DCL-05-090, Response to NRC Request for Additional Information Regarding License Amendment Request 05-01 Re Revision to Technical Specification 5.5.9, Steam Generator Tube Inspection Report.

Response to NRC Request for Additional Information Regarding License Amendment Request 05-01 Re Revision to Technical Specification 5.5.9, Steam Generator Tube Inspection Report.

ML052440396
Person / Time
Site:	Diablo Canyon
Issue date:	08/25/2005
From:	Oatley D Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-05-090
Download: ML052440396 (61)

Text

Pacific Gas and ElectricCompany David H. Datley Diablo Canyon Power Plant Vice President and PO. Box 56 General Manager Avila Beach, CA 93424 August 25, 2005 '

805545.4350 Fax: 805.545.4234 PG&E Letter DCL-05-090 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Request for Additional Information Regarding License Amendment Request 05-01. "Revision to Technical Specification 5.5.9. 'Steam Generator (SG) Tube Surveillance Program,' and 5.6.10. 'Steam Generator (SG)

Tube Inspection Report,' to Allow Use of the W* Alternate Repair Criteria for Indications in the Westinghouse Explosive Tube Expansion (WEXTEX) Region on a Permanent Basis" PG&E Letter DCL-05-018, dated March 11, 2005, submitted License Amendment Request (LAR) 05-01, "Revision to Technical Specification 5.5.9, 'Steam Generator (SG) Tube Surveillance Program,' and 5.6.10, 'Steam Generator (SG) Tube Inspection Report,' to Allow Use of the W* Alternate Repair Criteria for Indications in the Westinghouse Explosive Tube Expansion (WEXTEX) Region on a Permanent Basis.' LAR 05-01 proposes to allow use of the SG tube W star (W*) alternate repair criteria (ARC) on a permanent basis and to revise the Technical Specification (TS) 5.6.1 0.d NRC notification requirements for the voltage-based ARC.

On June 10, 2005, the NRC staff requested additional information required to complete the review of LAR 05-01. PG&E's responses to the staffs questions are provided in Enclosure 1. Enclosure 2 provides marked-up TS pages. Enclosure 3 provides retyped TS pages. Enclosure 4 provides Westinghouse proprietary technical information, which supports the responses. Enclosure 5 provides a nonproprietary version of the technical information contained in Enclosure 4.

The technical information contained in Enclosure 4 contains information proprietary to Westinghouse Electric Company LLC (Westinghouse). Accordingly, Enclosure 4 includes a Westinghouse authorization Letter, CAW-05-2034, an accompanying affidavit, a Proprietary Information Notice, and a Copyright Notice. The affidavit is signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the Westinghouse proprietary information contained in Enclosure 4 may be withheld from public disclosure by the Commission, and it addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390 of the A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak

• Diablo Canyon . Palo Verde

• South Texas Project . Wolf Creek ]}/pI

°

Document Control Desk PG&E Letter DCL-05-090 August 25, 2005 Page 2 Commission's regulations. PG&E requests that the Westinghouse proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390.

Correspondence with respect to the copyright or proprietary aspects of the application for withholding, related to the Westinghouse proprietary information or the Westinghouse affidavit provided in Enclosure 4, should reference Westinghouse Letter CAW-05-2034 and be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

TS 5.5.9.b.2.e, TS 5.5.9.d.1.h, TS 5.5.9.d.1.k (ii), TS 5.5.9.d.1.k (iv),

TS 5.5.9.d.1.k (v).2, TS 5.5.9.d.1.k (v).6, TS 5.6.10.d.2, and TS 5.6.10.e have been revised to respond to the request for additional information. TS 5.5.9.b.2.e has been clarified to require that tubes identified as W* tubes having a previously identified indication within the flexible W* length be inspected using a rotating pancake coil probe or equivalent for the full length of the W* region during all future refueling outages. The TS 5.5.9.d.1.h tube inspection definition has been revised to exclude the portion of the tube within the tubesheet below the flexible W* length or below 8 inches from the hot leg top of tubesheet, whichever is bounding, from the tube inspection. The TS 5.5.9.d.1.k (ii) W* length definition has been revised to provide clarification that the W* length is the distance in the tubesheet below the bottom of the WEXTEX transition (BWT). The TS 5.5.9.d.1.k (iv) W* plugging limit has been revised to remove the percent degradation (40 percent) from the definition. The TS 5.5.9.d.1.k (v).2 W* repair criteria have been revised to require that axial cracks in tubes returned to service using W* shall have the projected upper crack tip below the BWT at the end of the subsequent operating cycle by at least the nondestructive examination (NDE) measurement uncertainty and crack growth allowance. The TS 5.5.9.d.1.k (v).6 W* repair criteria has been clarified to allow any type or combination of tube degradation below the flexible W* length. The TS 5.6.1 0.d.2 voltage-based repair criteria reporting requirement has been clarified to require NRC notification prior to returning the SGs to service if outside diameter stress corrosion cracking indications are identified that extend beyond the confines of the tube support plate. The TS 5.6.10.e 90-day reporting requirements have been revised to require inclusion of the radial position of the tube within the tubesheet and the W*

Zone of the tube, verification that W* indications returned to service in the prior cycle have the upper crack tip below the BWT, including NDE measurement uncertainty, and to reference the leak rate methodology described in this letter. The marked-up TS pages in Enclosure 2 and the retyped TS pages in Enclosure 3 supersede Enclosures 2 and 3 of PG&E Letter DCL-05-018 in their entirety.

This information does not affect the results of the technical evaluation or the no significant hazards consideration determination previously transmitted in PG&E Letter DCL-05-018.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak

• Diablo Canyon

• Palo Verde

• South Texas Project

• Wolf Creek

Document Control Desk PG&E Letter DCL-05-090 August 25, 2005 Page 3 PG&E requests that the license amendment be effective as of its date of issuance and to be implemented prior to startup of Cycle 14 for each unit. PG&E is making this request since the enclosed TS changes result in more conservative W* repair criteria and revised 90-day reporting requirements, which cannot be applied until cycle 14 for each unit.

If you have any questions or require additional information, please contact Stan Ketelsen at 805-545-4720.

Sincerely, 7

David H. Oatley Vice President and General Manager kjse/4328 Enclosures cc: Edgar Bailey, DHS Terry W. Jackson Bruce S. Mallett Diablo Distribution cc/enc: Girija S. Shukla A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak

• Diablo Canyon

• PaloVerde

• SouthTexas Project

• Wolfcreek

PG&E Letter DCL-05-090 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

) Docket No. 50-275 Inthe Matter of ) Facility Operating License PACIFIC GAS AND ELECTRIC COMPANY) No. DPR-80

)

Diablo Canyon Power Plant ) Docket No. 50-323 Units I and 2 ) Facility Operating License

) No. DPR-82 AFFIDAVIT David H. Oatley, of lawful age, first being duly sworn upon oath says that he is Vice President and General Manager of Pacific Gas and Electric Company; that he has executed this response to the NRC request for additional information on License Amendment Request 05-01 on behalf of said company with full power and authority to do so; that he is familiar with the content thereof; and that the facts stated therein are true and correct to the best of his knowledge, information, and belief.

David H. Oatley Vice President and General Manager Subscribed and sworn to before me this 25th day of August, 2005, by David H. Oatley, personally known to me or proved to me on the basis of satisfactory evidence to be the person who appeared before me.

Notary Pubiccom

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1. 1397547 NotoiyKttil- caltoinl County of San LLdiS Obispz~aR son Lub Obipo Coun State of California _Comm. ExPW Feb . 2007

Enclosure 1 PG&E Letter DCL-05-090 PG&E Response to NRC Request for Additional Information Regarding License Amendment Request 05-01, "Revision to Technical Specification 5.5.9, 'Steam Generator (SG) Tube Surveillance Program,' and 5.6.10, 'Steam Generator (SG)

Tube Inspection Report,' to Allow Use of the W* Alternate Repair Criteria for Indications in the Westinghouse Explosive Tube Expansion (WEXTEX) Region on a Permanent Basis" NRC Question 1:

The staff notes that the constrained crack leak rate model was developed using a simultaneous one-sided 95 percent confidence on the arithmetic average value of the leak rate. Please discuss the rationale for selecting a 95 percent confidence bound on the average and not a 9 5 th percentile prediction interval for the constrained crack leak model. Provide a graph showing leak rate as a function of contact pressure (Enclosure 5, Figure 6) which shows the 9 5 th percentile prediction interval line on the graph.

PG&E Response:

The rationale for using the confidence bound is based on the consideration that the leak rate limit is based on the sum of the individual leak rates and the prediction interval is based on estimating a bound for the leak rate from a single indication. The expected total leak rate from a group of indications is the sum of the expected leak rates from each individual indication. A simultaneous bound approach using the F-distribution was proposed since the usual confidence and prediction bound formulae using variates from the Student-t distribution are intended to apply to a single future value of the independent variable.

Figure 1 in Enclosure 4 of this letter shows the steam line break (SLB) leak rate 95th percentile one-sided prediction bound using the Student-t distribution, as well as the 95t percentile confidence bound. A comparison of the leak rate prediction results reveals that the 95th percentile prediction bound leak rate would be slightly more conservative than the 95 percent confidence bound on the average leak rate, although the expected differences would be minimal. Therefore, PG&E will apply the 95 percent prediction bound leak rate instead of the originally proposed 95 percent confidence bound on the arithmetic average value of the leak rate.

To implement this change in leak rate method, the following changes to Enclosure I of PG&E Letter DCL-05-018, uRevision to Technical Specification 5.5.9, 'Steam Generator (SG) Tube Surveillance Program,' and 5.6.10, 'Steam Generator (SG) Tube Inspection Report,' to Allow Use of the W* Alternate Repair Criteria for Indications in the Westinghouse Explosive Tube Expansion (WEXTEX) Region on a Permanent Basis,"

dated March 11, 2005, are made:

1

Enclosure 1 PG&E Letter DCL-05-090

1) Leak rate equations (2) and (4) in Section 4.1.2.1 of DCL-05-018 are no longer applicable. Equation (4) was for a simultaneous confidence bound, and is replaced by the following equation to calculate the prediction bound,

• i1 (PI - P Qw=exp[aO+aIPdi +tn-2,.S I- +_+ (d)] l where t is the Student-t variate for the fraction a in the upper tail area (e.g., 1.69 for a 1--a value of 95 percent for 34 degrees of freedom).

2) Section 4.1.2.2 of DCL-05-018 is revised to define a new bounding constrained crack leak rate for undetected indications between 8 to 12 inches below the top of tubesheet (TTS). A revised leak rate of 0.0033 gpm will be assigned to each postulated undetected indication between 8 to 12 inches below the TTS, replacing the prior value of 0.0028 gpm. This leak rate is derived from Figure 1 in Enclosure 4 of this response. The Zone B1 contact pressure at a depth of 8 inches into the tubesheet is 2557 pounds per square inch (psi). From Figure 1 in Enclosure 4 of this response, the corresponding 95 percent prediction bound leak rate at this contact pressure is 0.0033 gpm, which is slightly higher than the 0.0028 gpm leak rate based on the 95 percent confidence bound on the arithmetic average.

3) As a result of the two changes above, Section 4.1.4 of DCL-05-018 is revised to describe the potential impact of the revised leak rate method on Diablo Canyon Power Plants Units 1 and 2 (DCPP). The following table replaces Table 3 of Enclosure 1 of DCL-05-018 to provide a comparison of projected end of cycle (EOC) 13 leak rates using the 95 percent prediction bound method and the existing DENTFLO method. For Units 1 and 2, the total SG leak rates using the 95 percent prediction bound method are slightly higher than the DCL-05-018 Table 3 leak rates, which used the 95 percent confidence bound on the arithmetic average. Combining the 95 percent prediction bound method with the tube sever model, the projected EOC 13 SLB W* leak rates for the worst case Unit 1 and Unit 2 SGs are about 0.6 gpm and 2.3 gpm, respectively, compared to about 0.5 gpm and 2.1 gpm using the 95 percent confidence bound on the arithmetic average.

The TS 5.6.10.e 90-day reporting requirements have been revised to reference the leak rate methodology described in this letter. The marked-up TS pages in Enclosure 2 and the retyped TS pages in Enclosure 3 are revised to include this TS revision.

Note: For actual EOC 13 condition monitoring SLB leak rate calculations, the new leak rate methods proposed in DCL-05-018 and as supplemented in this letter are not applicable. The EOC 13 condition monitoring leak rate calculations will use the existing DENTFLO leak rate methods.

2

Enclosure 1 PG&E Letter DCL-05-090 EOC 13 Projected Operational Assessment Leak Rates (gpm at Room Temperature) - Comparison of Existing W* Leak Model and Proposed W* Leak Model 95 percent Prediction 95 percent DENTFLO Bound - Severed Prediction model - Constrained Tube Bound -

Existing Crack Leak Model Constrained W* Leak Model for Assuming Total Crack Leak Method for Detected 360 Constrained Model for Detected Indications Degree Crack and SG Non-Axial in Flexible Tube Severed detected PWSCC W* Length Severance Tube Indications Indications and at 12 Models Between 8 in Flexible between inches and 12 W* Length Flexible W* Below inches Length and TTS Below TTS TTS - 12 inches 11 0.028 0.011 0.040 0.288 0.339 12 0.042 0.244 0.040 0.280 0.564 13 0.030 0.104 -0.040 0.298 0.441 14 0.006 0.031 0.040 0.290 0.360 21 0.289 1.675 0.040 0.294 2.009 22 0.140 0.633 0.040 0.283 0.956 23 0.604 1.938 0.040 0.294 2.271 24 0.484 1.702 0.040 0.275 2.016 NRC Question 2:

In Section 4.1.2.1, you indicate the 95th percentile F value is 2.221 given 2 regression coefficients and 36 data pairs. Please confirm this value and provide a reference.

PG&E Response:

As discussed in response to NRC question 1, PG&E is no longer applying the F-distribution, so the following clarification is provided for information only.

In Section 4.1.2.1 of DCL-05-018, the statement that reads "95th percentile value of Ffor 2 regression ..." would have been clearer if it read "95 th percentile value for the term containing F for 2 regression coefficients and 36 data pairs is 2.221." The value of 3

Enclosure I PG&E Letter DCL-05-090 2.221 is obtained from the F-distribution as, (2 _,'2)y =(2.2.466)y -2.221, (2) where 0.10 is the fraction of the population in two tails for a 95 percent one-sided confidence bound, 2 is number of parameters in the regression equation, and 36 is the number of data pairs. The actual value of F used was 2.466. This is easily verified by interpolation from standard tables of the F-distribution, although, depending on the author, the table may have to be entered using the cumulative value of 0.90 instead of the tail value of 0.10. See Chapter 3 of the reference book by Miller, R. G.,

"Simultaneous Statistical Inference," Second Edition, Springer-Verlag, New York, New York, dated 1981.

NRC Question 3:

Figure 1 in Enclosure 5 shows the leak rate versus the contact pressure from the constrainedcrack specimens. Although one relationship was developed for all the data, it appears there may be two populations of data (e.g., less than 1200 psi contact pressure, greater than 1200 psi contact pressure). Please provide an analysis which demonstrates there is not two or more populations of data. Alternatively, this analysis need not be provided if it can be shown the leak rates developed from the existing treatment of the data is more conservative than leak rates developed using more than one data population.

PG&E Response:

This question requests that the data be postulated to consist of independent sets drawn from two, or more, separate populations and that a statistical test be performed to examine whether or not such a postulate is warranted. While a statistical analysis cannot prove that the data are from the same population, the hypothesis that they are from the same population (the null hypothesis) can be tested at a specified level of confidence. An analysis was performed, which resulted in the conclusion that the null hypothesis that the two sets of data are from the same population is not rejected at a 5 or 10 percent probability level. Furthermore, an evaluation of the predicted leak rates indicates that it is conservative to treat the data as a single date set. Therefore, PG&E will treat the data as one set for determining the 95 percent prediction bound leak rates.

A detailed discussion follows.

The segregation of the data for the comparison requested was set at 1332 psi. This is midway between the contact pressures of the two data points adjacent to the selection.

The data and individual regression lines are illustrated on Figure 2 contained in . Since the data are being arbitrarily separated by the contact pressure, it is not appropriate to make a comparison of the intercepts without adjusting the pressure data to account for the segregation. Once this is done, the two regression lines intercept the ordinate located at the separation value with no meaningful difference, as seen on Figure 2. The shift in the origin along the abscissa has no effect on the slopes 4

Enclosure 1 PG&E Letter DCL-05-090 of the individual regression lines. A comparison of the regressions was performed by comparing two models, the first that the slopes and intercepts are the same and the second that the intercepts are the same and the slopes are different. These two models are the null and alternative hypotheses, respectively. The statistical comparison is made based on the decrease in the error sum of squares brought about by using the alternative hypothesis model and considering the degrees of freedom lost by increasing the complexity of the analysis based on two data sets. The resulting F-statistic is 0.956 with a cumulative probability of occurrence of 33.5 percent, leading to the conclusion that the probability is 66.5 percent that the two data sets are from different populations.

If the probability of occurrence of the F-statistic was on the order of 10 percent, it would be assumed that the improvement was borderline meaningful, and for a probability of 5 percent or less the improvement would be considered to be significant and the data should be treated as coming from two different populations. There is no statistical basis to reject the null hypothesis model in favor of the alternate hypothesis model (i.e., the models are not statistically different at a high level of confidence). As a check on the result, a second order fit of the data was performed with the result that there was no meaningful improvement in the model by adding a polynomial term, as seen in Figure 3 contained in Enclosure 4.

A scatter plot of the residuals from the regression analysis is presented in Figure 4 in , and the results support the observation that the data may be from separate underlying populations. This information is not conclusive and is not confirmed by the numerical analysis described later. A normal probability plot of the residuals is shown on Figure 5 in Enclosure 4. An apparent outlier is indicated at the lower left of the plot.

It is not a conclusive observation and physical information is not available to make a determination that it is indeed an outlier. Therefore, the data must be retained in the regression analysis. However, the effect of removing the data point is examined further as described later.

Comparison of the Individual Data Sets The variance of the individual sets of data about their mean values (i.e., without consideration of performing the regression analyses), indicates a cumulative probability of their having been drawn from the same population of about 11 percent. Although not very large, this number is too big to reject the null hypothesis that they were drawn from the same population. Normally, the rejection criterion would be set at about 5 percent, but could be set as high as 10 percent. This comparison ignores the reduction of the variance of leak rate predictions that can be achieved by performing the regression analysis or analyses and is a conservative screening evaluation.

Comparison of the Individual Regression Line Results A comparison of the variances about the individual regression lines indicates they are not statistically significantly different. The probability that they are from different populations is on the order of 82 percent. However, the criterion to reject the null hypothesis that they are from the same population would normally be on the order of 5

Enclosure 1 PG&E Letter DCL-05-090 95 percent, with 90 percent being considered as a warning level that perhaps more analysis should be performed. It is noted that this result is due to the influence of a single potentially outlying result from the lower contact pressure tests, evident on Figures 1, 2, and 3 of Enclosure 4 at a contact pressure of about 600 psi. If the suspected outlier is omitted, the likelihood that the data were drawn from the same population is effectively zero.

Consideration of Separate Regression Lines A plot of the 95 percent prediction bounds based on consideration of the individual regression lines is provided on Figure 6 of Enclosure 4. Additional 95 percent prediction curves are illustrated on Figure 7 of Enclosure 4 based on omitting the potentially outlying data point from the analysis of the small contact pressure test results.

Conclusions

• If the apparent outlying data point is retained in the analysis, the variances of the segregated data sets are not confirmed to be different at a 95 percent level of confidence. The F-statistic is 1.61 for 17 and 15 degrees of freedom respectively.

The associated probability that the data sets are from the same population is 18 percent, and the probability that they are from different populations is 82 percent (i.e., the variance of the second data set is greater than that of the first).

• A comparison of the prediction bounds associated with the data sets being from different populations (even though this is not indicated by the data) is illustrated on Figure 6 of Enclosure 4. The prediction bound leak rates are not meaningfully different from those considering a single combined data set.

• If the outlying data point is omitted from the analysis, the F-statistic is on the order of 13 for 17 and 14 degrees of freedom respectively. The associated probability that the data sets are from different populations is effectively 100 percent.

• A comparison of the prediction bounds associated with the data sets being segregated based on omitting the potential outlying data point is illustrated on Figure 7 of Enclosure 4. There are distinct differences in the leak rate predictions between the merged data set and the segregated data sets. The segregated data set leak rate predictions are contrary to engineering expectation, in that indications with high contact pressures could have higher leak rates than indications with lower contact pressures. The merged data set provides more conservative leak rates for indications with low contact pressures (near the TTS). Since the majority of indications are located near the TTS with attendant lower contact pressures, it can be expected that the use of a merged data set would be conservative relative to consideration of the segregated data sets for total SG leak rate analyses. In addition, since the exclusion of the potential outlying data point cannot be justified and overall leak rates could be higher (i.e., the single regression prediction curve in 6

Enclosure 1 PG&E Letter DCL-05-090 Figure 7 of Enclosure 4, which includes the outlier, is considerably higher at low contact pressures than the segregated regression prediction curve without the outlier) when including this data point, the omission of the data should not be further considered.

Finally, one additional conservative element to note in the evaluation is that the leak rate data are from specimens for which the initial contact pressure is constant over the length of the crack. In the SG, the contact pressure is a linearly increasing function of depth from the tip of the crack. This means that the leak rate associated with the data prediction is conservative relative to that from an analysis that would account for the increase in contact pressure along the length of the crack.

NRC Question 4:

You indicated that two of the leak rates from Table 6.3-3 in WCAP-14797-P were determined to be lower than the value shown since fluid collection times were actually greater than was listed in the data summary. You also indicated that the results were rounded and that for specimens reported with a zero contact pressure there was a gap between the tube and tubesheet, therefore a negative contact pressure was calculated.

Please provide an updated Table 6.3-3. In addition, please discuss how the negative contact pressures were determined and the basis for the adjustments.

PG&E Response:

The leak rate values are shown on Figure 1 contained in Enclosure 4, and can be used to update the information in Table 6.3-3 of WCAP-14797-P Revision 2. The three values identified with a zero contact pressure in the WCAP table but depicted with negative contact pressures have their effective contact pressure values listed on the figure. The effective negative contact pressures, P,, are calculated from the reported gap values, A,as, IC r.r E,,'7 iEt

+VJ

-A

+-~ C

'2 22 i,,-

(3) where the subscripts c and t indicate the tube and tubesheet collar respectively, subscripts o and i are for the outside and inside radii respectively, r is the radius, and E is the elastic modulus. Note that when there is a gap, the term involving the tubesheet collar would not normally be employed, however, in order to retain continuity of the prediction equation it was retained for this analysis. The net effect of including the collar in the analysis is small, about 20 percent of the resulting value.

7

Enclosure 1 PG&E Letter DCL-05-090 NRC Question 5:

Section 4.1.2.5 states that if greater than or equal to 75 percent of the length of an axial indication is below the BWT, the indication will be considered as constrained and 100 percent through-wall, and a 95 percent confidence bound leak rate for zero contact pressure will be assigned using the constrained crack model. Given that tubesheet dilation resulting from a main steam line break could produce a gap between the tube and tubesheet, the basis for the model is not clear.

Please provide the technical basis for this approach. Include in this basis a discussion of (1) the effects of tubesheet bow on the length of crack that could be non-constrained and compare this to the constrained crack test data; and (2)any data supporting the leak rates that could be experienced given the gaps between the tube and tubesheet during a steam line break for these non-constrained crack lengths. With respect to degradation near the BWT or top-of-tubesheet, discuss the effects of severe accident temperatures and pressures on leakage integrity. Include in your response a discussion of potential degradation of the joint or tubesheet related to ajet emanating from a non-constrained crack. Alternatively, modify your technical specifications to indicate that all flaws within the tubesheet that will not be constrained during postulated accidents will be plugged on detection. (The determination of whether a flaw would be non-constrained should address both NDE uncertainty and crack growth during the operating cycle.) The reporting requirements would also need to be modified to reflect this approach.

PG&E Response:

Section 4.1.2.5 of DCL-05-018 provided methods considerations for evaluating primary water stress corrosion cracking (PWSCC) indications near or above the BWT, including a discussion of the 75 percent criterion. Based on PG&E's commitments to revise the TS as discussed below, Section 4.1.2.5 is no longer applicable for LAR 05-01.

Nonetheless, a basis for the 75 percent criterion is provided below for completeness.

The technical basis for the 75 percent length below the BWT is that crack opening will be limited by the tubesheet if the center or slightly beyond the center of the crack is within the fully expanded tube length. The leakage model application of zero contact pressure implies no contact between the tube and the tubesheet. From Figure 1 of of DCL-05-018, the leak rate for constrained cracks near zero contact pressure has a very small slope indicating that the constrained crack leakage is not sensitive to no contact pressure or small gaps. The test data include a few data points with negative contact pressures or small gaps and these test data do not indicate a significant difference in leakage between small gaps (negative contact pressure in Figure 1 of Enclosure 4) and zero contact pressure. Under SLB conditions, the tube-to-tubesheet diametral gap near the TTS is only about one millimeter (mil). This small gap does not permit significant crack opening so that the constrained crack model is applicable and tubesheet bow does not affect the length of the crack that could be nonconstrained for cracks dominantly (75 percent criterion applied) within the fully 8

Enclosure 1 PG&E Letter DCL-05-090 expanded region of the tubesheet. It is concluded that the zero contact pressure assumption is a reasonable approximation for crack opening limited by the fully expanded tube length.

Due to the higher temperatures during a postulated severe accident as compared to a SLB event, the tube to tubesheet contact pressures in the fully expanded tube length can be expected to increase and any gaps near the BWT would be less than the one mil estimated for the SLB event. The tubesheet constraint in a severe accident limits any tube deformation that might occur, compared to freespan cracks, due to the high temperature affects on material properties. Consequently, severe accidents would not be expected to impact leakage integrity for cracks dominantly within the fully expanded tube length. Similarly, no tubesheet gap or a small tubesheet gap does not permit jets emanating from the crack and thus jets are not an issue for cracks dominantly within the fully expanded tube length.

Per existing TS 5.5.9.d.l.k (v).2, any indication found above the BWT, including nondestructive examination (NDE) uncertainty, or predicted to project above the TTS in the next cycle, including NDE uncertainty and crack growth allowance, are repaired.

Due to NRC concerns of severe accident considerations and to provide further margin, the TS W* repair limits will be revised (reference Section 4.2.1 of DCL-05-018).

TS 5.5.9.d.1.k (v).2 will be revised to require that any indication found above the bottom of the BWT, including NDE uncertainty, or predicted to project above the BWT in the next cycle, including 95 percent confidence NDE uncertainty and crack growth allowance, are repaired. The TS 90-day reporting requirements (reference Section 4.1.6 of DCL-05-018) and W* performance criterion (reference Section 4.2.5 of DCL-05-018) also require revision due to this change. The TS 5.6.10.e 90-day reporting will be revised to require verification that the upper crack tip of W* indications returned to service in the prior cycle remain below the BWT by at least the 95 percent confidence NDE uncertainty on locating the crack tip relative to the BWT. The W*

performance criterion will be revised to require NRC notification prior to returning the SGs to service if condition monitoring determines that the upper crack tip of W*

indications returned to service in the prior cycle do not remain below the BWT by at least the 95 percent confidence NDE uncertainty on locating the crack tip relative to the BWT. These changes are more conservative than existing requirements, and provide added assurance that cracks returned to service will remain constrained at the end of the subsequent operating cycle.

The marked-up TS pages in Enclosure 2 and the retyped TS pages in Enclosure 3 are revised to include this TS revision.

For end of cycle (EOC) 13 reporting, the new proposed 90-day reporting requirement and performance criterion reporting requirement are not applicable because the new proposed (more conservative) repair criterion was not in effect in cycle 13. For example, several W* indications that were returned to service in the current Unit 2 Cycle 13 would have been required to be plugged using the new criterion. Therefore, 9

Enclosure 1 PG&E Letter DCL-05-090 for EOC 13 reporting, PG&E will retain the existing requirements, that is, NRC notification prior to returning the SGs to service if condition monitoring determines that the upper crack tip of W* indications returned to service in the prior cycle do not remain below the TTS by at least the 95 percent confidence NDE uncertainty on locating the crack tip relative to the TTS. The 90-day report will also provide written verification that this condition was satisfied.

Since the start of W* ARC at DCPP Units I and 2, due to very small growth rates, only one W* indication that was left in service has been found to be above BWT, including NDE uncertainty, at the end of the next cycle. As discussed in the DCPP Unit 2 Refueling Outage 11 (2R1 1) 90-day report (Enclosure I of PG&E Letter DCL-03-076),

this indication (SG 2-3 indication R1 7C72 in 2R1 1) was located entirely below BWT.

Because of 0.28 inch NDE uncertainty for locating the crack tip relative to the BWT, the crack tip was slightly above BWT and, therefore, the tube was plugged. To avoid confusion in light of indication RI 7C72, the statement on page 29 of DCL-05-018 that

'No axial PWSCC indications have been left in service that, in subsequent inspections, failed the W* ARC ..." should be restated as: "No axial PWSCC indications have been left in service that, in subsequent inspections, failed the W* ARC performance criteria, that is, had the UCT extend above the TTS (allowing for NDE uncertainty), thereby meeting the performance criteria and validating the slow growth rate of axial PWSCC."

NRC Question 6:

Section 4.14 compares the leak rates obtained with the proposed methodology (based on constrained crack testing) to the existing W*leak methodology (based on the DENTFLO model). In general, the constrained crack methodology calculates higher crack leak rates than the DENTFLO model. In SG 1-1, however, the constrained crack leak rate is less than the DENTFLO leak rate due to an indication near the BWT that accounts for 95 percent of the DENTFLO leakage but has negligible leakage using the constrained crack methodology. This was attributed to a Row 3, Column 2 indication that had negligible constrained crack model leakage using the individual tube contact pressure. If the contact pressure for this particular tube were used in a DENTFLO calculation, would the DENTFLO leakage model calculate less leakage for this indication relative to the constrained crack leak rate m6thodology? If not, explain why the DENTFLO model leakage is higher than the constrained crack model leakage.

PG&E Response:

The Row 3, Column 2 indication is located with its upper crack tip 0.84 inch below the TTS, after accounting for 0.22 inch NDE uncertainty. The length was reported to be 0.38 inch. The contact pressure for that tube at that elevation is on the order of 2130 to 2250 psi. From Figure 6.4-2 of WCAP-14797-P Revision 2, the 95 percent confidence loss coefficient value for this contact pressure is conservatively bounded by 1.0x1 014 in4, with an attendant DENTFLO leak rate calculated to be on the order of 1.83x1 04 gpm. This DENTFLO leak rate is about one order of magnitude less than the 10

Enclosure 1 PG&E Letter DCL-05-090 95 percent confidence value of 3.6x1 0-3 gpm calculated from the constrained crack model.

NRC Question 7:

The tube inspection definition provided in the Technical Specifications 5.5.9.d. 1.i does not appear to be consistent with implementation of a W*criteria within the hot leg tubesheet. Please clarify (e.g., the inspection distance will be the greater of 8 inches below the top-of-tubesheet or the distance needed to satisfy a W*(or flexible W*) length below the BWT). Any revisions to the tube inspection definition should be consistent with all aspects of your structural and leakage integrity models. The staffs concern is that the existing wording implies that probes capable of finding cracks are used throughout the tubesheet.

PG&E Response:

Per TS 5.5.9.d.1 .h, "Tube Inspection means an inspection of the SG tube from the tube end (hot leg side) completely around the U-bend to the top support of the cold leg."

However, W* criteria in TS 5.5.9.d.1.k (iii) states uThe Flexible W* Length is the total RPC-inspected length as measured downward from the BWT, and includes NDE uncertainties and crack lengths within W* as adjusted for growth." Therefore, requirements to limit the tubesheet inspection to the flexible W* length are explicitly defined in TS 5.5.9.d.1.k (iii). As such, no changes to TS 5.5.9.d.1.h have ever been proposed by PG&E nor the NRC in prior licensing of W* ARC at DCPP Units 1 and 2.

In LAR 05-01, PG&E committed to apply a minimum rotating pancake coil (RPC) inspection extent of 8 inches in the hot leg, and 8.5 inches in the cold leg should cold leg inspections be required. To address the staffs concern that the existing TS wording implies that probes capable of finding cracks are used throughout the tubesheet, PG&E's will revise TS 5.5.9.d.1.h to state: 'Tube Inspection means an inspection of the SG tube from the tube end (hot leg side) completely around the U-bend to the top support of the cold leg, excluding the portion of the tube within the tubesheet below the Flexible W* Length or below 8 inches from the hot leg top of tubesheet, whichever is bounding." This TS revision clarifies that probes capable of finding cracks (e.g., rotating probes and array probes) are limited to the portion of the tubesheet region included in the Tube Inspection definition. The marked-up TS pages in Enclosure 2 and the retyped TS pages in Enclosure 3 are revised to include this TS revision.

NRC Question 8:

Clarify if the W*Length definition provided in Specification 5.5.9.d. 1.k (ii) should read "W*Length is the distance in the tubesheet below the BWK "

11

Enclosure 1 PG&E Letter DCL-05-090 PG&E Response:

TS 5.5.9.d.1 .k (ii) which states: 'W* Length is the distance to the tubesheet below the BWT..." should state: "The W* Length is the distance in the tubesheet below the BWT..." In order to provide clarification of the W* Length, TS 5.5.9.d.1 .k (ii) will be revised to state: "The W* Length is the distance in the tubesheet below the BWT..."

The marked-up TS pages in Enclosure 2 and the retyped TS pages in Enclosure 3 are revised to include this TS revision.

NRC Question 9:

Please confirm that the leakage model of Section 4.1.2.1 would be applied to all detected degradation below the BWT to 12 inches below the TTS.

PG&E Response:

The constrained crack leakage model of Section 4.1.2.1 of DCL-05-018 would be applied to all detected degradation below the BWT to 12 inches below the TTS.

NRC Question 10:

Clarify, (Specification 5.5.9.d. 1.k (iv)) if all tubes containing degradation within or below the W* length that are left in service and degraded within the limits specified in Specification 5.5.9.d. 1.k (v) are considered W* tubes. If not, please describe the methodology used to determine the percent through-wall for indications within or below the W* length. Please modify the specification as appropriate.

PG&E Response:

TS 5.5.9.d.1.k (iv) states: `W* Tube is a tube with equal to or greater than 40%

degradation within or below the W* length that is left in service, and degraded within the limits specified in Specification 5.5.9d.1.k)(v)." This means that all tubes containing degradation (as confirmed by qualified probes) within or below the W* length that are left in service under W* ARC are considered W* tubes. Determination of percent through-wall for indications within or below the W* length is not necessary, as PG&E assumes any degradation confirmed by RPC or equivalent is greater than 40 percent through-wall. For clarity, PG&E will revise TS 5.5.9.d.1.k (iv) by deleting the words "equal to or greater than 40%". Also, an editorial change is made to revise the reference to TS "5.5.9d.1.k)(v)" to TS "5.5.9.d.1.k (v)".to be consistent with the format used for other current TS references to TS 5.5.9 subsections. The revised TS 5.5.9.d.1.k (iv) will state: UW* Tube is a tube with degradation within or below the W* length that is left in service, and degraded within the limits specified in Specification 5.5.9.d.1.k (v)." This definition assumes that degradation is confirmed by RPC or equivalent probe. The marked-up TS pages in Enclosure 2 and the retyped TS pages in Enclosure 3 are revised to include this TS revision.

12

Enclosure 1 PG&E Letter DCL-05-090 NRC Question 11:

On page 12, you develop a relationship for determining the contact pressure as a function of tubesheet radius (equations 5 and 6). Please discuss the uncertainties in these relationships and their effects on the leakage estimate if a 95% prediction interval curve was used to determine the contact pressure.

PG&E Response:

There are effectively no uncertainties in the relationships. In Figure 5 in Enclosure 5 of DCL-05-018, plots are linear except for two points for zone B1 at depths less than one inch from the TTS. The index of determination, R2, values are unity to four decimal places, indicating that the errors of the regression are negligible. For example, the standard error of the intercept is about 0.6 for values ranging from -754 psi to 2275 psi.

The corresponding standard error of the slope prediction is 0.48 for slope values ranging from 414 psi/inch to 133 psi/inch. Therefore, the effect of including uncertainties is expected to be insignificant.

NRC Question 12:

On page 13, it was indicated that Figure I shows the leak rate data as a function of

'Yotal" contact pressure. Please clarify what is meant by "total" contact pressure.

Please clarify whether the same leakage value (0.0028 gpm) is obtained when evaluating the 95% confidence bound on the data in Figure 4 at an 8 inch distance below TTS. Please explain any differences.

PG&E Response:

The discussion in Section 4.1.2.2 of DCL-05-018 was intended to draw attention to the difference between the reporting protocol for the crevice test results and for the constrained crack test results. The data in Table 6.2-3 of WCAP-14797-P, Revision 2,

'Generic W* Tube Plugging Criteria for 51 Series Steam Generator Tubesheet Region WEXTEX Expansions," dated March 2003 (crevice leak test results), do not include the installation residual contact pressure in the reported values and are not the "total" contact pressures. There was no residual installation contact pressure associated with the test data in Table 6.3-3 of WCAP-14797-P, Revision 2, (constrained crack leak test results), and the reported values are the "total" contact pressures.

Figure 4 of Enclosure 5 of DCL-05-018 was developed for illustrative purposes only using the depth and Zone B contact pressure information from Table 4.3-9 of WCAP-14797-P, Revision 2, which was originally calculated solely for W* pullout length determinations for bounding faulted conditions at 4600F. Figure 4 of Enclosure 5 of DCL-05-018 was developed from a linear transformation of the contact pressure scale of Figure 1 of Enclosure 5 of DCL-05-018 to depth-below-the-TTS scale using a regression equation from the pullout data table.

13

Enclosure 1 PG&E Letter DCL-05-090 Figure 8 of Enclosure 4 of this response presents a re-plot of the data using the Zone B1 contact pressure data developed for the SLB leak rate analysis at 6000 F and reported in Table 4.3-11 of WCAP-14797-P, Revision 2. Figure 1 of Enclosure 4 of this response is a duplicate plot of Figure 1 of Enclosure 5 of DCL-05-018 with respect to the 95 percent simultaneous confidence curve. Therefore, comparisons between Figure 1 and Figure 8, both of Enclosure 4, of this response are valid. For example, from Figure 8 of Enclosure 4 of this response, at a depth of 8 inches, the 95 percent simultaneous confidence limit leak rate is 2.8x1 0-3 gpm, and the 95 percent one-sided prediction bound value is 3.3x1 0,3 gpm. The corresponding Zone B1 total contact pressure is 2557 psi from Figure 2 of Enclosure 5 of DCL-05-018. From Figure 1 of Enclosure 4 of this response, at these leak rates, the total contact pressure is also 2557 psi.

In Table 4.3-11 of WCAP-1 4797-P Revision 2, there is a typographical error for the contact pressure for Zone B4 at 2 inches below the TTS. The negative contact pressure listed should be a positive contact pressure. This has no affect on the leak rate assessment, as the corrected contact pressure was plotted in Figure 5 of Enclosure 5 to DCL-05-018.

Considerations of Temperature and Pressure for W* SLB Leakage Analysis SLB leak rates have been traditionally evaluated at a reference set of conditions corresponding to a pressure differential of 2405 psi (based on power-operated relief valve setpoint) or 2560 psi (based on safety valve setpoint) and a temperature of 6000F.

The reference conditions provide a consistent analysis basis with a bounding pressure differential and avoid time dependent analyses for varying temperature effects, which generally have a modest influence on leakage. The pressure differential is the maximum pressure differential at a later time in the event (4200 seconds per Table 3.2-1 of WCAP-14797-P, Revision 2, for a four loop Model 51 SG plant). The 6000 F temperature is representative of temperatures in the early part of the event and also has been a common temperature for leak rate measurements so that the leak rate database supports the analysis. During a SLB event, the primary to secondary pressure differential and temperature are time dependent, and the longer term values are dependent upon the time required for plant operator actions. From Table 3.2-1 of WCAP-14797-P, Revision 2, the SLB temperature at 4200 seconds is about 460 0F.

In PG&E Letter DCL-03-139, "PG&E Response to NRC Questions on 2R11 Steam Generator Tube Inspections," dated October 31, 2003, it has been shown that modest differences in pressure for a constant contact pressure have negligible influence on W*

leak rates. Per DCL-03-139, the data presented in Table 6.2-2 of WCAP-14797-P, Revision 2, clearly show that there is no trend for the leak rate to decrease as a function of the increase in internal pressure in the tube in tests ranging from 1620 to 2650 psi.

These results would indicate that the increase in leakage with increasing pressure differential across the crack face is more influential on W* leak rate than the associated increase in contact pressure between the tube and tubesheet. Moreover, a 14

Enclosure 1 PG&E Letter DCL-05-090 specimen-by-specimen comparison of the individual data demonstrates a marked reduction of leak rate with an increase in temperature from 700F to 6000F.

This assessment therefore focuses on effects of a 4600F temperature. As noted in Section 4.3.3 of the First Energy Nuclear Operating Company (FENOC) letter to the NRC, L-04-089, "Beaver Valley Power Station, Unit No. 1, Docket No. 50-334, License No. DPR-66, License Amendment Request No. 328, Revised Steam Generator Inspection Scope for One Cycle of Operation," dated June 28, 2004, the leakage hydraulic effects of only a change in temperature from 6000F to 4600F would be negligible. Flashing would occur under both conditions near or at the exit of the crevice between the tube and the tubesheet or the exit of the crack flanks. Since viscosity increases at lower temperatures, this effect would tend to lower the leak rate for lower temperatures. For W* considerations, the primary influence of lowering the temperature for a given pressure differential is a decrease in contact pressure. Comparisons of leak rates calculated using contact pressure data from WCAP-14797-P, Revision 2, Table 4.3-11 (6000F), as shown in Figure 8 of Enclosure 4 of this letter, and WCAP-14797-P, Revision 2, Table 4.3-9 (460 0F), as shown in Figure 4 of Enclosure 5 of DCL-05-018, show that the lower contact pressures at 4600 F lead to about a factor of two increase in the leak rate. However, this modest effect is more than offset by the conservatisms in the W* analysis as discussed below.

W* leak rates can be substantially reduced by increasing the complexity of the leak rate analysis. Figure 2 of the FENOC letter to the NRC, L-04-121, "Beaver Valley Power Station, Unit No. 1, Docket No. 50-334, License No. DPR-66, Response to Request for Additional Information in Support of License Amendment Request No. 328, Revised Steam Generator Inspection Scope for One Cycle of Operation," dated September 3, 2004, shows that correlating all the crevice leak rate data with both crevice depth and contact pressure decreases the leak rates by approximately an order of magnitude compared to correlating only with crevice depth. The resulting leak rates from the crevice depth/ contact pressure crevice model in Figure 2 of FENOC Letter L-04-121 are more than two orders of magnitude less than the PG&E constrained crack leak rates from Figure 8 of Enclosure 4 of this letter. For example, the median leak rate corresponding to Figure 2 of FENOC Letter L-04-121 at 2 inches is 1x104 gpm. From Figure 8 of Enclosure 4 of this letter, the median leak rate at 2 inches is 4x1 03 gpm, which supports the more than 2 orders of magnitude difference. Thus applying a slightly more complex model and using the crevice leak rate data for W* rather than the constrained crack model would lead to much larger leak rate reductions than the factor of two potential sensitivity to temperature. The complexity of the analysis could be further increased by considering the constrained crack and crevice leak rate correlations as resistances in series to account for both effects. This would combine the true crevice effects but avoid use of the DENTFLO model for combining the effects.

Overall, it is concluded that the use of 6000F for the W* analysis with the constrained crack model includes quantifiable conservatisms much greater than the more modest 15

Enclosure 1 PG&E Letter DCL-05-090 potential effect of a factor of two change in leak rate due to the lower contact pressures at lower temperatures.

NRC Question 13:

Page 14 indicates that the resistance to flow from the crevice is significantly greater than that of the restrained crack. However, evaluation of the "crevice data" at 8 inches into the tubesheet results in a leak rate of approximately 0.0045 gpm whereas an evaluation using the "constrained crack data" for this same location results in a leak rate of 0.0028 gpm. Please clarify this apparent difference in trends.

PG&E Response:

Section 4.1.2.3 of DCL-05-018 notes that the leak rate from the crevice tests is for an entrance length equivalent to the circumference of the expanded tube (about 2.8 inches). The entrance length for the constrained crack tests was 0.33 to 0.59 inch long. The leak rates are similar while the entrance length for the crevice tests is about 7 times that of the constrained crack tests. Figure 6 of Enclosure 5 of DCL-05-018 was included to illustrate the approximate equivalence of the test results. Since the crevice lengths in the SG would be significantly greater in practice than those in the test specimens, it is expected there will be significantly less leakage than is being used for the analysis and that both approaches to calculating the leak rate are conservative. It is also noted that the information on Figure 5 of Enclosure 5 of DCL-05-018 for the contact pressures as a function of depth indicates that the test results from the 2 inch nominal (1.28 inch actual) engagement specimens (2020 psi on the figure) would be expected to be more accurate than using the 1.25 inch nominal (0.61 inch actual) engagement length specimen data for the prediction of the leak rates. If the 2 inch nominal specimens had been used in the analysis, the predicted leak rate from the model based on the crevice test data would be reduced by a factor of 5. In summary, the resistance of the actual crevice in the SG would be expected to be significantly greater than that of the constrained crack.

NRC Question 14:

Pages 16 and 17 discuss a methodology for projecting the number of indications located 8 to 12 inches below the TTS (i.e., the more conservative of 2 methods is used to project the number of indications). This methodology is based on historical data.

Please discuss whether your proposed requirement to assess whether the results were consistent with expectations (Insert A for TS Page 5.0-30) includes describing the corrective actions should the number of indications detected during an outage be greater than the number of indications projected at the end of the previous cycle. If such an assessment is not included in this proposed requirement, please modify the requirement to include it.

16

Enclosure I PG&E Letter DCL-05-090 PG&E Response:

The proposed Insert A for TS Page 5.0-30 states that the 90-day report shall include "the cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet" and an "assessment of whether the results were consistent with expectations and, if not consistent, a description of the proposed corrective action.' By providing the cumulative number of indications detected in the tubesheet region as a function of elevation, the number of indications located 8 to 12 inches below the TTS that were detected during the outage would also be provided, along with corrective actions if the projections of the number of indications located 8 to 12 inches below the TTS were non-conservative. Since the proposed TS already require corrective actions to be provided in the 90-day report if the results are not consistent with expectations, no TS change is required.

17

Enclosure 1 PG&E Letter DCL-05-090 Additional Miscellaneous TS Changes In addition to the proposed TS changes described in PG&E's response to questions 1, 5, 7, 8, and 10, the following miscellaneous changes are being proposed to provide further clarity. The marked-up TS pages in Enclosure 2 and the retyped TS pages in are revised to include these TS changes.

TS 5.5.9.b.2.e has been revised to add the word "flexible" and "or equivalent," such that the revised TS would state: "Tubes identified as W*tubes having a previously identified indication within the flexible W* length shall be inspected using a rotating pancake coil (RPC) probe or equivalent for the full length of the W* region during all future refueling outages." Addition of "flexible" provides a more conservative TS. Addition of "or equivalent" allows for use of other qualified probe technology.

TS 5.5.9.d.1 .k (v).6 has been revised to add the word "flexible," change "of" to "or," and delete "the," such that the revised TS would state: "Any type or combination of tube degradation below the flexible W* length is acceptable." Addition of "flexible" provides a more conservative specification.

TS 5.6.10.d.2 (used to be TS 5.6.10.d.3) has been revised to add the word "ODSCC."

The revised TS would require (for implementation of voltage-based repair criteria) NRC notification prior to returning the SGs to service "if ODSCC indications are identified that extend beyond the confines of the tube support plate." Addition of "ODSCC" clarifies that this reporting requirement only pertains to ODSCC indications that are applicable to voltage-based repair criteria.

The TS 5.6.1 0.e 90-day reporting requirements have been revised to require, for tubes with detected degradation, identification of the radial position of the tube within the tubesheet and the W* Zone of the tube. Radial position of the tube is a necessary input in determining the leak rate using the new proposed leakage method.

18

Enclosure 2 PG&E Letter DCL-05-090 Proposed Technical Specification Changes (marked-up)

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Tube Surveillance Proqram SG tube integrity shall be demonstrated by performance of the following augmented inservice inspection program.

The provisions of SR 3.0.2 are applicable to the SG Tube Surveillance Program test frequencies.

a. SG Sample Selection and Inspection - SG tube integrity shall be determined during shutdown by selecting and inspecting at least the minimum number of SGs specified in Table 5.5.9-1.

b. SG Tube Sample Selection and Inspection - The SG tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 5.5.9-2. The inservice inspection of SG tubes shall be performed at the frequencies specified in Specification 5.5.9.c and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.5.9.d. The tubes selected for each inservice inspection shall include at least 3%

of the total number of tubes in all SGs; the tubes selected for these inspections shall be selected on a random basis except:

1. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

2. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each SG shall include:

a) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

b) Tubes in those areas where experience has indicated potential problems, c) A tube inspection (pursuant to Specification 5.5.9.d.1.h) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube quivalent inspection, this shall be recorded and an adjacent tube shall be lore elected and subjected to a tube inspection, d) Indions left in service as a result of application of the tube fIble support te voltage-based repair criteria shall be inspected by n cle be during all future refueling outagesHoz .- I e) Tubes iden' d a W*tubes having a previously identified L indication within

• length shall be inspected using a rotating pancake coil (RPC) prob for the full len th of the W* region during all future refueling outages.

(continued)

Applicable for Units 1 and 2, Cycles 10, 11, 12, and 13 only

• • In-Situ Testing will be performed in accordance with PG&E letters DCL 98-148 dated October 22, 1998, and DCL 01-052 dated May 4, 2001, for Cycles 10 and 11 and letter DCL 01-095 dated V September 13, 2001, for Cycles 12 and 13.

DIABLO CANYON - UNITS 1 & 2 5.0-10 Unit 1 - Amendment No. 43A, 464, TAB5.doc- R12 1 Unit 2 - Amendment No. 435, 45-1-,

Programs and Manuals Note: There are no Changes on this Page. Page included for Information Only 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

3. The tubes selected as the second and third samples (if required by Table 5.5.9-2) during each inservice inspection may be subjected to a partial tube inspection provided:

a) The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and b) The inspections include those portions of the tubes where imperfections were previously found.

4. Implementation of the steam generator tube/tube support plate repair criteria requires a 100% bobbin coil inspection for hot-leg and cold-leg support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersection having ODSCC indications shall be based on the performance of at least a 20% random sampling of tubes inspected over their full length.

5. Inspection of dented tube support plate intersections will be performed in accordance with WCAP-15573, Revision 1, to implement axial primary water stress corrosion cracking (PWSCC) depth-based repair criteria. The extent of required inspection is:

a) 100 percent bobbin coil inspection of all tube support plate (TSP) intersections.

b) Plus Point coil inspection of all bobbin coil indications at dented TSP intersections.

c) Plus Point coil inspection of all prior PWSCC indications left in service.

d) If bobbin coil is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all TSP intersections having greater than 2 volt dents up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20%

of greater than 2 volt dents at the next higher TSP. If a circumferential indication is detected in a dent of ux" volts in the prior two inspections or current inspection, Plus Point inspections will be conducted on 100% of dents greater than "x - 0.3" volts up to the affected TSP elevation in the affected SG, plus 20% of dents greater than "x - 0.3".volts at the next higher TSP. "x" is defined as the lowest dent voltage where a circumferential crack was detected.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-11 Unit 1 - Amendment No. 435,152 TAB5.doc - R1 2 2 Unit 2 - Amendment No. 435,152

Programs and Manuals 5.5 Note: There are no Changes on this Page. Page included for information Only 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued) e) If bobbin coil is not relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all dented TSP intersections (no lower dent voltage threshold) up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of all dents at the next higher TSP.

f) For any 20% dent sample, a minimum of 50 dents at the TSP elevation shall be inspected. If the population of dents is less than.

50 at the TSP elevation, then 100% of the dents at the TSP elevation shall be inspected.

The results of each sample inspection shall be classified into one of the following three categories:

Cateqory Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

c. Inspection Frequencies - The above required inservice inspections of SG tubes shall be performed at the following frequencies:

1. The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality.

Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; (continued)

DIABLO CANYON - UNITS 1 & 2 5.0-11 a Unit 1 - Amendment No. 435, 152 TAB5.doc- R12 3 Unit 2 - Amendment No. -35, 152

Programs and Manuals Note: There are no Changes on this Page. Page included for Information Onlyr 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2. If the results of the inservice inspection of a SG conducted in accordance with Table 5.5.9-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 5.5.9.c.1. The interval may then be extended to a maximum of once per 40 months; and

3. Additional, unscheduled inservice inspections shall be performed on each SG in accordance with the first sample inspection specified in Table 5.5.9-2 during the shutdown subsequent to any of the following conditions:

a) Reactor-to-secondary tube leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.13; or b) A seismic occurrence greater than the Double Design Earthquake, or c) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or d) A main steam line or feedwater line break.

d. Acceptance Criteria

1. As used in this Specification:

a) Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; b) Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube; c) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation; d)  % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

e) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; f) Plugging Limit means the imperfection depth at or beyond which the tube shall be removed from service and is equal to 40% of the nominal tube wall thickness.

1) This definition does not apply to tube support plate intersections for which the voltage-based repair criteria are being applied.

Refer to 5.5.9.d.1 .j for the repair limit applicable to these intersections.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-12 Unit 1 - Amendment No. 435 142 TAB5.doc -R12 4 Unit 2 - Amendment No. 435 142

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2) This definition does not apply to the portion of the tube within the tubesheet below the W* length. Acceptable tube wall degradation within the W* length shall be defined as in 5.5.9.d.1 .k. ( - l

3) This definition does not apply to axial PWSCC indications, or portions thereof, which are located within the thickness of dented tube support plates which exhibit a maximum depth greater than or equal to 40 percent of the initial tube wall thickness. WCAP-1 5573, Revision 1, provides repair limits applicable to these intersections.

4) A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be removed from service.

g) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of a Double Design Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 5.5.9.c.3, above; h) Tube Inspection means an inspection of the SG tube from the tube end (hot leg side) completely around the U-bend to the top support of the cold le i) Preservice Inspection means an inspection of the full length of each tube in each SG performed by eddy current techniques prior to service to Insert B establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial Power*

Operation using the equipment and techniques expected to be used during subsequent inservice inspections; j) Tube Support Plate Plugging Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging limit is based on maintaining steam generator tube serviceability as described below:

(i) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to the lower voltage repair limit (NOTE 1), will be allowed to remain in service.

(ii) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1), will be repaired or plugged, except as noted in 5.5.9.d.l j (iii) below. 9

,- -- '>(continued)

• AplicbleforUnis 1and2, ycls 1, 1, 1, ad 13 only.

DIABLO CANYON - UNITS 1 & 2 5.0-13 Unit 2 - Amendment No. 4635,151,452, TAB5.doc - R12 5Unit 2- Amendment No. 435,5,452,

Prnnrnrmq nrl M~ni inl Note: There are no Changes on this Pagehowever the requirement5.5.9.d.1j (ii) on Page 5.0-13 is rolled onto this Page due to Insert B on Page 5.0-13 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

(iii) Steam generator tubes, with indication of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1) but less than or equal to the upper voltage repair limit (NOTE 2), may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit (NOTE 2) will be plugged or repaired.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-13a Unit 1 - Amendment No. 435,151,452, TAB5.doc - R12 6 Unit 2 - Amendment No. 135,151,452,

Programs and Manuals

5.5 lNote

There are no Changes on this Page. Page included for Information Only 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance ProQram (continued)

(iv) Certain intersections as identified in PG&E Letter DCL-03-174, dated December 19, 2003, will be excluded from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA + SSE event.

(v) If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 5.5.9.d.1.j (i), 5.5.9.d.lj (ii), and 5.5.9.d.1.j (iii). The mid-cycle repair limits are determined from the following equations:

VSL VMURL =

(CL - At) l.0 +NDE +Gr CL VMLRL = VMURL - (VURIL - VLRL) (CL t) where:

VURL = upper voltage repair limit VLRL = lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMURL and time into cycle At = length of time since last scheduled inspection during which VURL and VLRL were implemented CL = cycle length (the time between two scheduled steam generator inspections)

VSL = structural limit voltage Gr = average growth rate per cycle length NDE = 95% cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20% has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 5.5.9.d.lj (i), 5.5.9.d.lj (ii), and 5.5.9.d.1lj (iii).

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-14 Unit 1 - Amendment No. 135, 176 TAB5.doc - R12 7 Unit 2 - Amendment No. 135, 178

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

NOTE 1: The lower voltage repair limit is 2.0 volts for 7/8 inch diameter tubing at DCPP Units 1 and 2.

NOTE 2: The upper voltage repair limit is calculated according to the

_ methodology in Generic Letter 95-05 as supplemented.

k Plini Limit is used for disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented inside diameter stress corrosion cracking confined within the tubesheet, below the bottom of the WEXTEX transition (BWT). As used in this specification:

(i) Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet as determined by eddy cur nt testing. i_

(ii) W* Length is the distan es ne Delow the BWT that precludes tube pull out in t~e event of the complete circumferential separation of the tube below the W* length. The W* length is conservatively set at: 1)an undegraded hot leg tube length of 5.2 P inches for Zone A tubes and 7.0 inches for Zone B tubes, and 2) an 2 undegraded cold leg tube length of 5.5 inches for Zone A tubes and

7. hes for Zone B tubes. Information provided in WCAP-14797 Revision 4, defines the boundaries of Zone A and Zone B.

(iii) Flexible W* Length is the W* length adjusted for any cracks found within the W* region. The Flexible W* Length is the total RPC-inspected length as measured downward from the BWT, and includes NDE uncertainties and crack lengths within W* as adjusted for growth.

(iv) W* Tube is a tube with equal o orgreatet han -4Ggradation within or below the W* length that is left in service, &id degraded within the limits specified in Specification 5.5.9d.1.V(v).

(v) Within the tubesheet, the plugging (repair) Ii it is based on maintaining steam generator serviceability a described below:

1) For tubes to which the W* criteria are a plied, the length of non-degraded tube below BWT shall be gr ater than or equal to the W* length plus NDE uncertainties and crack growth for the operating cycle.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-15 Unit 1 - Amendment No. i-a5, TAB5.doc- R12 8 Unit 2 - Amendment No. 435,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2) Axial cracks in tubes returned to service using W* shall have the upper crack tip-below the BWT by at least thie NFmnrm uncertaintnd below the top of tube st kNDE ~measureme~nt unc~ertaiin~tgand crack growth allowance, suchI that at the end of the subsequent operating I the entire crack remains below t ushZct zccondarth ic

3) Resoll e,single axial indications (multiple indications must urn to the null point between individual cracks) within the BWT ~flexible W* length can be left in service. Alternate RPC coils or an ultrasonic test (UT) inspection can be used to demonstrate return to null point between multiple axial indications or the absence of circumferential involvement between axial indications.

4) Tubes with inclined axial indications less than 2.0 inches long (including the crack growth allowance) having inclination angles relative to the tube axis of < 45 degrees minus the NDE uncertainty, ANDEcA, on the measurement of the crack angle can be left in service. Tubes with two or more parallel (overlapping elevation), inclined axial cracks shall be plugged or repaired. For application of the 2.0 inch limit, an inclined indication is an axial crack that is visually inclined on the RCP C-scan, such that an angular measurement is required, and the measured angle exceeds the measurement uncertainty of ANDECA.

or 5) Circumferential, volumetric, and axial indications with inclination angles greater than (45 degrees - ANDECA) within the flexible W*

l shall be plugged or repaired.

9-z

6) Any type combination ofbtube degradation below th W*

length is acceptable. flexible

2. The SG tube integrity shall be determined after completing the corresponding actions (plug all tubes exceeding the plugging limit) required by Table 5.5.9-2.

e. Reports The contents and frequency of reports concerning the SG tube surveillance program shall be in accordance with Specification 5.6.10.

(continued)

DIABLO CANYON - UNITS I & 2 5.0-16 Unit 1 - Amendment No. 45, 442, TAB5.doc - R1 2 9 Unit 2 - Amendment No. 1435, 442,

Programs and Manuals 5.5

• Applicable for Units 1and 2, Cycles 10,11,12, and 13 only.

tTHIS PAGE NOT USED I DIABLO CANYON - UNITS I & 2 5.0-17 Unit 1 - Amendment No. 435, 454, TAB5.doc - R12 10 Unit 2 - Amendment No. 435,454,

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of SG tubes, the number of tubes plugged in each SG shall be reported to the Commission.

b. The complete results of the SG tube inservice inspection shall be submitted to the Commission in a report within 12 months following completion of the inspection. This Special Report shall include:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged.

c. Results of SG tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations sources ( conducted to determine cause of the tube degradation and corrective measures tken to prevent recurrence.

d. For ementation of the voltage-based repair criteria to tube support plate intersect inotify the NRC prior to returning the steam generators to service should any of flowing arise: land non-alternate repair criteria indications), l

1. If estimated lea kaased on the projected end-of-cycle (Q/r if not practical, using the actual measu end-of-cycle) voltage dtribui- by estimated leakage by all oth alternate repair cri exceeds th eak limit determieform the licensing basis dose calculati'n for the postul ed eamline break for the next operating cycle. , increased Ifcircumferentialcrack-like indications are detected at the tube support plate' frm2oneretos

'pw indications are identified that extend beyond the confines of the tube 00CC support plate.

inications n__I are identified at the tube support plate elevations that are

-I

\ attributable to primary water stress corrosion cracking.

If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-29 Unit 1 - Amendment No. 435, TAB5.doc- R12 11 Unit 2 - Amendment No. 435,

Reporting Requirements 5.6 5.6 Reporting Requirement 5.6.10 Steam Generator (SG) Tube Inspection Reor l

e. he results of the inspection of W*tubes shall be reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators. This report shall include: 5

~<<, Identification of W* tubes.

iace \ d2) W* inspection distance measured with respect to the BWT or the top of

}d the tubesheet, whichever is lower.

3) Elevation and length of axial indications within the flexible W*distance and the angle of inclination of clearly skewed axial cracks (ifapplicable).

4) The total steam line break leakage for the limiting steam generator per WCAP-14797.

f.(*)_jThe aggregate calculated steam line break leakage from application of all alternate repair criterW shall be reported to the Commission pursuant to 10 CFR I 50.4 within 90 da ollowing return to service of the steam generators.

g. For imple tation of the repair criteria for axial PWSCC at dented TSPs, the NRC I be notified prior to startup, pursuant to 10CFR50.72, of the following c itions that indicate a failure of performance criteria:

I __1 ___ -IL-. _.

1) The calculated SG probability of burst for condition monitoring exceeds 1 X10.

repair criteria

2) The calculated SG leakage for condition monitoring from all sources (all alternate repair criteria and non-alternate repair criteria indications) exceeds the leakage limit determined from the licensing basis steam line break dose calculation.

h. For implementation of the repair criteria for axial PWSCC at dented TSPs, the results of the condition monitoring and operational assessments will be reported to the NRC within 120 days following completion of the inspection. The report will include:

1) Tabulations of indications found in the inspection, tubes repaired, and tubes left in service under the ARC.

2) Growth rate distributions for indications found in the inspection and growth rate distributions used to establish the tube repair limits.

3) Plus Point confirmation rates for bobbin detected indications when bobbin is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents.

4) For condition monitoring, an evaluation of any indications that satisfy burst margin requirements based on the Westinghouse burst pressure model, but do not satisfy burst margin requirements based on the combined ANL ligament tearing and throughwall burst pressure model.

(continued)

Applicable for Units 1 and 2, Cycles 10, 11, 12, and 13 only.

DIABLO CANYON - UNITS 1 & 2. 5.0-30 Unit 1 - Amendment No. 135,151,452, TAB5.doc- R12 12 Unit 2 - Amendment No. 435,451,452,

Ppnnrtinn RPamirPmontcz Note: The requirements on Page 5.0-30a are rolled onto new Page 5.0-30b due to Insert A on Page 5.0-30.

5.6 Reporting Requirement ( u )

5.6.10 Steam Generator (SG) Tube Inspection RenoR'

5) Performance evaluation of the operational assessment methodology for predicting flaw distributions as a function of flaw size.

6) Evaluation results of number and size of previously reported versus new PWSCC indications found in the inspection, and the potential need to account for new indications in the operational assessment burst evaluation.

7) Identification of mixed mode (axial PWSCC and circumferential) indications found in the inspection and an evaluation of the mixed mode indications for potential impact on the axial indication burst pressures or leakage.

8) Any corrective actions found necessary in the event that condition monitoring requirements are not met.

i. For implementation of the probability of prior cycle detection (POPCD) method, for the voltage-based repair criteria at tube support plate intersections, if the end-of-cycle conditional main steamline break burst probability, the projected main steamline break leak rate, or the number of indications are underpredicted by the previous cycle operational assessment, the following shall be reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators:

1) The assessment of the probable causes for the underpredications, proposed corrective actions, and any recommended changes to probability of detection or growth methodology indicated by potential methods assessments.

2) An assessment of the potential need to revise the alternate repair criteria analysis methods if: the burst probability is underpredicted by more than 0.001 (i.e., 10% of the reporting threshold) or an order of magnitude; or the leak rate is underpredicted by more than 0.5 gpm or an order of magnitude.

3) An assessment of the potential need to increase the number of predicted low voltage indications at the beginning of cycle if the total number of as-found indications in any SG are underestimated by greater than 15%

or by greater than 150 indications.

DIABLO CANYON - UNITS 1 & 2 5.0-30a Unit 1 - Amendment No. 135,151,152,417, TAB5.doc- R12 13 Unit 2- Amendment No. 435,151-,452,4-9,

Technical Specification Inserts TS Insert A for TS Page 5.0-30

1) Identification of W* tube indications and indications that do not meet W*

requirements and were plugged or repaired, including the following information:

the number of indications, the location of the indications (relative to the BWT and TTS), the orientation (axial, circumferential, volumetric, inclined), the radial position of the tube within the tubesheet, the W* Zone of the tube, the severity of each indication (estimated depth), the side of the tube in which the indication initiated (inside or outside diameter), the W* inspection distance measured with respect to the BWT or TTS (whichever is lower), the length of axial indications, the angle of inclination of clearly skewed axial cracks (if applicable), verification that the upper crack tip of W* indications returned to service in the prior cycle remain below the BWT by at least the 95% confidence NDE uncertainty on locating the crack tip relative to the TTS, updated 95% growth rate for use in operational assessment, the cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet, and the condition monitoring and operational assessment main steamline break leak rate for each indication and each SG in accordance with the leak rate methodology described in PG&E Letter DCL-05-018, dated March 11, 2005, as supplemented by PG&E Letter DCL-05-090 dated August 25, 2005.

2) Assessment of whether the results were consistent with expectations and, if not consistent, a description of the proposed corrective action.

TS Insert B for TS Page 5.0-13 excluding the portion of the tube within the tubesheet below the Flexible W* Length or below 8 inches from the hot leg top of tubesheet, whichever is bounding

Enclosure 3 PG&E Letter DCL-05-090 Proposed Technical Specification Changes (retyped)

Remove Page Insert Page 5.0-10 5.0-10 5.0-13 5.0-13 5.0-1 3a 5.0-1 3a 5.0-15 5.0-15 5.0-16 5.0-16 5.0-17 5.0-17 5.0-29 5.0-29 5.0-30 5.0-30 5.0-30a 5.0-30a 5.0-30b

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Tube Surveillance Program SG tube integrity shall be demonstrated by performance of the following augmented inservice inspection program.

The provisions of SR 3.0.2 are applicable to the SG Tube Surveillance Program test frequencies.

a. SG Sample Selection and Inspection - SG tube integrity shall be determined during shutdown by selecting and inspecting at least the minimum number of SGs specified in Table 5.5.9-1.

b. SG Tube Sample Selection and Inspection - The SG tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 5.5.9-2. The inservice inspection of SG tubes shall be performed at the frequencies specified in Specification 5.5.9.c and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 5.5.9.d. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all SGs; the tubes selected for these inspections shall be selected on a random basis except:

1. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

2. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each SG shall include:

a) All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

b) Tubes in those areas where experience has indicated potential problems, c) A tube inspection (pursuant to Specification 5.5.9.d.1.h) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection, d) Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages, e) Tubes identified as W* tubes having a previously identified indication within the flexible W* length shall be inspected using a rotating pancake coil (RPC) probe or equivalent for the full length of the W* region during all future refueling outages.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-10 Unit 1 - Amendment No. 435,454, TAB 5 RAI.DOC 10 Unit 2 - Amendment No. 435,454,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2) This definition does not apply to the portion of the tube within the tubesheet below the W* length. Acceptable tube wall degradation within the W* length shall be defined as in 5.5.9.d.1.k.

3) This definition does not apply to axial PWSCC indications, or portions thereof, which are located within the thickness of dented tube support plates which exhibit a maximum depth greater than or equal to 40 percent of the initial tube wall thickness. WCAP-1 5573, Revision 1, provides repair limits applicable to these intersections.

4) A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be removed from service.

g) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of a Double Design Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 5.5.9.c.3, above; h) Tube Inspection means an inspection of the SG tube from the tube end (hot leg side) completely around the U-bend to the top support of the cold leg, excluding the portion of the tube within the tubesheet below the Flexible W* Length or below 8 inches from the hot leg top of tubesheet, whichever is bounding; i) Preservice Inspection means an inspection of the full length of each tube in each SG performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial Power Operation using the equipment and techniques expected to be used during subsequent inservice inspections; j) Tube Support Plate Plucicina Limit is used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the plugging limit is based on maintaining steam generator tube serviceability as described below:

(i) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to the lower voltage repair limit (NOTE 1), will be allowed to remain in service.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-13 Unit 1 - Amendment No. 135,451,452, TAB 5 RAI.DOC 14 Unit 2-Amendment No. 135,451,412,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

(ii) Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1), will be repaired or plugged, except as noted in 5.5.9.d.1.j (iii) below.

(iii) Steam generator tubes, with indication of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1) but less than or equal to the upper voltage repair limit (NOTE 2), may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit (NOTE 2) will be plugged or repaired.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-1 3a Unit I - Amendment No. 135,151,452, TAB 5 RAI.DOC 15 Unit 2 - Amendment No. 435,151,452,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

NOTE 1: The lower voltage repair limit is 2.0 volts for 7/8 inch diameter tubing at DCPP Units 1 and 2.

NOTE 2: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.

k) W* Plugginq Limit is used for disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented inside diameter stress corrosion cracking confined within the tubesheet, below the bottom of the WEXTEX transition (BWT). As used in this specification:

(i) Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet as determined by eddy current testing.

(ii) W* Length is the distance in the tubesheet below the BWT that precludes tube pull out in the event of the complete circumferential separation of the tube below the W* length. The W* length is conservatively set at: 1) an undegraded hot leg tube length of 5.2 inches for Zone A tubes and 7.0 inches for Zone B tubes, and 2) an undegraded cold leg tube length of 5.5 inches for Zone A tubes and 7.5 inches for Zone B tubes. Information provided in WCAP-14797-P, Revision 2, defines the boundaries of Zone A and Zone B.

(iii) Flexible W* Length is the W* length adjusted for any cracks found within the W* region. The Flexible W* Length is the total RPC-inspected length as measured downward from the BWT, and includes NDE uncertainties and crack lengths within W* as adjusted for growth.

(iv) W* Tube is a tube with degradation within or below the W* length that is left in service, and degraded within the limits specified in Specification 5.5.9.d.1.k (v).

(v) Within the tubesheet, the plugging (repair) limit is based on maintaining steam generator serviceability as described below:

1) For tubes to which the W* criteria are applied, the length of non-degraded tube below BWT shall be greater than or equal to the W* length plus NDE uncertainties and crack growth for the operating cycle.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-15 Unit 1 - Amendment No. 435, TAB 5 RAI.DOC 17 Unit 2 - Amendment No. 435,

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2) Axial cracks in tubes returned to service using W* shall have the upper crack tip below the BWT by at least the NDE measurement uncertainty and crack growth allowance, such that at the end of the subsequent operating cycle the entire crack remains below the BWT.

3) Resolvable, single axial indications (multiple indications must return to the null point between individual cracks) within the flexible W* length can be left in service. Alternate RPC coils or an ultrasonic test (UT) inspection can be used to demonstrate return to null point between multiple axial indications or the absence of circumferential involvement between axial indications.

4) Tubes with inclined axial indications less than 2.0 inches long (including the crack growth allowance) having inclination angles relative to the tube axis of < 45 degrees minus the NDE uncertainty, ANDECA, on the measurement of the crack angle can be left in service. Tubes with two or more parallel (overlapping elevation), inclined axial cracks shall be plugged or repaired. For application of the 2.0 inch limit, an inclined indication is an axial crack that is visually inclined on the RCP C-scan, such that an angular measurement is required, and the measured angle exceeds the measurement uncertainty of ANDEcA.

5) Circumferential, volumetric, and axial indications with inclination angles greater than (45 degrees - ANDEcA) within the flexible W*

length shall be plugged or repaired.

6) Any type or combination of tube degradation below the flexible W*

length is acceptable.

2. The SG tube integrity shall be determined after completing the corresponding actions (plug all tubes exceeding the plugging limit) required by Table 5.5.9-2.

e. Reports The contents and frequency of reports concerning the SG tube surveillance program shall be in accordance with Specification 5.6.10.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-16 Unit 1 - Amendment No. 4-35, 442, TAB 5 RAI.DOC 18 Unit 2 - Amendment No. I43, 442,

Programs and Manuals 5.5

--I THIS PAGE NOT USED I DIABLO CANYON - UNITS 1 & 2 5.0-17 Unit 1 - Amendment No. 435, 451-,

TAB 5 RAI.DOC 19 Unit 2 - Amendment No. 4-35, 414,

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report

a. Within 15 days following the completion of each inservice inspection of SG tubes, the number of tubes plugged in each SG shall be reported to the Commission.

b. The complete results of the SG tube inservice inspection shall be submitted to the Commission in a report within 12 months following completion of the inspection. This Special Report shall include:

1) Number and extent of tubes inspected,

2) Location and percent of wall-thickness penetration for each indication of an imperfection, and

3) Identification of tubes plugged.

c. Results of SG tube inspections, which fall into Category C-3, shall be reported in a Special Report to the Commission within 30 days and prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

d. For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC prior to returning the steam generators to service should any of the following arise:

1. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution, increased by estimated leakage by all other sources (alternate repair criteria and non-alternate repair criteria indications), exceeds the leak limit determined from the licensing basis dose calculation for the postulated main steamline break for the next operating cycle.

2. If ODSCC indications are identified that extend beyond the confines of the tube support plate.

IH

3. If the calculated conditional burst probability based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x 10-2, notify the NRC and provide an assessment of the safety significance of the occurrence.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-29 Unit 1 - Amendment No. 4-35, TAB 5 RAI.DOC 33 Unit 2 - Amendment No. 435,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued) I

e. The results of the inspection of W* tubes shall be reported to the Commission I pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators. This report shall include:

1) Identification of W* tube indications and indications that do not meet W*

requirements and were plugged or repaired, including the following information: the number of indications, the locations of the indications (relative to the BWT and TTS), the orientation (axial, circumferential, (

volumetric, inclined), the radial position of the tube within the tubesheet, the W* Zone of the tube, the severity of each indication (estimated depth), the side of the tube in which the indication initiated (inside or outside diameter), the W* inspection distance measured with respect to the BWT or TTS (whichever is lower), the length of axial indications, the angle of inclination of clearly skewed axial cracks (if applicable),

verification that the upper crack tip of W* indications returned to service in the prior cycle remain below the BWT by at least the 95% confidence NDE uncertainty on locating the crack tip relative to the TTS, updated 95% growth rate for use in operational assessment, the cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet, and the condition monitoring and operational assessment main steamline break leak rate for each indication and each SG in accordance with the leak rate methodology described in PG&E Letter DCL-05-018, dated March 11, 2005, as supplemented by PG&E Letter DCL-05-090, dated August 25, 2005.

2) Assessment of whether the results were consistent with expectations and, if not consistent, a description of the proposed corrective action. I

f. The aggregate calculated steam line break leakage from application of all alternate repair criteria and non-alternate repair criteria shall be reported to the I Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators.

g. For implementation of the repair criteria for axial PWSCC at dented TSPs, the NRC shall be notified prior to startup, pursuant to 10CFR50.72, of the following conditions that indicate a failure of performance criteria:

1) The calculated SG probability of burst for condition monitoring exceeds 1 x 10-2

2) The calculated SG leakage for condition monitoring from all sources (all alternate repair criteria and non-alternate repair criteria indications) exceeds the leakage limit determined from the licensing basis steam line break dose calculation.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-30 Unit 1 - Amendment No. 435,451,452, TAB 5 RAI.DOC 34 Unit 2 - Amendment No. 435,454,452,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

h. For implementation of the repair criteria for axial PWSCC at dented TSPs, the results of the condition monitoring and operational assessments will be reported to the NRC within 120 days following completion of the inspection. The report will include:

1) Tabulations of indications found in the inspection, tubes repaired, and tubes left in service under the ARC.

2) Growth rate distributions for indications found in the inspection and growth rate distributions used to establish the tube repair limits.

3) Plus Point confirmation rates for bobbin detected indications when bobbin is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents.

4) For condition monitoring, an evaluation of any indications that satisfy burst margin requirements based on the Westinghouse burst pressure model, but do not satisfy burst margin requirements based on the combined ANL ligament tearing and throughwall burst pressure model.

(continued)

-1 DIABLO CANYON - UNITS 1 & 2 5.0-30a Unit 1 - Amendment No. 4135,451-,452, TAB 5 RAI.DOC 35 Unit 2 - Amendment No. 4135,441,4-52,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

5) Performance evaluation of the operational assessment methodology for predicting flaw distributions as a function of flaw size.

6) Evaluation results of number and size of previously reported versus new PWSCC indications found in the inspection, and the potential need to account for new indications in the operational assessment burst evaluation.

7) Identification of mixed mode (axial PWSCC and circumferential) indications found in the inspection and an evaluation of the mixed mode indications for potential impact on the axial indication burst pressures or leakage.

8) Any corrective actions found necessary in the event that condition monitoring requirements are not met.

i. For implementation of the probability of prior cycle detection (POPCD) method, for the voltage-based repair criteria at tube support plate intersections, if the end-of-cycle conditional main steamline break burst probability, the projected main steamline break leak rate, or the number of indications are underpredicted by the previous cycle operational assessment, the following shall be reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators:

1) The assessment of the probable causes for the underpredications, proposed corrective actions, and any recommended changes to probability of detection or growth methodology indicated by potential methods assessments.

2) An assessment of the potential need to revise the.alternate repair criteria analysis methods if: the burst probability is underpredicted by more than 0.001 (i.e., 10% of the reporting threshold) or an order of magnitude; or the leak rate is underpredicted by more than 0.5 gpm or an order of magnitude.

3) An assessment of the potential need to increase the number of predicted low voltage indications at the beginning of cycle if the total number of as-found indications in any SG are underestimated by greater than 15%

or by greater than 150 indications.

DIABLO CANYON - UNITS 1 & 2 5.0-30b Unit 1 - Amendment No. 435,45-145277, TAB 5 RAI.DOC 36 Unit 2 - Amendment No. 435,454,4524-79,

Enclosure 4 PG&E Letter DCL-05-090 WARNING - PROPRIETARY INFORMATION ENCLOSED Proprietary Notice and Application for Withholding for Westinghouse Electric LLC Proprietary Information to Respond the NRC Request for Additional Information on PG&E Letter DCL-05-018, Diablo Canyon Units 1 and 2 License Amendment Request 05-01, TAC Nos. MC6409 and MC6410 Westinghouse Electric LLC Information to Respond the NRC Request for Additional Information on PG&E Letter DCL-05-018, Diablo Canyon Units 1 and 2 License Amendment Request 05-01, TAC Nos. MC6409 and MC6410 (proprietary)

Westinghouse Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 3744643 Document Control Desk Direct fax: (412) 3744011 Washington, DC 20555-0001 e-mail: greshajawestinghouse.com Our ref: CAW-05-2034 August 3, 2005 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-CDME-05-129, Rev. 1, P-Attachment, "Response to NRC Request for Additional Information on PG&E Letter DCL-05-018, License Amendment Request 05-01 Revision to Technical Specification 5.5.9, 'Steam Generator (SG) Tube Surveillance Program,' and 5.5.10,

'Steam Generator (SG) Tube Inspection Report,' to Allow Use of the W* Alternate Repair Criteria for Indications in the Westinghouse Explosive Expansion (WEXTEX) Region on a Permanent Basis" The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-05-2034 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Pacific Gas and Electric Company.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-05-2034, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

Very truly yours, J. A. Gresham, Manager Regulatory Compliance and Plant Licensing Enclosures cc: B. Benney L. Feizollahi A BNFL Group company

CAW-05-2034 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared J. S. Galembush, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

J. S. Galembush Customer I"5 Leader Sworn to and subscribed before me this day of ,2005 Notary Public NotSea Sharon L Frod, Notary PuA*

Monroevfle BAro, A County dey MyComrnsslon Eores Jauay29.2007 Member, Pensyntva Assiagjn Of olbaries

2 CAW-05-2034 (1) I am Customer I"' Leader in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) 1am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse "Application for Withholding" accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.

The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

3 CAW-05-2034 (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

4 - CAW-05-2034 (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-CDME-05-129, Rev. 1, P-Attachment, "Response to NRC Request for Additional Information on PG&E Letter DCL-05-0 18, License Amendment Request 05-01 Revision to Technical Specification 5.5.9, 'Steam Generator (SG) Tube Surveillance Program,' and 5.5.10, 'Steam Generator (SG) Tube Inspection Report,' to Allow Use of the W* Alternate Repair Criteria for Indications in the Westinghouse Explosive Expansion (WEXTEX) Region on a Permanent Basis," dated July 2005 (Proprietary). The information is provided in support of a submittal to the Commission, being transmitted by Pacific Gas and Electric Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted for use by Westinghouse for Diablo Canyon Units I and 2 is expected to be applicable for other licensee submittals in support of implementing the W* inspection methodology addressing service induced degradation in the tube joint region of steam generators.

This information is part of that which will enable Westinghouse to:

(a) Provide documentation of the analyses, methods, and testing for the implementation of the W* tube inspection methodology.

5 CAW-05-2034 (b) Provide contact pressure as a function of depth by zone for W* leak rate determination for Diablo Canyon Units I and 2.

(c) Provide a bounding W* potential steam line break leakage evaluation from within and below the W* depth for Diablo Canyon Units I and 2.

(d) Assist the customer to respond to NRC requests for information.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for licensing documentation.

(b) Westinghouse can sell support and defense of this information to its customers in the licensing process.

(c) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar licensing support documentation and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Vestinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Enclosure 5 PG&E Letter DCL-05-090 Westinghouse Electric LLC Information to Respond the NRC Request for Additional Information on PG&E Letter DCL-05-018, Diablo Canyon Units 1 and 2 License Amendment Request 05-01, TAC Nos. MC6409 and MC6410 (nonproprietary)

Enclosure 5 PG&E Letter DCL-05-090 ac,e l

-Figure 1: SLB Leak Rate from Constrained Cracks Prediction Bound. (RAI #1 and #4) a,ce Figure 2: Segregated SLB Leak Rates. (RAI #3) 1

Enclosure 5 PG&E Letter DCL-05-090 ac,e Figure 3: 2nd Versus I't Order SLB Leak Rates. (RAI #3) a,c,e Figure 4: Scatter Plot of the Residuals from the Regression Analysis (RAI #3) 2

Enclosure 5 PG&E Letter DCL-05-090 a,c,e Figure 5: Normal Plot of the Residuals from the Regression Analysis (RAI #3) a,c,e Figure 6: Prediction Bounds from Separate Regression Lines (RAI #3) 3

Enclosure 5 PG&E Letter DCL-05-090 ac,e Figure 7: Prediction Bounds With Outlier Removed (RAI #3) a,ce Figure 8: Leak Rate Versus Distance Below TTS. (RAI #12) 4