ML053110148

Response to Request for Additional Information for End-of-Cycle 6 Steam Generator Inservice Inspection Reports

ML053110148
Person / Time
Site:	Watts Bar
Issue date:	11/02/2005
From:	Pace P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC7485
Download: ML053110148 (17)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 NOV 0 2 2005 10 CFR 50.4 U. S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C. 20555-0001 Gentlemen:

In the Matter of

)

Docket No. 50-390 Tennessee Valley authority

)

WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 -

REQUEST FOR ADDITIONAL INFORMATION FOR END-OF-CYCLE 6 STEAM GENERATOR INSERVICE INSPECTION REPORTS (TAC NO. MC7485)

The purpose of this letter is to respond to NRC's request for additional information dated September 7, 2005, concerning the subject steam generator reports.

The enclosure provides TVA's responses to NRC's questions as requested.

There are no regulatory commitments associated with this submittal.

If you have any questions concerning this matter, please call me at (423) 365-1824.

Sincerely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosure cc:

See Page 2 Pmwted on mcA paWs

U.S. Nuclear Regulatory Commission Page 2 NOV 0 2 2005 Enclosure cc (Enclosure):

NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS NRC QUESTION 1 Subsequent to your 2003 steam generator tube inspections (End-of-Cycle 5), you identified five tubes which were left in service despite having indications in the parent tube below the location where a sleeve was installed during the 2003 outage.

Please confirm that these five tubes (row 18 column 35; row 19 column 32; row 22 column 31; row 22 column 37; row 42 column 55) were plugged during your 2005 outage (end-of-cycle 6).

In addition, please confirm that no similar tubes were left in service following your 2005 steam generator tube inspection (i.e.,

confirm that F* was not applied to a sleeved tube.)

TVA RESPONSE The steam generator tubes listed above were plugged during the End-of-Cycle (EOC)-6 outage.

F* was not applied to tubes if sleeves were installed.

NRC QUESTION 2 In your June 28, 2005 letter, it was indicated that several of the tubes remaining in service contained indications located where the tube passes through the flow distribution baffle.

Please confirm that your amendment request for implementing the voltage-based tube repair criteria discussed in Generic Letter (GL) 95-05 addressed the conditions in Section 2.a.3 of to GL 95-05 (since specific U.S. Nuclear Regulatory Commission (NRC) approval is needed to apply the voltage-based limits to flow distribution baffle intersections).

a)

That is, confirm that your amendment request addressed the causal factors for high voltage growth at flow distribution baffle intersections and the applicability of these conditions at your plant. Also, b) please discuss whether the average growth rates for the flow distribution baffle indications were less than that observed in steam generator 3 (although page 3-18 indicates that the average growth rates are less than that seen in steam generators 1 and 2, flow distribution baffle indications were found in steam generators 1, 2, and 3).

If not, please discuss the implications.

E-1

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS TVA RESPONSE a)

TVA requested a change to Technical Specifications allowing the application of voltage based criteria at tube support plates and flow distribution baffle plates in a letter dated April 10, 2000, and provided additional information in letters dated September 18, 2000, August 22, 2001, November 8, 2001 and January 15, 2002.

NRC approved the change to Technical Specifications in a letter dated February 26, 2002.

The NRC safety evaluation attached to the February 26, 2002 letter approved the amendment as applicable to both tube support plates (TSPs) and flow distribution baffle (FDB) plates.

b)

The indications detected at the FDB that were used in the growth analyses are shown below in a copy of Table 3-10 of TVA's June 28, 2005 letter:

Change EOC6 EOC5 in SG Row Column Support Plugged Volts Volts Volts 1

8 75 H01 Yes 0.31 0.19 0.12 2

18 79 H01 Yes 0.46 0.29 0.17 2

27 85 H01 0.38 0.19 0.19 2

38 77 H01 Yes 0.27 0.22 0.05 3

14 86 H01 0.55 0.69

-0.14 Average change = growth / EOC-5 volts = 25 percent The average growth rates (growth/EOC-5 volts) for each of the four steam generators is:

Average SG Growth Reference*

1 26%

Table 3-5 2

27%

Table 3-6 3

14%

Table 3-7 4

9%

Table 3-8

• Reference is found in TVA's letter to NRC dated June 28, 2005 The statement that the average growth rate of the FDB indications is less than the average of Steam Generator 1 and Steam Generator 2 was made to indicate that the growth rate of the FDB indications is in the range of the rate observed for all indications in the steam generators with the highest rate and is therefore, not considerably E-2

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS different from the growth rate of indications in the other TSPs.

In the predictive analysis the growth rates of individual steam generators was not used.

Instead, the bounding growth rate curve shown in Figure 3-13, Figure 3-14, and Table 3-9 of TVA's June 28, 2005 letter is a result of the growth in Steam Generator 1 and Steam Generator 2. Thus, while the average growth rate observed in the FDBs exceed the growth in Steam Generator 3, the growth rate in Steam Generator 3 is bounded by the growth observed in Steam Generator 1 and Steam Generator 2, and therefore, there are no implications.

A better demonstration of the observation that the growth rates are not considerably different is a comparison with the bounding cumulative distribution of the growth rate in the four steam generators shown in Figure 1 below and in Figure 3-13 in the June 28, 2005 report.

Also shown is the cumulative distribution of the FDB indications conservatively adjusted by Benard's approximation equation (Note 1) because of the small number of data points.

Note 1 Benard's approximation equation, i.e.,

Median Rank Fraction =Actual Rank -0.3 NT + 0.4 E-3

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS FIGURE 1 Growth Comparison, Volts per cycle 0.9 0.8 0.7 0.6

-Bounding Figure 3-13 0.5 4-0.3 0.2 0.1 0

• 0 2

4 Volts /Cycle From the figure, it is clear that the growth rate of the FDB indications is not considerably different from the growth rate of the TSPs.

NRC QUESTION 3 One tube was identified with a 6.32 volt indication.

Please discuss what actions, if any, were taken to ensure that this tube had adequate structural integrity since the voltage exceeded the structural limit of 5.65 volts.

RESPONSE

There is no requirement to do single-tube structural integrity calculations.

The requirement in GL 95-05 is for a probability of burst for the bundle.

This indication was included in the bundle analysis, and the steam generator met structural E-4

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS performance criteria.

Therefore, no additional actions were taken.

If a single tube is of interest, according to Figure 6-2 of Addendum 6 of EPRI Report NP-7480-L shown below, an indication greater than 12 volts would be expected to meet the steam line break pressure with 95 percent probability using 95/95 lower tolerance limit material properties.

The 5.65 limit in the Addendum is used as an upper bound repair limit.

Above this limit, rotating coil examination cannot be used to keep the indication in service.

Below this limit, if rotating pancake coil (RPC) does not confirm the indication, the tube can be left in service.

FIGURE 6-2 M.-INCH BURST PRESSURE CORRELATION Burst Pressure vs Bobbin Amplitude 3/4" x 0.043" Alloy 600 MA SG Tubes Database @ 650'F, Sy +Su = 143.13 ksi 12.0 10.0 8.0 0.0 4.0 2.0 0.0 0.1 1

10 Bobbin Amplitude (Volts) 100 NRC QUESTION 4 On page 5-7 of your June 28, 2005, letter, you indicated that the end-of-cycle 6 predicted values of the probability of burst and leakage were conservative because they were based on a very conservative industry voltage growth rate.

However, your January 15, 2004, letter (ML040220171) indicates that Table 3-9 was used E-5

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS for the End-of-Cycle 6 projections.

Table 3-9 provided a bounding growth rate from all four steam generators at Watts Bar.

Please confirm whether an industry voltage growth rate or a plant-specific growth rate was used in the calculations.

RESPONSE

The statement "The predicted values of the probability of burst and leakage were conservative because they were based on a very conservative industry voltage growth rate in Reference 5." is incorrect.

This is an editorial error.

A plant specific growth rate was used in the calculations.

Table 3-9 of the Cycle 5 90-day Report provided a bounding growth rate of all four steam generators at Watts Bar as NRC indicated.

NRC QUESTION 5 The largest voltage indication observed at the End-of-Cycle 6 (2005) was not predicted.

This resulted in exceeding the limiting projection on probability of burst.

In addition, the most limiting accident induced leak rate (0.175 gallons per minute (gpm)) was exactly predicted (although for a different steam generator).

Given that the maximum voltage indication was not predicted and that your probability of burst projections for End-of-Cycle 7 (8.65 x 10-3) are near the limits (10-2), discuss what corrective actions were taken to ensure such an under-prediction in the maximum observed voltage (and probability of burst) does not occur for the End-of-Cycle 7. The NRC recognizes that tube inspections will not be performed at the End-of-Cycle 7 due to the planned replacement of the Watts Bar steam generators.

RESPONSE

TVA has followed industry guidance at Sequoyah for preventing large growth for outside diameter stress corrosion cracking (ODSCC) at support plates.

This guidance suggests that an indication between 1 volt and 2 volts may experience large growth and recommends plus point inspections of indications greater than 1 volt instead of the required 2 volts for Model 51 steam generators with preventive plugging.

The requirement for plus point examination for Model D steam generators at Watts Bar is already 1 volt and the repair limit is 1 volt, therefore, no additional examinations were performed.

All indications except for the one 6.32 volt indication were E-6

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS under the predicted maximum voltage.

The one 6.32 volt indication significantly affected the predictions for Unit 1 Cycle 7.

Since performance criteria were met and predicted to be met for the entire cycle, no additional actions were taken for indications less than 1 volt.

NRC QUESTION 6 Figure 3-17 indicates that several of the larger voltage indications exhibited negative growth.

Although the staff is aware that this does occur at other plants, the large number of negative growths (resulting in a decreasing trend in growth rate as a function of beginning-of-cycle voltage) does not appear to be consistent with that observed at other plants.

Please discuss any insights on this trend including a discussion of whether these results draw into question the non-destructive examination uncertainty models applied at Watts Bar.

RESPONSE

The trend of decreasing growth with increasing beginning-of-cycle (BOC) voltage has been observed and reported in other plants, including Sequoyah Unit 1 EOC 11.

This is not unexpected when the average growth rate is small.

This phenomenon is likely caused by small indications with larger non-destructive examination (NDE) error in one cycle growing to an easier to characterize size in the next inspection, thus less NDE error, giving the appearance of negative growth.

NRC QUESTION 7 In section 4.6, the upper voltage repair limit was calculated.

Please discuss the purpose for the "(518/482)" adjustment in calculating this limit.

RESPONSE

The upper voltage repair limit is based on the structural limit in Table 4-1 of 5.65 volts for accident pressure of 2405 pounds per square inch (psi).

It must be reduced by considering the projected voltage growth during the next cycle and NDE uncertainty.

The maximum average percentage growth rate for any steam generator is seen from Table 3-6 (SG 2) to be 27 percent per 482 day Cycle 6. According to Reference 1 in TVA's June 28, E-7

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS 2005 letter, the minimum growth adjustment is 30 percent per effective full power year (EFPY)(42.5 percent per cycle for the anticipated 518 effective full power days (EFPD)Cycle 7).

Therefore, the specific maximum value of 42.5 percent and 20 percent for NDE uncertainty will be used to estimate the voltage repair limit.

This results in an upper voltage repair limit of 5.65 / (1 + 0.425 + 0.20) = 3.47 volts.

No indications equal to or greater than this voltage were left in service.

The limiting free span burst pressure is three times normal operating differential pressure (NODP).

Reference 2 in TVA's June 28, 2005 letter, notes a reactor coolant system (RCS) pressure of 2235 pounds per square inch gauge (psig) and steam pressure of 934 psig, making 3NODP = 3903 psi.

The upper voltage repair limit for FDB intersections is based on the structural limit from Figure 6-2 of Reference 6 in TVA's June 28, 2005 letter, of 3.29 volts for a free span burst pressure of 3903 psi.

It must be reduced by considering the projected voltage growth during the next cycle and NDE uncertainty.

The average percentage growth rate for the FDBs, from Table 3-10, is 25 percent during the 482 day Cycle 6. According to Reference 1 in TVA's June 28, 2005 letter, the minimum growth adjustment is 30 percent per EFPY (42.5 percent per cycle for the anticipated 518 EFPD Cycle 7).

Therefore, the specific maximum value of 42.5 percent and 20 percent for NDE uncertainty will be used to estimate the voltage repair limit.

This results in an upper voltage repair limit of 3.29 / (1 + 0.425 + 0.20) = 2.02 volts.

No FDB indications equal to or greater than this voltage were left in service.

The small difference in the upper voltage repair limit is conservative and does not change the results.

NRC QUESTION 8 Section 5.0 describes the condition monitoring assessment.

Figures 5-1 through 5-4 depict the distribution of end-of-cycle voltages adjusted by the non-destructive examination uncertainty distribution.

Please discuss whether the discrete distributions in these figures (which may have been truncated/adjusted for fractional indications) were used in the condition monitoring assessment or whether the condition monitoring assessment utilized a non-truncated/adjusted distribution of indications.

E-8

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS

RESPONSE

Condition monitoring assessments use the as-found-distribution with size adjusted for NDE uncertainties.

These distributions are not truncated or adjusted otherwise.

The fractional indications are a result of the 0.6 probability of detection and are used in operational assessments.

The Monte Carlo analysis for the condition monitoring assessment calculation predicts some indication trial voltages in excess of 8.2 volts in the case of Steam Generator 2, where the values are truncated in Figure 5-2 at 6.9 volts.

The values are truncated for graphical display only by integrating the upper tail of the Monte Carlo trial results to 0.3 and 0.7 of an indication.

NRC QUESTION 9 Section 4.7 provides an assessment of the probe wear criteria used at Watts Bar.

In a portion of this assessment, the ratio of the current number of indications greater than 1 volt to the total number of these indications that were inspected with a worn probe in the previous inspection was compared to the ratio of the number of indications greater than 1 volt to the total number of indications in the current inspection.

Based on similar ratios, a conclusion was drawn that there was no significant effect of probe wear on the population of indications.

The staff notes that such a comparison is only valid if the number of tubes inspected with both a worn and good probe is comparable.

That is, if the number of tubes inspected with worn probes is significantly different from the number of tubes inspected with a good probe, an erroneous conclusion may be made with respect to the adequacy of the probe wear criteria.

As a result, please compare the percentage of new indications at the end-of-cycle (EOC) 6 that were inspected with a worn probe during the EOC-5 inspection to the percentage of new indications that were inspected with a good probe during the EOC-5 inspection.

In addition, please compare the percentage of new indications greater than or equal to 0.5 volts during the EOC-6 inspection that were inspected with a worn probe during the EOC-5 inspection to the percentage of new indications greater than or equal to 0.5 volts during the EOC-6 inspection that were inspection with a good probe during the EOC-5 inspection.

If there are significant differences, please provide an assessment of the adequacy of the probe wear criteria and its impact on your operational assessment for EOC-7.

Calculation of the above percentages requires the E-9

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS total number of tubes inspected with a good and worn probe during the EOC-5 inspections.

A value of 0.5 volts was chosen to be consistent with the NRC staff's approval of the alternate probe wear criterion (refer to NRC letter to Nuclear Energy Institute dated February 9, 1996).

RESPONSE

A review of the data from the 2003 EOC-5 inspection was conducted to identify the tubes that were inspected with a worn probe and not reinspected with a good probe.

The list of calibration groups that failed calibration was identified.

The following information was developed by interrogating the data:

TABLE 1 Identification of Tubes Tested with Worn Probe in 2003 EOC-5 and Categorization of New Indications Number of Number of new (2005) new (2005)

Tested Number of Number of indications indications Plug with Tested indications new in tubes in tubes prior Worn with Good in 2005 indications with worn tested with original to probe Probe EOC-6 in 2005 probe in good probe SG tubes 2003 (HL) (Difference) (HL and CL)

EOC-6 2003 in 2003 SG1 4674 49 1974 2651 330 122 59 63 SG2 4674 57 2727 1890 193 85 38 47 SG3 4674 48 1776 2850 201 76 23 53 SG4 4674 98 1883 2693 216 94 17 77 Total I _

8360 10084 940 377 137 240 NOTE:

The number of new indications in tubes tested with a worn probe in 2003 of 137 is higher than the number of 106 reported in the 90 Day Report.

This is a consequence of a more thorough search for such indications.

This small difference does not impact the conclusions.

E-10

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS TABLE 2 Ratio of New Indications in Tubes with Worn/Good Probe to Number of Tubes Tested with Worn/Good Probe.

(Data from Table 1.)

Ratio of new indications in Ratio of new indications in tubes with worn probe to number tubes with good probe to number SG of tubes tested with worn probe of tubes tested with good probe SG1 0.0299 0.0238 SG2 0.0139 0.0249 SG3 0.0130 0.0186 SG4 0.00903 0.0286 SUM 0.0658 0.0958 Average 0.0165 0.0240 TABLE 3 Number of New Indications Equal to or Greater than 0.5 Volts Number of new (2005)

Number of new Number of new (2005) indications greater indications in indications greater or or equal to 0.5 2005 greater or equal to 0.5 volts in volts in tubes equal to 0.5 tubes with worn probe tested with good SG volts in 2003 probe in 2003 SG1 82 44 38 SG2 45 19 26 SG3 51 15 36 SG4 63 10 53 Total 241 88 153 TABLE 4 Ratio of New Indications in Tubes with Worn/Good Probe to of Tubes Tested with Worn/Good Probe (indications equal to or greater than 0.5 volts)

Number Ratio of new indications Ratio of new indications in in tubes with worn probe tubes with good probe to to number of tubes with number of tubes with good SG worn probe probe SG1 0.0223 0.0143 SG2 0.00697 0.0138 SG3 0.00845 0.0126 SG4 0.00531 0.0197 SUM 0.0430 0.0604 Average 0.0108 0.0151 E-ll

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS The voltage distributions of the new indications are shown in Figures 1 through 4 for Steam Generators 1 through 4 respectively.

FIGURE 1 New Indications, SG 1 1

ad-0~~~~ ~

~ 8.

-=

oo rb 0.9 0.8 0.7 -

__Good probe (2003) 63 points 0.6 0.5

• Worn Probe (2003) 59 points 0.4 i

I F

0.3 0.2 0.1 0

0.5 1

1.5 2

2.5 3

EOC-6 Volts E-12

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS FIGURE 2 New Indications, SG 2 0.9 -

m

=

0.8 -

0.7 -_

0.6 -

Good Probe (2003)

__0_5

_ _47 points

-*Worn Probe (2003) 0.--

38 points 0.3 0.2 0.1 0~

0 0.5 1

1.5 EOC-6 Volts FIGURE 3 New Indications, SG 3 1 -

0.9 _

0.8 -

0.7 -

t 0.6 -

Good Probe (2003) 0.5 53 points 0.4

-Worn Probe (2003) 0.3 - 323 points 0.3 -

0.2 -

0.

0 1

2 3

EOC-6 Volts E-13

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS FIGURE 4 New Indications, SG4 1 '--------

0.9 0.8-0.7-0.6 Good Probe (2003)

L0.

77 points 0Q.5*

04

Worn Probe (2003) 0'

___7points 0.2-0.1 XL

=

=

=

0 0

0.5 1

1.5 2

2.5 3

EOC-6 volts Discussion From Table 1 it is seen that the number of new indications found in tubes that had previously been tested with a good probe is slightly greater than the number found in tubes that had been previously tested with a worn probe.

Table 1 also shows that the number of tubes tested with a worn probe is only slightly less than the number of tubes tested with a good probe.

From the new indication totals in Table 1, 36.3 percent of the new indications were in tubes previously tested with a worn probe and 63.7 percent were in tubes previously tested with a good probe.

From Table 3, the corresponding percentages for new indications equal to or greater than 0.5 volts are essentially the same, 36.5 percent and 63.5 percent respectively.

The ratio of new indications to the number of tubes tested is given in Table 2. There is some variation among the different steam generators as would be expected from random events.

The totals, however, show that the ratio of new indications in tubes previously tested with worn probes to the total number of tubes tested with worn probe is 0.0165, or 1.7 percent.

The ratio of new indications in tubes previously tested with good probes to the total number of tubes tested with good E-14

ENCLOSURE WATTS BAR NUCLEAR PLANT (WBN)

UNIT 1 REQUEST FOR ADDITIONAL INFORMATION END-OF-CYCLE 6 STEAM GENERATOR 15 AND 90 DAY INSERVICE INSPECTION REPORTS probes is 0.0240, or 2.4 percent. For new indications equal to or greater than 0.5 volts, the corresponding totals from Table 4 show that the ratio of new indications in tubes previously tested with worn probes to the total number of tubes tested with worn probe is 0.0108, or 1.1 percent. The corresponding ratio of new indications in tubes previously tested with good probes to the total number of tubes tested with good probes is 0.0151, or 1.5 percent.

These observations indicate that there is not a significant difference in the rate of detection of new indications in tubes previously tested with worn probes or good probes.

The voltage magnitude of the new indications in tubes previously tested with worn probes and good probes is essentially the same as seen in Figures 1 through 4. This further supports the position that significant differences in results are not occurring due to previous testing.

E-15