ML20236K326

SER Accepting Licensee Response Related to Revised Feedwater Nozzle Analysis to License NPF-43 for Enrico Fermi Nuclear Power Plant,Unit 2

ML20236K326
Person / Time
Site:	Fermi
Issue date:	07/01/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20236K316	List: ML20236K313 ML20236K326
References
GL-81-11, NUDOCS 9807090277
Download: ML20236K326 (5)

Text

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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO REVISED FEEDWATER NOZZLE ANALYSIS 1

TO FACILIW OPERATING LICENSE NOS. NPF-43

' DETROIT EDISON COMPANY ENRICO FERMI NUCLEAR POWER PLANT. UNIT 2 DOCKET NO. 50-341

1.0 INTRODUCTION

By letter dated December 1,1997, as supplemented by letter dated June 24,1998 the licensee submitted confirmation of a revised feedwater nonle crack growth analysis for Fermi Unit 2 (Fermi 2). The information was in response to an NRC letter dated December 8,1992 which requested that the licensee confirm the revised analysis with new operating data on thermal cycles. The NRC also requested that the licensee submit the analysis for review six months prior to the end of 12 operating years.

The original analysis submitted by [[letter::NRC-89-0187, Submits Justification for RHR HX Tube Leakage Test Commitment Clarification.Commitment to Perform Periodic Leakage Test on RHR HX Tubes Necessary|letter dated November 2,1989]] demonstrated that the postulated crack would grow to 1.0 inch in 8.9 operating years. The licensee used estimated or assumed thermal cycles since, at the time, actual data were unavailable. The crack growth analysis did not satisfy the Generic Letter (GL) 81-11 criterion of limiting growth of a postulated 0.25 inch deep crack to a depth no greater than 1.0 inch in 40 years.

The revised analysis submitted by Vs Generic Ltr|letter dated July 29,1992]] demonstrated that the postulated cLck would grow to a 1.0 inch depth in 38.3 years. Since the analysis results were close, but did not satisfy the criterion in GL 81-11, the staff determined that the overall methodology was acceptable; but confirmation of the analysis is required. The licensee also committed to follow the feedwater nonle inspection schedule and examination specified in NUREG-0619,"BWR Feedwater Nonle and Control Rod Drive Retum Line Nonle Cracking

• In order to monitor the structuralintegrity of the feedwater nonles in the interim.

Feedwater is distributed through spargers that deliver the flow evenly to assure proper jet pump subcooling and help maintain proper core power distribution. The thermal sleeve, which projects into the nonle bore, is intended to prevent the impingement of cold feedwater on the hot nonle surface. The incoming feedwater is colder than the reactor vessel during normal operation. The feedwater is much colder during startup and shutdown when feedwater heaters are not in service. Turbulent mixing of the hot water retuming from the steam separators and dryers and the incoming cold feedwater causes thermal stress on the noule bore if it is not protected by a thermal sleeve.

ATTACHMENT 1 Wa7 EE W62 P

in the late 1970's, inspections at BWR plants disclosed cracks in feedwater nonles for those I

plants that have a loose-fit sparger/ thermal sleeve design. The loose-fit design allows leakage

' past the area where the thermal sleeve and the nonle safe-end meet. This bypass leakage is the primary source of cold water impingement on the nonle bore. Bypass leakage past a loose thermal sleeve causes fluctuations in the metal temperature of the feedwatar nonle and can result in metal fatigue and crack initiation. The flow of cold feedwater into the vessel during startup, shutdown, and hot standby conditions can induce crack growth if feedwater additions are not modulated smoothly. '

General Electric (GE) performed an extensive feedwater nonle/sparger testing and analysis j

program, and the results of this program were reported to the staff in several documents. The j final document, which incorporates the information from all earlier submittals, is topical report l NEDE-21821-A, "BWR Feedwater Nonle/Sparger Final Report, February 1980." l l In November 1980, the NRC issued NUREG-0619, "BWR Feedwater Nonle and Control Rod l- Drive Retum Line Nonle Cracking," recommending that BWR owners,1) remove feedwater

nonle cladding, 2) install modified sparger/thermai sleeves, 3) change operating procedures, l

4) modify the feedwater' control system with a low-flow controller, and 5) follow the NRC's  !

inspection program.

l l Comments received from GE and BWR owners after the publication of NUREG-0619 noted the difficulties in meeting the requirements for a low-flow controller having the six characteristics i described in GE report NEDE-21821-A. The comments also stated that an existing controller l

t may not meet the dix characteristics, but the system may still meet the criterion of the crack '

growth analysis from which the characteristics were derived. The staff recommended the use l of the low-flow controller, as opposed to an on-off flow control system, in order to modulate l feedwater additions.

On February 20,1981, the NRC issued GL 81-11 to amend NUREG-0619 and to allow for a plant specific fracture mechanics analysis in lieu of replacing the existing controller. The analysis must show that the growth of a postulated 0.25 inch crack does not exceed 1.0 inch in

40 years. GL 81 11 stated that the analysis should be submitted as part of the reporting

. requirements specified in NUREG-0619.

2.0 EVALUATION I

in accordance with the proposed solutions in NUREG-0619, Fermi 2 uses a triple-sleeve sparger design which provides an acceptable improvement over previous designs, in addition, the Fermi 2 vessel was manufactured with unciad feedwater nonles. The presence of stainless steel cladding on nonle surfaces contributes to fatigue cracking because thermal stresses from the cycling are higher in the stainless steel than they would be in the unciac' base metal. In addition, the thermal expansion coefficients of the base metal and the clad are different. Fermi 2 uses a plant-specific low-flow controller which is different from the one recommended in the GE analysis described in GE report NEDE-21821-A. The licensee opted to perform a plant-specific fracture mechanics analysis in lieu of replacing the existing controller. This option is recommended in GL 81-11.

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3 GE performed botn the original and the revised feedwater nozzle analyses which consisted of thermal cycle definition, plant operating history, finite element analysis, and crack growth analysis. Thermal cycles were assumed in the original analysis since actual data were 1 unavailable. The revised analysis used thermal cycles that were based on plant operating data with actual cycle counts from 1986-1990. By letter dated December 8,1992, the staff determined that the fracture mechanics analysis method was acceptable. The current confirmation of the revised analysis adds thermal cycles that are based on operating data from 1991-1996, excluding 1994 since very little plant operation occurred in 1994.

Thermal cycles for feedwater nozzles can occur as a result of several different normal and l upset events. GE assessed these events as either startup, shutdown, or SCRAM to low and high pressure hot standby followed by a retum to full power. The startup, shutdown, and SCRAM cycles for 1991-1996 (excluding 1994) were 17,17, and 16, respectively. The licensee also counted power reductions to less than 50% as SCRAMS which resulted in 10 additional j SCRAM cycles. Adding the data from 1986-1990 resulted in 46 startups and shutdowns, and l 78 SCRAMS for the confirmation analysis. These cycles were projected to 40 vaars for a total of 496 thermal cycles of startups, shutdowns, and SCRAMS. The licensee s% n that the confirmation of the revised analysis is conservative because the number of theni.al cycles for the first 10 years is assumed to repeat four times for the projection to 40 years of operation.

In the revised analysis, GE used the fatigue crack growth rate for low alloy steel from the 1989 Edition of Appendix A to Section XI of the American Society of Mechanical Engineers (ASME)

Code to calculate crack growth. For each thermal cycle, the maximum and minimum stress intensity factor and the number of occurrences were calculated. The stress intensity factor range and the corresponding R-ratio' were calculated for each cycle. Using the calculated information described above and the crack growth data in the ASME Code, the incremental crack growth was calculated for each cycle. This process was repeated for all cycles until all events had been analyzed.

The confirmation of the revised analysis result shows that a postulated 0.25 inch crack is estimated to grow to approximately 0.8 inches depth in 40 years. This result satisfies the crack i growth criterion in GL 81-11.

The staff confirmed inat the methodology used in the revised crack growth analysis remains valid when considering the new thermal cycle count data from 1991-1996 (excluding 1994). In addition, the staff determined that the confirmation of the revised analysis is acceptable, and satisfies the criterion in GL 81-11.

1 The R ratio (KA) is defined as the algebraic ratio of two specified stress intensities in a stress cycle, used for prediction of fatigue crack growth.

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3.0 QQfiCLUSION The staff has determined that the confirmation of the revised feedwater nonle crack growth analysis for Fermi 2 is acceptable, and satisfies the Generic Letter (GL) 81-11 criterion of limiting growth of a postulated 0.25 inch deep crack to a depth no greater than 1.0 inch in 40 years. 1

4.0 REFERENCES

1. Generic Letter 81-11, To All Power Reactor Licensgs and License Applicants, (No Title), February 20,1981.

2. NUREG-0619, "BWR Feedwater Nonle and Control Rod Drive Retum Line Nonle Cracking" November 1980.

3. December 1,1997, Letter from D. R. Gipson to U.S. Nuclear Regulatory Commission  !

Document Control Desk,

Subject:

Confirmation of Feedwater Nonle Crack Growth Analysis.

. 4. December 8,1992, Letter from T. G. Colburn (USNRC) to W. S. Orser,

Subject:

Ferml-2 Revised Feedwater Nonle Crack Growth Analysis Per Generic Letter 81-11 (TAC # ,

M84140).

5. July 29,1992, Letter from W. S. Orser to U.S. Nuclear Regulatory Commission Document Control Desk,

Subject:

Revised Feedwater Nonle Crack Growth Analysis.

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I SALPINPUT j FACILITY NAME: Fermi Unit 2

)

SUMMARY

OF REVIEW ACTIVITIES The staff reviewed the licensee's confirrnation of a revised feedwater nozzle crack growth analysis. The infom1ation was in response to an NRC letter dated December 8,1992 which requested that the licensee confirm the revised analysis when new operating data on thermal l cycles become available. The result of the analysis was compared to the Generic Letter

, (GL) 81-11 criterion of limiting growth of a postulated 0.25 inch deep crack to a depth no

! greater than 1,0 inch in 40 years. The staff verified the licensee's methodology and the new l

thermal cycle count data from 1991-1996. Data from 1994 were excluded since very little plant ;

operation occurred in 1994, L fBBRATIVE DISCUSSION OF LICENSEE PERFORMANCE-FUNCTIONAL AREA i ENGINEERING / TECHNICAL SUPPORT The licensee's original submittal did not contain the actual thermal cycle count data for 1991-1996, which constituted the basis for confirmation of the previous revised analysis.

This information was provided by letter dated June 24,1998. The licensee did not submit the complete set of data that the staff needed to confirm the revised feedwater nozzle L crack growth analysis in a timely manner, The staff reviewer had to make repeated inquires in order to obtain the docketed information needed to complete the safety evaluation report.

Author: A.D. Lee l

(301)415-2735 ATTACHMENT 2 s

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