DCL-08-008, License Amendment Request 08-01, Revision to Technical Specification 5.5.16, Containment Leakage Rate Testing Program

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License Amendment Request 08-01, Revision to Technical Specification 5.5.16, Containment Leakage Rate Testing Program
ML080390454
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 02/01/2008
From: Becker J
Pacific Gas & Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-08-008, LAR 08-01
Download: ML080390454 (41)
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Text

PacificGas and Electric Company' James R. Becker Diablo Canyon Power Plant Vice President P. 0. Box 56 Diablo Canyon Operations and Avila Beach, CA 93424 Station Director February 1, 2008 '

805.545.3462 Fax: 805.545.4234 PG&E Letter DCL-08-008 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 License Amendment Request 08-01 Revision to Technical Specification 5.5.16, "Containment Leakage Rate Testing Program"

Dear Commissioners and Staff:

In accordance with 10 CFR 50.90, enclosed is a License Amendment Request (LAR) for Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively. The enclosed LAR proposes to revise Technical Specification (TS) 5.5.16.a to add an exception to Regulatory Guide 1.163 to allow use of Standard ANSI/ANS 56.8-2002, and to revise TS 5.5.16.b to specify both a lower peak calculated containment internal pressure following a large-break loss-of-coolant accident (LOCA) and the containment design pressure.

The proposed revision to TS 5.5.16.a will allow performance of Types A, B, and C containment leak-tests in accordance with the guidance provided in ANSI/ANS-56.8-2002. Use of the 2002 standard clarifies requirements in the 1994 version and is expected to result in the performance of fewer Type C as-found tests for those penetrations that require testing on a fixed refueling outage frequency at DCPP.

The proposed revision to TS 5.5.16.b will bring the specification into conformance with the NUREG-1431, Revision 3, Standard TS 5.5.16.b, Option B, which will allow all DCPP containment leak-rate tests to be performed at a slightly lower pressure, and will also provide a wider acceptable pressure test range for containment integrated leak-rate tests.

The proposed changes to TS 5.5.16.b in this LAR are consistent with similar changes approved by the NRC for Entergy's Indian Point Unit 3 in Amendment 225 to Facility Operating License No. DPR-64, "Indian Point Nuclear Generating Unit A member of the STARS (Strategic Teaming and Resource Sharing) Atliance "P F7 Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde ° South Texas Project
  • Wolf Creek ., /,,')

Document Control Desk PG&E Letter DCL-08-008 February 1,2008 Page 2 No. 3 - Issuance of Amendment Re: 4.85 Percent Stretch Power Uprate and Relocation of Cycle-Specific Parameters," dated March 24, 2005, and Union Electric's Callaway Plant Unit 1 in Amendment 168 to Facility Operating License No. NPF-30, "Callaway Plant, Unit 1 - Issuance of Amendment Regarding the Steam Generator Replacement Project," dated September 29, 2005. contains a description of the proposed changes, the supporting technical analyses, and the no significant hazards consideration determination.

Enclosures 2 and 3 contain marked-up and retyped (clean) TS pages, respectively.

There are no TS Bases changes required for this LAR. Enclosure 4 contains, "For Information Only," the DCPP Final Safety Analysis Report Update (FSARU) section that describes the containment integrity analysis performed for the large-break LOCA using the GOTHIC code supporting installation of the replacement steam generators.

Pacific Gas and Electric Company (PG&E) has determined that this LAR does not involve a significant hazard con7sideration as determined per 10 CFR 50.92.

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The changes in this LAR are not required to address an immediate safety concern.

PG&E requests approval of this LAR by January 23, 2009, to support performance of leak-rate tests during the Unit 1 Fifteenth Refueling Outage, which is scheduled to begin on January 26, 2009. PG&E requests the license amendment(s) be made effective upon NRC issuance, to be implemented within 120 days.

There are no new or revised regulatory commitments in this letter.

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

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

Executed on January 31, 2008.

Sincerely James R. Be ker Vice President- Diablo Canyon Operationsand Station Director A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

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Document Control Desk PG&E Letter DCL-08-008 February 1,2008 Page 3 kjse/4328 Enclosures cc: Gary W. Butner, DPH Elmo E. Collins, NRC Region IV Michael S. Peck, DCPP NRC Senior Resident Inspector Diablo Distribution cc/enc: Alan B. Wang, NRC Project Manager A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
  • Palo Verde
  • Wolf Creek

Enclosure 1 PG&E Letter DCL-08-008 EVALUATION

1.0 DESCRIPTION

This letter is a request to amend Operating Licenses DPR-80 and DPR-82 for Units I and 2 of the Diablo Canyon Power Plant (DCPP), respectively.

The proposed changes would revise Technical Specification (TS) 5.5.16, "Containment Leakage Rate Testing Program," Part a, to add Exception 4 to Regulatory Guide 1.163 that would specify performance of Types A, B, and C containment leak-tests in accordance with ANSI/ANS-56.8-2002 (Reference 1),

and Part b, to specify the peak calculated containment internal pressure, Pa, as 43.5 pounds per square inch gage (psig) following a large-break loss-of-coolant accident (LOCA) in Units 1 and 2, and to specify the containment design pressure as 47 psig.

The proposed change to TS 5.5.16.a to allow performance of containment leak-rate tests in accordance with ANSI/ANS-56.8-2002 is expected to reduce the number of Type C as-found leak-rate tests for those penetrations that require testing on a fixed refueling outage frequency at DCPP. The proposed change to TS 5.5.16.b will bring the specification into conformance with the NUREG-1431, Revision 3, Standard TS (STS) 5.5.16.b, Option B (Reference 2). The revised TS 5.5.16.b will allow Type A, B, and C leak-rate tests in DCPP Units 1 and 2 to be performed at a lower pressure and provide a wider acceptable pressure test range for containment integrated leak-rate tests (ILRTs) without exceeding the containment design pressure.

2.0 PROPOSED CHANGE

S TS 5.5.16.a currently specifies performance of the leakage rate testing program in accordance with the guidelines of Regulatory Guide 1.163 (Reference 3),

which refers to use of ANSI/ANS-56.8-1994 (Reference 5) for guidance related to test methods and techniques. This License Amendment Request (LAR) proposes to add Exception 4 to TS 5.5.16.a to specify Types A, B, and C containment leak-tests will be performed in accordance with the guidance provided in ANSI/ANS-56.8-2002 (Reference 1).

TS 5.5.16.b currently states:

"The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa, is 47 psig."

This LAR proposes to change TS 5.5.16.b to the following:

1

Enclosure 1 PG&E Letter DCL-08-008 "The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa, is 43.5 psig. The containment design pressure is 47 psig."

The proposed change to TS 5.5.16.b is consistent with that recommended in STS 5.5.16.b, Option B. The actual peak calculated containment internal pressure is 41.4 psig for DCPP Unit 2 (see analysis in Enclosure 4), which bounds the actual calculated value of 41.2 psig for DCPP Unit 1. To ensure the peak calculated containment internal pressure value specified in TS 5.5.16.b contains adequate margin to the Final Safety Analysis Report Update (FSARU) containment internal pressure analysis value, a conservative value of 43.5 psig is proposed as the peak calculated containment internal pressure in TS 5.5.16.b for Units 1 and 2.

The containments for Units 1 and 2 are designed and constructed to withstand a maximum internal pressure of 47 psig, which is specified as the containment design pressure in the proposed TS 5.5.16.b.

The proposed changes to TS 5.5.16 are shown in the marked-up TS page provided in Enclosure 2. The proposed retyped TS are provided in Enclosure 3.

There are no TS Bases changes required as a result of this proposed TS change.

3.0 BACKGROUND

Purpose for Chanqe TS 5.5.16 describes the program for leakage rate testing of the DCPP containments as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The performance-based leak-test program, test methods, procedures, and analyses developed by DCPP to meet the requirements of Appendix J, Option B, comply with the guidance provided in Regulatory Guide 1.163 (Reference 3), with exceptions as specified in TS 5.5.16.a. Regulatory Guide 1.163 endorses NEI 94-01 (Reference 4),

subject to several conditions related to test intervals, visual examinations, and valve testing. Regulatory Guide 1.163, in turn, references ANSI/ANS-56.8-1994 (Reference 5) for detailed descriptions of the technical methods and techniques for performing Type A, B, and C tests under Appendix J to 10 CFR 50.

This LAR proposes to add Exception 4 to TS 5.5.16.a to specify performance of leak-tests in accordance with ANSI/ANS-56.8-2002. The 2002 version of ANSI/ANS-56.8 clarifies requirements contained in the 1994 version. It also consolidates the requirements detailed in NEI 94-01, Revision 0, and Regulatory Guide 1.163 into one standard, and corrects the NRC exceptions to NEI 94-01, which are listed in Regulatory Guide 1.163. It is expected that use of the 2002 2

Enclosure 1 PG&E Letter DCL-08-008 standard will result in fewer Type C as-found tests for those penetrations that require testing on a fixed refueling outage frequency at DCPP.

Regarding TS 5.5.16.b, the current DCPP specification states that the peak calculated containment internal pressure for the design basis LOCA, Pa, is 47 psig, which is the containment design pressure value. The use of the containment design pressure as Pa in TS 5.5.16.b is very conservative since it results in containment leak-rate testing being performed at a pressure well above the current peak calculated containment internal pressure of 41.53 psig following the design basis LOCA, as delineated in the FSARU. In accordance with Section 3.2.11 of ANSI/ANS 56.8-1994, compliance with this TS for an ILRT requires that the ILRT test pressure shall not be less than 0.96 Pa (45.12 psig) nor exceed P-design (47 psig).

Pacific Gas and Electric Company (PG&E) is concerned that meeting the requirement to test in the range of 0.96 - 1.0 Pa (45.12 - 47 psig) in future ILRTs could cause refueling outage durations to be extended because of the narrow acceptable test pressure range, which makes it more likely the containment pressure will fall outside the acceptable pressure range during testing. Also, testing at higher than necessary pressures unnecessarily challenges the containment. The proposed DCPP TS 5.5.16.b change to specify Pa at 43.5 psig and P-design at 47 psig is intended to incorporate values based on the FSARU containment internal pressure analysis and containment design pressure, and to reduce the outage duration risk due to meeting the acceptable test range for ILRTs. Also, the change will have the added benefit of allowing Type B and C leak-rate tests to be performed at a lower pressure.

Containment Internal Pressure Analyses The current FSARU analysis of record for the peak containment internal pressure following the design basis LOCA with the original Model 51 steam generators (SGs) is contained in Appendix 6.2C of the DCPP FSAR Update. Using the COCO digital computer code, the peak containment internal pressure was calculated to be 41.53 psig.

The Model 51 SGs will be replaced with Model Delta 54 SGs, termed replacement SGs, in the Unit 2 Fourteenth Refueling Outage, currently scheduled to begin in February 2008, and in the Unit 1 Fifteenth Refueling Outage, currently scheduled to begin in January 2009. The TS,change proposed in this LAR will be applicable once installation of the Model Delta 54 SGs is completed in both units.

As described below, a new containment internal pressure analysis has been performed using the GOTHIC code to support installation of the Model Delta 54 SGs. The GOTHIC analysis is being implemented for DCPP under the 3

Enclosure 1 PG&E Letter DCL-08-008 provisions of 10 CFR 50.59, and is described for information only in Enclosure 4 of this LAR. The basis for implementation under 10 CFR 50.59 is that the GOTHIC methodology has been accepted by the NRC (Reference 8). As such, PG&E is not requesting NRC approval of the revised containment pressure analysis using the GOTHIC code.

4.0 TECHNICAL ANALYSIS

The 2002 version of ANSI/ANS-56.8 (Reference 1) has been revised to consolidate into one document the guidelines for containment leak-rate testing under Option B, "Performance-Based Requirements," of 10 CFR 50, Appendix J.

The 2002 standard also corrects the four exceptions to NEI 94-01, Revision 0, listed in Regulatory Guide 1.163. The technical bases for the testing guidelines are contained within the text of the standard and are not reiterated in this LAR.

One of the areas where additional detail is provided in ANSI/ANS-56.8-2002 is for test interval extensions. For Type A tests described in Section 3.2.13.3, it states that the containment ILRT test interval may be extended by up to 15 months. This extension should be used only in cases where refueling schedules have been changed to accommodate other factors. In addition, for Type B and C tests described in Section 3.3.4.1, it states that the test interval may be extended by no more than 25 percent, not to exceed 15 months, due to unplanned or unusual events that necessitate such an extension. The scheduling guidance was less restrictive in ANSI/ANS-56.8-1994 and NEI 94-01, Revision 0.

Regarding the proposed changes to TS 5.5.16.b, a new containment integrity analysis was performed for the DCPP replacement SG program to evaluate the bounding peak pressure and temperature of a design basis LOCA inside containment, and to demonstrate the ability of the containment heat removal systems to mitigate the accident. The analysis is described in Enclosure 4 of this document, which contains the DCPP FSARU section that will be implemented to support the replacement SGs. A brief summary of that analysis and its results are provided below:

Calculation of the containment response following a postulated LOCA was analyzed by use of the digital computer code GOTHIC version 7.2. The GOTHIC Technical Manual (Reference 6) provides a description of the governing equations, constitutive models, and solution methods in the solver. The GOTHIC Qualifications Report (Reference 7) provides a comparison of the solver results with both analytical solutions and experimental data.

The GOTHIC modeling for DCPP is consistent with the NRC-approved Kewaunee evaluation model (Reference 8). Kewaunee and DCPP both have large dry containment designs with similar active heat removal capabilities. The 4

Enclosure 1 PG&E Letter DCL-08-008 latest code version is used to take advantage of the diffusion layer model heat transfer option. This heat transfer option was approved by the NRC (Reference 8) for use in the Kewaunee analyses with the condition that the effect of mist be excluded from what was earlier termed as the mist diffusion layer model. The GOTHIC containment modeling for DCPP has followed the conditions of acceptance placed on the Kewaunee evaluation model. The differences in GOTHIC code versions are documented in Appendix A of the GOTHIC User Manual Release Notes (Reference 9). Version 7.2 of GOTHIC is used consistently with the restrictions identified in Reference 8; none of the user-controlled enhancements added to Version 7.2 were implemented in the DCPP containment model. A description of the DCPP GOTHIC model is contained in Enclosure 4 of this document, including:

" Input parameters and assumptions

" Acceptance criteria

  • Description of analyses and evaluations o Noding structure o Volume input o Initial conditions o Flow paths o Heat sinks o Heat and mass transfer correlation o Sump recirculation
  • Boundary conditions o Mass and energy release o Containment fan coolers o Containment spray system o Accumulator nitrogen gas modeling The results of the GOTHIC analysis show that the analysis margin is maintained for DCPP with replacement SGs (analysis margin is the difference between the peak calculated pressure and temperature and the acceptance limits). The peak calculated containment internal pressure using the GOTHIC code is 41.2 psig for Unit 1, and 41.4 psig for Unit 2. (Previously, the peak containment pressure following a double ended hot-leg [DEHL] LOCA with the original SGs was calculated to be 41.53 psig using the COCO code.) To ensure the peak calculated containment internal pressure value specified in TS 5.5.16.b contains adequate margin to the FSAR Update analysis value, a value of 43.5 psig is proposed as the peak calculated containment internal pressure in TS 5.5.16.b for both Units 1 and 2. This margin also allows for a small increase in the FSARU 5

Enclosure 1 PG&E Letter DCL-08-008 containment internal pressure analysis calculated pressure without requiring a TS 5.5.16.b change to be made. Changes to containment internal pressure analysis calculated pressure can occur due to plant design modifications, plant operational parameter changes, and correction of errors that may be identified in the analysis. The use of a TS 5.5.16.b Pa, which includes 4 percent margin in the Pa value with respect to the actual analysis peak calculated containment pressure (analysis value of 46.25 psig versus TS 5.5.16.b Pa value of 48.1 psig),

has been previously approved for the Callaway plant in Amendment 168 to Facility Operating License No. NPF-30, "Callaway Plant, Unit 1 - Issuance of Amendment Regarding the Steam Generator Replacement Project," (TAC No.

MC4437) dated September 29, 2005 (Reference 11).

The proposed TS 5.5.16.b value of 43.5 psig does not contain instrument uncertainties, and is based on the FSARU containment internal pressure analysis calculated pressure with margin incljded. The measurement and test equipment instrument uncertainties that are applicable while performing containment leak-rate testing are incorporated into the test acceptance pressure values in the leak-rate testing test procedures, which ensures the TS 5.5.16.b peak calculated containment internal pressure value is bounded during all Type A, B, and C leak-rate testing.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change to TS 5.5.16.a adds an exception to Regulatory Guide 1.163 to specify use of Standard ANSI/ANS-56.8-2002, rather than ANSI/ANS-56.8-1994.

The proposed change to TS 5.5.16.b specifies both the peak calculated containment internal pressure with margin following a large-break LOCA and the containment design pressure.

6

Enclosure 1 PG&E Letter DCL-08-008 These changes only affect the applicable version of the standard (2002 in place of 1994) and the test pressures for containment leak-rate tests, and do not involve the modification of any plant equipment or have any affect on plant operation. The changes are made based on the safety analysis and containment design, and do not have any adverse affect on accidents previously evaluated.

N Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different accident from any accident previously evaluated?

Response: No.

The proposed changes do not involve a physical alteration to the plant or a change in the methods governing normal plant operation. The changes are made based on the safety analysis and containment design, and do not affect any previously evaluated accidents.

Therefore, the proposed change does not create the possibility of a new or different accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by these changes, and the changes will not result in plant operation in a configuration outside the design basis.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluation, PG&E concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

7

Enclosure 1 PG&E Letter DCL-08-008 5.2 Applicable Requlatory Requirements/Criteria 10 CFR 50, Appendix A, General Design Criterion 16, "Design," requires that reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment, and to assure that the containment design conditions important to safety are not exceeded for as long as the postulated accident conditions require.

A program for leakage rate testing of containments is required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The DCPP performance-based leak-test program, test methods, procedures, and analyses meet the requirements of Appendix J, Option B, and comply with the guidance of Regulatory Guide 1.163 (Reference 3). Regulatory Guide 1.163 endorses NEI 94-01 (Reference 4), subject to several TS 5.5.16.a exceptions related to test intervals and visual examinations. Regulatory Guide 1.163, in turn, references ANSI/ANS 56.8-1994 (Reference 5) for detailed descriptions of the technical methods and techniques for performing Types A, B, and C tests under Appendix J to 10 CFR 50.

The proposed changes revise the applicable testing standard from the 1994 to the 2002 version of ANSl/ANS-56.8, and specify pressure conditions used to perform tests to verify leak-tightness of the containment. The changes do not physically affect the leak-tightness of the containment, and are consistent with 10 CFR 50.54(o) and 10 CFR 50 Appendix J Option B.

In conclusion, based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the NRC's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

PG&E has evaluated the proposed amendment and has determined that it does not involve: (1) a significant hazards consideration, (2) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (3) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or 8

Enclosure 1 PG&E Letter DCL-08-008 environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 References

1. American National Standard ANSI/ANS-56.8-2002, "Containment System Leakage Testing Requirements."
2. NUREG-1431, Revision 3.0, "Standard Technical Specifications Westinghouse Plants," dated March 31, 2004.
3. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.
4. Nuclear Energy Institute (NEI) Report NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," dated July 26, 1995.
5. American National Standard ANSI/ANS 56.8-1994, "Containment System Leakage Testing Requirements."
6. NAI 8907-06, Revision 15, "GOTHIC Containment Analysis Package Technical Manual, Version 7.2," dated September 2004.
7. NAI 8907-09, Revision 8, "GOTHIC Containment Analysis Package Qualification Report," dated September 2004.
8. License Amendment No. 169 for the Kewaunee Nuclear Power Plant, "Authorizing Changes to the UFSAR to Allow Use of an Upgraded Computer Code (GOTHIC) for Design-Basis Accident Containment Integrity Analysis," Operating License No. DPR-43 (TAC No.

MB6408), dated September 29, 2003.

9. NAI 8907-02, Revision 16, "GOTHIC Containment Analysis Package User Manual, Version 7.2," dated September 2004.
10. Entergy Nuclear Northeast letter No. NL-04-069, "Proposed Changes to Technical Specifications: Stretch Power Uprate (4.85%) and Adoption of TSTF-339," for Indian Point Unit No. 3, Docket No. 50-286, dated June 3, 2004, approved by the NRC in Amendment 225 to Facility Operating License No. DPR-64, "Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: 4.85 Percent Stretch Power Uprate and Relocation of Cycle-Specific Parameters,"

(TAC No. MC3552) dated March 24, 2005.

11. Union Electric letter ULNRC-05056, "Technical Specification Revisions Associated with the Steam Generator Replacement Project," for Callaway, Docket No. 50-483, dated September 17, 2004, approved by the NRC in Amendment 168 to Facility Operating License No. NPF-30, "Callaway Plant, Unit 1 - Issuance of Amendment Regarding the Steam Generator Replacement Project (TAC No.

MC4437)", dated September 29, 2005.

9

Enclosure 1 PG&E Letter DCL-08-008 7.2 Precedent The change to TS 5.5.16.b proposed in this LAR to include Pa and the containment design pressure is similar to a part of the TS 5.5.16.b changes proposed by Entergy Nuclear Northeast in their LAR titled, "Proposed Changes to Technical Specifications: Stretch Power Uprate (4.85%) and Adoption of TSTF-339," for Indian Point Unit No. 3, Docket No. 50-286, Letter No. NL-04-069, dated June 3, 2004. The Indian Point Unit 3 LAR was approved by the NRC in Amendment 225 to Facility Operating License No. DPR-64, "Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: 4.85 Percent Stretch Power Uprate and Relocation of Cycle-Specific Parameters," (TAC No. MC3552) dated March 24, 2005.

The proposed change to the TS 5.5.16.b Pa in this LAR, which includes margin in the Pa value with respect to the actual analysis peak calculated containment pressure, is similar to a change proposed by Union Electric for the Callaway plant in their LAR titled, "Technical Specification Revisions Associated with the Steam Generator Replacement Project,"

Docket No. 50-483, Letter ULNRC-05056, dated September 17, 2004.

The Callaway LAR, included a revised peak calculated containment internal pressure analysis value of 46.25 psig which was less than the TS 5.5.16.b Pa value of 48.1 psig (4 percent margin-to the TS 5.5.16.b Pa value), and a change to the TS 5.5.16.b Pa value of 48.1 was not proposed. This was approved by the.NRC in Amendment 168 to Facility Operating License No. NPF-30, "Callaway Plant, Unit 1 - Issuance of Amendment Regarding the Steam Generator Replacement Project," (TAC No. MC4437) dated September 29, 2005.

10 J

Enclosure 2 PG&E Letter DCL-08-008 Proposed Technical Specification Changes (marked-up)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exceptions:
1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.
2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by ASME Section XI code, Subsection IWE, except where relief has been authorized by the NRC.
3. The ten-year interval between performance of the integrated leakage rate (Type A) test, beginning May 4, 1994, for Unit 1 and April 30, 1993, for Unit 2, has been extended to 15 years.
4. Types A, B, and C containment leak tests will be performed in accordance with the guidance provided in ANSI/ANS-56.8-2002.
b. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 43.5 47 psig. The containment design pressure is 47
c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day (continued)

DIABLO CANYON - UNITS 1 & 2 5.0-24 Unit 1 - Amendment No. 4-35, 4-50, 1-72, 1-97, Unit 2 - Amendment No. 1-35, 1-50, 4-74, 1-98,

Enclosure 3 PG&E Letter DCL-08-008 Proposed Technical Specification Changes (retyped)

Remove Page Insert Page 5.0-24 5.0-24

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exceptions:
1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.
2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by ASME Section XI code, Subsection IWE, except where relief has been authorized by the NRC.
3. The ten-year interval between performance of the integrated leakage rate (Type A) test, beginning May 4, 1994, for Unit 1 and April 30, 1993, for Unit 2, has been extended to 15 years.
4. Types A, B, and C containment leak tests will be performed in accordance with the guidance provided in ANSI/ANS-56.8-2002.
b. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 43.5 psig. The containment design pressure is 47 psig.
c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.10% of containment air weight per day.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-24 Unit 1 - Amendment'No. 4-35, 451, 4-7-2, 4-97, Unit 2 - Amendment No. 4-35, -5G, 47-4, 1-98,

Enclosure 4 PG&E Letter DCL-08-008 LOCA Containment Integrity Analysis Using GOTHIC Code Diablo Canyon Power Plant Final Safety Analysis Report Update Section to be Implemented to Support Steam Generator Replacement (For Information Only)

(

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 6.X.1 CONTAINMENT INTEGRITY ANALYSIS 6.X.1.1 Loss-of-Coolant Accident The purpose of the LOCA containment integrity analysis performed to support the DCPP replacement steam generator program is to evaluate the bounding peak pressure and temperature of a design basis LOCA event inside containment and to demonstrate the ability of the containment heat removal systems to mitigate the accident. The impact of LOCA mass and energy releases on the containment pressure and temperature are assessed to ensure that the containment pressure and temperature remain below their respective design limits. The containment heat removal systems must also be capable of maintaining the environmental qualification (EQ) parameters to within acceptable limits.

The DCPP LOCA containment response analysis considers a spectrum of cases that address differences between the individual DCPP Units, LOCA break locations, and postulated single failures (minimum and maximum safeguards). The limiting cases that address the containment peak pressure case and limiting long-term EQ temperature are presented in this section.

Calculation of the containment response following a postulated LOCA was analyzed by use of the digital computer code GOTHIC version 7.2. The GOTHIC Technical Manual (Reference 1) provides a description of the governing equations, constitutive models, and solution methods in the solver. The GOTHIC Qualifications Report (Reference 2) provides a comparison of the solver results with both analytical solutions and experimental data.

The GOTHIC containment modeling for Diablo Canyon is consistent with the NRC approved Kewaunee evaluation model (Reference 3). Kewaunee and Diablo Canyon both have large dry containment designs with similar active heat removal capabilities. The latest code version is used to take advantage of the diffusion layer model heat transfer option. This heat transfer option was approved by the NRC (Reference 3) for use in Kewaunee containment analyses with the condition that the effect of mist be excluded from what was earlier termed as the mist diffusion layer model. The GOTHIC containment modeling for Diablo Canyon has followed the conditions of acceptance placed on Kewaunee. The differences in GOTHIC code versions are documented in Appendix A of the GOTHIC User Manual Release Notes (Reference 4). Version 7.2 is used consistently with the restrictions identified in Reference 15; none of the user-controlled enhancements added to version 7.2 were implemented in the Diablo Canyon containment model. A description of the Diablo Canyon GOTHIC model is provided later in this section.

6.X.1.1.1 Input Parameters and Assumptions The major modeling input parameters and assumptions used in the DCPP LOCA containment evaluation model are identified in this section. The assumed initial conditions and input assumptions associated with the fan coolers and containment sprays are listed in Table 6.X-1.

The containment spray flow data used in the analysis are presented in Table 6.X-2. The primary function of the residual heat removal system (RHR) is to remove heat from the core by way of the ECCS. The recirculation system and CCW system parameters are outlined in 1

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Table 6.X-1. The containment structural heat sink input is provided in Table 6.X-3, and the corresponding material properties are listed in Table 6.X-4.

The LOCA containment analysis described here uses revised input and assumptions in support the DCPP replacement steam generator program, while addressing analytical conservatisms.

The following summarized assumptions are areas where known differences exist between the current licensing analysis and the replacement steam generator program containment integrity analysis.

The mass and energy releases are calculated specifically for the DCPP plant conditions with replacement steam generators.

Decay heat steaming mass and energy release rates, after the end of the sensible heat release from the RCS and steam generators, are calculated each time step by GOTHIC using the transient containment pressure and recirculation safety injection water temperature.

Non-condensable accumulator gas addition is modeled in the GOTHIC model; no accumulator gas addition is considered in the FSAR Update current licensing analysis.

A recirculation system model that couples the RHR, CCW, CFCUs and auxiliary service water systems was developed for the replacement steam generator program. More detailed accounting of CCW flow rates through the containment heat removal systems was used for the CFCUs, RHR heat exchangers, and miscellaneous CCW heat loads.

The DCPP LOCA containment response analysis considered a spectrum of cases for the replacement steam generator program. The cases address break locations, and postulated single failures (minimum and maximum safeguards) for each DCPP unit. Only the limiting cases, which address the containment peak pressure and limiting long-term EQ temperature, are presented in this section. The LOCA pressure and temperature response analyses were performed assuming a loss of offsite power and a worst single failure (loss of one solid state protection system [SSPS] train, i.e., loss of one containment cooling train). The active heat removal available in the long term cooling case is:

0 One containment spray pump during injection-phase only

  • Two containment fan cooler units
  • One RHR pump and one RHR heat exchanger
  • Two CCW pumps and one CCW heat exchanger

The double-ended hot leg break produces the peak pressure at the end of the blowdown. The calculation for the DEPS case was performed for a 116 day (lx1 07 second) transient in support of long term EQ temperatures. The sequence of events for the DEHL containment peak pressure case is shown in Table 6.X-5 and the DEPS long term EQ temperature case for Units 1 and 2 is shown in Tables 6.X-6 and 6.X-7, respectively.

2

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 6.X.1.1.2 Acceptance Criteria The containment response for design basis LOCA containment integrity is an ANS Condition IV event, an infrequent fault. The relevant requirements to satisfy NRC acceptance criteria are as follows:

GDC-16 and -50: In order to satisfy the requirement of GDC-16 and -50, the peak calculated containment pressure should be less than the containment design pressure of 47 psig.

GDC-38: In order to satisfy the requirement of GDC-38, the calculated pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> should be less than 50 percent of the peak calculated value. (This is related to the criteria for containment leakage assumptions as affecting doses at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.)

6.X.1.1.3 Description of Analyses and Evaluations Noding Structure The Diablo Canyon GOTHIC containment model is comprised of one control volume with separate vapor and liquid regions. Mass and energy releases, containment spray injection, and sump water recirculation are modeled using boundary conditions. A cooler component is used to model containment fan cooler units (CFCUs) heat removal. Injection of accumulator nitrogen during the event is modeled with a boundary condition.

The component cooling water system model is comprised of three control volumes (CFCU cooling water, the hot side of the CCW system, and the cold side of the CCW system) and uses GOTHIC component models for the RHR and CCW heat exchangers. A heater component models the CFCU heat transfer to the CCW water. Boundary conditions model the CCW flow through the CFCUs, RHR heat exchangers, and miscellaneous CCW heat loads.

Volume Input Values for the volume, height, hydraulic diameter, and elevation are input for each node. The containment is modeled as a single control volume. The lower bound free volume is 2,550,000 ft3 . The hydraulic diameter, height, and floor elevation input values are 24.1 ft, 166 ft, and 91 ft, respectively.

A conservatively calculated pool surface area is used to model interfacial heat and mass transfer to liquid pools on the various floor surfaces in the containment volume. The conductor representing the floor is essentially insulated from the vapor region after the sump pool develops; however, there can still be condensation or evaporation from the surface of the liquid pools. Using this method to model the interfacial heat and mass transfer between the pools and the atmosphere was previously approved by the NRC for the Kewaunee containment DBA and equipment qualification analyses (Reference 3).

3

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Initial Conditions The containment initial conditions for containment integrity cases are:

  • Pressure: 16.0 psia
  • Relative Humidity: 18 percent
  • Temperature: 120°F The LOCA containment response model contains volumes representing the CCW system. The system volumes are water solid and assumed to be initially at 50 psia and 90 0F.

Flow Paths Flow boundary conditions linked to functions that define the mass and energy release model the LOCA break flow to the containment. The boundary conditions are connected to the containment control volume via flow paths. The containment spray is modeled as a boundary condition connected to the containment control volume via a flow path.

The flow rates through the flow paths are specified by the boundary conditions, so the purpose of the flow path is to direct the flow to the proper control volume. The flow path input is mostly arbitrary. Standard values are used for the area, hydraulic diameter, friction length, and inertia length of the flow path. Since this is a single volume lumped parameter model, the elevation of the break flow paths is arbitrarily set to 100 ft and the elevation of the spray flow paths is arbitrarily set to 70 ft above the containment floor.

Heat Sinks The structural heat sinks in the containment are modeled as GOTHIC thermal conductors. The heat sink geometry data is based on conservatively low surface areas and is summarized in Table 6.X-3. A thin air gap is assumed to exist between the steel and concrete for steel-jacketed heat sinks. A gap conductance of 10 Btu/hr/ft 2/OF is conservatively assumed between steel and concrete. The volumetric heat capacity and thermal conductivity for the heat sink materials are summarized in Table 6.X-4.

Heat and Mass Transfer Correlation GOTHIC has several heat transfer coefficient options,that can be used for containment analyses. For the Diablo Canyon GOTHIC model, the direct heat transfer coefficient set is used with the diffusion layer model mass transfer correlation for the heat sinks inside containment.

This heat transfer methodology was reviewed by the NRC and approved for use in containment DBA analyses in the Kewaunee analysis (Reference 3). The diffusion layer model correlation does not require the user to specify a revaporization input value, as was done in previous analyses using the Uchida correlation.

Split heat transfer coefficients are used for the heat sinks representing walls and floors. The split coefficient allows one thermal conductor to model heat transfer to both the water and vapor regions. The submerged portions of conductors are essentially insulated from the vapor after the pool develops. The fraction of the wall that is not submerged uses the vapor heat transfer 4

DCPP UNITS-1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 coefficient as described above. GOTHIC calculates the fraction of the walls that are submerged in the sump water. The floors are submerged quickly.

Sump Recirculation The calculated containment peak pressure and temperature occur before the transfer to cold leg recirculation. However, a sump recirculation model comprised of simplified RHR and CCW system models was added to the Diablo Canyon containment model for the long-term LOCA containment pressure and temperature response calculation.

The recirculation system is actuated after a low RWST level signal and the ECCS takes suction from the containment sump. The RHR heat exchanger cools the water before it is injected back into the reactor vessel. The RHR heat exchanger is cooled by CCW water and service water provides the ultimate heat sink, cooling the CCW heat exchangers.

Switchover to hot leg recirculation is assumed to occur at 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

6.X.1.1.4 Boundary Conditions Mass and Energy Release The LOCA mass and energy release rates are generated using the Westinghouse methodology (Reference 6). Mass and energy releases are calculated for both sides of the double-ended break in the coolant loop: the vessel side of the break and the steam generator side of the break. The mass and energy releases are input to the GOTHIC containment model as mass flow rates and enthalpies via boundary conditions connected to the containment volume with flow paths.

During blowdown, the liquid portion of'the break flow is released as drops with an assumed diameter of 100 microns (0.00394 inches). This is consistent with the methodology approved for Kewaunee (Reference 3) and is based on data presented in Reference 5. After blowdown, the liquid release is assumed to be a continuous pour into the sump.

GOTHIC uses the mass and energy release tables from the time of accident initiation to 3,600 seconds, the time at which all energy in the primary heat structures and steam generator secondary system is'assumed to be released/depressurized to atmospheric pressure, (i.e., 14.7 psia and 212'F). After primary system and secondary system energy have been released, the mass and energy releases to the containment are due to long-term steaming of decay heat.

GOTHIC calculates the decay heat steaming mass and energy releases within user defined control variables. The steaming calculations incorporate the transient containment pressure and RHR recirculated ECCS enthalpy to calculate the mass and energy release. The calculations are essentially the same as the Westinghouse methodology previously approved by the NRC, except the calculations are performed within the GOTHIC code.

The ANS Standard 5.1 decay heat model (+2a uncertainty) is used to calculate the long-term boil-off from the core. All the decay heat is assumed to produce steam from the recirculated ECCS water. The remainder of the ECCS water is returned to the sump region of the 5

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 containment control volume. These assumptions are consistent with the long-term mass and energy release methodology documented in Reference 6.

Containment Fan Coolers The CFCUs are modeled with a GOTHIC cooler component. There are a total of five CFCUs in three trains. In all cases, two CFCUs are assumed to be out of service for maintenance. An inherent assumption in the LOCA containment analysis is that offsite power is lost with the pipe rupture. This results in the actuation of three emergency diesel generators (EDGs), powering the two trains of safeguards equipment. Startup of the EDGs delays the operation of the safeguards equipment that is required to mitigate the transient. There are two trains of the SSPS that actuate the two trains of emergency safeguards. The failure of one train of SSPS will fail one train of safeguards. A minimum of two CFCUs are available and a maximum of three CFCUs are assumed to be available based on the single failure assumptions.

Three long term cases are analyzed to assess the effects of single failures. The first case assumes minimum safeguards based on the postulated single failure of an SSPS train. This assumption results in the loss-of-one train of safeguards equipment. The operating equipment is conservatively modeled as: two CFCUs, one containment spray pump, one train of RHR, and one CCW heat exchanger. The other two cases assume maximum safeguards, in which both trains of SSPS are available. With the maximum safeguards cases, the single failure assumptions are the failure of one containment spray pump or the failure of one CFCU. The analysis of these three cases provides confidence that the effect of credible single failures is bounded.

The fan coolers in the containment evaluation model are modeled to actuate on the containment high pressure setpoint with uncertainty biased high, (5 psig), and begin removing heat from containment after a 48-second delay.

The CFCUs are cooled by CCW. The heat removal rate per containment fan cooler is calculated as a function of containment steam saturation temperature, the CCW inlet temperature and flow rate, and input to the GOTHIC cooler model. The heat removal rate is multiplied by the number of CFCUs available. The heat removed from the containment control volume is transferred to the CCW control volume receiving the flow through the CFCUs using a coupled heater model.

Containment Spray System The containment spray is modeled with a boundary condition. DCPP has two trains of containment safeguards available, with one spray pump per train. An inherent assumption in the LOCA containment analysis is that offsite power is lost with the pipe rupture. This results in the actuation of the three EDGs powering the two trains of safeguards equipment. Startup of the EDGs delays the operation of the safeguards equipment that is required to mitigate the transient.

Relative to the single failure criterion with respect to a LOCA event, one spray pump is considered inoperable due to the SSPS failure (minimum safeguards case) or as a single failure 6

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 in a maximum safeguards case. In the maximum safeguards case, in which the single failure is assumed to be one CFCU, two spray pumps are available.

The containment spray actuation is modeled on the containment high-high pressure setpoint with uncertainty biased high (24.7 psig). The sprays begin injecting 90'F water after a specified 80 second delay. The spray flow rate is a function of containment pressure and is presented in Table 6.X-2. The containment spray is credited only during the injection phase of the transient and is terminated on a refueling water storage tank empty alarm after switchover to cold leg recirculation at a time based on the number of SI and spray pumps operating. The timing of recirculation and spray termination assumed in the LOCA containment analysis are presented in Table 6.X-1.

Accumulator Nitrogen Gas Modelinq The accumulator nitrogen gas release is modeled with a flow boundary condition in the LOCA containment model. The nitrogen release rate was conservatively calculated by maximizing the mass available to be injected. The nitrogen gas release rate was used as input for the GOTHIC function, as a specified rate over a fixed time period. Nitrogen gas was released to the containment at a rate of 327.4 Ibm/s. The release begins at 51.9 seconds, the minimum accumulator tank water depletion time.

6.X.1.1.5 LOCA Containment Integrity Results Plant input assumptions (identified in Section 6.X.1.1.1) are the same as, or slightly more restrictive, than in the licensing-basis analyses performed with the COCO code (Reference 7).

Benchmarking between the Diablo Canyon COCO and GOTHIC models was performed to confirm consistency in the implementation of the plant input values.

The containment pressure, steam temperature, and water (sump) temperature profiles of the DEHL peak pressure case are shown in Figures 6.X-1 through 6.X-3. Table 6.X-5 provides the transient sequence of events for the DEHL transient.

The containment pressure, steam temperature, and water (sump) temperature profiles of the DEPS long-term EQ temperature transient are shown in Figures 6.X-4 through 6.X-6 1. Tables 6.X-6 and 6.X-7 presents the sequence of events for the Unit 1 and Unit 2 DEPS transients, respectively. The peak pressure (Figure 6.X-4) for the DEPS case occurs at 24.1 seconds after the end of the blowdown. The fans begin to cool the containment at 48.7 seconds.

Containment sprays begin injecting at 88.01 seconds. The pressure comes down as the steam generators reach equilibrium with the containment environment, but spikes up again at recirculation when the CCW temperature increases and the CCW flow rate to the CFCUs decreases. The sensible heat release from the steam generator secondary system and RCS metal is completed at 3600 seconds, but at 3798 seconds, the RWST reaches a low level alarm and spray flow is terminated. The containment pressure increases for a time and then begins to decrease over the long term as the RHR heat exchangers and CFCUs remove the heat from the containment.

1The peak DEPS values are from Unit 2.

7

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Table 6.X-8 summarizes the containment peak pressure and temperature results and pressure and temperature at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for EQ support and the acceptance limits for these parameters.

A review of the results presented in Table 6.X-8 shows that the analysis margin (analysis margin is the difference between the peak calculated pressure and temperature and the acceptance limits) is maintained for Diablo Canyon with replacement steam generators. From the GOTHIC analysis performed in support of the Diablo Canyon replacement steam generator program the containment peak pressure is 41.4 psig. At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the maximum containment pressure is 8.9 psig and the maximum temperature is 167.540 F.

6.X.1.1.6 Conclusion The DCPP containment can adequately account for the mass and energy releases that would result from the replacement steam generator program. The DCPP containment systems will continue to provide sufficient pressure and temperature mitigation capability to ensure that containment integrity is maintained. The containment systems and instrumentation will continue to be adequate for monitoring containment parameters and release of radioactivity during normal and accident conditions and will continue to meet the DCPP licensing basis requirements with respect to GDC -13, -16, -38, -50, and -64 following installation of the replacement steam generators 6.X.2 REFERENCES

1. GOTHIC Containment Analysis Package Technical Manual, Version 7.2, NAI 8907-06, Rev. 15, September 2004.
2. GOTHIC Containment Analysis Package Qualification Report, Version 7.2, NAI-8907-09, Rev. 8, September 2004.
3. License Amendment No. 169 for Kewaunee Nuclear Power Plant, Operating License No. DPR-43 (TAC No. MB6408), September 29, 2003.
4. GOTHIC Containment Analysis Package User Manual, Version 7.2, NAI 8907-02, Rev. 16, September 2004.
5. Brown and York, "Sprays formed by Flashing Liquid Jets", AICHE Journal, Volume 8, #2, May 1962.
6. Westinghouse LOCA Mass and Energy Release Model for Containment Design

- March 1979 Version, WCAP-1 0325-P-A (Proprietary), WCAP-1 0326-A (Nonproprietary), May 1983.

7. F. M. Bordelon and E. T. Murphy, Containment Pressure Analysis Code (COCO), WCAP-8327 (Proprietary), WCAP-8326 (Non-Proprietary), July 1974.

8

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-1 Sheet 1 of 2 DIABLO CANYON CONTAINMENT LOCA INTEGRITY ANALYSIS PARAMETERS Parameter Value Auxiliary Service Water Temperature (OF) 64 RWST Water Temperature (OF) 90 Initial Containment Temperature ('F) 120 Initial Containment Pressure (psia) 16.0 Initial Relative Humidity (%) 18 Net Free Volume (ft3) 2,550,000 Reactor Containment Fan Coolers Total CFCUs 5 Analysis Maximum 3 Analysis Minimum 2 Containment High Setpoint (psig) 5.0 Delay Time (sec)

Without Offsite Power 48.0 CCW Flow to the CFCUs (gpm)

During Injection 8,000 During Recirculation 7,450 Containment Spray Pumps Total CSPs 2 Analysis Maximum 2 Analysis Minimum 1 Flowrate (gpm)

During Injection Table 6.2.D-18 During Recirculation 0 Containment High High Setpoint (psig) 24.7 Spray Delay Time (sec)

Without Offsite Power 80 Containment Spray Termination Time, (sec)

Minimum Safeguards 3,798 Maximum Safeguards (1 CSP) 3,018 Maximum Safeguards (2 CSPs) 1,824 9

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008

)

TABLE 6.X-1 Sheet 2 of 2 ECCS Recirculation ECCS Cold-Leg Recirculation Switchover, sec Minimum Safeguards 1,678 Maximum Safeguards (1 CSP) 1,033 Maximum Safeguards (2 CSPs) 829 Containment ECCS Cold-Leg Recirculation Flow, (gpm)

Minimum Safeguards (1 RHR train) 3,252.3 Maximum Safeguards (2 RHR trains) 8,082.4 ECCS Hot-Leg Recirculation Switchover, sec 25,200 Containment ECCS Hot-Leg Recirculation Flow, (gpm)

Minimum Safeguards (1 RHR train) 3,071.7 Maximum Safeguards (2 RHR trains) 4,576.8 Component Cooling Water System Total CCW Heat Exchangers 2 Analysis Maximum 2 Analysis Minimum 1 CCW Flow Rate to RHR Heat Exchanger (gpm per available HX) 4,800 ASW Flow Rate to CCW Heat Exchanger (gpm per available HX) 10,300 CCW Misc. Heat Loads (MBTU/hr)

During Injection 1.0 During Recirculation 2.0 CCW Flow Rate to Misc. Heat Loads (gpm)

During Injection 2,500 During Recirculation 500 10

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-2 CONTAINMENT SPRAY FLOW RATES AS A FUNCTION OF CONTAINMENT PRESSURE 1 CSP 2 CSPs Containment Spray Flow Rate Spray Flow Rate Pressure (psig) (gpm) (gpm) 0 3036 6142 10 2926 5922 20 2806 5692 30 2686 5442 40 2546 5182 47 2456 4992 11

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-3 GOTHIC THERMAL CONDUCTOR MODELING No. Description Materials Surface Area Thickness Initial Temp 2 (in) (F)

(ft )

1 Concrete Interior Walls Paint 79965 0.0075 120 Concrete 12 2 Concrete Floor Paint 13012 0.0075 120 Concrete 24 3 SS Fuel Transfer Tube Stainless Steel 8852 0.144 120 4 SS Structures Stainless Steel 857 0.654 120 5 CS Structures Paint 48024 0.0075 120 Carbon Steel 0.0815 6 CS Structures Paint 60941 0.0075 120 Carbon Steel 0.133 7 CS Lined Containment Paint 90560 0.0075 120 Concrete Shell Carbon Steel 0.375 HGap = 10 0.0168 Concrete 35.6007 8 CS Structures Paint 42517 0.0075 120 Carbon Steel 0.567 9 CS Structures Paint 56494 0.0075 120 Carbon Steel 0.738 10 CS Structures Paint 31902 0.0075 120 Carbon Steel 1.355 11 CS SG Snubbers Paint 522 0.0075 120 Carbon Steel 3.0 12 CS RCP Motors Paint 1610 0.0075 200 Carbon Steel 6.99 12

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-4 MATERIAL PROPERTIES. FROM REFERENCE 15 GOTHIC MODEL Material Thermal Conductivity Vol. Heat Capacity (BTU/hr-ft-°F) (BTU/ft 3-OF)

Paint 0.2083 35.91 Carbon Steel 28 58.8 Air Gap 0.0148 0.018 Concrete 1.04 23.4 Stainless Steel 8.6 58.8 13

DCPP UNITS 1 &2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-5 DOUBLE-ENDED HOT-LEG BREAK SEQUENCE OF EVENTS Time (sec) Event Description 0.0 Break Occurs 1.1 Reactor Trip Occurs on Compensated Pressurizer Pressure Setpoint of 1859.7 psia and SG Throttle Valves Closed 4.0 Low Pressurizer Pressure Sl Setpoint = 1694.7 psia Reached (Safety Injection begins without any delay and feedwater control valve starts to close) 13.0 Main Feedwater Control Valve Fully Closed 15.5 Broken Loop Accumulator Begins Injecting Water 15.6 Intact Loop Accumulator Begins Injecting Water 24.4 End of Blowdown Phase - Transient Modeling Terminated 14

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-6 DIABLO CANYON UNIT 1 DOUBLE-ENDED PUMP SUCTION BREAK SEQUENCE OF EVENTS (MINIMUM SAFEGUARDS)

Time (s) Event Description 0.0 Break Occurs and Loss-of-Offsite Power is Assumed 1.0 Reactor Trip Occurs on Compensated Pressurizer Pressure Setpoint of 1,859.7 psia and SG Throttle Valves Closed 4.0 Low Pressurizer Pressure SI Setpoint = 1,694.7 psia Reached (SI begins after a 27-second delay and feedwater control valve starts to close) 13.0 Main Feedwater Control Valve Closed 16.4 Broken-Loop Accumulator Begins Injecting Water 16.8 Intact-Loop Accumulator Begins Injecting Water 25.6 End of Blowdown Phase 31.1 Pumped Safety Injection Begins 48.7 CFCUs On 51.9 Broken Loop Accumulator Water Injection Ends 53.2 Intact Loop Accumulator Water Injection Ends 87.6 Containment Sprays Begin Injecting 193.7 End of Reflood for Minimum Safeguards Case 508.8 Mass and Energy Release Assumption: Broken-Loop SG Equilibration to 61.7 psia 889.2 Mass and Energy Release Assumption: Broken-Loop SG Equilibration to 40.7 psia 1,495.2 Mass and Energy Release Assumption: Intact-Loop SG Equilibration to 61.7 psia 1,678.0 Cold-Leg Recirculation Begins 1,695.6 Mass and Energy Release Assumption: Intact-Loop SG Equilibration to 39.7 psia 3,600.0 End of Sensible Heat Release from RCS and SGs 3,798.0 Containment Sprays Terminated 25,200.0 Switchover to Hot-Leg Recirculation 2

15

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-7 DIABLO CANYON UNIT 2 DOUBLE-ENDED PUMP SUCTION BREAK SEQUENCE OF EVENTS (MINIMUM SAFEGUARDS)

Time (sec) Event Description 0.0 Break Occurs and Loss-of-offsite Power is assumed 1.2 Reactor Trip Occurs on Compensated Pressurizer Pressure Setpoint of 1859.7 psia and SG Throttle Valves Closed 4.2 Low Pressurizer Pressure SI Setpoint = 1694.7 psia Reached (Safety Injection begins'after a 27 second delay and feedwater control valve starts to close) 13.2 Main Feedwater Control Valve Closed 18.1 Broken Loop Accumulator Begins Injecting Water 18.6 Intact Loop Accumulator Begins Injecting Water 26.2 End of Blowdown Phase 31.3 Pumped Safety Injection Begins 48.7 CFCUs On 52.7 Broken Loop Accumulator Water Injection Ends 53.7 Intact Loop Accumulator Water Injection Ends 88.0 Containment Sprays Begin Injecting 203.5 End of Reflood for Minimum Safeguards Case 568.5 Mass and Energy Release Assumption: Broken Loop SG Equilibration to 61.7 psia 829.3 Mass and Energy Release Assumption: Broken Loop SG Equilibration to 40.7 psia 1,536.5 Mass and Energy Release Assumption: Intact Loop SG Equilibration to 61.7 psia 1,678.0 Cold-Leg Recirculation Begins 1,717.8 Mass and Energy Release Assumption: Intact Loop SG Equilibration to 39.7 psia 3,600.0 End of Sensible Heat Release from Reactor Coolant System and Steam Generators 3,798.0 Containment Sprays Terminated 25,200.0 Switchover to Hot-Leg Recirculation 16

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 TABLE 6.X-8

SUMMARY

OF LOCA PEAK CONTAINMENT PRESSURE AND TEMPERATURES Peak Peak Press @ Temp @

Break Pressure Time Gas Temp Time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Location (psig) (sec) (OF) (sec) (psig) (OF)

DEHL 41.4 23.8 261.8 23.4 - -

DEPS min SI 39.8 24.1 259.3 24.1 8.9 167.5 Acceptance <47 <50% of peak Criteria pressure 17

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Diablo Canyon Unit 2 LOCA Containment Analysis Double Ended Hot Leg Breaks Containment Pressure 45 40 35

" 30 -

- 25 200 a 15 -

10 5

0 1 10 100 Time (seconds)

Figure 6.X-1 18

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Diablo Canyon Unit 2 LOCA Containment Analysis Double Ended Hot Leg Break Containment Gas Temperature 260 240 S220

°1 1 E18 c200 160 I-140 120 10 100 Time (seconds)

Figure 6.X-2 19

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Diablo Canyon Unit 2 LOCA Containment Analysis Double Ended Hot Leg Break Containment Sump Temperature 280 260 240 0

.. 220 200 E 180 I-160 140 120 1 10 100 Time (seconds)

Figure 6.X-3 20

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Diablo Canyon LOCA Containment Analysis DEPS Break with Minimum Safeguards Containment Pressure 45.0

- Unit 1 40.0 35.0

.0 30.0 215.0 10.0 5.0 0.0 .

1 10 100 1000 10000 100000 1000000 1E+07 Time (seconds)

Figure 6.X-4 21

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Diablo Canyon LOCA Containment Analysis DEPS Break with Minimum Safeguards Containment Gas Temperature 280 A 260 - Unit1 240 -Unit 2

, 220 200 180 E

1 I-- 160 140 -

120 100 1 10 100 1000 10000 100000 1000000 1E+07 Time (seconds)

Figure 6.X-5 22

DCPP UNITS 1 & 2 FSAR UPDATE Enclosure 4 PG&E Letter DCL-08-008 Diablo Canyon LOCA Containment Analysis DEPS Break with Minimum Safeguards Containment Sump Temperature 280 21.. ...

. 180 E 160 140 120 100 1 10 100 1000 10000 100000 1000000 1E+07 Time (seconds)

Figure 6.X-6 23

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