties for the weld procedures qualifica~ions for FC~Lid and SM-Lid are adequate..

There will be no.increase in probability of any of the chapter 11 accidents analyzed in VSC SAR by the exemption to post weld lieat treatment for the MSB lid welds.

There will be no increase in the consequences of any accident analyzed in chapter 11. A hypothetical breach of the containment boundary by way of the MSB lid welds is

enveloped by the accident analysis described in section 11.2. l.1.

Section 11.2.1.1 of the VSC SAR identifies that the worst accidental release would be one where there is a hypothetical failure of all the.fuel pins within the MSB and a breach of the* MSB containment. The failure of the welds due to no preheating is very small and if there was a failure of the weld the subsequent release would be within the accident analyzed in section 11.2. 1. 1.

Section 11.2.1.1 of the VSC SAR describes a breach of the MSB and envelops the hypothetical accidental failure of the MSB Lid welds. No new types of accidents or malfunction will be created. The margin of safety will not be reduced as the ~elease associated with a failt.Jre of the lid welds is less than the_ radioactive release described in section 11.2.1. 1.

The resulting exposure due to a hypothetical crack in the MSB lid welds will not significantly increase occupational exposure. *

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CONSUMERS POWER COMPANY -

PALISADES NUCLEAR PLANT Facility Changes, Specification Changes,

. Temporary Modifications a*nd Other Documents The environmental impact of a postulated breach of the MSB lid welds due to a crack is bounded by the environmental impact related to the accident described in section 11.2. 1. 1 of the VSC SAR.

C-136A*

SE93-0604 10CFR72.48 REVIEri -

FABRICATION SPECIFICATION FOR THE MULTI_.ASSEMBLY SEALED BASKET A new specification is required to identify the design requirements for the multi-assembly sealed basket (MSB).. This Safety Evaluation also reconciles various changes from the SAR descriptions. Most of the changes were made by the cask vendor prior to NRC

issuance of the C of C.

Safety Analysis Summary None of the changes could result in ~n unreviewed Safety Question. Specification CMSB-89-001, Rev. 4 provides for the fabrication of the MSB.

SAR Section 1.1 identifies that the MSB is fabricated to Specification MSB-87-001. The MSBs were actually fabricated in accordance with Specification CMSB-89-001.

For acceptability of the changes, reference Docket # 72-1007, SNC letter to the NRC dated 11/25/91 and EA-FC-864-02.

SAR Section 3.1 identifies material requirements for the MSB shell, bottom plate, structural lid, shield lid top plate, and shield lid bottom plate as A-516 Grade 70 or A36.

Drawing CMSB-24-002 S 1 /2 and CMSB-24'-003 S 1 /2 identify this material as SA-516 Grade 70.

Section 3.1 of the SER dated 4/28/93 identifies the material requirements of the MSB as SA-516 Grade 70. This discrepancy between the specification and the SAR was identified by the NRC as a design requirement in a letter from R.E. Cunningham to J.V.

Massey on 8/26/91. As this is a change from the SAR which has been identified by the NRC in its SER and correspondence with the MSB designer, no further evaluation is required and no unreviewed safety question exists.

SAR Section 7. 1. 1 identifies that the MSB is designed by analysis to meet the requirements of Section Ill of the ASME Code and fabricated to Specification MSB-87-001. The actual fabrication specification is CMSB-89-001.

For acceptability of the changes, reference Docket # 7 2-1 007, SNC letter to the NRC dated 11 /25/91 and EA-FC-864-02.

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PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents SAR Section 7. 1. 1.2 identifies that the MSB assembly will be fabricated and assembled in accordan_ce with the drawings in App. 1 and MSB-24-001, Rev. 4; MSB-24-002, Rev:

3; MSB-24-003, Rev. 3; MSB-24-004 (Sheet 1 ), Rev. 3; and MSB-24-004 (She_ets 2&3),

Reiv. 1. The actual assembly was in accordance with drawings as referenced in CMSB-.89-001.

For acceptability of the changes, reference Docket # 7 2-1007, SNC letter to the NRC dated 11 /25/91 and EA-FC-864-02.

SAR Section 7.1.3.2 identifies that all welding is in accordance with the referenced drawings. Welding was actu~lly performed per CMSB-89-001.

For acceptability of the changes, reference Docket # 7 2-1007, SNC letter to the NRC dated 11 /25/91 and EA-FC-864-02.

Section 3.9 of the fabrication specification identifies that all exterior surfaces of the MSB.

shall be coated with Keeler & Long epoxy white primer (no. 6548/7107) a polyamide epoxy, and applied in accordance with the manufacturer's specification.

For documentation of the acceptability of the changes, reference Docket # 7 2-1007, SNC.letter to the NRC dated 11/25/91 and EA-FC-864-23.

SAR Section 7. 1.3.3 and Table 1.2-3 identify visual welding inspections.in accordance with AWS D1.1. Section 3.7.8 of the fabrication specification identifies th_at all welds

shall be visually examined to ASMf:,Section V, Article 9.

For acceptability of the changes, reference Docket # 72-1007, SNC letter to the NRC dated 11 /25/91 and EA-FC-864-02.

The Certificate of Compliance was reviewed to dete.rmine requirements which would be appropriate for inclusion in the MSB fabrication specification. Sections 1.2.2 and 1.2.9 of the C of C identify the leak tightness requirement for the MSB. This requirement is covered in Section 3.11.3 of the specification. Section 1.2.13 of the C of C requires

that, during fabrication, each MSB shell material will have shown by Charpy test (per ASTM A370) that it has 15 ft-lb of absorbed energy at -50° F. This requirement is covered in Section 3.4.4 of the specification.

SM-LID SE93-1374 10CFR72.48 REVIEfl ~ MANUAL WELDING PROCEDURE FOR MSB.

Page 5-4 of the SER, dated April 28, 1993 lists the exemptions to the ASME Section Ill code requirements for the MSB closure welds. During the review of the SER it was identified that the NRC exemption did not address the ASME code requirements for post weld heat treatment for the MSB shield lid to shell weld and the three valve access cover 173 -

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PALISADES NUCLEAR PLANT Facility Changes, Specification Changes, Temporary Modifications and Other Documents welds. An exception to the code can be taken due to the charpy V-notch impact testing performed for the welds. The impact testing showed that the lowest energy absorbed

.for any of the tests was 35 ft-lbs, which was a specimen removed from the unaffected base. metal. The lowest energy absorbed by a specimen from either the weld or base metal heat affected zone was 43ft-lbs.

The probability of the MSB lid weld failing will be no greater due to the lack.of preheat since the A 516 Gr. 70 material is not significantly hardenable, the three welds are multi-

. pass welds, and the impact properties for the weld *procedures qualifications for FC-Lid and SM-Lid are adequate.

There will be no increase in probability of any of the chapter 11 accidents analyzed in VSC SAR by the exemption to post weld heat treatment for the MSB lid welds.

There will be no increase in the consequences of any accident analyzed in chapter 11. A hypothetical breach of the containment boundary by way of the MSB lid welds is enveloped by the accident analysis described in section 1J.2.1.1.

Section 11.2.1. 1 of the VSC SAR identifies that the worst accidental release would be one where there is a hypothetical failure of all the fuel pins within the MSB and a breach of the MSB containment. The failure of the welds due to no preheating is very small and if there was a failure of the weld the subsequent release would be within the accident analyzed in section 11.2. 1.1.

No new accidents or malfunctions would be postulated as a result of these charge.

The margin of safety will not be reduced as the release associated with a failure of the lid welds is less than the radioactive release described in SAR section 11.2.1.1.

The resulting exposure.due to a hypothetical crack MSB lid welds will not be significant as the increase occupational exposure.

The environmental impact of a postulated breach of the MSB lid welds due to a crack is bounded by the environmental impact related to the accident de~cribed in section 11.2.1. 1 of the VSC SAR.

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ML18059B126

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SUMMARY

GENERAL COMMENT

SUMMARY

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