ML20069B476

Non-proprietary Rev 0 to BAW-10196, W-D4 F* Qualification Rept

ML20069B476
Person / Time
Site:	Byron, Braidwood
Issue date:	03/31/1994
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML19304C174	List: ML20069B441 ML20069B471 ML20069B476
References
BAW-10196, BAW-10196-R, BAW-10196-R00, NUDOCS 9405270243
Download: ML20069B476 (37)

Text

_ - _ _ _ _ _ _ _ _ _ ..

'fr BAW-10196 REVISION 00 -

MARCH 1994

-e W-D4 F* QUALIFICATION REPORT BWNT NON-PROPRIETARY B&W NUCLEAR TECHNOLOGIES COPY _

g -

K J

I

-F

'l This document is the non-proprietary version of ' the proprietary!

document BAW-10196P-01. In order-for this document to meet the non-proprietary . criteria, certain blocks of information wereLwithheld.

The basis for determining what information to withhold wasLbased on the two criteria listed below. Depending upon the: applicable criteria,,

the criteria code, (c)' or . (d) , represents the withheld information. -

(c) -.The;use of.the information by'a. competitor would-decrease-his expenditures', in time or resources,'in-designing,-

producing cn marketing a similar. product.

(d) - The information consists of test data or other similaridatat concerning a process, method or' component,;the1 application of which results in a competitive advantage to BWNT..

1

1 i;q .

BAW-10196 REVISION 00.

-MARCH 1994 W-D4 F* QUALIFICATION REPORT

.1

]

1 B&W~ NUCLEAR TECHNOLOGIES:

.PO BOX 10935 :

LYNCHBURG, VA 24506-0935

CONTENTS PAGE

1.0 INTRODUCTION

. . . . . . . . . . . . . . . . . 1-1 1.1 Background . . . . . . . . . . . . . . . . . 1-1 1.1 Scope of Report . . . . . . . . . . . . . . . 1-2 2.0

SUMMARY

. . . . . . . . . . . . . . . . . 2-1 3.0 DESIGN REQUIREMENTS . . . . . . . . . . . . . . . 3-1 3.1 General Requirements . . . . . . . . . . . . 3-1 3.2 Functional Requirements . . . . . . . . . . . 3-1 3.3 Design and Operational Loading Conditions . . 3-1 3.4 Corrosion . . . . . . . . . . . . . . . . . 3-2 4.0 F* CRITERIA DEVELOPMENT . . . . . . . . . . . . . 4-1 4.1 Structural Justification . . . . . . . . . . 4-1 4.2 Establishing F* Criteria . . . . . . . . . . 4-1 5.0 QUALIFICATION ANALYSES AND TESTS . . . . . . . . . 5-1 5.1 Analyses . . . . . . . . . . . . . . . . . 5-2 5.1.1 Radial Stresses . . . . . .. . . 5-3 5.1.2 Axial Loading . . . . . . . . . . . 5-4 5.1.3 Locked Tube Loading . . . . . . . . 5-4 5.1.4 F* Determination and Correction . . 5-5 5.2 Mechanical Testing . . . . . .. . . . . . . 5-7 ,

5.2.1 Specimen Description . . . . . . . 5-7 5.2.2 Leak Tests . . . . . . . . . . . . 5-8 5.2.3 Tensile Tests . . . . . . . . . . . 5-10 5.3 NDE Measurement Testing . . . . . . . . . . . 5-12 5.3.1 Test Equipment . . . . . . . . . . 5-12 5.3.2 F* Length Verification Methodology 5-13 5.3.3 Results . . . . . . . . . . . . . . 5-14 5.4 Determining Final F* Criteria . . . . . . . . 5-15 5.5 Boric Acid Corrosion Within the Tubesheet . . 5-16

6.0 CONCLUSION

S . . . . . . . . . . . . . . . . . 6-1 B&W NUCLEAR TECHNOLOGIES

.l.

7.0 REFERENCES

. . . . . . . . . . . . . . . . . 7-1 8.0 APPENDICES . . . . . . . . . . . . . . . . . 8-1 Appendix A NO SIGNIFICANT HAZARDS REVIEW . . . 8-1 FIGURES 1.1 W-D4 TUBESHEET ROLL EXPANSION PROFILE . . . . 1-3 3.3.1 W-D4 RSG GENERAL ARRANGEMENT . . . . . . . . 3-4 5.2.1 MOCKUP BLOCK LAYOUT . . . . . . . . . . . . . 5-19 4

5.3.2.1 SAMPLE BOBBIN PLOT SHOWING SELECTION POINTS . 5-23 5.3.2.2 SAMPLE MRPC PLOT SHOWING SELECTION POINTS . . 5-26 TABLES 3.3.1 W-D4 DESIGN AND OPERATING CHARACTERISTICS . . 3-3 5.1.1 RADIAL STRESS AND AXIAL LOADING

SUMMARY

. . . 5-18 5.2.1 QUALIFICATION SPECIMEN INSTALLATION

SUMMARY

. 5-20 5.2.2 LEAK TEST RESULTS . . . . . . . . . . . . . . 5-21 5.2.3 TENSILE TEST RESULTS . . . . . . . . . . . . 5-22 5.3.3 ECT MEASUREMENT ACCURACY COMPARISONS . . . . 5-14 a

L I

r B&W NUCLEAR TECHNOLOGIES

1.0 INTRODUCTION

1.1 Backctround Primary Water Stress Corrosion Cracking (PWSCC) has been found s during routine inspections of the tubesheet roll transitions in u-tube steam generators (7.9]. Without an alternate plugging criteria, such as F*, it is necessary to remove the tube from service when the indication exceeds 40% of the tube wall thickness. However, in some instances these defects occur in areas significantly below the tube expansion transition at the secondary face of the tubesheet. Since the tubesheet provides structural support for the tube in this area, plugging these tubes is overly conservative. The F* plugging criteria discussed in this report was established as a *means to justify leaving tubes in service which have PWSCC type indications within the rolled region of the tubesheet.

Westinghouse D4 (W-D4) series recirculating steam generators (RSGs) were constructed with 0.750" OD x 0.042" wall mill annealed (MA) alloy 600 tubing. During installation, the tubing was roll expanded into the tubesheet with a tack roll and then seal welded at the primary face of the tubesheet. Step rolls were then performed to close the crevice between the tube and the tubesheet to minimize the possibility of secondary side crevice corrosion.

Occasionally, roll expanders were stepped in such a manner that skip roll areas were created (Figure 1.1).

These skip roll areas and roll transitions contain high residual tensile stresses which accelerate the initiation of PWSCC. If this PWSCC occurs within the tubesheet region, then there is a length of tubing roll expanded into the tubesheet above the defect p location. This rolled length of tubing above the defect provides -

structural support for the tube and limits primary to secondary leakage, and is thus the basis for the F* criteria. Thus the F*

criteria is the minimum length of undegraded expanded tube within B&W NUCLEAR TECIINOLOGIES 1-1

the tubesheet, below which, a tube defect can exist and remain.in service. Thia F* length must be shown to: -

o Exhibit a joint strength sufficient to carry normal operating l and faulted loads with an acceptable margin.of safety.

(

l-I o Demonstrate a leak rate at the normal operating primary-to-secondary differential pressure which is acceptable for plant I- operation and within technical specification limits.

The final F* criteria must be verified using standard steam generator eddy current inspection techniques (ECT). Thus any.

errors which are inherent with remote ECT measurements must also be. factored into the final F* values.

1.2 Scone of ReDort l

This document summarizes the qualification of an alternate plugging criteria, F*, for application in W-D4 series RSGs at Byron Unit 1 and Braidwood Unit 1. This report contains summaries of the design requirements, design verification-testing results, j analysis results, ECT verification testing, and tubesheet. d d

corrosion evaluation performed to justify the use of F*. In 1 1

addition, a "No Significant Hazards Review" per 10CFR50.92 (c) is I included as an appendix.

1

)

1 B&W NUCLEAR TECIINOLOGIES 1-2 l

_ _ _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _- _ _ __ _ _ x

BWNT Proprietary FIGURE 1.1 W-D4 TUDESHEET ROLL EXPANSION PROFILE SKIP ROLL TUBE

[UNEXPANDED REGION)

HARD ROLL CONTACT POINT (HRCP]

TUBESHEET SECONDARY FACE ij e N!~p .

') <~ /'

ROLL

.g p _- TRANSITION

REGIONS t

j .

i i

f TYPICAL 3 STEP ROLLS j

4 l

l 3

i

! TACK

- ROLL

(.

SEAL / ESHEET PRIMARY FACE l WELD I

B&W NUCLEAR TECIINOLOGIES 1-3 L_ _ _ _ _ _ - - _ - _ - - _ - - _ - -

2.0

SUMMARY

The F* alternate plugging criteria has been qualified for use in the W-D4 series steam generators at Byron Unit 1 and Braidwood Unit 1. The use of the F* criteria will allow tubes with otherwise pluggable ECT indications to remain in service as long as the indications are a minimum distance below an undegraded expanded region within the tubesheet. This minimum length,-

referred to as F* distance, was determined to be [c] inches through a combination of analysis, mechanical testing, and evaluation of ECT measurement accuracy.

An initial analysis was performed to determine the normal operating and faulted loads imposed on the tubes for Byron Unit 1 and Braidwood Unit 1. The NRC Regulatory Guide 1.121 safety factors of 3 for normal operation and 1.43 for faulted conditions were also used in developing the loads (7.1]. In addition, the effects of tubes locked into the tube support plates were considered. Conservative loads were used for the final qualification testing.

e The joint strength and leakage of various lengths of the existing tube-to-tubesheet roll expansions were then tested under these conditions. Leak testing, load testing, pressure cycling, and ultimate pull testing were performed on a variety of samples to simulate the actual installed rolled joint and loading conditions within the W-D4 RSGs.

Additional analyses were performed to calculate the effects that operating and f aulted pressure, thermal effects, and tubesheet bow have on the tube OD radial stress, and thus their effect on the rolled joint's strength. The F* value qualified by testing was verified by analysis to be adequate for all of these various conditions.

B&W NUCLEAR TECHNOLOGIES 2-1

y 1 i.:

Eddy current testing was performed on a number of F* specimens to -

.. determine measurement accuracy and repeatability. Both bobbin and MRPC were used in this testing. Based ' on the - - ECT tect results, an additional length of (d] inches was added to the o ,

tested F* length to account for ECT uncertainty.

")

The effects of boric acid corrosion on the carbon steel'tubesheet~

were examined as part of the qualification program. In the event-that the defect in the tube went 100%. through wall, 'a small. region-of the tubesheet could be exposed to ' primary side fluid. 'At' _.

worst, small amounts of localized tubesheet' degradation, on the 1 '

E order of a few mils, could occur. Such shallow attack represents no structural concerns for the tubesheet or the F* joint.

The qualified F* distance applies to.all tube ends within the

(. steam generator.- In addition, the use of F* to maintain tubes ~in service does not represent an unanalyzed safety concern.

Furthermore, its use does not increase the risk of an unanalyzed accident nor does it reduce the margin of safety.

B&W NUCLEAR TECHNOLOGIES 2-2 2

3.O DESIGN REQUIREMENTS 3.1 General Reauirements The ASME Boiler and Pressure Vessel Code and US NRC Regulatory Guide 1.121 were used to establish the safety factors for evaluating the roll expanded tube-to-tubesheet interface associated with F* [7.1,7.2]. The safety factors correspond to 3 for normal operating conditions and 1.43 for faulted loading conditions. The applicable design conditions used for F* criteria evaluation are given in Reference 7.3 and are summarized in this section.

3.2 Functional Reauirements The F* design criteria, which is based on the original tube roll, shall provide a mechanical leak limiting seal between the tube and tubesheet above the degraded location. It shall be assumed that the tube severs circumferentially for 360* and that the remaining joint carries all anticipated loading conditions, including the ,

margins of safety described above. In addition, primary to secondary leakage cannot exceed the station Technical Specification limits.

3.3 D_esian and Operational Loadina Conditions i

The design and operating conditions for the steam generator used to evaluate the F* plugging criteria are summarized in Table 3.3.1. Table 3.3.1 reflects combinations of the worst case conditions conservatively selected from both Byron Unit 1 and Braidwood Unit 1 Tw and Tw % operating design data.

Figure 3.3.1 illustrates the key steam generator geometry and material constraints for evaluating F*.

A significant requirement added to the F* design criteria is the assumption that the tube is not free to move through- the first B&W NUCLEAR TECHNOLOGIES 3-1 l

I 4

tube support plate (TSP). This " locked tube" condition imparts.

axial loads on the tube, resulting in a conservative design. The loading imparted by the locked tube condition is displacement limited, such that as the rolled tube joint slips, the applied load is reduced. [

(d) ] The locked tube loading condition is discussed further in Section 5.1.3.

3.4 Corrosion The W-D4 tubesheet is made of SA-508 Class 2A carbon steel clad with Inconel. In the steam generator design, the tubesheet is isolated from the primary coolant by the cladding, the alloy 600 tubing and the tube-to-tubesheet weld at the primary face of the tubesheet. Any breach of these boundaries, such as through PWSCC cracks in the tubing, may initiate corrosion of the tubesheet.

Therefore, the effects of boric acid corrosion from primary system fluid in contact with the carbon steel tubesheet through F* type cracks shall be considered.

s B&W NUCLEAR TECHNOLOGIES 3-2

4 TABLE 3.3.1 W-D4 DESIGN AND OPERATING CHARACTERISTICS

[

(c) l 3

l B&W NUCLEAR TECHNOLOGIES 3-3

.s Figure 3.3.1 W-D4 RSG GENERAL ARRANGEMENT Y'

a f-i r-TOP SUPPORT PLATE TIE ROD SPACER LE 2nd TSP 9

g ,

g.750 x .042 WALL TUBE ALLOY 600 - 4578 TUBES g ist TSP FLOW DISTRIBUTION BAFFLE n

TUBESHEET 21.03" Mat'I. SA-508 Cl. 2A

! rs v

^)

• CLADDING

^

.15"-

14.3" min FLUSH \ BOWL RADIUS TUBESHEET HOLE GEOMETRY 62.62" IN D-4

\( 0.767"  ; i 0.762" _; _,

v 1.0625"

....Q ._ .Q ,

i -

< l

/

B&W NUCLEAR TECIINOLOGIES 3-4 l

4.O F* CRITERIA DEVELOPMENT 4.1 Structural Justification An analysis was performed which evaluated the joint pullout strength for a degraded tube in which the defect propagated into a full 360 degree circumferential sever at the F* distance (7.4].

This analysis utilized the normal operating and faulted condition loadings as well as Reg. Guide 1.121 and ASME Code safety factors.

Tubesheet bow, pressure effects, thermal effects, seismic and flow loading effects were considered relative to their impact on reducing the holding power of the rolled tube-to-tubesheet interface. A secondary loading condition for locked tubes was also considered.

Room temperature mechanical testing was performed on qualification mockups to the loadings described above at various F* lengths.

Primary to secondary leakage of the various F* lengths was also determined. Finally, the qualification tubes were pulled to failure to determine the structural adequacy of the rolled tube-to-tubesheet joint over the F* length.

4.2 Establishina F* Criteri_a_

An analytical technique was developed to determine the required F*

length for the actual steam generator tubes based on the measured joint strength determined by room temperature mechanical testing.

The F* length is determined by ratios that correct for the differences between the mechanical test conditions of the mockups and the actual steam generator conditions. The equation used to calculate the required F* length is: m

( (d) J B&W NUCLEAR TECIINOLOGIES 4-1 _

D

4 where:

(

(d) 1 The above equation establishes the minimum F* longth for structural adequacy to resist imposed axial loads. In addition, the minimum F* longth must limit primary to secondary leakage to within allowable limits. The leak rates for the F* length were determined through testing mockups representative of the steam generator.

l B&W NUCLEAR TECIINOLOGIES 4-2 l

1

5.o QUALIFICATION ANALYSES AND TESTS The qualification analyses and testing program for the F* criteria focused on satisfying the following objectives:

o Establish tube loads based on operating and faulted conditions for evaluating F* lengths.

o Perform mechanical tests necessary to verify the F* criteria as a structurally sound, leak limiting joint which meets Reg.

Guide 1.121 margins of safety. '

o Analytically adjust mechanical test condition results for actual steam generator conditions.

o Perform ECT verification testing to determine the accuracy associated with length measurements for final F* criteria determination.

The analytical approach used to determine tube loads was discussed in Se-tion 4.1 and is detailed below in Section 5.1.

Mechanical testing was performed on mockups designed to represent the range of conditions existing in the steam generators. The tubes in these mockups had full 360* severs at the F* length being tested. Testing included press'.re cycling, thermal evaluation, locked tube load tests, ultimate joint strength tests, and leak tests. These tests are described in Section 5.2 below.

The F* length to satisfy structural requirements was calculated using the equation of Section 4.2 and the mechanical test results.

These results were adjusted for the operating conditions analyzed in Section 5.1.

The ECT measurement accuracy testing was performed using multiple probe types in multiple mockups. A statistical evaluation of the

! B&W NUCLEAR. TECIINOLOGIES 5-1 l

g.

results was performed to establish the final F* correction factor for measurement accuracy.

5 . l' ' Analyses j Analyses were performed to determine axial tube loads for operating and faulted conditions for use in the mechanical testing  !

f described in Section 5.2. The combined radial stresses imposed.on

(- the installed tube-to-tubesheet joint determine the axial strength of the joint and thus determine the required tube engagement

]

length (F*). The following parameters were included in - the 1

! analyses:

l Radial preload stress from tube installation 1 Thermal effect Internal (primary) pressure effect Tubesheet bow effect i 1

[

(d)

-)

t

] 2 The axial load that the joint must resist varies depending on the design condition being evaluated. Thus multiple cases were .)

analyzed and the testing performed to encompass the worst case.

The calculations (radial stress and axial load) were performed for four different cases and are summarized in Table 5.1.1:

normal operating condition faulted condition locked tube condition tested mockup configuration B&W NUCLEAR TECIINOLOGIES 5-2

......,,A

i'

-4 5.1.1 Radial Stress 1.

The radial preload stresses were determined by testing mockups with tubing installed in the same manner as the steam generators at Byron Unit 1 and Braidwood Unit 1. After tube l installation, the tubesheet was cut away from the tubing and l the expanded tube OD measured. By comparing the measured tube OD with the tubesheet bore, the tube springback was I

determined. [ (d)

] The radial stress equivalent to this springback was then calculated [7.4] and is presented in Table 5.1.1.

The differential thermal growth between- the tube and tubesheet increases the tube OD radial stress and thus serves to strengthen the tube-to-tubesheet joint. For conservatism, the effect of d.ifferential thermal growth is calculated for the cold leg, since the higher temperature in the hot leg gives a higher radial stress and thus a stronger joint. g

[

(d)

]

B&W NUCLEAR TECIINOLOGIES 5-3

-\

Because of the analysis model used, the " Total Radial Stress" does not equal the sum of the individual radial stresses.

The ring model geometry used in the analysis changes when the residual radial stress is set to zero to quantify the other individual effects. Thus, the individual radial stresses are close approximations of the actual stress.

5.1.2 Axial Loading

• The axial loads imposed on the tubes for the four cases are summarized in Table 5.1.1 [7.4]. The normal operating load is determined by the end force applied to a tube from three times normal operating differential pressure. The faulted load was derived by applying a safety factor of 1.43 to the force generated during faulted conditions. The derivation of the locked tube loading is discussed in section 5.1.3.

5.1.3 Locked Tube Loading s

(

(c)

]

B&W NUCLEAR TECHNOLOGIES 5-4 I

_m___________.___

-f I'

(

l' i

(c)

)

[ (d)

] As discussed earlier this load is displacement limited. If the tube-to-tubesheet joint slips then the load reduces proportionally by the ,

amount of joint movement.

5.1.4 F* Determination and Correction By analyzing the three steam generator loading conditions summarized in Table 5.1.1, (

(d)

) The F* equation (Section 4.2) used to correct for differences between the testing mockups and actual steam generator conditions can be reduced to:

l

[ (d) )

I

\

B&W NUCLEAR TECHNOLOGIES 5-5 I'

's where:

[

(d)

]

The following values are contained in Table 5.1.1:

[

} (d) 3 Thus the F* equation further reduces to:

[

l (d) l 3

Where:

I

[

(d)

]

The above equation will be used to correct mockup test conditions for actual steam generator conditions. The values for [(c)] and [(c)] are determined from the tests presented in Section 5.2.

I l

l B&W NUCLEAR TECIINOLOGIES 5-6

5.2 Mechanical Testina The structural adequacy of the tube-to-tubesheet joint was evaluated by testing different F* lengths for joint strength and leak tightness. The effects of different rolled tube lengths, tubing yield strength, pressure and thermal cycling, tubesheet bore surface finish, and tubesheet bore diameter were included.

Normal operation, faulted, and locked tube conditions were tested.

~

5.2.1 Specimen Description The F* qualification specimens consisted of mockup blocks fabricated from material with the same material properties as the W-D4 tubesheet material. [

(d) l

]

B&W NUCLEAR TECHNOLOGIES 5-7 u _._.___._ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

g After the perimeter and primary side tube sections were expanded into the . block, the F* test tube opecimens were installed. These tubes were inserted through the bore in the top of the block until contact was made with the primary side-tube section. The tube was restrained from moving and rolled in place from the primary side. The physical separation.

between tubing sections represented a full 360 degree sever at the F* distance. Roll expansion lengths of 1 inch and 1 1/2 inches were tested.

Various installation parameters such as tubesheet bore diameter, tubesheet bore surface finish, and tubing yield strength were evaluated to address a wide range of potential steam generator conditions. [

(d) 3 Table 5.2.1 provides a summary of the qualification specimen installation parameters.

5.2.2 Leak Tests The leak rate was determined by maintaining the test assembly at test pressure witn a calibrated pressure generator, and measuring the volume of makeup water injected to maintain the test pressure over the test interval. The leak rate of the rolled tubesheet joints was determined at room temperature for normal operating differential pressure (1430 psi) and maximum faulted differential pressure (2750 psi). . The leak test at 2750 psi was repeated after specimens were subjected to 400 pressure cycles (Section 5.2.3) to simulate normal startup and shutdown transients. :q The acceptance criteria for leakage was based on the j technical specification limit of 1 gpm. This limit was. j divided by the number of tube ends to be evaluated against the F* plugging criteria. With four steam generators per

' B&W NUCLEAR TECHNOLOGIES 5-8 L

i

.s.

plant, 4578 tubes per steam generator, and two tube ends per tube, [

(c) 3 The technical specification leakage limit is based on the maximum allowed primary to secondary leakage for continued plant operation. Thua the leakage limit only applies to normal operating differential pressure. The test specimens were also leak tested at faulted differential pressure to insure that excessive primary to secondary leakage would not occur in the event of a faulted transient.

The results of the leak tests are summarized in Table 5.2.2.

Several observations from these leak rates are discussed below.

[

i.

(d)

Y r

l B&W NUCLEAR TECHNOLOGIES 5-9

3

[

(d)

]

5.2.3 Tensile Tests l

A series of tensile tests were performed to determine the strength of the rolled tube-to-tubesheet joint. First the joints were subjected to the maximum loading from Table 5.1.1. The joints were then subjected to locked tube loadings and to pressure and axial load cycling. Finally the joints were pulled to ultimate load. Table 5.2.3 provides a summary of the specimens, the tests performed, and the results.

From Table 5.1.1, the largest axial load [(d)] which comes from 3 times the normal operating pressure differential of 1423 psi. [

(d) .

]

A second test evaluated the locked tube condition described in Section 5.1.3. [

(d)

] As discussed in Section 5.1.3, the locked tube loading is displacement limited. This means that as the joint moves', the applied load . reduces linearly with the movement. Thus during testing, the applied B&W NUCLEAR TECHNOLOGIES 5-10

T load was reduced when joint movement was detected to simulate locked tube loading in the steam generator.

The third test was load cycling to simulate normal plant transients. Normal startup and shutdown transients were simulated by pressure cycling specimens from 0 to 1430 psi pressure differential for 400 cycles. [

(d) -

]

Joint slippage was monitored for both cycling tests and >

leakage rates were measured after the pressure cycling.

[

(d)

]

The final test was an ultimate load test where joints were loaded until failure. [

(d)

] The axial load was applied by combining 2750 psi pressure differential while axially pulling the tubes.

The acceptance criteria for the load tests was no excessive-slippage under operating and faulted condition loads.

[

(d) l ] However, excessive movement would I

l l

f B&W NUCLEA.R TECHNOLOGIES 5-11

indicate that'the joint had little or no structural integrity and could eventually lose much of its loak tightness. The movement criteria does not apply to locked tube loading since this is a secondary load and is displacement limited.

The results of the tensile tests are summarized in Table 5.2.3 and discussed below.

[

(d)

]

5.3 HpE Measurement Testina The F* lengths tested were measured-in a laboratory environment i with precise equipment. Applying the F* criteria in the steam generator will be based on a length measured by ECT. Any errors associated with ECT measurement of the- F* rolled _ tube' length beyond an ECT indication . must be included in the final F*

criteria. Thus, testing was performed to determine the accuracy of ECT measurement techniques.  ;

.j

. B&W NUCLEAR TECIINOLOGIES 5-12

5.3.1 Test Equipment Standard ECT equipment and techniques that are commonly used during normal in-service inspections were used to measure the F* 1ength of the mockups used during the mechanical joint testing described in Section 5.2. [

(c)

)

5.3.2 F* Length Verification Methodology

[

(c)

]

s The testing of MRPC probes was performed with frequencies of

[

(c)

] All measurements were made from the initial excursion of the tubesheet signal. Distances were then measured to the initial excursions of the roll signal and the crack signal.

The ECT data was then analyzed to determine the F* length of each specimen. Physical measurements of the same lengths were taken for comparison using calibrated digital calipers.

Figures 5.3.2.1 and 5.3.2.2 provide sample plots showing l

B&W NUCLEAR TECIINOLOGIES 5-13

1 where the key points were selected for the ECT measurement of the F* length.

5.3.3 Results Four test blocks with four F* tubes per block were pulled 3 times each with bobbin and MRPC probes and the ECT measurements for each specimen were averaged [7.6]. The differences between the ECT F* lengths from the various ECT techniques and the actual measured F* lengths are summarized in Tabla 5.3.3 below:

Table 5.3.3 ECT Measurement Accuracy Comparisons

[

(d)

)

As illustrated above, [

(d) l 3

l l B&W NUCLEAR TECIINOLOGIES 5-14

io 5.4 Determininct Final F* Criteria The final F* length is deterinined by combining the FA equation derived in Section 5.1 with the mechanical test results and with the uncertainty associated with ECT measurement:

( (d) )

Where:

[

(d) ,

]

Several mockup blocks were heated to determine what effect, if any, plant operating temperature had on the rolled tube-to-tubesheet joints. Since the heated blocks more accurately represent the conditions expected in actual steam generator conditions, the test results from these samples were used to determine the required F* length.

The tensile test results for the heated tube samples are conservatively analyzed to determine the F u with a 95% tolerance limit. (

(d)

]

[

(c)

]

B&W NUCLEAR TECIINOLOGIES 5-15

\

4

( (c)

]

5.5 Boric Acid Corrosion Within the Tubesheet The effects of boric acid corrosion on the carbon steel tubesheet were examined as part of the F* qualification program. In the l

event that the defect in the tube went 100% through wall, the tubesheet bore would be exposed to primary side fluid. At low temperatures with aerated boric acid solutions, some corrosion may be expected.

[

(c)

]

The defects associated with PWSCC in the tubesheet region are typically minute which limits the amount of " flowing solution" available to replenish boric acid at the tubesheet. Furthermore, dissolved hydrogen in the primary chemistry acts as an oxygen scavenger to minimize corrosion throughout the primary bystem.

These two factors make boric acid attack on the tubesheet an unlikely scenario.

Some RSGs utilize small concentrations of boric acid in the secondary water chemistry to help mitigate caustic IGA in the crevices. Thus, all of the carbon steel surfaces on the secondary side become exposed to some level of boric acid.

For the reasons discussed above, there is a very low probability of any significant corrosion of the tubesheet bore associated with L

boric acid corrosion. [ (c) ) ,

B&W NUCLEAR TECHNOLOGIES 5-16

( (c)

, ]Such a small level of degradation would have no impact on the F* joint nor the structural adequacy of the tubesheet.

i f-l t

I l

l B&W NUCLEAR TECHNOLOGIES 5-17 i

s i.

TABLE 5.1.1 RADIAL STRESS AND AXIAL LOADING

SUMMARY

[

l l

l (c) 4

]

B&W NUCLEAR TECIINOLOGIES 5-18

\

FIGURE 5.2.1 MOCKUP BLOCK LAYOUT

[

(c) v

]

l l

l l

B&W NUCLEAR TECIINOLOGIES 5-19

TABLE 5.2.1 QUALIFICATION DPECIMEN INSTALLATION

SUMMARY

(

(d) 3 1

B&W NUCLEAR TECIINOLOGIES 5-20 i

TABLE 5.2.2 LEAK TEST RESULTS

(

l' (d) 3 B&W NUCLEAR TECHNOLOGIES 5-21

TABLE 5.2.3 TENSILE TEST RESULTS

[

(d)

.]

t a

B&W NUCLEAR TECIINOLOGIES 5-22

7 4

FIGURE 5.3.2.1 BAMPLE DOBBIN PLOT SHOWING SELECTION POINTS (Sheet 1 of 3)

(

s (d)

]

B&W NUCLEAR TECIINOLOGIES 5-23

FIGURE 5.3.2.1 t.-

SAMPLE BOBBIN PLOT SHOWING SELECTION POINTS (Sheet 2 of 3)

[

(d) l 1

1

-. -1 l

I ]

l B&W NUCLEAR TECHNOLOGIES 5-24 l-.

FIGURE 5.3.2.1 8 AMPLE BOEDIN PLOT BHOWING SELECTION POINTS (Sheet 3 of 3)

[

1 l

l l

l (d)

]

B&W NUCLEAR TECIINOLOGIES 5-25

4 FIGURE 5.3.2.2 BAMPLE MRPC PLOT BHOWING SELECTION POINTS

[

1 1

l l

(d) t

]

B&W NUCLEAR TECIINOLOGIES 5-26

___1______________--_---_-------___________________--___----_-_-_______ - - _ _ _ _ _ _ _ _

6.O CONCLUSIONS l

l Based on the design verification analyses and testing performed, the following conclusions are provided:

o A total F* length of ((d)] inches is structurally adequate to satisfy all of the requirements for normal operating conditions with a safety factor of 3, faulted loading conditions with a safety factor of 1.43, and locked tube loading conditions for Byron Unit 1 and Braidwood Unit 1 W-D4 series steam generators.

o The primary to secondary leakage (

(d)

] technical specification limit for normal operation, o Considerable conservatism exists in the derivation of the F*

criteria. Specifically,

1) The joint strength was conservatively determined for use in developing the F* criteria.

2) The factor for ECT uncertainty is based.on the least accurate NDE technique tested, o The application of the F* plugging criteria at Byron Unit 1 and Braidwood Unit 1 does not raise any concerns over boric acid attack of the tubesheet.

Il&W NUCLEAR TECIINOLOGIES 6-1

s 9 -

7.O REFERENCFE 7.1 NRC Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes".

7.2 "ASME Boiler and Pressure Vessel Code",Section III, Subsection NB j and Division I Appendices, 1989 Edition.

l '.

1 l 7.3 BWNT Document 51-1227908, " Technical Requirements for F* and L*

Qualification".

I

-7.4 BWNT Document 32-1228356, "F* Calc for W-D RSG's".

1 7.5 BWNT Document 51-1228675, " Summary of Springback Test Results for Byron F*".

l 7.6 BWNT Document 51-1228688, " Summary of ECT Verification Testing for )

Byron 1/Braidwood 1 F*".

l 7.7 BWNT Document 51-1206178, " Boric Acid Corrosion of Oconee 1 Upper .)

Tubesheet". I i

7.8 BWNT Document 02-1189609, " Bobbin Coil Probe Speed / Data Sampling l Rate Test".

i 7.9 EPRI Report NP-6864-L, COMMITTEE FOR ALTERNATE REPAIR LIMITS FOR EZ PWSCC,.Rev.1.

I j

1 i

i B&W NUCLEAR TECHNOLOGIES 7-1

)

s APPENDIX A NO SIGNIFICANT HAZARDS REVIEW An evaluation is provided which concluder, in accordance with 10CFR50.92 (c), there are no significant haLard considerations for F* criteria application at the Commonwealth Edison Byron Unit 1 and Braidwood Unit 1 Stations. Byron Unit 1 and Braidwood Unit 1 are both four loop Westinghouse NSSS's with Mode.1 D4 steam generators.

The F* criteria maintains the structural integrity of the degraded tube as the primary pressure boundary and allows the tube to remain in-service for heat transfer and core cooling. S/G tubing I NDE is performed on a scheduled basis an therefore additional 1

! degradation is trended accordingly which allows for tube repair or plugging at a later date should that need arise.

According to 10CFR50.92(c), a proposed amendment to an operating license involves no significant hazards if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The F* criteria would require an amendment to the Station Operating Technical Specifications for in-service inspection of the reactor coolant system steam generators. However, the incorporation of this criteria into the Technical Specifications, changes none of the original plant design conditions or performance characteristics.

l I ,,

i B&W NUCLEAR TECIINOLOGIES s-1 l

a ________ .

l --

l 1

e.

Rescriotion of Chance .

1 The F* criteria defines a length of expanded tube engagement ]'

within the tubenheet bore (F* distance), below which, defects can exist and remain in service. The F* criteria includes an overall length of undegraded tube providing sufficient structural strength to withstand normal operating and faulted condition loadings with an acceptable margin of safety. The F* distance also.provides resistance to leakage to remain well within plant technical specification leakage limits. The criteria assumes that the defect is a full 360 degree circumferential sever at the overall-

)

engagement distance.

Affected Systems The steam generator is the key affected system. However, there is no reduction of fluid flow or heat transfer surface in the steam generator associated with implementing the F* criteria.

RCP pump / motor performances and turbine valve settings are not affected by the implementation of F* criteria. j Reactor core nucleonic computations (i.e., for boric acid calculations or control rod positioning) are also not affected by F*.

I Secondary side operation for feedwater, auxiliary feedwater,.

blowdown, and outage maintenance are not;affected by F* criteria j implementation.

Oualification Summarv 1

The F* plugging criteria has been fully qualified for use in-the Byron Unit 1 and Braidwood Unit 1 W-D4 series steam generators.

The use of the F* criteria will allow tubes with otherwise pluggable indications to remain in service as long as they are B&W NUCLEAR TECIINOLOGES 8-2 l

i

I below a minimum distance from the hard roll contact point near the tubesheet secondary face. This distance, referred to as F*

distance, was established at [(d)] ir'hes through analysis and testing. The F* criteria is adequate to meet technical specification requirements for leakage and maintain adequate joint strength for normal operating and faulted conditions.

l Qualification specimens were prepared simulating the same installation conditions as the original tubing. These specimens were subjected to tensile tests and leak rate tests. The loads and pressures for the testing included safety factors of 3 for l

normal operating differential pressure and 1.43 for faulted conditions. In addition, ECT was performed on a number of F*

specimens to determine measurement accuracy and repeatability.

Bobbin and MRPC techniques were used for this testing. The results of the ECT testing were factored into the F* criteria.

The effects of boric acid corrosion on the carbon steel tubesheet were also examined as part of the qualification program. In the event that the defect in the tube went 100% through wall a small region of the tubesheet could be exposed to primary side fluid.

At worst, a small amount of localized tubesheet degradation, on the order of a few mils, could occur. Such shallow attack represents no structural concerns for the tubesheet or the F*

joint.

Impact on Accidents Evaluated as the Desian Basis Since F* utilizes the "as rolled" tube configuration that exists as part of the original steam generator design, all of the design and operating characteristics of the steam generator and connected systems are preserved. The F* joint has been analyzed and tested for design, operating, and faulted condition loadings in accordance with Regulatory Guide 1.121 safety factors. At worst case a tube leak would occur with the result being a primary to secondary system leak.

Il&W NUCLEAR TECIINOLOGIES 8-3

i .

I 9

e l

l l Should a tube leak occur, the impact is bounded by the ruptured I tube evaluation submitted by the utility for the operating license. No new or unreviewed accident conditions are created by the use of F* criteria. The potential for a tube rupture is not increased from the original submittal, thus there is no impact on I accidents evaluated as the design basis.

Thus, 10CFR50. 92 (c) (1) is satisfied.

Potential for Creatina an Unanalyzed Event The failure of a tube which remained unplugged in accordance with l

the F* criteria would result in a tube leak, which is a previously analyzed condition. Since this leak would occur below the secondary face of the tubesheet its leak rate would be limited by the tube-to-tubesheet interface. Qualification testing and previous experience indicates that normal and faulted leakage would be well below technical specification limits creating no threat associated with tube rupture type leakages. Since the normal and faulted leak rates are well within the 1 GpM normal operating limit, the UFSAR analyzed accident scenarios are still bounding. This conclusion is consistent with previous F* programs approved and used at other operating plants. .

However, in the unlikely event the failed tube severed completely at a point below the F* region, the remaining F* joint would retain engagement in the tubesheet due to its length of expanded contact within the tubesheet bore, preventing any interaction with neighboring tubes. If the tube severs at a point above the F*

region, then it is covered by the tube rupture evaluation performed as part of the UFSAR. Therefore, there is not .a potential for creating an unanalyzed event.

Thus, 10CFR50. 92 (c) (2) is satisfied.

B&W NUCLEAR TECIINOLOGIES s-4

4 i

Y Imoact on Marain of Safp_tiy Based on previous responses, the protective boundaries of the steam generator are preserved. A tube with degradation can be kept in service through an F* criteria which provides an undegraded expanded interface with the tubesheet and which satisfies all of the necessary structural and leakage requirements per Reg. Guide 1.121 and the Station Technical Specifications.

j Since the joint is constrained within the tubesheet bore, there is l no additional risk associated with tube rupture. Since the UFSAR analyzed accident scenarios remain bounding, the use of an F* -

criteria does not reduce the margin of safety.

Thus, 10CFR50. 92 (c) (3 ) is satisfied.

Conclusion The use of the F* criteria described herein, to maintain tubes in service, does not represent an unanalyzed safety concern.

Furthermore, its use does not increase the risk of creating an unanalyzed accident nor does it reduce the margin of safety.

B&W NUCLEAR TECIINOLOGIES 8-5

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