ML042430214

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Draft Information for June 2004 - Email from Hicks to Thadani - 06/09/04
ML042430214
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 06/09/2004
From: Hicks J
TXU Energy
To: Thadani M
Office of Nuclear Reactor Regulation
Thadani M, NRR/DLPM, 415-1476
Shared Package
ML042450085 List:
References
Download: ML042430214 (17)
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CA\TEMP\GW}00001 .TMP Page -1 ,

C:\TEMP\GWIOOOO1 .TMP Pae1 Do 5ki-p ,S a0 - D44S Mail Envelope Properties (40C75148.884: 1: 63620)

Subject:

Response to Question on LAR 03-004 (TXX-04090)

Creation Date: 6/9/04 1:57PM From: <ihicks I @txu.com>

Created By: ihicksl @txu.com Recipients nrc.gov owf4_po.0WFN_DO MCT (Mohan Thadani) txu.com dbuschbl CC Post Office Route owf4_po.0WFNDO nrc.gov txu.com Files Size Date & Time

,MESSAGE 221 06/09/04 01:57PM T04090.doc 91136 Mime.822 126846 Options Expiration Date: None Priority: Standard Reply Requested: No Return Notification: None Concealed

Subject:

No Security: Standard

Mohan Thadani - Response to Question on LAR 03-004 (TXX-04090) Page 111 From: <jhicksl @txu.com>

To: <mct@nrc.gov>

Date: 6/9/04 2:05PM

Subject:

Response to Question on LAR 03-004 (TXX-04090)

Attached are the responses to the qustions on LAR 03-004, Revision to TS3.6.3 to extend SR 3.6.3.7 frequency for Containment Purge, Hydrogen Purge and Containment Pressure Relief Valves.

(See attached file: T04090.doc)

CC: <dbuschbl @txu.com>

Mohan Thadani - T0409O.doc Page1 iI Mohan Thadani T04090.doc 4- -

-- ---.- -----..- ----- Paael Ref: 10CFR50.90 CPSES-200401509 Log # TXX-04090 File # 00236 June 9, 2004 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50445 AND 50-446 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST (LAR)03-004 REVISION TO TECHNICAL SPECIFICATION (TS) 3.6.3 TO EXTEND SURVEILLANCE REQUIREMENT (SR) 3.6.3.7 FREQUENCY FOR CONTAINMENT PURGE, HYDROGEN PURGE AND CONTAINMENT PRESSURE RELIEF VALVES WITH RESILIENT SEATS (TAC NOS. MC0911/MC0912)

REF: 1) TXU Energy Letter, logged TXX-03078, from C. L. Terry to the U. S. Nuclear Regulatory Commission, dated September 23, 2003 Gentlemen:

In Reference 1, TXU Generating Company LP (TXU Power) submitted a proposed change to the Technical Specification (TS) associated with containment isolation valves (LAR 03-004). The proposed change will revise TS 3.6.3 entitled "Containment Isolation Valves," to extend the frequency of Surveillance Requirement (SR) 3.6.3.7 for the containment purge, hydrogen purge and containment pressure relief valves with resilient seats.

Based on conversations with the NRC staff, TXU Power provides the following information regarding LAR 03-004. The attachment contains the NRC questions and TXU Power's response immediately following each question.

This communication contains no new or revised commitments.

Mohan Thadani - T0409O.doc Page Paae 2l1 211 Mohan Thadani T04090.doc Should you have any questions, please contact Mr. Jack Hicks at (254)897-6725 or email (jhicksl @tux.com).

I state under penalty of perjury that the foregoing is true and correct.

Executed on June 9, 2004.

Sincerely, TXU Generation Company LP By: TXU Generation Management Company LLC Its General Partner Mike Blevins By:

Fred W. Madden Regulatory Affairs Manager JCH/jch Attachment TXU Power Responses to RAI c - B. S. Mallett, Region IV W. D. Johnson, Region IV M. C. Thadani, NRR Resident Inspectors, CPSES Ms. Alice Rogers Bureau of Radiation Control Texas Department of Public Health 1100 West 49th Street Austin, Texas 78756-3189 A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway . Comanche Peak . Diablo Canyon . Palo Verde . South Texas Project - Wolf Creek

Mohan Thadani - T0409O.doc -- Page 3Q1; i

ATTACHMENT to TXX-04090 TXU Power Responses to RAI

Mohan Thadani - T040'0.doc Page 4 NRC Question 1:

Describe the two methods mentioned in Section 4.4 of the license amendment request that were used to evaluate the risk significance of the proposed change (ISLOCA methodology and RI-IST methodology). For each methodology, provide calculation details and results that support the conclusion that the proposed change is not risk significant.

TXU Power Response:

The potential of increase in LERF due to the extension in the LLRT test interval was evaluated by two different methodologies. The first used the approach developed by Brookhaven National Laboratory for their NUREG on ISLOCA. The second followed the methodology used for the Risk-Informed IST program implemented at CPSES. These methodologies were chosen because the PRA model does not explicitly consider containment isolation failures for penetrations that have two normally closed valves in series since the failure of spuriously opening or failure to remain close were deemed to be low probability events. The valves under consideration are normally closed and remain closed during power operation. The exceptions to this are the containment pressure relief valves, which are periodically opened. It is their failure to re-close following an event that is the significant containment isolation failure contributor to LERF. This failure mode is unaffected by the requested extension since it is not confirmed by the LLRT.

ISLOCA Methodology The evaluation for leakage through containment penetrations was accomplished by adapting the generic system failure models found in NUREG/CR-5 102, "ISLOCA Evaluation Guidelines," to the specific valve arrangements found at CPSES. The formulas obtained for the generic case of two valves in series were applied to the valve arrangement for these penetrations.

The basic failure model for two valves in series described in the NUREG, can be applied to calculate the average failure frequency of this arrangement provided the proper valve failure modes are selected based on valve arrangements and test practices. The frequency of "excessive leakage" is determined by use of the NUREG equation 9 (page 4-20, derived from equation 12a, Section B.1.2 of Appendix B). The reference equation is adapted to the specific failure modes identified below:

I (CIV Bypass) = A

  • B [X 1 7T+ XRXLT+ 2XLd ((dT+1)/2)]

Where, XL.XR.Xd denote the mean rate of the 1) internal leakage, 2) disk ruptures and 3) valve fails A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway - Comanche Peak . Diablo Canyon . Palo Verde

  • Wolf Creek

Paae 511 Thadani - T04090.doc Mohan Thadani Mohan - T0409O.doc ~Page 5 d Attachment to TXX-04090 Page 5 of 12 open, while indicating closed due to failure of internal connections or the limit switch itself.

The disk rupture value (XR) was taken from the NUREG (A.2.1) based on generic industry data for valve disk failures; specific industry data for PWR MOVs was not identified in the NUREG. The value used in the NUREG is 1.20E-03 per year for disk rupture and that value was used in this evaluation. The valve fails open value (Xd) used in the NUREG represents a MOV failing open while indicating closed. The value in the NUREG for a MOV is approximately a factor of 5 less than the MOV transfer value for a year value (1.07E-4 vs.

4.71E-4). Therefore, the CPSES PRA value for an AOV transferring over a period of a year will be affected by a similar ratio. This modified value of 4.28E-02 (1.1 1E-6

  • 8760 *

(4.71E-04/1.07E4)) was used in this evaluation. Leakage probabilities of an AOV are also not readily available. However, the NUREG provides a value based on MOV leakage.

Since the initial basis for increased testing was due to industry poor performance of these kinds of containment isolation valves, the value for MOVs in the NUREG was increased by a factor of 3 to account for the past industry poor performance. The factor of 3 was chosen based on CPSES past performance of these, which indicated that these penetration perform adequately from a LERF perspective. The revised value for (XL) is 1.46E-02 (4.85E-03

  • 3) per year.

"d" denotes the demand rate of the valves between test ("d" is assumed to be constant with respect to test interval change, as the number of demands remain unchanged. Therefore for the purge valves d = 1 was assumed and for the relief valves a value of d = 6 for once per 92 days was assumed).

T is the period of time between leak tests and T = 1/2/2, every 6 months, for purge valves and T = 1/4, every 92days, for the relief valves for the original test interval. T represents the stroke test performed at each refueling outage (TR = 1.5 years) for the new test interval.

A and B are unitless factors that denote the number of similar lines (i.e., 2 for the purge lines and 1 for the relief line) and the unitless capacity factor determined from the plants operating history. This evaluation assumes a conservative value of 1.0 for capacity factor.

Substituting the values into the equation, the base case for each type of penetration (purge or relief) for the likelihood of the penetration to leak given a core damage event would be:

I (avPurge) = 2*1 [((1.23E-02) 2

  • 0.5) + (1.20E-03
  • 1.23E-02* 0.5) + (2*1.23E-02* 4.86E-
02) * (((1
  • 0.5) +1) -2))] = 4.181E-04 I (V~eMief) = 1*1 [((1.23E-02) 2
  • 0.25) + (1.20E-03
  • 1.23E-02* 0.25) + (2
  • 1.23E-02*

(4.86E-02) * (((6

  • 0.25) +1) .2))] = 2.139E-04 The change in test interval as outlined in the LAR would change "T" to be 1.5 years and "d" to be 1 for purge valves and 6 for relief valves between tests. Substituting this value into the equations above, the new likelihood of the penetration to leak given a core damage event would be:

Mohan Thadani - T04090.doc Page6Id Attachment to TXX-04090 Page 6 of 12 I (av Purge) = 1.046E-03 I (civ Rlief) = 1.056E-03 Using the results above with the current CPSES CDF value of 2E-05, the change in LERF can be estimated to be:

L&ERF= [(I (av Purgep)cw

  • CDF) + (I (cIv Relief)new
  • CDF)] - [(I (CIv Purge) base

(I (pvRelie)ba.e *CDF)]

= [(1.046 -03* 2E-05) + (1.056E -03* 2E-05)] - [(4.18 1E-04* 2E-05) + (2.139E-04

= 2.939E-08 This is less than Reg. Guide 1.174 (ALERF < IE-07) and Reg. Guide 1.177 (ICLERP < 5E-

08) guidelines for risk significance.

Risk-inform IST Methodology The methodology employed by the Risk-Informed IST program simply assumes a linear increase in failure rate of a component associated with test frequency. For this LAR request, the bounding change in test interval would be from every 92 day (quarterly) to every refueling outage (18 month) or a factor of 6 increase in the failure rate. Since the LAR pertains to a change in LLRT test interval, the failure mode of interest would be that of the normally close penetration leaking a significant amount such that the resulting bypass would be considered large. As seen from the earlier evaluation, no specific or generic leak rate information was found on these types of valves. Therefore, based on past service history, a failure rate of 4.18E-03 was used for each valve. It needs to be noted that the failure rate calculated is for a penetration failing to meet leakage criteria and not on an individual valve basis to meet leakage rate criteria. The one failure noted in the CPSES service history for these valves indicated that upon detail inspection of the 2 series valves, that only one failed to meet the specified limit and that the second valve did meet its acceptance criteria.

Therefore, the failure rate of the penetration is the calculated failure rate of a single valve squared as each penetration contains two valves in series.

The change in release frequency can than be calculated using the following formula:

ALERF = LERF,,w - LERForignal

= CDF * (2~ -

= 1.69E-05 (6.29E 1.75E-05) = 1.03E-8 Where, X...v = (4.18E-03) 2 = 1.75E-05 and X, = (6

  • 4.18E-3)2 = 6.29E-04 Reg. Guide 1.174 provides risk-acceptance guidelines for risk-informed regulations. It A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
  • Comanche Peak
  • Diablo Canyon . Palo Verde . South Texas Project
  • Wolf Creek

Mohan Thadani - T04090.doc Page 7; Attachment to TXX-04090 Page 7 of 12 establishes regions and acceptance criteria for each region based on a plant's baseline LERF and the change in LERF of the proposed request. The guide sets a threshold of IE-07 for delta LERF as being risk insignificant. Reg. Guide 1.177 provides risk-acceptance guidelines for risk-informed regulations associated with Technical Specifications. It provides a threshold of less than 5E-08 for CLERP as being risk insignificant. Therefore based on the calculation above, which assumes that the failure rate of the valves to prevent leakage increase proportionally with the increase in test interval, the change is found to be not risk significant.

Conclusion Two methods of determining the potential change in LERF are provided in the preceding sections. Each method has been used at CPSES in support of risk informed applications.

The ISLOCA methodology is part of the current CPSES PRA model and the RI-IST methodology is part of the CPSES Risk informed IST program. Both methods provide insight into the potential increase in penetration leak rate that may occur with an extension in test interval. Each method provides similar results with respect to risk significance of the proposed change. The increase in test interval from the current 3 month or 6 month to each refueling outage is supported using this risk informed approach. Extending the interval for these valves is not risk significant. The two different methods presented above, which both assumed an increased failure rate of the valves, found the increase in risk (LERF) to be less than Reg. Guide 1.174 (ALERF < 1E-07) and Reg. Guide 1.177 (ICLERP < 5E-08) guidelines and the extension in test frequency is considered to be not risk significant.

NRC Question 2:

Identify the approach used by TXU to estimate LERF (e.g., a complete Level 2 PRA, the simplified approach in NUREG/CR-6595, etc.). Provide the LERF prior to implementing the proposed change (baseline LERF), the percentages of its major contributors, and the percentage attributed to containment isolation failure.

TXU Power Response:

The back-end (sometimes called containment) analysis of the CPSES PRA utilizes the approach of a Level II probabilistic risk assessment (PRA). A Level II analysis involves two types of considerations: (1) analyses of physical processes during severe accidents, where degraded core and containment thermo-hydraulic variables are determined along with source terms for the accident progressions, and (2) a probabilistic component where the likelihood of the various outcomes is assessed. The Level I core melt sequences are collected or binned into a Plant Damage State (PDS), such that sequences in a given bin have similar accident progressions and outcomes of similar likelihood. For each PDS there is a containment event tree (CET). The path through each CET begins with a PDS and ends with a CET end-state which is defined by a containment failure mode, time and release

Mohan Thadani - T0409O.doc Pacle 8 l MohaThaaniT0090;o Pce 11 Attachment to TXX-04090 Page 8 of 12 fractions. These CET end-states are later binned into release categories. The primary considerations for defining the CPSES release categories are whether: (1) the containment is intact at the time of core melt or whether (2) the release involves a bypassed or unisolated containment. For CPSES, LERF is made up of those release categories that were considered to be large and early as well as those that led to direct containment by-pass (i.e., V-sequences, Containment Isolation failures and SGTR sequences).

The CPSES Large Early Release Frequency (LERF) is 5.3 1E-07 per year assuming average test and maintenance. The dominant initiating event contributors to LERF are containment failure events (i.e., Loss of Offsite Power/SBO induced events including RCP Seal LOCAs and containment overpressurization events), and containment bypass events (i.e., ISLOCA, SGTR sequences). Loss of Offsite Power/SBO induced events account for approximately 45%, ISLOCAs account for approximately 29% and SGTR accounts for approximately 18%

of the total large early release frequency.

The Loss of Offsite Power/SBO initiator results in the loss of systems providing RCP Seal cooling or a loss of RCS primary and secondary cooling as well as the loss of containment protection systems. Both the RCS overpressurization and the induced RCP Seal LOCA scenario involve a loss of inventory inside containment (e.g. inventory in the containment sumps) with a subsequent loss of containment from various failure mechanisms.

Containment Isolation valve failures were found to be negligible contributors to containment failure associated with the Loss of Offsite Power/SBO initiator results. The loss of containment provides the release path to the environment.

By definition, the ISLOCA scenarios involve a direct path outside containment for radionuclide release. A failure of the valves on the intermediate pressure injection lines provides a direct path from the RCS to the safeguards building due to the postulated break in the SI piping outside containment. A similar scenario exists for the RHR ISLOCA. For this scenario, a failure of the valves on the RHR suction lines from the RCS hot legs provides a direct path from the RCS to the safeguards building due to the postulated break in the RHR piping outside containment.

NRC Question 3:

In the Level 2 PRA, what is the definition of containment isolation failure? Relate the size of the isolation failure assumed to cause a large release in the PRA to the administrative limits for measured leakage through the valves with resilient seats (12,500 sccm for the containment purge and hydrogen purge valves, and 15,100 sccm for the containment A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon . Palo Verde - South Texas Project . Wolf Creek

Mohan Thadani - T04090.doc _ _ Page99 Mohan T n -

Attachment to TXX-04090 Page 9 of 12 pressure relief valves, as stated in Section 4.2 of the license amendment request).

TXU Power Response:

The CPSES PRA models the various containment penetrations with respect to the function of closing and remaining closed (when isolation is demanded). The failure probability used in the analysis is a combination of generic and plant specific information. Plant specific data is based in part, on acceptance criteria defined by the plant's design/limits. The failure of any one penetration (loss of penetration integrity) is considered a direct by-pass of containment. The Level II analysis evaluates complementary attributes of containment performance, that is, those attributes not (previously) explicitly modeled as part of the penetration integrity function (e.g. containment/liner integrity). These attributes can be described as those that establish the physical integrity of the containment building itself.

The CPSES Level II PRA analysis has an end state of radioactive release following core damage. The Large Early Release Frequency (LERF) is defined as a rapid, unscrubbed release of airborne fission products from the plant to the environment that occurs before the effective implementation of the off-site emergency response protective actions. This has been further refined to be release of one containment volume within one hour, which occurs before or within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of vessel breach. The PRA failure rates are based on the component's ability to perform its function. In this case, the components either would have failed to closed, if opened for the existing modeled failure mode or they would have failed to meet their LLRT requirements for the leakage failure rate used in the evaluation.

For CPSES the overall Containment leakage limit is 1.0 L. This includes all sources of leakage. The total of LLRT leakage is 0.6 La. The limit is < 0.05 L for the 48" and 12" penetrations and < 0.06 Lfor the 18" penetration. These are the bases for the administrative limits referred to in the question. Using the EPRI PSA Applications Guide criteria, to have a containment volume (2,985,000 ft3) escape within one hour of the event equates to a leakage rate of -2.1 8E+8 sccm. This is significantly larger than administrative limits (1.2 to 1.5E+4 sccm) for measured leakage through these valves with resilient seats.

In addition, based on the draft SDP for containment (appendix H), a significant leakage rate for LERF has been defined as 100 volume percent per day leakage. The leakage associated with the LLRT acceptance criteria is also significantly less than this threshold.

NRC Question 4:

Describe the significant causes of containment isolation failure (e.g., isolation valves fail to close on demand, isolation valve leakage, etc.) and indicate the approximate percentage that each failure cause contributes to the overall containment isolation failure probability.

TXU Power Response:

Mohan Thadani - T04090.doc Page 10 Attachment to TXX-04090 Page 10 of 12 The CPSES LERF results indicate that the significant cause of containment isolation failure can be attributed to common cause failure of air-operated valves failing to close on demand.

Isolation valve leakage mode is not considered due to the low probability of significant leakage from two normally closed valves in series. As indicated in the response to question number one, containment isolation valve failures are not significant contributors (less than 1%) to LERF.

NRC Question 5:

Describe the time-dependent leakage rate through valves with resilient seats, citing relevant studies or research. How much increase in leakage rate can be expected after extending the current surveillance test interval (3 months or 6 months, depending on valve type) to 18 months? Since (a) the total containment leakage rate is the sum of individual component leakage rates, and (b) the proposed increase in test interval may cause increased leakage rates for valves with resilient seats, the licensee should demonstrate that the proposed 18-month test interval does not increase the total containment leakage rate to the extent that it approaches the size of the isolation failure assumed to cause a large release in the PRA.

TXU Power Response:

As provided in the response to question number one, the leakage rates through these valves were increased based on the increase in the requested test interval. With the proposed extension of test intervals, leakage rates are not expected to change significantly; they should remain at or below the existing administrative limits. This expectation is substantiated by past performance that demonstrates leakage at current test intervals has consistently been observed to be below the administrative limit (see the tabular results from past LLRT).

When the administrative limit is not exceeded, no action is required and the valve is typically unaltered between outages when routine maintenance is carried out. This indicates that future leakage rates should be consistent with past experience even if testing frequency is reduced.

Question number three indicated the isolation failure rates assumed to cause a large release are larger than the leakage corresponding to the current LLRT administrative limits by orders of magnitude. The comparison shows PRA assumptions would remain valid even when summing individual component leakage rates. The magnitude of PRA assumptions A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway - Comanche Peak . Diablo Canyon - Palo Verde . South Texas Project

  • Wolf Creek

- . ; J - . . - .- -- I .

Mohan Thadani - T0409O.doc Page 1111IA Page Mohan Thadani T04090.doc Attachment to TXX-04090 Page II of 12 relative to leakage also allows for evaluation of failures and associated increases should administrative limits be exceeded. Therefore even assuming a linear increase in the leakage limits proportional to the increase in test frequency, these larger leak rates in combination with other existing leak paths would not approach the PRA limit.

CPSES's corrective action and feedback plans will ensure that testing failures are evaluated to determine whether adjustment to the component's test strategy is appropriate.

Component corrective action procedures will be in place before any test intervals are extended.

When a component on the extended test interval fails to meet established test criteria, corrective actions will be taken in accordance with the CPSES corrective action program.

For components not meeting the acceptance criteria, a SmartForm or equivalent will be generated. This document initiates the corrective action process. Also, the initiating event for a SmartForm may be from causes other than an unacceptable LLRT test. Programs exist that provide timely information to the LLRT coordinator that the performance of a reliable component may have degraded. The recorded information could then be used to assess whether a significant change in component reliability has occurred such that the component would merit a change in test interval.

The initiating event (for the preparation of a SmartForm) could be any other indication that the component is in a non-conforming condition. The unsatisfactory condition will be evaluated to achieve the following objectives:

(1) Determine the impact on system operability and take appropriate action.

(2) Review the previous test data for the component and all components in the group.

(3) Perforrn a root cause analysis.

(4) Determine if this is a generic failure. If it is a generic failure whose implications affect a group of components, initiate corrective action for all components in the affected group.

(5) Initiate corrective action for failed LLRT components.

(6) Evaluate the adequacy of the test strategy. If a change is required, review the LLRT test schedule and change as appropriate.

Past Penetration History The most recent set of test history (from 1999 to present) indicates that there was one failure in the 160 tests performed on the containment penetrations associated with this LAR. A cursory review of the earlier service history indicates that there was no failure of the penetration to meet their LLRT requirements in the additional 79 test performed. Therefore, the revised history would be indicates that there was one failure in 239 tests. It should be

Mohan Thadani - T0409O.doc Page -12 dI Paoel2 Mohan Thadani T04090.doc Attachment to TXX-04090 Page 12 of 12 noted that the LLRT is a test of the penetration and not of the individual valves in the penetration. The one failure that did occur was found to have been in one of the two valves and that the other valve did maintain pressure.

Unit I Penetration MII-19 (12" penetration, Admin. limit of 12,500 SCCM) Data (Leakage is in SCCM):

Date Leakage Date Leakage Date Leakage 01/29/03 2590 09/23/99 1625 11/10/96 2500 10/17102 2900 05/30199 2420 07/10/95 2410 10/09/02 2450 12113/98 2690 03/30/95 2560 08/18/02 2500 06/27/98 2400 08/26/94 2470 02/28/02 2580 04/11/98 2510 02/24/94 2480 09/15/01 2550 10/17/97 2510 11/26/93 2530 04/11/01 2750 04/25/97 2490 6/18/93 *3530 11/12100 2640 10/05/96 2520 _

04/30/00 2340 07/12/96 2360 1 _

Date Leakage Date Leakage Date Leakage 03/01/03 1524 03/12/99 1430 06/06/95 1523 09/13/02 1603 11/06/98 1406 11/09/94 1833 04/23/02 1776 05/11/98 1497 05/23/94 *4490 01/14/02 3350 11/27/97 1638 01/18/94 3090 03/22/01 1346 09/26/97 1368 06/22/93 2260 02/04/01 1510 03/27/97 1502 02110/93 2470 10/23/00 1280 09/27/96 1265 _

08/20/00 1350 03/27/96 1496 03/03/00 1360 11/07195 1669

Date Leakage Date Leakage Date Leakage 05/03/03 5040 08/29/99 4700 01/31/96 5720 11/07/02 3820 03/14/99 855 01/11/96 2930 04/13/02 3400 09/20/98 4570 07/11/95 5580 A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon . Palo Verde
  • South Texas Project - Wolf Creek

Mohan Thadani - T0409O.doc Page 1-3 1 Mohan Thadani T04090.doc

- Paae 13d Attachment to TXX-04090 Page 13 of 12 09/23/01 1840 04/21/98 3650 04/09/95 5400 04/17/01 4050 01/16/98 4192 09/08/94 5770 01/12/01 3930 08/08/97 4700 03/08/94 5490 07/30/00 667 02/07/97 4810 12/05/93 5600 02/13/00 4400 10/05/96 5680 03/18/93 *11250 10120199 4280 08/16/96 5270 12/17/92 10430

Date Leakage Date Leakage Date Leakage 03/29103 1122 04/16/99 714 08/18/95 20 10110/02 1150 11/07/98 1265 02/14/95 393 04/29/02 2150 05/01/98 844 11/14/95 603 08/18/01 16 12/01/97 22 05/13/94 *9720 03/03/01 1100 08/01/97 101 03/30/94 24 10/27/00 1960 01/31/97 20 10/01/93 702 09/17/00 1375 08/16/96 34 06/21/93 766 04/02/00 289 05/01/96 38 02/26/93 703 10/17/99 217 04/06/96 28

Date Leakage Date Leakage Date Leakage 03/08/03 1540 06/06/99 1009 10/10/95 998 11/06/02 5700 12/19/98 719 04/09/95 2750 09/20/02 1310 07/10/98 1180 12/08/94 877 04/05/02 975 04/21/98 3550 06/07/94 1462 10/18101 840 10117/97 992 12/05/93 4000 03/22/01 1092 05/09/97 1240 11/26/93 4960 10/22/00 1317 10/05/96 1305 06/18/93 635 05/06/00 893 05/30/96 1353 12/11/92 **6000 09/23/99 769 01/31/96 *99,999/1224 1

  • Initial failure assigned value of 99,999 - after corrective maintenance of one of the two penetration valves (stop nut loose) value of 1224 was measured.

Mohan Thadani - T04090.doc Page 14 1l Mohan Thadani T04090.doc

- Paael4 II Attachment to TXX-04090 Page 14 of 12 is in SCCM):

Date Leakage Date Leakage Date Leakage 03/21/03 520 04/16/99 842 11/22/96 490 10/03/02 480 04/02199 291 05/02/96 667 04/30/02 682 03/14/99 320 04/06/96 *981 09/21/01 2 02/06/99 410 12/11/95 652 03/18/01 675 07/26/98 772 06/05/95 733 10/27/00 425 02/03/98 660 11/14/94 766 06/25/00 674 12/01/97 833 08/12/94 569 01/22/00 104 10/10/97 456 05/12/94 196 07/25/99 313 05/02/97 672 12/14/93 803

  • High value for this penetration Unit 1Penetration MV-14 (18" penetration, Admin. limit of 15,100 SCCM) Data (Leakage is in SCCM):

Date Leakage Date Leakage Date Leakage 05/10/03 200 01/21/01 209 12/13/98 339 02/14/03 235 11/18/00 330 09/20/98 298 08/30/02 689 08/06/00 317 06/27/98 295 06/09/02 260 05/12/00 279 03/24/98 295 03/16/02 290 02/19/00 195 01/09/98 219 12/23/01 326 11/28/99 1152 10/17/97 251 09/28/01 390 10/21/99 *1320 08/01/97 260 07/08/01 330 09/03/99 419 05/02/97 79 05/06/01 244 03/14/99 187 01/31/97 279

Date Leakage Date Leakage Date Leakage 06/06/03 1717 05/12/01 1735 04/23/99 1875 03/16/03 2100 02/18/01 2440 01/24/99 3000 12/21/02 2600 11/26/00 2210 11/01/98 1869 09/28/02 1650 09/03/00 1775 08/15/98 2220 07/06/02 1430 06/10/00 1730 05/10/98 2240 04/04/02 2260 03/19/00 2530 10/26/97 2720 01/20/02 2810 01/08/00 *3330 10/10/97 2010 10/27/01 1618 10/02/99 1910 07/11/97 1890 08/04/01 1560 07/11/99 1447 04/11/97 2480

  • High value for this penetration A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway - Comanche Peak
  • Diablo Canyon
  • Palo Verde - Soulh Texas Project - Wolf Creek

Mohan Thadani - T04090.doc Page 15 Attachment to TXX-04090 Page 15 of 12

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