ML20055E251

Trip Rept of 900521-24 Visit to Plant Site to Review Status of Steam Generator Tube Insp

ML20055E251
Person / Time
Site:	Millstone
Issue date:	07/09/1990
From:	Vissing G Office of Nuclear Reactor Regulation
To:	Stolz J Office of Nuclear Reactor Regulation
References
TAC-76448, NUDOCS 9007110217
Download: ML20055E251 (31)

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UNITED STATES -

NUCLEAR REGULATORY COMMISSION

& WASHINGTON, D C. 20666 h ..... p July 9, 1990 Docket 50-336 MEMORANDUMUM FOR: John F. Stolz, Director Project Directorate I-4 Division of Doctor Projects - 1/II FROM: Guy S. Vissing, Senior Project Manger Project Directorate I-4 Division of Reactor Projects - 1/II

SUBJECT:

REPORT OF VISIT TO MILLSTONE 2 0F MAY 21 THROUGH MAY24,1990(TACNO.76448)

INTRODUCTION This visit was a regular scheduled project manager quarterly one week site visit. Activities included attending the Millstone Station and Millstone 2-morning status meetings, attending three PORC meetings, attending a briefing on the steam generator replacement ' project, reviewing the status of the steam generator tube inspection, observing and discussing with contractor staff the steam generator tube inspection actdvities and results touring the plant site including entry into the containment, observing control room activities, reviewing operating logs and plant incident reports and performing a review of the licensee's 10 CFR 50.59 determinations on plant modifications and changes l of 1989.

1 i An entrance meeting was held with Mr. Jeffery Smith, Unit 2 Operations

( Supervisor who was acting for the Unit 2 Director in his absence. During the week, meetings were held with the following plant personnel:

R. Bates Engineering L T. Blanchard, Engineering W. Hutchings, Licensing  ;

CE contractor personnel L

Other plant personnel An exit meeting was held with J. Smith on May 24, 1990.

DISCUSSION Millstone 2 Steam Generator Replacement Project A briefing on.the steam generator replacement project was provided by members of the-dedicated project team. Enclosure 1 provides an overview of the scope <

of the project.

Steam Generator Tube Inspection The licensee had video displays of both steam generators tube sheets with a 180 t lb. nitrogen over pressure. The displays showed leaks in one welded plugged t tube and three other plugged tubes. The welded plug tube, a tube that

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2 was plugged in 1978, showed the greatest lesk at the weld of the plug. The licensee concluded that this leak was the primary leak that resulted in shutting down the reactor. The licensee was performing an inspection with the rotating pancake coil of affected tubes from 1 inch below the top of the tube sheet to 2 inches above the top of the tube sheet. Several suspect indications were identified but not confirmed. The inspection was to continue through Wednesday, May 30, 1990, and repairs would follow.

The contractors operations were viewed both in the on site trailers where the.

data was being collected and analyzed and in the containment at the steam generators where the actual operations were conducted.

Review of the licensee's 10 CFR 50.59 determinations of modifications made in 1989 The licensee's annual report for January 1,1989 to December 31, 1989 for Millstone, Unit 2, was reviewed. The report identified 24 plant design changes, 34 plant design change evaluations, 14 procedure changes 22 jumper-lifted lead-bypass changes, 11 set point changes, and 6 tests. As required by 10 CFR 50.59, the report contained, for each change except one, a brief description of the change, including a sumary of the safety evaluation. Plant Design Change Report MP-2-030-88, Replacement of Service Water Piping, was identified but no  !

summary was included. However, a safety evaluation was provided in the licensee's record. The licensee concluded for each change that the modifi-cation did not constitute an unreviewed safety question per the criteria of 10 CFR 50.59. In many cases the sumary of safety evaluation, as provided in the annual report, did not record thorough justifications for each of the seven criteria of 10 CFR 50.59 to support the conclusion that the changes did not '

constitute unreviewed safety questions.

A sample of 3 plant design change reports and 9 plant design change evaluations identified as follows were reviewed in depth to determine if acceptable determinations were performed.

PDCR Number Title MP-2-034-87 Reactor Coolant System (RCS) Vent Upgrade MP-2-011-88 Secondary Side Safety Relief Yalve(SRV) Position Indication HP-2-030-88 Replacement of Service Water Piping ,

PDCE Number Title MP-2-89-012 Terminations of Containment Air Rectreulation (CAR) Fan A .

MP-2-89-020 ContainmentAirRecirculation(CAR)FanOutlet Duct Expansion Joint MP-2-89-023 ReactorBuildingClosedCoolingWater(RBCCW)

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PDCE Number Title  !

MP-2-89-024 Modification of Pressurizer Safety Valve Support MP-2-89-030 Pressurizer Proportional Heater removal from Operation Due to Low Insulation Resistance MP-2-89-035 Power Supply X-1113A/B Replacement  ;

MP-2-89-070 Fire Pump Pressure Switch Isolation i MP-2-89-083 Battery Charger Power Failure Relay  ;

MP-2-89-114 ReactorCoolantPump(RCP)Conduitimprovements  ;

Each of the above' files centained a safety evaluation that concluded that the -

change did not constitute an unreviewed safety question. The document NSAC/125, June 1989, Guidelines for 10 CFR 50.59 Safety Evaluations, prepared by the Nuclear Management and Resources Council of the Nuclear Safety Analysis Center and operated by the Electric Power Research Institute was used in the review of the safety evaluations. This document indicates, for the purposes of performing safety evaluations, that the three 50.59 criteria for the determination of an unreviewed safety question can be broken down into seven separate questions as follows: l l

1. May the proposed activity increase the probability of ,

occurrence of an accident previous',y evaluated in the SAR?

2. May the proposed activity increase the consequences of an accident previously evaluated in the SAR?

, 3. May the proposed actnity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR?

4 May the proposed activity increase the consequences of a L malfunction of-equipment important to safety previously evaluated in the SAR?

5. May the proposed activity create the possibility of an accident of a different type than any previously evaluated in the SAR7

6. May the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the SAR?

7. Does the proposed activity reduce the margin of safety as defined in the basis for any technical specification?

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Although the answers to the above questions can be simply "yes" or "no", for each safety evaluation, there should be an accompanying explanation providing justification for the answer. Consistent with the intent of 10 CFR 50.59, these explanations should be complete in the sense that a qualified independent reviewer could draw the same conclusion. Restatement of the question in a negative sense or making a simple statement of conclusion is not sufficient.

Each of the safety evaluations were reviewed consistent with these guidelines. )

PDCR MP-2-034-87 Reactor Coolant System (RCS) Vent Upgrade  !

This PDCR installs manual isolating valves downstream of the reactor head and pressurizer remotely operated vent valves. Two safety evaluations (SE) were prov.ided, one developed by Plant Engineering ano another developed by the Reactor Engineering Branch. The Reactor Engineering Branch SE is more compre- i hensive and provides justification for each of the the seven questions except question 6. "May the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously

. evaluated in the SAR7"' No explanation was provided to answer question 6.

Although justification for answers to each of the other six question were  ;

provided, they were not clearly discussed on a one for one basis. The SE by ,

Plant Engineering provides a good description of the change and the reasons for l the change. It draws the proper conclusions. However, it does not explicitly support the conclusions in the discussions.

PDCR MP 2-011-88 Secondary Side Safety Relief Valve (SRV) Position '

Indication This PDCR installs instrumentation required to monitor the Safety Relief Yalve positions (16 total valves). The SE draws the proper conclusions except for answering question 6. As in the case for PDCR MP-2-034-88, no explanation was l provided to answer question 6. Although justification for answers to each of ,

the other six question were provided, they were not clearly discussed on a one for one basis.

PDCR MP-2-030-88 Replacement of Service Water Piping .

This PDCR replaces portions of the service water system with upgraded or e equivalent materials. Carbon steel epoxy-lined pipe is replaced with carbon steel PVC-lined pipe and the stainless steel spool pieces on the discharge side of all three RBCCW heat exchangers were replaced with carbon steel PVC lined pipe. The SE draws proper conclusions on questions 1, 2, 5 and 7. It does not address questions 3, 4, and 6. It concludes that the modification will not produce an unreviewed safety question. -

l This conclusion is drawn because the change is an improvement in protecting '

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the piping from corrosion. The modification does not change the configuration of the piping system.

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l PDCE MP-2-89-012, Terminations of Containment Air Recirculation (CAR) Fan A j

This PDCE evaluates the removal of an existing electrical connection box on the "A" CAR fan motor and replacing it with a new connection box in order to ,

provide an environmentally qualified electrical connection to the motor. The  ;

safety evaluation provides a good description and reason for the change. It '

makes the proper conclusion on the basis that the change does not change the design functional requirements of the CAR cooling system and the operation of the CAR fan motor. An appropriate justification for each conclusion is not  :

explicitly provided.

PDCE MP-2-89-020, Containment Air Recirculation (CAR) Fan Outlet Duet Expansion Joint This PDCE evaluates the replacement of the the fan outlet duct asbestos type  !

expansion joint to a non-asbestos material joint. The SE provides a good description, the reason for the changes and concludes that the change does not  ;

constitute an unreviewed safety question. It does not provide appropriate justification for each conclusion.

PDCE MP-2-89-023 ReactorBuildingClosedCoolingWater(RBCCW) Baseplate Modifications *

-This PDCE evaluates the modifications of RBCCW heat exchangers support assemblies to correct the effects of corrosion to the existing base plates and ,

anchor bolts. The SE provides a good description of the modification and concludes that the change does not constitute an unreviewed safety question with appropriate justification for each of the conclusions.  ;

PDCE MP-2-89-024, Modification of Pressurizer Safety Valve Support This PDCE evaluates the change involving increasing the bolt hole diameter '

of the pressurizer safety valve inlet support plates and the use of hardened steel washers with the valve bolting. The SE provides a restatement of the questions in a negative sense and thus concludes the change is not an unreviewed safety question.

PDCE MP-2-89-030, Pressurizer Proportional Heater removal from Operation Due to Low Insulation Resistance This PDCE evaluates the removal of the electrical energy from a pressurizer proportional heater. The SE provides a good description of the modif.ication and concludes that the change does not constitute an unreviewed safety question with appropriate justification for each of the conclusions.

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I PDCE MP-2-89-035 Power Supply X-1113A/B Replacement This PDCE evaluates the replacement of the original equipment 125 YDC power  ;

supplies for the Auto Auxiliary Feedwater Initiation System and the Anticipated Transient Without Scram System with Acopian Model #A120HT350. The SE concludes

.that the change does not constitute an unreviewed safety question with good i justification for each conclusion except question 7. Although it concludes that the margin of safety of the Technical Specifications would not be reduced, the SE does not support the conclusion in the discussion. It must be implied that, because there was a one for one replacement with equipment of equal or better quality, there would be no reduction of margin of safety.

PDCE HP-2-89-070 Fire Pump Pressure Switch Isolation This change installs an isolation valve in the instrument supply piping to the Fire Pump pressure control switch. The SE concludes that the change does not constitute an unreviewed safety question. It addresses affects on accidents which provides justification for questions 1 and 2. It bases conclusions <

relating to questions 3 through 7 on a general discussion of effect of the implementation of the change on the fire suppression water system by removing  ;

it from service in approximately_4 hours which is less than 7 days as allowed '

by the Technical Ssecifications. The justification to support the conclusions have been implied but not fully developed. ,

PDCE MP-2-89-083 Battery Charger Power Failure Relay I t L This change replaces a Furnas relay with a similar General Electric relay in l

the "B" Battery Charger. The relay is a power failure r91ay. The SE concludes '

that the change does not constitute an unreviewed safety question and provided for the proper conclusions based on the fact that the component is a one for L one electrical and seismic replacement. The SE does not fully develop justifi-p cation to support the responses to each of the seven questions.  ;

l l PDCE MP-2-89-114 Reactor Coolant Pump (RCP) Conduit improvements

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This change involves the replacement of 1-1/2 inch EHT conduit that extends into the motor cavity on all four Reactor Coolant Pumps with 1-1/2 inch Rigid conduit with approsriate seismic supports. The conduit is for the RCP speed sensor circuit. T11s is an improvement in that it provides seismic qualified supports for the RCP speed sensor circuit. The SE concludes that the change l- does not constitute an unreviewed safety question and justifies the response to the 7 questions based on the improvement in the seismic qualification of the speed sensor circuit. It fully develops the conclusion relating to effects of analyzed accidents and the margin of safety. The justification to support the responses relating to the other questions is implied. L i:

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7 July 9, 1990 Conclusions On Review of 10 CFR 50.59 Determinations The licensee has a good 10 CFR 50.59 determination process in place and improvement is evident but, as shown in the above discussions, there is room for improvement. This review used a process that was more rigorous than in the past and thus found evidence that the detail criteria of 10 CFR 50.59 was not fully supported in every case. .I indicated this to the licensee and encouraged the use of the guidance in NSAC/125 in developing future safety evaluations.

Because of the nature of the modifications I conclude that there is reasonable assurance that the changes meet the criteria of 10 CFR 50.59.

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Guy S. Vis',ing Senior Project Manager Project Directorate 1 4 Divisir,n of Reactor Projects . 1/11 Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosures See next page DISTRIBUTION (Beehoeftles NRCT 1.ocal'PDRs PDI.4(MemoFile)

S. Norris G. Vissing 0FC :LA:PDI.4 :PM:PDI.4 :PD:PDI  :  :  :

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NAME :SN rr s. m sing:gy :JStolz  :  :  :

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OFFICIAL RECORD COPY Document Name: MEMOT76448

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Conclusions On Review of-10 CFR 50.59 Determinations

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The: licensee has.a good 10_CFR 50.59 determination process in place and improvement is evident but, as shown in the above discussions, there is room.

for improvenent. This review used a process that-was more rigorous than in the a

n past and thus found evidence that the detail criteria of'10 CFR 50.59 was not j fully supported in every case. I indicated this to'the licensee and encouraged , ,

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-the.use of the guidance in NSAC/125 in developing future safety evaluations._  !

Because of the nature of the modifications, I conclude that there is reasonable-  !

Av. arsurance that the changes _ meet ~the criteria of 10 CFR 50.59.

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uy'S, Vissing, eniorProjectManager

' Project Direct ate ~I-4 .;

Division of Reactor Projects - I/II .

Office of Nuclear Reactor Regulation

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Enclosure:

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MILLSTONE 2 4

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STEAM GENERATOR REPLACEMENT PROJECT v ,", -OVERVIEW PRESENTATION q MAY 22, 1990 i ;

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AGENDA-N i.

INTRODUCTION

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REPLACEMENT SG SUBASSEMBLY INFORMATION 9

' SCHEDULE OF-REPLACEMENT--ACTIVITIES

• '1990 REFUEL-OUTAGE

-* MID-CYCLE-: ACTIVITIES

• '1992 REFUEL OUTAGE e:

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• PURCHASE-ORDER ISSUED TO BABCOCK & WILCOX CANADA.

IN FEBRUARY, 1988-FOR REPLACEMENT SGSA's!

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• DECISION TO PROCEED WITH ALL ACTIVITIES REQUIRED FOR-S/G REPLACEMENT.MADE:IN SEPTEMBER, 1989 L

'* t DEDICATED PROJECT TEAM ESTABLISHED - OCTOBER,~1989 L '

• CONTRACT AWARDED FOR INSTALLATION-SERVICES APRIL, 1990 i

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• A/EL (FLUOR DANIEL) ON SITE MAY, 1990 TO SUPPORT U

. REPLACEMENT ACTIVITIES '

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MILLSTONE 2 .

I STEAM GENERATOR REPLACEMENT

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PROJECT ORGANIZATION PROJECT MANAGER g- - - - - - - - - - i .

s a R.P. NECCI e MP2 e s- UNIT a esAlurry GUPERINTENDENTB ASSURAlWCg SECRETARY W. J. DfETZ 8 8 C.F. IJBBY E.D. VA11JERE mg K.A. ASURfHY . .-u.---.. - - - - - - . - - . .

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8 COST /8CIEEDUE2 FROJECT CONSTRUCTBON OPERATIO688 WOGEWEER REANAGER BEARAGER 8 B.A. KRAlfTH --J F.J. TANAFA S. OREFICE RA. WAY . J.G. RESETAR A.K. GUIESSERIAN g l - PLAfEENGINEERING

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- SECURTIY pg 'ENSTAIJATBON CIVIL - STAKr.UP - - - R.E. LEFEBVRE

-TURNOVER gg M.F. AHERN F.A. KOCON - OPERA'llONS 2/15/90

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. TOM BEASLEY CONSTRUCTION MANAGER BOB LAMMER GENERAL SUPERINTENDENT BILL BRENNAN ENGINEERING' MANAGER RICHARD BIELOWICZ STRUCTURAL ENGINEERING KEN: ROBERTS MECHANICAL ENGINEERING JOHN MADISON HEALTH PHYSICS

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• IMPROVED OPERATIONS AND MAINTENANCE FEATURES BLOWDOWN BELOW TUBESHEET FOR SECONDARY SIDE DRAINAGE MATERIAL CHANGES FOR A REDUCTION IN PERSONNEL RADIATION EXPOSURE NEW SINGLE NOZZLE DAM DESIGN LARGER PRIMARY MANWAYS AND SECONDARY SIDE MANDHOLES PRIMARY MANWAY HANDLING MECHANISM

• WIDE RANGE LEVEL INDICATION WET LAYUP RECIRC CONNECTIONS ON -22'6" LEVEL OF CONTAINMENT

• TWO LARGER SIZE (4") BLOWDOWN CONNECTIONS NOTE: BLOWDOWN PIPE EXITING CONTAINMENT WILL REMAIN AT PRESENT SIZE - 2" MANWAY OPENING LOCATED IN THE LOWER SECTION OF THE BOWL MODIFICATIONS TO STAGING AND SHIELDING FOR FUTURE OUTAGES

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DESIGN OBJECTIVE FOR IMPROVING TUBE! BUNDLE = RELIABILITY .

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, SELECTION OF COMPATIBLE-HIGHLY-CORROSION RESISTANT TUBE ANDLSUPPORT MATERIALS y:

i TUBE BUNDLE' DESIGN'WHICH-AVOIDS: TUBE VIBRATION AND-

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MATERIAL SELECTION =

TUBING

,-THERMALLY TREATED ALLOY 690 (existing-Mill annealed

'A1.loy,600).-

SELECTION BASED ON:

HIGHER-CORROSION RESISTANCE OVERALL TO THE'

-VARIOUS FORMS OF ATTACK; CURRENT. INDUSTRY STANDARD FOR REPLACEMENT

. DESIGNS--(exceeds industry-standard in areas of y ECT noise and-heat treatments);

RECOMMENDED BY EPRI

TUBE SUPPORTS e

TEhPERED;MARTENSITIC STAINLESS STEEL - MODIFIED 4]OS (existing-carbon steel)

E- SELECTION BASED ON:

HIGHER. CORROSION RESISTANCE; LOW TUBE WEAR POTENTIAL; AND STRENGTH

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.SELECTIONiOF COMPATIBLEEHIGHLY CORROSION RESISTANT' TUBE.ANDISUPPORT MATERIALS g.7 LTUBE' BUNDLE DESIGN WHICH AVOIDS-TUBE VIBRATION AND-9 MINIMIZES 1 SLUDGE DEPOSITION HIGH: SECONDARY SIDE CIRCULATION TUBELBUNDLELSUPPORT SYSTF.M TO ACCOMMODATE HIGH

-CIRCULATION

. TUBE-AND SUPPORT PROCESSING CONTROLS SECONDARY SIDE CHEMISTRY CONTROL FEATURES.

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DBJECTIVE:

AVOID CONCENTR ATION OF CHEMICAL IMPURITIES

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MINIMIZATION OFLCREVICE DEPTH'BETWEEN TUBE AND TUBESHEET ON SECONDARY SIDE.(s%-INCH) 2.

' LOW RESISTANCE' TUBE SUPPORT SYSTEM 3.

AVOIDANCE-0F FLUID CHANNELING IN U-BENDS

14. : HIGH:. CIRCULATION RATES 5.

SUBCOOLED' LIQUID ENTERING TUBE BUNDLE 6.

FLOW DISTRIBUTION OF ENTRANCE LIQUID 7.

VERIFICATION OF DESIGN WITH A 3-D THERMAL HYDRAULI MODEL (ATHOS, PATHOS, THIRST) i nae... is.. -2.asi 3,,,,,,,

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0BJECTIVE: IMPROVE ABILITY TO CONTROL' CHEMISTRY F .

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. INCREASED CAPACITY BLOWOOV!N' SYSTEM- (UP TO 7%)

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INTEGRAL COLD SHUT 00WN RECIRCULATION SYSTEM (one:

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BPSCIFICAT1'M CP REPLACDGNT STEAM GENERATORS

, aT MILLSTONE IDGT NO.-2 Design Feature Change '

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1.- Secondary Bandholes - - Increase number froml

' Improve access-for-2 to 4 ' sludge. lancing and-Increase s.ise.from inspections 6" to 8" L2. Primary Manvays ,

Increase sise from Improve' access s'nd.

16" to-16" final closure of

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Relocate for ease of

< manvays access t Provide accommodations-for aulti-bolt stud .

tensioner Include aanvay cover

-sanipulator 3.- Cladding Preparation Issrove surface' finish Reduce shutdovn by mechanical and/or radiation levels' electropolishing-

4. ~ Tubing & Cladding -

_Specify_both' low cobalt Chemistry Reduce'shutdovn tubing (0.015% saximum) radiation levels and cladding (0.1%

saximum)

5. . Tubing-Inspectability - Specify a signal to noise Ensure inspectable-ratio.(5:1 minimum at tubing-500 kEz based on ASME-calibration standard)

6. < Primary Noszle Dans -

1-2 minute installation / Reduce personnel

. removal time. radiation exposure and outage time

7. Secondary Shell-Drain' Provide complete draining -Eliminate the need a ' capability for pump:dovn=

operation

. 8. LTubesheet Narking -

Add tube location - Improve tube reference markings identification and reduce personnel radiation exposure.

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DIAMETER:0F CUT-16'-10" (202. INCHES) 10' .y Pb I REMOVE S/G IN.TWO: PIECES 5

f STEAM. DRUM WEIGHT -197 TONS.

, s- LENGTH 273 INCHES (22.75 FT) )

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. . DIAMETER 239 INCHES-0.D._(20 FT) J 4 I

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• UTILIZE TRUNNIONS AT SECONDARY MANWAY i- 1* INVERT DRUM '

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• PLACE ON 38'6" LEVEL -

• JREMOVE EXISTING MOISTURE SEPARATORS AND' DRYERS-  :

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• INSTALL NEW: SEPARATORS AND DRYERS EVAPORATOR AND-PRIMARY HEAD *

-* ATTACH BELLY BAND WITH TRUNNIONS, WELD TRUNNIONS TO SG SHELL OR, CUT LIFTING ATTACHMENTS THRU SG l SHELL-

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-*-LIFT, DEVICE LOWER ONTO 3B'6" AND ATTACH'TO-UPENDING- u '

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• TRANSLATE FROM' VERTICAL TO HORIZONTAL POSITION.

L AND-PLACE SG:ON MOVING SKID- ~

• LOWER TO 14'6" MOVE SG THROUGH EQUIPMENT HATCH in 3

d C'ONTAINMENT MODIFICATIONS '

h UPGRADE' POLAR CRANE WITH TEMPORARY LIFT TROLLEY g

• STRENGTH 14'6" FLOOR NEAR EQUIPMENT HATCE

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• INSTALL FLOOR OVER THE REFUEL POOL i^.

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,\_~ t RIGGING. METHODS AND CONTAINMENT MODIFICATIONS (CONTINUED)

KNOWN-CONTAINMENT INTERFERENCES

• 'A'. CAR FAN

• HYDROGEN RECOMBINER.ON 38'6" LEVEL

• MISSILE SHIELD

• NEUTRON SHIELD TANK AND SUPPORTS

• CEDM DUCTWORK

• HACSS

'* REACTOR HEAD _ STUDS AND STANDS

• '38'6" DECK GRATING-AND SUPPORT STEEL NEAR EQUIPMENT HATCH <

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'*. BARGE SLIP UPGRADE r

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• MAUL ~ ROUTE UPGRADE

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• -OFFICE SPACE - WHSE,7, BLDG 447

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• CRAFT,' ASSEMBLY FACILITY a

• MOCKUP, TP.AINING AND WELDING FACILITY, ~!

• WAREHOUSE,: STORAGE AND' FABRICATION SHOP i 4

• . CONTAINMENT ACCESS' BUILDING

• -INTERIM STORAGE FACILITY (IF REQUIRED).

• NDE FACILITY, LAYDOWN-AND STORAGE AREAS.

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. *. UPGRADE OF POLAR CRANE 2.

'- INSTALL RELIANCE TROLLEY ASSEMBLY g

- SEISMICALLY: SECURE' TROLLEY. '

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? . - NOTE:-POLAR 1 CRANE WILL REMAIN' OPERABLE '

l UTILIZING NORMAL CONTROL FEATURES 1'

- NEW~END TRUCK ~ MODIFICATION ON EXISTING 1 POLAR CRANE 3 ,

. - NEWLTROLLEY' SUPPLY RAIL INSTALLATION AND POWER' i-

• STRUCTURAL MODIFICATIONS '

- INSTALL. SUPPORTS FROM -2286"'TO'l6" 4 I NEAR EQUIPMENT. HATCH 'i I'

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-JPOSSIBLE BASE 1 PLATE" INSTALLATION ON -22'6"

'UNDERLEQUIPMENT HATCH '

- POSSIBLE'.BASEcPLATE' INSTALLATIONS OR'HILTI BOLT INSTALLATIONS FOR' PULL, POINTS AND/OR 1

PIPING RESTRAINTS

• 1 ENGINEERING WALKDOWNS  :

PLATFORM.AND STAGING

- PDCR AS BUILT' DIMENSION CHECKS

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• INTERFERENCE CHECKS

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4 REPLACEMENT

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-* AWARD INSTALLATION CONTRACT 4/90

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< *LENGINEERING PACKAGES.FOR-1990 8/90 '

LREFUEL' OUTAGE (3 MONTH PREP)

• -1990 REFUEL OUTAGE 10/90

'* NEW SG'. DELIVERY 4/91 l

• ' COMPLETE REPLACEMENT ENGINEERING. 6/91  !

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• APPROVE' PDCR s :FOR REPLACEMENT - 9/91

. TRAINING AND QUALIFICATION 3/92 (6, MONTHS REQUIRED)~. .c

• -PLANNED REPLACEMENT OF SG's-

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(DURATION.4'TO 6 MONTHS) /

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