ML18058B169

Forwards Proposed Changes to FSAR Sections 4.2 & 4.3 to Comply W/Requirement of License Amend 135.Changes Correct & Clarify Cycle Design Load Descriptions Resulting from Replacement of Steam Generators

ML18058B169
Person / Time
Site:	Palisades
Issue date:	10/20/1992
From:	Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9210280063
Download: ML18058B169 (17)

Text

consumers Power GB Slade General Manager POWERING MICHIGAN'S PROGRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 October 20, 1992 Nuclear Regulatory Commission

-Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - SUBMITTAL OF PROPOSED CHANGES TO THE PALISADES FSAR AS REQUIRED BY LICENSE AMENDMENT 135 The enclosed Final Safety Analysis Report (FSAR) change has been developed to comply with the requirements of Amendment 135 to the Palisades Operating License, dated February 11, 1991, which included a change to Technical Specification 5.3.la, Primary Coolant System. The Safety Evaluation Report (SER) for Amendment 135 included a requirement that changes to Section 4.2 of the FSAR be made through a formal amendment process.

The proposed FSAR change includes the following: Deletion of a design load in FSAR Section 4.2.2, since this was not treated as a necessary design condition in the new steam generators; a change in the feedwater temperature from 70°F to 40°F, since this assumption was changed in the analysis for the replacement steam generators; and editorial changes that are covered by a 10 CFR 50.59 review for Facility Change 909, "Steam Generator Replacement."

These changes are required as a result of the steam generator replacement project at Palisades, which was conducted under the provisions of 10 CFR 50.59. It is requested that this change be made effective upon approval.

In addition to the FSAR change we request NRC response with regard to three conditions contained in the SER for Amendment 135 as described below:

First, on page 3 of the February 11, 1991 SER under Change No. 3, the third paragraph states " ... a sentence shall be added to the end of the first paragraph of Section 4.2." The sentence was to read "Replacement parts and 270027 .Ll~of 9210280063 ~ig5g55 rr~;,

PDR AOOCK POR p A CMS' ENERGY COMPANY

2 components will satisfy the requirements of the original Plant construction code in a manner that is consistent with 10 CFR 50.55a, and the rules and requirements specified in ASME B&PV Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components", Article IWA-7000." This sentence was not added to the end of the first paragraph of Section 4.2 but was instead added to FSAR Section 4.2.4. We believe this is a proper placement of this sentence. Although this sentence was added, we consider it to be unnecessary (and could lead to eventual confusion) to stipulate within any one particular FSAR section that 10 CFR 50.55a must be followed.

Furthermore, compliance with 10 CFR 50.55a is promulgated through a condition of the Palisades License and Technical Specifications 4.0.5. We, therefore, request NRC concurrence to allow removal of this sentence at our convenience.

Second, on page 4 of the SER, in the final paragraph under Change No. 3, it is stated that " ... since the FSAR is referenced in the Technical Specifications, any changes to the referenced FSAR section shall require a formal amendment requires [sic] instead of only a 10 CFR 50.59 review." (underlining added).

We do not believe that "any" (meaning all) changes to FSAR Section 4.2 require NRC review via a formal amendment process. We, therefore, added an explanation in the FSAR Section 4.2 stating, " ... Administrative or editorial changes to Section 4~2, which do not change design parameters, may still be processed without explicit NRC approval." We request NRC concurrence that this explanation fulfills the intent of the SER.

The third condition requiring NRC response is the process of applying for a "formal amendment" to facilitate changes to FSAR Section 4.2. We believe this to be unnecessary. In many recent changes brought about by the Technical Specifications Improvement Program (e.g., the Fire Protection Program and RETS) specific requirements have been removed from the Technical Specifications and references or license conditions have been added to maintain the program requirements. These changes were made in part to relieve the regulatory burden and allow changes to these programs to be made through the 10 CFR 50.59 process instead of through the license amendment process. We believe that the 10 CFR 50.59 process is applicable here in that we removed the specific requirement from the Technical Specifications and inserted in its place a reference to FSAR Section 4.2.

In discussions with the Palisades Project Manager, we understand that the NRC staff position on the intent of the SER to require a "formal amendment" is that changes to PCS design parameters and ASME Code references are the types of changes that the NRC expects to review and approve prior to their incorporation into FSAR Section 4.2. This is not consistent with the fact that this particular FSAR change we are requesting herein is a result of a plant modification which was conducted via the 10 CFR 50.59 process. No unreviewed safety question was identified in the 10 CFR 50.59 review and, as a .

result, no amendment request was made. We concluded in our 10 CFR 50.59 process that, through the code reconciliation process, consistent with 10 CFR 50.55a, the replacement components were made to standards that were no less restrictive than those used in the original construction of the plant. Our

3 conclusion is that if a modification does not identify an amendment as being necessary, then the FSAR change process being invoked via the Amendment 135 SER is not within the present rules and regulations as we understand them.

We, therefore, ask the NRC to modify the SER to be consistent with 10 CFR 50.59 by removing the statement which requires " ... any changes to the referenced FSAR section shall require a formal amendment instead of only a 10 CFR 50.59 review."

Action taken as requested with respect to condition 3 will supersede any action needed on condition 2.

Gerald B Slade General Manager CC Administrator, Region III, USNRC NRC Resident Inspector - Palisades Attachment

CONSUMERS POWER COMPANY To the best of my knowledge, information and belief, the contents of this submittal are truthful and complete.

President Sworn and subscribed to before me this ;?o<<,day of (j~ 1992.

[SEAL]

NOtaryUbli c Michigan My commission expires BEVERLY ANN AVERY NOTARY PUBLIC-JACKSON COUNTY, Ml MY COMMISSION EXPIRES 12-7-92

ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 Proposed Changes to FSAR Sections 4.2.2 and 4.2.3 October 20, 1992 5 Pages

1 CONSUMERS POWER COMPANY Docket 50-255 License DPR-20 I. Changes This proposed FSAR change to Section 4.2 and Section 4.3 corrects and clarifies the cyclic design load descriptions resulting from the replacement of the Palisades steam generators. The changes, which are shown in marked-up pages of FSAR Section 4.2 and Section 4.3, are contained in Attachment 1 and are described below. The numbers used below correspond to the change numbers in the left-hand column of the marked-up pages in Attachment 1. Further justification of the change~

is contained in the no significant hazards section of this submittal.

1. In FSAR Section 4.2.2, Item 3, the design load of 15,000 power change cycles with a ramp load change of 15% of full load per minute, i~ being deleted since the replacement steam generators were not analyzed for this load change rate. No analyzed accident considers this case as an initial condition. Also FSAR Section

~.2.2, Item 2,* is more restrictive than Item 3. *

2. In FSAR.Section 4.2.2, Item 4, a change to 10 cycles of hydrostatic testing adds the reference for assumptions on steam generator differential pressure. After tube degradation occurred on the original steam generators, a limit on differential pressure was incorporated into the Technical Specifications and was subsequently deleted in Amendment 134. If tube wastage occurs, this assumption will have to be verified to determine if ASME Code limits for the ..

tubes becomes more limiting than the limits in {added) reference 37.

3 &4. In FSAR Section 4.2.2, Items 6 and 7, changes to the primary leak testing design load are added to show the assumptions used in the design analysis. This also applies to the normal operating pressure variation design load.

5. In FSAR Section 4.2.2, in the middle of page 4.3-6 following Item 7, the edition of the ASME Code is deleted since the replacement steam generators are not built to Section III, Class A of the 1965 edition. FSAR Section 4.2.4 describes the applicable edition.

6. In FSAR Section 4.2.2, in the middle of page 4.2-2,* a change is made to better define the abnormal transient loads. The description of one cycle of loss of secondary system pressure is clarified. This definition was moved fr.om FSAR Section 4.3.

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7. In FSAR Section 4.3.4, in the fourth paragraph in the middle of page 4.3-6, the word "accident" is deleted since all of the design load cycles listed are not accident conditions.

8. In FSAR Section 4.3.4, on page 4.3-6, Item 1, the first design cycle listed is deleted because it is a duplicate of the condition listed in the middle of page 4.2-2 for loss of secondary pressure. The wording in the paragraph above (" ... In addition to the cyclic transients listed in Subsection 4.2.2 ... ") falsely leads the reader to believe that the steam generator is designed for two of these events when it is not.

9. In FSAR Section 4.3.4, items 4 and 5, (re-numbered as Items 3 and 4) the assumptions used in the steam generator design analysis have been added.

10. In FSAR Section 4.3.4, Items 6 and 7, (re-numbered as Items 5 and 6) the feedwater temperature has been changed from 70°F to 40°F since this assumption was changed in the analysis for the replacement steam generators.

11. In FSAR Section 4.3.4, in the first paragraph at the top of page 4.3-7, the edition of the ASME Code was changed from 1965 to 1977 to be consistent with section 4.2.4.

12. In the Chapter 4 reference section, reference 37 is added to page 4-3 in conjunction with the change described in Item 2 above.

II. Discussion The changes to the FSAR that clarify the assumptions in the analysis of the replacement steam generators are editorial or change ASME Code references. The editorial changes clarify or state the assumptions used in the analysis that verified the design of the replacement steam generators. The design of the replacement steam generators to the 1977 ASME Code was addressed by the NRC in Amendment 135. All of the changes, except for the 153 per minute load changes and changing the feedwater temperature from 70°F to 40°F, are considered editorial and covered by a 10 CFR 50.59 review for Facility Change 909, "Steam Generator Replacement."

3 Analvsis of No Significant Hazards Consideration In summary, Consumers Power Company finds that activities associated with this change request include no significant hazards; and accordingly, a no significant hazards determination in accordance with 10 CFR 50.92(c) is justified. The following summary supports the finding that the proposed change would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; The probability of *an accident previously evaluated in the FSAR will not be increased by deleting the design load change of 15% per minute or decreasing the minimum feedwater temperature from 70°F to 40°F. There is no design requirement that the plant be capable of 15% per minute load changes. No accident ha~ as an initial condition of a 15% per minute load change taking place, and since this FSAR change is the*

result of the replacement steam generators design, no accident

• probabilities are increased. The 40°F feedwater temperature affects the steam generators, but nothing else is affected in the primary coolant system (PCS). The replacement steam generators have been shown by the design analysis report to be able to withstand the same number of cycles of the addition of 40°F water as the old steam generators could 70°F water.

The consequences of an accident previously evaluated in the FSAR are not increased by either of these two changes. Deleting the design load rate of 15% per minute deals with normal plant operation and would not affect the course of a Chapter 14 event since none of the Chapter 14 events involve power level changes with respect to the steam generators. Also, reducing the maximum design load change rate is a conservative change!

Lowering the feedwater temperature could increase the consequences of the main steam line break (MSLB) accident by increasing the likelihood of a return to power event caused by increased core cooling; however, the current FSAR analysis in Section 14.14 used 32°F as the auxiliary feedwater temperature and thus bounds 40°F.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The possibility of a new or different type of accident is not created by these FSAR changes. By deleting the 15% per minute load change rate from the FSAR, the operation of the plant is unaffected because the 5%

per minute limit on load rate change is more limiting. There is no license requirement to be able to change power at 15% per minute except as described in the proposed FSAR deletion. Furthermore, FSAR Section 4.3.7.2 states that the pressurizer heaters cannot be uncovered by the

4 outward surge of water following load increases; a 10% step increase and 15% ramp increase. FSAR Section 1.2.4.9.a states that the nuclear steam supply system (NSSS) is capable of a ramp change from 15% to 100% power at 5% per minute, and at a greater rate over smaller load changes up to a step change of 10%.

Another consideration is that the analysis for the original steam generators was not as detailed or exact as the analysis for the replacement steam generators. The thermal analysis section of the original steam generator design analysis report states for the three power change cases, 5% per minute, 15% per minute and a 10% step changes, that " ... the transient thermal effects of the power changes are small and neglected. The situations of significance are due to cycling between steady state conditions at different power levels."

Thus, the rate of change was not a consideration in the original design analysis. The replacement steam generators analysis calculated the transient temperature changes with respect to time, so the rate of change was considered. Therefore, the replacement steam generator analysis is more accurate, but does not consider a 15% per minute rate change. The original steam generators were not designed for 15% per minute power changes but could withstand power increases from 50% to 100% 15,000 times without considering the rate of power change.

Reducing the analyzed feedwater temperature from 70°F to 40°F does not change the possibility of whether another type of accident or malfunction can occur since the steam generator is analyzed for this.

3. Involve a significant reduction in a margin of safety.

The margin of safety as defined by plant licensing basis is not reduced due to the replacement steam generators not being analyzed for a 15% per minute power ramp because the 15% per minute ramp rate was not a licensing basis of the plant design. The original plant Safety Evaluation Report does not mention the design power ramp rates. The basis for Technical Specification 3.1.2 states that all components are designed to withstand the effects of cyclic loads due to primary coolant system temperature and pressure changes induced by load changes, trips, and start-ups and shutdowns. FSAR Section 4.2.2 is referenced. The change of eliminating the analyzed ability to make 15% per minute power changes does not reduce the margin of safety because:

a. the plant is not operated in a manner wherein 15% per minute power increases are made. Rapid power decreases during emergency conditions are not covered by this analysis since they are not controlled to 15% per minute but should be considered analyzed by the 500 trips or 10% step change analysis and,

b. the original steam generator did not use the ramp rate in the analysis and,

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c. a 15% per minute power change from 50% to 100% power is a fairly benign change for the steam generator with respect to pressure and temperature changes as compared to heatups and cooldowns because the total changes are small!

The only requirements from the NRC with respect to the number and type of loads is contained in Section II of the NRC Standard Review Plan (SRP) 3.9.l which states " ... The section of the applicant's SAR which pertains to transients will be acceptable if the transient conditions selected for equipment fatigue evaluation are based upon a conservative estimate of the magnitude and frequency of the temperature and pressure conditions resulting from those transients." " ..... Transients and resulting loads and load combinations with appropriate specified design and service limits must provide a complete basis for design of the reactor coolant pressure boundary for all conditions and events expected over the service lifetime of the plant."

In the intervening years between design of the original steam generators *I and the replacement steam generators, Combustion Engineering (ABB-CE) decided that a 15% per minute power ramp rate was beyond what was necessary and expected to occur. This position was acceptable to the NRC since ABB-CE letter CPC-90-170, dated October 24, 1990 states that the replacement steam generators are identical in design to the Palo Verde (Arizona Public Service) steam generators. (The ABB-CE letter was concerned with the stress analysis for steam line breaks, therefore, the reference to being identical was with respect to that stress analysis.)

The change in feedwater temperature from 70°F to 40°F maintains the margin of safety because the replacement steam generators have been shown by the design analysis report to be able to withstand the same number of cycles of the addition of 40°F water as the old steam generators could 70~F water.

~:

...

DESIGN BASIS 4.2.1 PERFORMANCE OBJECTIVES AND PARAMETERS FOR NORMAL CONDITIONS The Primary Coolant System is designed to operate at a power level of 2,650 MWt. The present licensing limit is, however, 2,530 MWt core power plus 15 MWt for the primary coolant pump heat input for a total Primary Coolant System output of 2,545 MWt. The principal parameters for the Primary Coolant System are listed in Table 4-1. The design parameters for each of the major components are given under the individual component discussion later in this section. The Primary Coolant System is a CP Co Design Class 1 system per Section 5.2. The applicable stress and seismic criteria are given in Section *5.10. The primary system components and controls are also designed for cyclic transient conditions as listed in Subsection 4.2.2.

Amendment 135 to the Technical Specifications (Reference 32) added a reference to FSAR Section 4.2 in lieu of specifying PCS design parameters in the Technical Specifications. Accordingly, proposed changes to fundamental design parameters as specified in the following subsections will require NRC approval prior to implementation. It is the intent of this change to assure that any replacement parts and components will be held to standards no less restrictive than those used in the original construction of the plant. Administrative or editorial changes to Section 4.2 which do not change design parameters may still be processed without explicit NRC approval.

4.2.2 DESIGN CYCLIC LOADS The following design cyclic transients which include conservative estimates of the operational requirements for the components listed in Table 4-2 were used in the fatigue analysis required by the applicable code:

1. 500 h~atup and cooldown cycles during the system 40~year design life at a heating and cooling rate of l00°F/h. The pressurizer is designed for a cooldown rate of 200°F/h. *

2. 15,000 power change cycles over the range of 10% to 100% of full load with a ramp load change of 5% of full load per minute increasing or decreasing. * *

}Eiaa 15,000 cyeles of 10% of full load step power changes fncreasing from 10%*

to 9~ of full power and decreasing from 100% to 20% of full power.

10 cycles of hydrostatic testing the primary system at 3,110 psig and at a temperature at 1east 60° F above the Nil Duct i 1i ty Transition Temperature (NOTT) of the component having the hi3hest NOTT. ]ryij 4.2-1 Rev tt

~- ~20 cyc1e~ of leak testing at 2,485 psig and at a tempera~ure at least 60.F greater than the NOTT of the component havini the highest NOTT.

• -~~

11 '1 +§. PR

81. 500 reactor trips from 100% power.

In addition to the above list of normal design transients, the following abnormal transients were al so considered when arriving at a sa~.t.s..f.~.C:J()J'.'Y. ll~~ge lf .S iiii:!al:::1.~i.:l~:iiiltift;jhe ASHE Boiler and Pressure Vessel Code.:i::::::§!!l::}:!i,:)~J:lf.

1. 200 cycles of loss of turbine load from 100% power

2. 200 cycles of total loss of reactor coolant flow when at 100% power rlc; 3. liiiiiilli~

4.2.3 DESIGN SERVICE LIFE CONSIDERATIONS The major Primary Coolant System components are designed considering a 40-year service life. In order to achieve this, the strict quality control assurance standards as outlined in Subsections 4.5.4 and 4.5.5 were followed.

Component design has also considered environmental protection, adherence to established operating procedures and irradiation effects on the material.

The reactor vessel is the only component of the Primary Coolant System which is exposed to a significant level of neutron irradiation. The irradiation surveillance program is outlined in Subsection 4.5.3. To compensate for any increase in the NOTT shift caused by irradiation, the Plant operating procedures for the pressure-temperature relationship during heatup and cooldown will be periodically revised to stay within the stress limits.

The design of"the Primary Coolant System components allows for adequate inspection techniques to be applied over the lifetime of the Plant. All reactor internals are designed to be removable for inspection and to allow reactor vessel internal inspection. Insulation panels are removable for external inspection of selected highly stressed areas.

• 4.2.4 CODES ADHERED TO AND COMPONENT CLASSIFICATION The original design, fabrication, construction, inspection, testing and classification of all reactor coolant system components are in accordance with the ASHE Boiler and Pressure Vessel Code, Section Ill, 1965 edition, including 4.2-2

the -code for Pressure Piping, ASA 831.1, 1955. The replacement steam generators installed during 1990 meet ASME Code Section III 1977 edition.

The codes adhered to and component classifications are listed in Table 4-2.

Replacement parts and components will satisfy the requirements of the original plant construction code in a manner that is consistent with 10CFR50.55a, and the rules and requirements specified in ASME B&PV Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components", Article IWA-7000."

(Reference 32) 4.2.5 SAFETY CONSIDERATIONS OF DESIGN PARAMETERS Desian Pressure After establishing the normal operating pressure conditions of the Primary Coolant System, a minimum design pressure was determined which exceeds the normal operating pressure and anticipated operating transient pressure changes.

Major considerations employed i~ the determination of this selected minimum design pressure inc~ude: normal operating pressure, instrumentation and control response, reactor core thermal lag, coolant trarisport time, system pressure drop, and safety and relief valve characteristics. The design pressures for the individual reactor coolant system components are listed in their respective component description sections.

oesign Temperatyre The design temperature was selected to exceed the normal operating temperature and anticipated operating transient temperature changes for each primary coolant component. The design tempe~atures for the ptimary system components are listed in their respective component description sections. *

'Design Loads The Primary Coolant System was designed to the criteria for load combination and stresses 'as defined for a CP Co Design Class 1 system in Section 5.10.

These criteria assure the integrity of the Primary Coolant System to withstand the load imposed by the design basis accident simultaneously with the load imposed by the maximum seismic disturbance without loss of safety function.

4.2.6. PRIMARY CQQLANT SYSTEM ASYMMETRIC LOADS Pursuant to industry and NRC concerns for the potential effects of asymmetric loads on the Primary Coolant System components and supports, Consumers Power (in 1978) contracted with Combusti9n Engineering for a study to evaluate these concerns. A generic plant evaluation (see References 33 and 34 ) was

• completed by Combustion Engineering for Calvert Cliffs 1 and 2, Palisades, Millstone 2 and Fort Calhoun.

4.2-3 Rev--~

A further evaluation (see reference 35) was performed by Combustion

• - Engineering to show that a flaw in the Primary Coolant System will result in a detectable leak before a large guillotine break would occur. The analysis was reviewed by the NRC in an SER dated October 27, 1989 (Reference 36). The SER concluded.that, with the exception of concerns regarding seismic grid design, Palisades reactor system would withstand the effects of asynmetric LOCA loads

• and that the reactor could be brought to a cold shutdown condition safely.

4.2-4 Rev~

. --1

• ,

e same nominal opening pressure, but with staggered group opening pressures consistent with ASHE B&PV Code allowances. The valves can pass a steam flow equivalent to an NSSS power level of 2,650 MWt at the nominal 1,000 psia set pressure. Parameters for the secondary safety valves are given in Table 4-5.

Listed in Table 4.5 is a Set Point Tolerance (as-found testing) of +/-3% of set pressure for the Hain Steam Relief Valves. The basis for this value is documented in Technical Specification 3.1.7, "Primary and Secondary Safety Valves* and Amendment No 116 to Provisional Operating License No DPR-20, Secondary Safety Valves (TAC No 69225). In sunnnary, these documents allow a

+/-31 as-found set point tolerance for the Main Steam Safety Valves without the requirement for increasing testing scope per ASHE Code Subsection IWV.

However, all valves which are tested and found to be outside of +/-1% of set pressure shall be restored to within the 1% criteria ai required by Amendment No 116.

The steam generator shell is constructed of carbon steel. Manways and handholes are provided for access to the steam generator internals. The steam generators are mounted vertically on bearing plates to allow horizontal motion parallel to the hot leg due to thermal expansion of the primary coolant piping. Stops are provided to limit this motion in case of a coolant pipe rupture. The top of the unit is testrained from sudden lateral movement by energy absorbers mounted rigidly to the concrete shield.

In addition to the cyclic transients listed i.n....~.lJ.P~l:!~tion 4.2.2, each steam generator is also designed for the following @@:gl,:l!i:I conditions such that no component wil 1 fail either by rupture or by deiieloplng deformations {el ast; c or plastic) that will impair the function, performance or integrity of the steam gene~ator for further operatidn.

at.

• • .:

One cycle during which the steam on the shell side is at 900 psia and 532*F while tube (primary) side is depressuri zed to atmospheric pressure.

~* 2,400 cycles of transient pressure differentials of 85 psi across the primary head divider plate due to starting and stopping the primary coolant pumps.

I 4 .

• des of h ~~~~~;,~~ry I

~Ila) 15,000 cycles of adding 425 gpm of 79~0*F feedwater with the Plant in hot standby conditions. .......

8 cycles of adding a maximum of 300 gpm of 79~~*F feedwater with the steam generator secondary side dry and at soo*F.

4.3-6 Rev W'

• The unit is capabl~ of withstanding these conditions for the prescribed number of cycles in addition to the prescribed operating conditions without exceeding 11 the allowable cumulative g~:il~::I********f.:~5:~8:!: as prescribed in ASME B&PV Code, Section I/ II I? Glass A, 19&5, W&Sa l:~i~:t!Sl:!:!:PJJ*

4.3.4.l Steam Generator Tube Degradation The Palisades Plant experienced its first steam generator tube leak in early 1973.

• Eddy current examinations of the tubing detected general wastage attack in the U-bend area of tubes in the first eleven rows from the divider plate.

The attack was attributed to the use of a coordinated phosphate secondary water chemistry treatment for pH control. All tubes in these first eleven rows were plugged.

In 1974, further leakage led to discovery of increased tube wastage and evidence of intergranular attack. A flushing program was performed and subsequent plant chemical control was changed to all-volatile treatment.

Subsequent eddy current examinations in 1975 through 1981 showed that corrosion of the steam generator tubing had essentially ceased although minor tube denting was occurring as a result of the switchover to all-volatile treatment.

Continued examination over the next eight years revealed further IGA and other growing problems related to denting at the tube support plates. With excessive outage times and plant operation nearing the point of power limitation due to plugged tubes, the Steam Generator Replacement Project was initiated in mid 1989.

4.3.4.2 Steam Generator Replacement Due to the tube degradation problems noted above, replacement of both steam generators was undertaken in late 1990. The replacement steam generators are designed to physically match the essential parameters of the existing steam generators and to be compatible with the performance characteristics utilized in the FSAR and the license for operation at 2530 MWt. Consistent with other PCS equipment, the replacement steam generators are designed for operation at 2650 HWt should *an increase in the licensed power level be pursued in the future.

Some component design changes were made to improve the replacement units:

1. Tube wall thickness was reduced slightly (.042 vs .048) to improve heat transfer. Combined with 308 less tubes, the overall steam generator heat transfer effect is unchanged. *

2. Tube support design was changed from solid plate to eggcrate dividers and other features to minimize corrosion crevices and denting.

3. Slowdown capability was improved through an internal center duct and increase in blowdown nozzle size. Sampling improvements were also made.

4. Hand holes and inspection ports were added for future internals inspection. Manway sizes were also enlarged.

4.3-7 Rev 11'

.. .---.

34. - Combustion Engineering Report, "Response to Questions on the Reactor Coolant System Asy!Mletric Loads Evaluation Program Final Report,"

Submitted to the NRC on July 31, 1981.

35. Combustion. Engineering Owner's Group, "Leak-Before-Break Evaluation of Primary Coolant Loop Piping in Combustion Engineering Designed Nuclear Steam Supply Systems," CEN-367, November 1987.

36. DeAgazio, Albert, USNRC, "Safety Evaluation on Asy1M1etric LOCA Loads -

MPA D-010 - Palisades Plant (Tac No M08621}" to KW Berry, October 27, 1989.

PkJi¥?t!

4-3 Rev "H--.