ML19263D707

Generator Repair Rept, Revision 1

ML19263D707
Person / Time
Site:	Palisades
Issue date:	03/31/1979
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18043A612	List: ML18043A611 ML19263D707
References
NUDOCS 7904130173
Download: ML19263D707 (37)

Text

3 f CONSUMERS POWER COMPANY PALISADES PLANT STEAM GENERATOR REPAIR REPORT DOCKET NO 50-255 LICENSE NO DPR-20 At the request of the Commission and pursuant to the Atomic Energy Act of 1954 and the Energy Reorganization Act of 1974, as amended, and the Commission's Rules and Regulations thereunder. Consumers Power Company submits the

" Palisades Plant Steam Generator depair Report," original (submitted January 3, 1979), which describes the potential repair prcgram based on complete replacement of the existing steam generators if major repairs become necessary.

CONSUMERS POWER COMPANY t

By d C R Bilby, Vice Prgsi ent Production & Transk sion Sworn and subscribed to before me this 10th day of April 1979.

() ) )

w N ): W /2) / D . ( /l ,/2 / / 0 N Y Uinda R Thayer, Notar,y Public Jackson County, Mi%higan My commission expires July 9, 1979.

790113003!

i .

CONSUMERS POWER COMPANY PALISADES PLANT STEAM GENERATOR REPAIR REPORT DOCKET NO 50-255 LICENSE NO DPR-20 At the request of the Commission and pursuant to the Atomic Energy Act of 1954 and the Energy Reorganization Act of 1974, as amended, and the Commission's Rules and Regulations thereunder, Consumers Power Company submits the

" Palisades Plant Steam Generator Repair Report," Revision 1 March 1979, which describes the potential repair progra n based on complete replacement of the existing steam generators if major repairs become necessary.

CONSUMERS POWER COMPANY t

By C R Bilby, Vice Pres'ifent -

Production & Transmission ,

Sworn and subscribed to before me this 10th day of April 1979.

[?f)f//))/0? ' A/NA)

J.inda R Thayer, Notary Public Jackson County, Michigan My commission expires July 9, 1979.

REMOVE INSERT 1 Rev. O i Rev. 1 ii Rev. O ii Rev. 1 iv Rev. O iv Rev. 1 viii Rev. O viil Rev. 1 LOEP-1 Rev. O LOEP-1 Rev. 1 LOEP-2 Rev. O LOEP-2 Rev. 1 LOEP-5 Rev. O LOEP-5 Rev. 1 LOEP-6 Rev. 1 1-4 Rev. O l-4 Rev. 1 1-4a Rev. 1 1-6 Rev. 0 1-6 Rev. 1 1-7 Rev. 1 2-5 Rev. 0 2-5 Rev. 1 2-6 Rev. 0 2-6 Rev. 1 2-7 Rev. 0 2-7 Rev. 1 2-11 Rev. 0 2-11 Rev. 1 Table 2.1-1 Rev. O Table 2.1-1 Rev. 1 Table 2.1-2 Rev. O Table 2.1-2 Rev. 1 Table 2.2-1 Rev. O Table 2.2-1 Rev. 1 Figure 2.2-1 Rev. O Figure 2.2-1 Rev. 1 Figure 2.2-2 Rev. O Figure 2.2-2 Rev. 1 -

Figure 2.2-3 Rev. O Figure 2.2-3 Rev. 1 Figure 2.2-4 Rev. O Figure 2.2-4 Rev. 1 Figure 2.2-10 Rev. O Figure 2.2-10 Rev. 1 3-3 Rev. 0 3-3 Rev. 1 3-4 Rev. 1 6-1 Rev. 0 6-1 Rev. 1 6.la Rev. 1 6-lb Rev. 1 6-lc Rev. 1 6-ld Rev. 1 6-le Rev. 1 6-lf Rev. 1 6-lg Rev. 1 6-lh Rev. 1

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s PALISADES PIANT STEAM GENERATOR REPAIR REPORT TABLE OF CONTENTS Page LIST OF EFFECTIVE PAGES LOEP-1

1. 0 INTRODUCTION,

SUMMARY

, AND CONCLUSIONS 1- 1 1.1

SUMMARY

OF STEAM GENERATOR REPAIR PROGRAM 1- 2 1.1.1 REPAIR ALTERNATIVES 1-2 1.1.2 REMOVAL AND REPIACEMENT OF THE STEAM 1-3 GENERATORS FROM CONTAINMENT 1.1.3 STEAM GENERATOF CHARACTERISTICS 1-3 .

1.1.4 SAFETY-RELATED CONSIDERATIONS 1-4 C

1.1.5 ALARA CONSIDERATIONS 1-4 1.1.6 OFFSITE RADIOLOGICAL CONSEQUENCES 1-4 1.1.7 UNIQUE ASPECTS OF PROGPAM 1-4a 1.1.8 STEAM GENERATOR DISPOSAL 1-5

1. 2 IDENTIFICATION OF PRINCIPAL AGENTS 1-5 1.3 10 CFR 50. 59 CONSIDER ATIONS 1-6

1. 4 CONCLUSIONS 1-6 2.0 REPLACEMENT COMPONENT DESIGN 2-1 2.1 COMPARISON WITH EXISTING COMPONENT DESIGN 2-1 2.2.1 PARAMETRIC COMPARISON 2- 1 2.1.2 PHYSICAL COMPATIBILITY WITH EXISTING 2-2 STEAM GENERA'K;R AND SYSTEMS 2.1.3 ASME CODE APPLICATION 2-2 2.1.4 REGULATORY GUIDE APPLICATION 2-3 i MARCH 1979 REV. 1

t PALISADES PLANT SGPR 4.7 OUALITY ASSURANCE 4-47 4.7.1 CONSUMERS POWER COMPANY QUALITY 4-47 ASSURANCE PROGRAM 4.7.2 BECHTEL POWER CORPORATION QUALITY 4-47 ASSURANCE PROGPAM 4.7.3 COMBUSTION ENGINEERING POWER SYSTEM 4-47 GROUP NUCLEAR QUALITY ASSURANCE PROGPAM 4.8 REGULATORY GUIDE APPLICABILITY TO REPAIR 4-47 PROGRAM 4.9 SCALE MODEL OF THE PALISADES PLANT CONTAINMDIT 4-55 5.0 RETURN TO SERVICE TESTING 5-1 6.0 SAFETY EVALUATIONS 6-1 -

t 6.1 FSAR EVALUATIONS 6-1

(

6.

1.1 INTRODUCTION

6-1 6.1. 2 NON-LOCA ACCIDENTS 6-1 6.1.3 LOSS-OF-COOLANT ACCIDENT EVALUATION 6-1d 6.1.4 CONTAINMENT PRESSURE ANALYSIS 6-1f 6 .1. 5 FSAR EVALUATION CONCLUSION 6-1g 6.2 CONSTRUCTION-REL ATED EVALUATIONS 6-2 6.2.1 HANDLING OF HEAVY OBJECTS 6-2 6.2.2 OFFSITE RADIOACTIVE RELEASES AND 6-2 DOSE ASSESSMENT 6.3 FIRE PROTECTION 6-4 6.3.1 EXISTING FIRE PROTECTION 6-4 6.3.2 FIRE PROTECTION DURING THE REPAIR 6-5 PROGRAM 6.

3.3 CONCLUSION

6-8 iv MARCH 1979 REV. 1

PALISADES PLANT SGRR LIST OF FIGURES FIGURE NO. TITLE 2.2-1 REPLACEMENT STEAM GENERATORS 2.2-2 DELETED 2.2-3 DELETED 2.2-4 BOTTOM BLOWDOWN DUCT ASSEMBLY 2.2-5 TUBE SUPPORT TYPES 2.2-6 EGGCRATE ASSEMBLY 2.2-7 BEND REGION TUBE SUPPORT 2.2-8 TUBE SUPPORT 2.2-9 UPPER ASSEMBLY 2.2-10 STEAM GENERATOR - FLOW RESTRICTOR NOZZLE 2.2-11 PRIMARY HEAD DRAINS 3.1-1 EXISTING BLOWDOWN AND RECIRCULATION SYSTEM 3.1-2 MODIFIED BLOWDOWN AND RECIRCULATION SYSTEM 3.3-1 EXISTING TURBINE ANALYZER PANEL FOR SAMPLING SYSTEM -

3.3-2 MODIFIED TURBINE ANALYZER PANEL

( FOR SAMPLING SYSTEM 3.4-1 PRIMARY HEAD DRAIN SYSTEM 3.5-1 WIDE RANGE LEVEL TRANSMITTER 4.1-1 SITE PLAN 4.1-2 BARGE SLIP 4.1-3 OLD STEAM GENERATOR STORAGE FACILITY PLAN VIEW 4.1-4 OLD STEAM GENERATOR FACILITY SECTIONS AND DETAILS 4.1-5 CONTAINMENT LAYDOWN AREAS 4.1-6 GENERAL ARRANGEMENT PLAN VIEW, SH.1 4.1-7 GENERAL ARRANGEMENT PLAN VIEW, SH.2 4.1-8 GENERAL ARRANGEMENT SECTION A-A 4.1-9 GENERAL ARRANGEMENT SECTION B-B 4.1-10 DOWN-ENDING STEAM GENERATOR ONTO SLEDS 4.1-11 LOWERING STEAM GENERATOR FROM ELEVATOR PLATFORM ONTO TRANSPORTERS 4.1-12 STEAM GENERATOR IN HOISTED POSITION, SECTION VIEW 4.1-13 STEAM GENERATOR ON TRANSPORTER BETWEEN STORAGE AND CONTAINMENT 4.3-14 STEAM GENERATOR ON TRANSPORTER BARGE TO STORAGE 4.1-15 OFF LOADING STEAM GENERATOR FROM BARGE, PLAN VIEW 4.1-16 OFF LOADING STEAM GENERATOR FROM BARGE, viii MARCH 1979 REV. 1

PALISADES PLANT STEAM GENERATOR REPAIR REPORT LIST OF EFFECTIVE PAGES Page Latest Identification Amendment i 1 il 1 iii 0 iv g 1

v 0 vi 0 vii ,

0 v111 1 l

ix 0 x 0 LOEP-1 1 LOEP-2 1 LOEP-3 0 IDEP-4 0 LOEP-5 1 LOEP-6 1 1-1 0 1-2 0

_1-3 _

o 1-4 1 i 1-4a 1 1 1-5 0 1-6 (

1-7 1 2-1 0 2-2 0 2-3 0 2-4 0 2-5 1 2-6 3 2-7 4 2-8 0 2-9 0 2-10 0 2-11 1 l 2-12 0 2-13 0 Tb1 2.1-1 1 Tb1 2.1-2 1 Tb1 2.2-1 1 Fig. 2.2-1 1 Fig. 2.2-2 1 Fig. 2.2-3 1 Fig. 2.2-4 1 MARCH 1979 LOEP-1 REV. 1

PALISADES PLANT SGRR Page latest Identification Amendment Fig. 2.2-5 0 Fig. 2.2-6 0 Fig. 2.2-7 0 Fig. 2.2-8 0 Fig. 2.2-9 0 Fig. 2. 2- 10 1- l Fig. 2.2-11 0 3-1 0 3-2 0 3-3 1 3-4 1 Fig. 3.1-1 0 Fig. 3.1-2 0 Fig. 3.3-1 0 Fig. 3.3-2 0 Fig. 3.4-1 0 Fig. 3.5-1 0 4-1 0 4-2 0 4-3 0 4-4 0 4-5 0 4-6' 0 4-7 0 4-8 0 4-9 0 4-10 0 4-11 0 4-12 0 4-13 0 4-14 0 4-15 0 4-16 0 4-17 0 4-18 0 4-19 0 4-20 0 4-21 0 4-22 0 4-23 0 4-24 0 4-25 0 4-26 0 4-27 0 4-28 0 4-29 0 EP-2 MARCH 1979 REV. 1

PALISADES PLANT SGRR Page Latest Ide nt i fication Amendment Fig. 4.3-3 0 Fig. 4.3-4 0 Fig. 4. 3 - 5 0 Fig. 4.3-6 0 Fig. 4.3-7 0 5-1 0 5-2 0 6-1 1 6-1a 1 6-1b 1 6-1c 1 6-1d 1 6-le 1 6-1f 1 6-1g 1 6-lh 1 ,

6-2 0 6-3 0 .

6-4 0 6-5 0 6-6 0 6-7 0 6-8 0 Table 6.2-1 0 Table 6.2-2 0 Table 6.2-3 0 Table 6.2-4 0 Table 6.2-5 0 7-1 0 7-2 0 7-3 0 7-4 0 7-5 0 8-1 0 8-2 0 823 0 8-4 0 8-5 0 8-6 0 LOEP-5 MARCH 1979 REV. 1

PALISADES PLANT SGRR Page Latest Identification Amendment 8-7 0 8-8 0 Table 8.4-1 0 Table 8.8-1 0 9-1 C 9-2 0 9-3 0 10-1 0 LOEP-6 MARCH 1979 REV. 1

PALISADES PLANT SGRR techniques and American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) editions.

1.1.4 SAFETY-RELATED CONSIDERATIONS The potential impacts of the replacement steam generators on the design basis events, analyzed in the FSAR, have been evaluated and are described in Section 6.1. Those evaluations conclude that the replacement steam generators will have no significant adverse impact on the design basis events.

Construction-related incidents pertaining to the transportation and handling of the steam generators have been evaluated and are described in Sections 4.1.4 and 6.2.

Those evaluations conclude that the constr"ction activities will have no significant adverse impact on the safety of the Palisades Plant.

The fire prevention and protection program to be in operation during the steam generator repair program is described in Section G.3. It is concluded that these measures provide reasonable assurance that tne Palisades Plant is protected from significant damage due to fire during the repair activities.

1.1.5 ALARA CONSIDERATIONS Personnel exposure will be maintained "as low as is reasonably achievable" (ALARA) th roughou t the steam generator repair program.

Estimates of the exposures to personnel involved in the various repair alternatives have been developed using projections of work activity durations, manpower levels, and expected radiation levels.

1.1. 6 OFFSITE RADIOLOGICAL CONSEQUENCES Radiological evaluations of the gaseous and liquid releases attributable to the steam generator repair have been conducted. The effects of the releases are less than those associated with normal operation of the f acility on the basis of the discussion in Section 6.2.2.

1-4 MARCH 1979 REV. 1

PALISADES PLANT SGRR 1.1.7 UNIQUE ASPECTS OF PROGRAM As presently contemplated, there are no unique engineering or construction aspects of the Palisades Plant s team generator repair program. The repair program, including the fabrication of replacement units, will utilize conventional nuclear industry manuf acturing and construction methods.

The shop fabrication of the steam generators will be conducted in accordance with standard shop practices. The closure of the temporary construction opening in the containment will be performed in a manner similar to that used to close the original containment construction opening.

The transport and rigging of *.he steam generator will utilize proven techniques. In short, the repair program will rely on f abrication and construccion practices or techniques which have been previously qualified for similar applications.

9 l-4a MARCH 1979 REV. 1

PALISADES PLANT SGRR 1.3 10 CFR 50.59 CONSIDERATIONS Repair or replacement of equipment at a power plant, performed in accordance with appropriate procedures, is a maintenance activity that is routinely conducted. Because of the scope of the steam generator repair, it was considered prudent to evaluate this activity to determine:

a. If the proposed repair activity would involve a change in the technical specifications incorporated in the licence.

b. If the proposed repair activity would involve an unreviewed safety question per the requirements of 10 CFR 50. 59 (a) (2) .

Each design basis event FSAR accident analysis has been evaluated, and it has been concluded that the replacement steam generators would not alter the conclusions reached in the FSAR. The evaluation indicates that the steam generator repair does not involve an unreviewed safety question. A change in the plant technical specifications is required to incorporate the main steam line isolation on high containment pressure discussed in Section 3.6 of this report.

The construction incident potential has been evaluated to determine the presence of any new or unique accidents and the potential impact on cooling spent fuel. The evaluation indicates that the steam cenerator repair activity does not involve an unreviewed saf 'y question.

Additionally, before replacement of the Steam generators, the ef fective Palisades Plant Technical Specifications will be reviewed again and revised as necessary.

1.4 CONCLUSION

S The fundamental conclusions reached are that the steam generator repair can be conducted utilizing proven manufacturing and construction techniques and that the repair program does not result in any significant adverse impact on the plant safety analysis and the ability to maintain a safe configuration and cool stored fuel.

Additionally, current FSAR safety analyses are applicable to the replaced steam generators. The detailed bases 1-6 MARCH 1979 REV. 1

PALISADES PLANT SGRR supporting these conclusions are provided in the report that follows.

O l-7 MARCH' 1979 REV.

PALISADES PLANT SGRR

j. Regulatory Guide 1.84, Code Case Acceptability -

ASME III Design and Fabrication (August 1977)

k. Regula tory Guide 1.8 5, Code Case Acceptability -

ASME III Materials (August 1977) 2.2 COMPONENT DESIGN IMPROVEMENTS The replacement steam generator design incorporates the traditional Combustion design features and design improvements that have evolved through several generations of steam generator designs in response to the operational steam generator - 'lems that have occurred in the nuclear industry.

The replacement steam generators will essentially duplicate the physical, thermal, and hydraulic characteristics of the original units while incorporating a combination of features proven in field operation and design imp rove men ts to mitigate operttional problems.

The heating surface has been selected to provide thermal performance which would match that presently installed and to respond to plant thermal transients in the same manner as does the existing unit. The design will provide improvements in thermal / hydraulics, notably in secondary flow distribution. These improvements are intended to minimize flow stagnation, steam blanketing, and harmful solids accumulations. It is important to avoid harmful solids deposits in contact with heat transfer tubing in the steam generators. The blowdown arrangements are designed to l take advantage of the improved flow distribution, making significant imp rove me nts in the effectiveness of blowdown in removing harmful solids deposits.

The tube support system utilizes the traditional Combustion eggcrate tube support, with its low flow res is tance , support against vibration and wear, and resistance to tube denting or lateral tube deformation.

The bend region tube support system also uses the standard Combustion approach, with double 90 degree bends and support assemblies of interlocking strips. The design provides positive restra int agains t vibration and resistance to tube deformations during LOCA, steam line break, and seismic events. The tube support system provides rugged, positive support while minimizing flow resistances and the possibility of local dryout of steam blanketed regi.ons.

2-5 MARCH 1979 REV. 1

PALIS ADES PLANT SGRR Access openings and inspection ports are provided to enable inspection of tubes, tubesheet, and support surfaces within the tube bundle as well as at the periphery of the bundle.

2.2.1 DESIGN FE ATURES TO IMPROVE PERFORMANCE 2.2.1.1 Thermal Performance In order to minimize the effect on plant transient performance, heat transfer tubes of 3/4 inch outside diameter and .042 inch average wall thickness (consistent with Combustion's System 80 design) will be provided in such quantities and lengths on the replacement units that the product of their areu (A) and heat transfer coefficient (U) equals the product of the original area and original coefficient, i.e., (UA) new = (UA) original. The total cross-sectional flow area of the heat transfer tubes will be equal to the cross-sectional flow area of the tubes in the original units. The cor. trol of these parameters on the replacement units allows the hydraulic impedance to primary flow to essentially correspond with that of the original steam generators.

2.2.1.2 Deleted 2-6 MARCH 1979 REV. 1

PALISADES PLANT SGRR 2.2.1.3 Blowdown Capability The potential blowdown capabilities for the replacement steam generators are designed to take advantage of the improvements in corpora ted , which contribute to the imp roved secondary flow distribution and hydraulics within the operating steam generator. The recirculating fluid exits from the downcomer and flows radially across the tube bundle. The lowest fluid. velocities occur near the center open region of the tube bundle where dropout of solid particles may occur.

In this region, free of heat transfer tubes, blowdown duct that takes suction in a circular pattern adjacent to the innermost tubes is provided. Figure 2.2-4 shows the schema tic arrangement of the blowdown duct and its relationship to the tube bundle. At the end of each circumferential section of the blowdown duct, a transport duct (with no blowdown openings) carries the blowdown fluid across ti divider lane discharging through intersecting holes drilled in the tubesheet to a 6-inch Schedule 80 blowdown nozzle. See Section 3.1 for a description of the connecting blowdown system. The internal blowdown duct and tubesheet blowdown connection for the replacement units have been sized to accommodate future higher blowdown capabilities than those available on the original steam generators. .

2-7 MARCH 1979 REV. 1

PALISADES PLANT SGRR 2.2.2 DESIGN FEATURES TO IMPROVE MAINTENANCE AND INSPECTION 2.2.2.1 Handholes The replacement steam generators will include four 6-inch handhole openings on the loser and intermediate shells to facilitate inspections. The lower two handholes will be positioned just above the tubesheet and have provisions for viewing through the tube bundle shroud. l The upper handholes are located just above the eggerate in the tube lane and are adjacent to the bend region of the tube bundle. These handholes will also incorporate the provision for viewing through the tube bundle shroud (see Figures 2.2-1 and 2.2-10).

2.2.2.2 Inspection Ports Two 2-inch inspection ports will be added to the replacement units just above the tubesheet secondary f ace to provide accessibility to the tubesheet surface and to allow use of an inspection device such as a boroscope to observe tubes on either side of an opening between two particular tube rows.

2.2.2.3 Deleted 2.2.2.4 Primary Head Drains To facilitate draining of the steam generator primary head before maintenance or inspection activities in this area, the replacement units include a drain nozzle (see Figure 2.2-11) on both the inlet and outlet plenums of the primary head. See Section 3.4 for a description of the connecting drain system.

2-11 MARCH 1979 REV. 1

PALIS ADES PLANT SGRR TABLE 2.1-1 STEAM GENERATOR COMPARISON DATA

• Original Replacement Steam S te am A. Primary Side Generators Generators

1. Thermal power, MWt 2450 2450

2. Design pressure, psi 2500 2500

3. Design temperature, 'F 650 650

4. Cold leg temperature, 'F 547.8 547.8

5. Hot leg temperature, *F 598.5 598.5

6. Coolant flow, 106 lb/hr 62.25 62.25 7 Calculated pressure drop, psid 30.5 29.5

8. Normal operating pressure, psi 2100 2100 B. Secondary Side

1. Design pressure, psi 1000 1000

2. Design temperature, 'F 550 550

3. Flow rate, 10" lb/hr 5.281 5.281

4. S team outlet pressure, psi 770 770 -

5. Fee,dwater temperature, 'F 429.1 429.1 C. Dimensions

1. Evapo,rator outs ide diame ter, in 164 164

2. Steam drum outside diame ter, in 239-3/4 239-3/4

3. Overall length, in 709.78 742.00 l

4. Tubing outside diame ter, in 0.750 0.750

5. Tubing wall thickness, in .048 .042 D. Hydrostatic Pressure

1. Primary, psia 3125 3125

2. Secondary, psia 1250 1250 E. Weights and Volumes

1. Complete vessel dry, lb 924,596 934,637

2. Vessel C. G. dry, in 345.32 344.02

3. Secondary fluid 0% power, lb 209,180 207,771

4. Secondary fluid 100%

power, lb 129,164 147,288 Note:

"' Values are per steam generator, except Item A.l .

MARCH 1979 REV. 1

PALISADES PLANT SGRR TABL E 2.1-2 REPLACEMENT STEAM GENERATOR DATA "'

--Safety Analysis- Design A. Primary Side

1. Thermal power, MWt 2530 2650

2. Design pressure, psi 2500 2500

3. Design temperature,*F 650 650

4. Cold leg temperature,'F 542.5 547.8

5. Hot leg temperature,*F 595.4 598.5

6. Coolant flow, 106 lb/hr 62.5 70.0 7 Calculated pressure drop, 29.6 36.1 psid

9. Normal operating 2100 2250 pressure, psi B. Secondary Side

1. Design pressure, psi 1000 1000

2. Design temperature,*F 550 550 -

3. Flowrate, 10 6 lb/hr 5.491 5.786

4. , Steam outlet pressure, 770 770 psi

5. Feedwater temperature,*F 435 438 C. Weights and Volumes

1. Complete vessel dry, lb 934,637 934,637

2. Vessel C.G. dry, in 344.02 344.02

3. Secondary fluid 0% power 207,771 207,771 lb

4. Secondary fluid 100% 145,969 143,934 power, 16 NOTE:

"' Values are per steam generator, except I tem A .l .

MARCH 1979 REV. 1

PALISADES PLANT SGRR TABLE 2.2-1 STEAM GENERATOR MATERI ALS Original Replacement Steam Generators Steam Generators Upper, inter- SA-302, Grade B SA-533, G rade A ,

mediate, and alloy steel Class I alloy cone shells steel Lower shell S A- 516, G rade 70 S A-5 3 3, G rade A, carbon steel Class I alloy steel Tubesheet forging SA-508, Class II SA-508, Class III alloy steel alloy steel Tube support plates SA-36 --

ca rbon steel Eggcrate tube supports A-570, Grade D/ A-176, Type 409 A-303-64, Grade D stainless steel carbon steel Primary head SA- 30 2, G rade B SA-5 3 3, G rade B ,

alloy steel Class I alloy steel Primary head clad Stainless steel Stainless steel Tubesheet clad Inconel Inconel Heat transfer tubing SB-163 SB-163 Inconel Inconel Secondary head SA-516, Grade 70/ SA-516, Grade 70 SA-302, Grade B carbon steel carbon steel /

alloy steel Nozzles / primary stay SA-508, Class II SA-508, Class III alloy steel alloy steel MARCH 1979 REV. 1

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PALISADES PLANT SGRR 3.5 WIDE RANGE LEVEL INDICATION A differential pressure type level transmitter will be added to each new steam generator to provide steam generator wide range level indication of about 44 feet. The top head of the new steam generators will have a 1-inch no: sie at el 671', which will be used for the low-pressure sensing line connection. The high-pressure sensing line will be connected to pressure taps at el 627' (see Figure 3.5-1).

With this addition, an operator can determine the water level in the secondary side of each steem generator during wet layup (or similar operations) beyond the ranle measurable with the present level indicators (about 15 feet).

The new level indicating system is functionally independent, both electrically and mechanically, of any safety-related systems. The sensing lines will be in accordance with ASME Code Section III, Class 2 and seismic Category I classifications. The transmitter will be located in a low radiation zone. The transmitter output will electrically connect to a new indicator in the main control room and will not be used to automatically initiate or terminate any action.

J . ,6 MAIN STEAM ISOLATION VALVE CLOSURE SIGNAL The existing circuitry provides for closure of the Main Steam Isolation Valves based on low steam generator pressure. This closure signal is provided to protect against excessively high releases of steam to the containment as a result of a s team line break.

The replacement steam generators have steam nozzle flow restrictors to restrict the blowdown rate following a steam line break. The effect of the flow restrictors is to reduce the rate of steam generator pressure change during blowdown.

Hence, following a design basis steam line break with the replacement steam generators, the high containment pressure trip setpoint will be reached befor' the low steam generator pressure trip is reached.

The Main Steam Isolation Valve Closore signal will be modified to be actuated f rom high containment pressure as well as low steam generator pressure. This will reduce the mass / energy release following a steam line break and result in lower containment peak pressures.

3-3 MARCH 1979 REV. 1

PALISADES PLANT SGRR The containment pressure instrumentation and circuitry are safety grade.

3-4 MARCH 1979 REV. 1

PALIS ADES PLANT SGRR 6.0 SAFETY EVALUATIONS 6.1 FSAR EVALUATIONS 6.1.1 INTRODUCfION The purpose of this section is to evaluate the impact, if any, of the replacement steam generators on the accident analysis trans ien ts for the Palisades Plant. Under the guidelines specified in 10 CFR 50.59, such an evaluation is required to verify that no unreviewed safety concerns or changes to the Palisades Plant Technical Specifications occur. This section provides a qualitative discussion of the effect on the accident analyses of steam generator parameter changes resulting f rom steam generator repair.

Conclusions are made in this section concerning the validity of the original FS AR to the repaired units. Consistent with the requirements of 10 CFR 50.59, licensing regulations and guidelines of the original licensing of the Palisades Plant are assumed to apply, and only changes in the saf ety analyses due to the equipment changes are considered.

The relevant plant operating parame ters and steam generator design parameters have been compared for the original and replacement steam generators in Section 2.1. While incorporating design imp roveme n ts that will improve the flow distribution and tube bundle accessibility and reduce secondary side corrosion, the replacement steam generators continue to match the design performance of the original steam generators. I t may be noted f rom Section 2.1 that there is very little effect on plant operating parameters due to the replacement of the steam generators. It is, therefore, to be anticipated that the impact on the accident analyses will be insignificant. The results of the accident evaluation show that no unreviewed safety concerns exist because of operation with the replacement steam generators.

6.1.2 NON-LOCA ACCIDENTS Analyses of the following non-LOCA design basis events were originally presented in the Palisades Plant FSAR. These events were evaluated to de te rmine the ef f ect, if any, of the replacement steam generators on the plant transient response,

a. Control rod withdrawal

b. 3oron dilution 6-1 MARCII 1979 REV. 1

PALISADES PLANT SGRR

c. Full-length control rod drop

d. Malpositioning of the part-length control rod group

e. Loss of coolant flow

f. Idle loop startup

g. Excessive feedwater

h. Excessive load increase

i. Loss of load

j. Loss of feedwater flow

k. Steam line rupture inside containment

1. Steam generator tube rupture

m. Control rod e jection The excessive feedwater, excessive load increase, loss of .

load, loss of feedwater flow, steam line rupture, and steam

- generator tube rupture events are discussed in the succeeding subsections. The remaining events are primarily core-related and are not significantly af fected by the replacement of the s team genera tors.

6.1.2.1 Excessive Feedwater An excessive feedwater transient may be caused by a decrease in feedwater temperature or by an increase in feedwater flow. These conditions primarily af fect reactor coolant parameters due to the resulting excessive heat removal froa.

the primary system. Since the feedwater system flowrates have not been altered, the replacement of the steam generators has no effect on the results for this t ra ns ien t and the consequences will be no more adverse chan those determined in analyses for the FSAR or the power uprating submittal.

6.1.2.2 Excessive Load Increase The excess load transient may be initiated by an inadve rt ent opening of the turbine control valves, atmospheric steam dump valves, and/or steam bypass valve. The ensuing tra nsient causes a high power level trip to protect the 6-la MARCH 1979 REV. 1

PALISADES PLANT SGRR reactor core. At hot standby conditions, there may be an excessive reduction in the steam generator water inventory.

However, the time required to boil the steam generators dry is in excess of that time predicted to empty the s team generators during a loss of feedwater flow tra ns ien t . Since there are no changes in the valves identified above as the potential initiating mechanisms, and since an excess load transient is much less severe than a steam line break, the consequences of this transient are no more adverse with the replacement steam generators than those results reported for previous analyses. In addi tion, the results of these analyses are bounded by those of the main steam line break and loss of feedwater flow events.

6.1.2.3 Loss of Load A loss of load transient leads to a rapid (and large) reduction in the power demand while the reactor is operating at full power. There is a corresponding reduction in the rate of heat removal f rom the primary coolant system. This leads to elevated pressurizer and steam generator pressures that cause the pressurizer and steam generator safety valves to open to minimize the peak primary and secondary pressures. Additional protection is provided by the high pressurizer pressure trip. As a result of the actuation of the reactor trip and the opening of the valves, the peak primary and secondary system pressures are no more adverse with the replacement steam generators than they were when the analyses were performed using the characteristics of the origins ~ steam generators.

6.1.2.4 Loss of Feedwater Flow A complete loss of feedwater flow may be initiated by a rupture of the feedwater crossover line downstream of the main feedwater pumps or a condensate pump failure, which would cause low suction pressure on both f eedwater pumps.

The prima ry consequence of this accident is the reduction in, and eventual loss of, the primary coolant system heat sink. An analysis of this transient assuming the installation of the replacement steam generators shows that the consequences of this accident are no more adverse than those reported for analyses using the original steam generator characteristics. In fact, the increase in secondary water inventory for the replacement steam generators increases the predicted time to empty the steam generators.

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PALISADES PLANT SGRR 6.1.2.5 Steam Line Break A rupture in a main steam line would increase the rate of heat extractions of the steam generators and cause a rapid temperature reduction in the primary coolant. The f as tes t blowdown and the most rapid reactivity addition are associated with a break located at a steam generator nozzle.

The replacement steam generators were analyzed for the full power Main Steam Line Break (MSLB) event using the NSSS simulation code version which is consistent with the analytical methods use for the corresponding FSAR analysis.

Although the replacement steam generators have the increased water inventory (see Tables 2.1-1 and 2.1-2), the effect of including the steam nozzle flow res trictors is to decrease maximum power during the full power MSLB for the replacement steam generators. Therefore, the results of the full power MSLB event with the replacement steam generators are no worse than the results reported in the FSAR except for a slight increase in the total blowdown flow. The effect of this slight increase of blowdown flow is discussed in Section 6.1.4.

6.1.2.6 Steam Generator Tube Rupture -

A steam generator tube rupture is a penetration of the barrier between the primary coolant system and the main steam system. A double-ended (g u illo tine ) rupture of a steam generator U-tube at the tube shee t is postulated.

Integrity of the barrier between the primary coolant system and main steam system is radiologically significant, since a leaking steam generator tube allows transport of primary coolant into the main s team system. Radioactivity contained in the reactor coolant can then mix with shell-side water in the af fected s team generator and be expelled to the atmosphere. During normal plant operations, some of this radioactivity is transported through the turbine to the condenser, where the noncondensable radioactive materials are released via the condenser air ejectors.

The steam generator tube rupture event has been analyzed because the tube inside diameter is larger for the replacement s team generators than for the original steam generators. For a double-ended tube rupture within the replacement s team generators , the larger break area could result in a higher primary-to-secondary leak rate. The re-analysis confirms that the fluid leak rate with the replacement s team generators is higher, during the initial stages of this transient. However, the escalated decrease 6-l e MARCH 1979 REV. 1

PALIS ADES PLANT SGRR in the primary coolant inventory leads to an earlier reactor trip on low-pressurizer pressure and an earlier emptying of the pressurizer for this transient. Because during this transient, the reactor remains at power for a considerable period of time, a noticeable reduction in the time to trip the reactor causes a reduction in the total primary coolant activity transferred to the secondary side of the steam generators. Thus, a reduction in time to trip the reactor also reduces the total curie content transferred from the secondary side of the steam generators to atmosphere via the atmospheric dump valves or s team genera tor saf ety valves.

The overall impact is that the radiological releases from the steam generator tube rupture event are no worse than the values reported in the corresponding FSAR analysis.

6.1.3 LOSS-OF-COOLANT ACCIDENT EVALUATION A major primary coolant system pipe break would result in a rapid depressurization of the primary coolant system and subsequently in reactor trip and safety injection system actuation on either low pressurizer pressure or high containment pressure. The reactor trip and safety injection systems serve to mitigate the consequences of the event in the f ollowing ways:

a. Reactor trip and borated water injection, in addition to void formation as a result of the depressurization, cause a rapid reduction in core power to the fission product decay heat level.

b. Water injected by the safety injection system provides for core cooling and prevents excessive fuel and clad temperatures.

Safety injection system water is supplemented by the injection of borated water from the safety injection bottles. The safety injection tanks passively actuate when the primary coolant system pressure drops below 200 psia (plus the elevation head in the injection lines and bottles). The safety injection tanks affect a rapid refilling of the reactor vessel due to large capacity and, hence, strictly limit the period of time during which the reactor core remains uncovered.

The emergency core cooling system (i.e. , the safety injection system in combination with the safety injection tanks) is designed so that the reactor can be safely shut down and the essential heat transfer geometry of the core 6-ld MARCH 1979 REV. 1

PALISADES PLANT SGRR preserved following the LOCA. More specifically, when the emergency core cooling system (ECCS) is degraded by the most severe active single failure, it is designed to meet the ECCS acceptance criteria as stated in Reference 4.

An evaluation was performed to determine the effects of the replacement steam generators on ECCS performance and is summarized below. The most recent loss-of-coolant accident analysis submitted for the Palisades Plant was used as the reference analysis in evaluating these effects. The mos t recent analysis, as documented in Reference 5, used the currently approved Exxon Nuclear Company WREM-II PWR Evaluation Model.

For this evaluation, sensitivity studies were conducted to determine'the effect of changes in significant primary system operating parameters and steam generator characteristics on peak cladding temperature (PCT) for the mos t limiting large break LOCA as determined by the reference analysis. The me thods used f or these s tudies are identical to those used in the reference analysis. The sensitivity of PCT to s, team generator tube plugging, primary system pressure, and core inlet tempera ture were evaluated. -

The results of this study are summarized below.

Change in Change in Parameter Parameter PCT Plugged tubes + 850 tubes +25F Core inlet + 8.5F -18F temperature Primary system + 90 psi + 54 F pressure Bas'ed on these sensitivity studies and noting that the reference analysis assumed a total of 4,175 plugged tubes, it can be deduced that replacement of the steam generators without changing plant operating conditions should result in a net reduction in PCT with respect to the reference case of about 125F. This improvement is primarily due to reduced steam generator flow resistance during the reflood phase of the transient and, hence, higher core reflooding rates.

However, because of higher expected primary system flowrates and higher expected secondary steam pressures with the new steam generators as compared to the referenced analysis, it is expected that the Palisades Plant will be able to operate 6-le MARCH 1979 REV. 1

PALISADES PLANT SGRR at a slightly higher core inlet temperature (+5F) and a slightly higher primary system pressure (+40 psi) as compared to the core inlet temperature and pressure at which the reference analysis was performed. The increase in core inlet temperature should result in a slight reduction in PCT (-10F), whereas the increase in pressure should result in a slight increase in PCT (+25F). Taking all of these changes into account, the replacement steam generators should result in a net improve me n t in PCT for the limiting large break LOCA of abou t 110F.

Change in Change in Parameter Parameter PCT Plugged tubes -4175 tubes -125F Core inlet +5F -10F temperature Primary system +40 psi +25F Net -110F Hence, it can be concluded that the replacement s team generators will have a beneficial ef fect on ECCS perf ormance and that the ECCS acceptance criteria (1) will be met with the new steam generators ins talled in the Palisades Plant.

6.1.4 CONTAINMENT PRESSURE ANALYSIS The effects of the replacement steam generators upon the containment pressure response analysis have been evaluated by assessment of the mass / energy releases to containment during the main steam line break (MSLB) and the loss-of-coolant accident (LOCA). FSAR Section 14.18 and Answer 14.11 of Amendment 14 (FSAR) indicated that for the original steam generators the MSLB at full load would be more severe from a containment pressure point of view than either the LOCA or the MSLB at no load.

The LOCA mass / energy release for the replacement steam generators at f ull power (2530 Mwt) was compared to the LOCA mass / energy for the original steam generators. It was concluded based on this comparative evaluation that the peak containment pressure following a LOCA would be slightly less than that predicted in the FSAR (51.0 psig).

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PALIFADES PLANT SGRR The full power MSLB mass / energy release for the replacement steam generators was analyzed using analytical methods comparable to those used in the preparation of the FSAR, except that credit was taken for the steam nozzle flow restrictors in the replacement s team generators. To obtain full benefit of the flow restrictors, the analysis included a Main Steam Isolation System (MSIS) actuation on high containment pressure (5.75 psig) as described in Section 3.6. The containment response to a full power MSLB was analyzed using the version of the COPATTA computer program which is described in FSan Section 14.18.1.

Containment initial conditions, engineered safeguard equipment actuation times and containment heat sink data used for this analysis were identical to those presented in FSAR Section 14.18.1. Although the mass / energy data were developed by conservatively assuming the availability of off-site power, the containment response analysis conservatively assumed the loss of of f-site power.

Consequently, the single active failure assumed for the containment response analysis was a diesel-generator failure. This postulated active f ailure minimizes the engineered safeguard equipment available during the accident and maximizes containment pressure.

A peak containment building pressure of 47.6 psig was calculated for the full load MSLB with the replacement steam generators; this value is less than that predicted in the FSAR (51.8 psig). Since the zero power inventory for the replacement steam generators is slightly less than that in the original steam generators, the peak containment pressure for the no-load MSLB would be less for the replacement steam generators than that predicted in the FSAR.

Based on the foregoing, it is concluded that the peak containment pressure following either a MSLB or a LOCA would be no more severe f~or the replacement steam generators than for the original steam generators.

6.1.5 FSAR EVALUATION CONCLUSIONS The conclusions based on the safety evaluation of the design basis events for the replacement steam generators are as follows:

a. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the saf ety analysis report is not increased.

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s ,ss PALIS ADES PLANT SGRR

b. The possibility for an accident or malf unction of a different type than any of those evaluated previously in the safety analysis report is not created.

c. The margin of safety as defined in the basis for any present technical specification is not reduced.

d. The Palisades Plant, equipped with the replacement steam generators, may be safely operated without presenting any undue hazard to the health and safety of the public.

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