ML18152A280

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Forwards Summary of Structural Integrity Evaluation of Thermally Induced Over Pressurization of Containment Penetration Piping During DBA for SPS & Naps,Units 1 & 2,per GL 96-06.Draft Proposed UFSAR Revised Pages,Encl
ML18152A280
Person / Time
Site: Surry, North Anna  Dominion icon.png
Issue date: 03/30/1999
From: Christian D
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
99-134, GL-96-06, GL-96-6, NUDOCS 9904120195
Download: ML18152A280 (23)
v • d • e


Text

e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 March 30, 1999 U.S. Nuclear Regulatory Commission Serial No.: 99-134 Attention: Document Control Desk NL&OS/GDM: RO' Washington, D.C. 20555 Docket Nos.: 50-280, 281 50-338, 339 License Nos.: DPR-32, 37 NPF-4, 7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY AND NORTH ANNA POWER STATIONS UNITS:1 AND 2*

SUPPLEMENTAL RESPONSE TO GENERIC LETTER (GL) 96-06 STRUCTURAL INTEGRITY EVALUATION OF THERMALLY INDUCED OVER PRESSURIZATION OF CONTAINMENT PENETRATION PIPING DURING DBA In a letter dated February 25, 1998 (Serial No. 96-516C), Virginia Electric and Power Company (Virginia Power) proposed acceptance criteria to be used for evaluating design adequacy for thermally induced overpressurization of piping systems that penetrate the containments of both North Anna and Surry Power Stations during. a postulated design basis accident (DBA). Specifically, we proposed to use the ASME Code Section Ill, Appendix F, "Rules for Evaluation of Service Loading with Level D Service Limits." The proposed acceptance criteria were provided to address NRC Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions." Detailed criteria for the use of the linear elastic analysis method of the code was proposed and was subsequently discussed in a telephone conversation with members of the NRC staff.

We have completed our structural integrity evaluations for the piping systems that are susceptible to such a postulated loading for both Surry and North Anna Power Stations to address the concerns raised in GL 96-06. Attachment 1 to this letter provides: (1) a summary of the method used for determining thermally induced *pressure in isolated piping with confined fluid, (2) identification of the piping with its associated containment penetration numbers, (3) therm~lly_induced pressur~ in the identified:_piping, and (4) a summary of stresses in the susceptible piping components and associated valves along with the Code allowable stress. The pressure determination in the pipe considered the differential expansion between the confined fluid and the pipe metal,* and also took credit for a limited amount of circumferential strain in the pipe wall due to pressure.

Detailed verification of structural integrity was not considered necessary when the .

faulted pressure was less !hail 1.2 limes ~e design pressure. Detailed linear e~a~ /) 1 JJ

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  • stress analysis was performed for the components when the faulted pressure exceeded 1.2 times*the design pressure.

The results of the analysis and determination of Code compliance were based upon the ASME Boiler and Pressure Vessel Code, Section Ill, Appendix F. The results demonstrate that the structural integrity of the containment penetration piping and associated valves will be maintained when subjected to thermally induced over-pressurization during postulated OBA conditions (i.e., Loss of Coolant Accident/Main Steam Line Break). Adequate margin exists between the applied stresses and the Appendix F allowable stresses. The deformation of the piping components will be limited to the amount of strain listed in the attachment, and a gross failure during this faulted event is not considered credible. Adequate technical basis exists to conclude that there are no safety significant issues that could affect containment integrity or equipment operability during OBA conditions. Thus, the results obtained provide adequate assurance of continued equipment operability and containment integrity during OBA conditions addressing the concern raised in GL 96-06.

Our review did not result in any physical modifications to our plant facilities. However, Surry and North Anna Power Stations were not originally licensed to use Appendix F of the ASME Section Ill Code. Consequently, we request your approval to use the 1989 version of the ASME Boiler and Pressure Vessel Code, Section Ill, Appendix F, as the applicable code for Surry and North Anna Power Stations for this particular faulted loading event as modified and detailed in Attachment, 1. Drafts of the proposed changes to the Updated Final Safety Analysis Reports (UFSAR) for Surry and North Anna Power Stations are provided in Attachments 2 and 3, respectively, for your information. A revision to the respective UFSARs for each station will be implemented upon NRC closeout of GL 96-06 in accordance with 10 CFR 50.71(e).

If you have any further questions or require additional information, please contact us.

Very truly yours, D. A. Christian Vice President - Nuclear Operations Attachments I ~

cc: U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. M. J. Morgan NRC Senior Resident Inspector North Anna Power Station Mr. R. A. Musser NRC Resident Inspector Surry Power Station Commitment Summary The following represents the specific commitment made by the subject correspondence (Serial No.99-134):

1. A revision to the respective UFSARs for each station will be implemented upon NRC closeout of GL 96-06 in accordance with 10 CFR 50.71 (e).
  • -*

Attachment 1 Structural Integrity Evaluation of Thermally Induced Over Pressurization of Containment Penetration Piping During a DBA Surry and North Anna Power Stations Units 1 and 2 Virginia Electric and Power Company

1.0 Introduction

  • e NRC Generic Letter 96-06: "Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions" was issued on September 30, 1996 to notify all holders of operating licenses about potentially safety significant issues that could affect containment integrity and equipment operability during design basis accident (DBA) conditions.

As a part of the actions associated with the NRC Generic Letter 96-06 (Ref. 6), it was considered necessary to perform analyses to provide assurance of equipment operability and containment integrity during DBA conditions. Specifically, it was considered necessary to analyze and establish that thermally induced overpressurization of isolated water-filled piping sections in the containment boundary could not jeopardize the ability of the accident mitigating systems to perform their safety functions and could not lead to a breach of containment integrity through bypass leakage. The maximum internal pressure developed inside the isolated containment piping penetrations during a design basis accident (i.e., LOCA or MSLB) was calculated taking into account the differences in the expansion of the fluid and the pipe, the temperature increase immediately following the DBA and crediting a limited amount of radial/circumferential strain of the piping material during the pressurization process.

Piping susceptibility to thermal overpressurization following a DBA was not specifically evaluated prior to the issue of NRC GL 96-06. It should be noted that no specific design criteria exist in the North Anna or Surry design basis or original piping codes for evaluating isolated pipe segments under faulted conditions.* Consequently, the loading conditions and the criteria for structural integrity evaluation of such a faulted event were not established or required during initial licensing. Also, the particular sections of pipe and the corresponding penetrations that may be susceptible to overpressurization were not specifically identified earlier. As a part of the response to GL 96-06, the susceptible piping sections were evaluated (Ref. 12). Many sections were not susceptible to overpressurization because of their configurations or the operating conditions during the DBA. Detailed thermal analysis of the remaining sections was performed (Ref. 7 & 8) to determine the extent of overpressurization. A detailed evaluation to establish structural integrity was not considered necessary for sections subjected to 1.2 times the design pressure or less during the event. Structural evaluation was performed for remaining sections using linear elastic analysis method stipulated in ASME Boiler and Pressure Vessel Code, Section Ill, Appendix F.

Page 1 of 11

2.0

  • Summary of the Method for Determination of Thermally Induced Pressure in Isolation Piping with Confined Fluid The heat transferred to isolated sections of containment piping during a Design Basis Accident (OBA) was conservatively evaluated using the post-accident containment bulk atmospheric temperature profile for minimum engineered safeguards response. It should be noted that the conservatism in the analysis that produces the bulk atmospheric temperature profile was appropriate for sizing the safeguards systems, the design of the containment structure and pressure boundary components. Lacking more realistic containment response analysis, these highly conservative temperature profiles were used. The location of the containment piping penetrations of concern, relatively low in the structure and outside the crane wall, would also result in their being exposed to lower peak temperatures. However, this analysis assumed no temperature reduction based on piping location.

The maximum internal pressure developed inside the isolated containment piping penetrations during a OBA was calculated as follows:

  • Determine piping parameters (length of piping inside and outside, system operating conditions at time of OBA, etc.)
  • Determine the time dependent, heat transfer to the piping/fluid inside containment, associated peak temperature, the heat content and mass of the water in the pipe
  • Determine the temperature distribution along the piping which passes through the containment wall by numerical relaxation method for two dimensional heat transfer in solids, and determine the heat content and mass of the water in the pipe
  • Model the outside containment penetration piping as a fin and calculate the temperature distribution along the piping, determine heat content and mass of the water in the pipe
  • Determine the total heat content and total mass of water in the entire length of isolated piping
  • Calculate the bulk temperature of the isolated piping
  • The expansion of the piping circumference and length resulting from the fluid temperature increase is applied to the piping circumference and length
  • A final pressure is assumed and the stress resulting from the pressure is calculated
  • The strain resulting from the calculated stress is then calculated and the strain is applied to the piping circumference to determine the increase in volume resulting from the increase in temperature (pressure)
  • Calculate the final pressure rise in the isolated piping based on the change in volume of an incompressible fluid
  • The final pressure *is checked aga1nst the assumed pressure
  • The iteration is repeated until the assumed final pressure and the calculated final pressure are equal; this results in the final equilibrium state for the piping.

Page 2 of 11

3.0

  • Identification of Piping with Its Identifying Containment Penetration Numbers The following penetrations have been evaluated using the methodology in ASME Section Ill, Appendix F, to determine whether any modifications are required. The thermally induced pressure on this piping was determined to be more than 1.2 times the design pressure.

Surry Power Station, Units 1 & 2 - Penetrations 20, 28 and 46.

North Anna Power Station, Unit 1 - Penetrations 5, 12, 13, 14, 20, 25 and 46.

North Anna Power Station, Unit 2 - Penetrations 5, 12, 13, 14, 20, 25, 26, 27, 46 and 106.

Page 3 of 11

e e 4.0 Thermally Induced Pressure in the Identified Piping

. Table 4.1- Pressure Loading Due to Temperature Increase in Isolated Pipe Penetration No. (1) Pressure Predicted Circumferential Psi Percent Strain (2)

NAPS #5 U1 1,247 0.57 NAPS#5 U2 1,242 0.68 NAPS #12 U1 2,176 0.09 NAPS #12 U2 2,555 0.10 NAPS #13 U1 2,765 0.11 NAPS #13 U2 2,499 0.10 NAPS #14 U1 2,574 0.10 NAPS #14 U2 2,499 0.10 NAPS #20 U1 5,503 1.20 NAPS #20 U2 5,337 1.20 NAPS #25 U1 2,069 0.33 NAPS #25 U2 2,347 0.57 NAPS #26 U2 2,230 0.22 NAPS#27 U2 1,664 0.08 NAPS #46 U1 7,582 0.99 NAPS #46 U2 7,936 0.90 NAPS #106 U2 3,171 0.44 SPS #20 U1 5,022 1.67 SPS#20 U2 5,022 1.67 SPS #28 U1 2,960 0.65 SPS #28 U2 2,960 0.65 SPS #46 U1 7,125 1.46 SPS#46 U2 7,125 1.46 Notes:

(1) NAPS - North Anna Power Station SPS - Surry Power Station U1 - Unit 1 U2 - Unit 2 (2) Pressure determination utilizes credit of a limited amount of strain in the pipe Page 4 of 11

e e 5.0 Summary of Stresses in the Susceptible Piping Components and Associated Valves Along with the Code Allowable Stress Table 5.1 - Pipe Stress Summary SPS-1 20 5022 46200 45000 28 2960 24900 46200 27170 45000 46 7125 25340 46200 21520 45000 SPS-2 20 5022 26360 46200 21620 45000 28 2960 24900 46200 27170 45000 46 7125 25340 46200 21670 45000 NAPS-1 5 1247 33700 42000 32870 60000 12, 13, 14 2765 36280 42000 24630 60000 20 5503 28890 46200 23400 45000 25 2069 30840 42000 28110 60000 46 7582 26970 48000 23460 46600 NAPS-2 5 1242 33570 42000 17430 60000 12, 13, 14 2555 33520 42000 18980 60000.

20 5337 28020 46200 22790 45000 25,26,27 2347 34980 42000 26160 60000 46 7936 28230 48000 24640 46600 106 3171 26140 46200 20990 45000 Notes:

(1) Linear Elastic Analysis Method of analysis is used.

(2) Pressure and dead weight loadings are used. Seismic loading is not considered concurrent with the event.

(3) Allowable membrane stress = 2.4Sm or 0. 7Su whichever is lower.

(4) Allowable membrane plus bending stress = 3.0Sm or 2Sy whichever is lower.

(5) Only the piping with pressure greater than 1.2 times the design pressure is listed.

(6) Pressure.increase .is _due_.to_.temperature...effecton .. confined Jluid.inside .. piping on both sides of the containment penetration.

(7) Allowable stresses are taken from ASME B & PV Code Section Ill, 1989 (Ref. 4).

(8) Adequate margins exist between applied stress and allowable stress.

Page 5 of 11

e Table 5.2 - Valve Stress Summary (applicable to weld end valves)

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NAPS #5 U1 1-CC-764 18,275 73,500 NAPS #12 U1 1-CC-568 20,262 73,500 NAPS #12 U1 1-CC-715 20,262 73,500 NAPS #12 U1 1-CC-718 20,262 73,500 NAPS #13 U1 1-CC-581 21,522 73,500 NAPS #13 U1 1-CC-721 21,522 73,500 NAPS #14 U1 1-CC-555 21,114 73,500 NAPS #14 U1 1-CC-710 21,114 73,500 NAPS#20 U1 1-Sl-58 16,726 72,000 NAPS #20 U1 1-Sl-59 19,875 72,000 NAPS #20 U1 1-Sl-111 12,942 72,000 NAPS#20 U1 1-Sl-110 16,725 72,000 NAPS #20 U1 1-Sl-245 14,451 72,000 NAPS #20 U1 1-SI-HCV-1851A 33,335 72,000 NAPS #20 U1 1-SI-HCV-1851B 33,335 72,000 NAPS #20 U1 1-SI-HCV-1851C 33,335 72,000 NAPS #25 U1 1-CC-108 20,033 73,500 NAPS#25 U1 1-CC-754 20,033 73,500 NAPS #46 U1 1-CH-FCV-1160 38,454 72,000 NAPS#46 U1 1-CH-330 30,606 72,000 NAPS#46 U1 1-CH-331 17,958 72,000 NAPS #46 U1 1-CH-332 17,958 72,000 NAPS #46 U1 1-CH-476 22,195 72,000 NAPS #46 U1 1-CH-488 22,195 72,000 NAPS#46 U1 1-CH-489 22,195 72,000 NAPS #46 U1 1-RC-HCV-1556A 38,454 72,000 NAPS #46 U1 1-RC-HCV-1556B 38,454 72,000 NAPS #46 U1 1-RC-HCV-1556C 38,454 72,000 NAPS #5 U2 2-CC-329 18,265 73,500 NAPS #5 U2 2-CC-705 18,265 73,500 NAPS#12 U2 2-CC-298 21,073 73,500 NAPS#12 U2 2-CC-750 21,073 73,500 NAPS#12 U2 2-CC-751 21 ;073 73,500 NAPS #13 U2 2-CC-311 20,953 73,500 NAPS #13 U2 2-CC-756 20,953 73,500 NAPS#14 U2 2-CC-284 20,953 73,500 NAPS #14 U2 2-CC-285 20,953 73,500 NAPS #20 U2 2-Sl-47 16,427 72,000 NAPS #20 U2 2-Sl-48 12,674 72,000 NAPS #20 U2 2-Sl-136 20,704 72,000 Page 6 of 11

NAPS #20 U2 NAPS #20 U2 e

2-Sl-137 2-Sl-243 14,284 12,674

  • 72,000 72,000 NAPS#20 U2 2-SI-HCV-2851A 32,883 72,000 NAPS #20 U2 2-SI-HCV-2851 B 32,883 72,000 NAPS #20 U2 2-SI-HCV-2851 C 32,883 72,000 NAPS#25U2 2-CC-103 20,628 73,500 NAPS #25 U2 2-CC-712 20,628 73,500 NAPS#26 U2 2-CC-177 20,378 73,500 NAPS #26 U2 2-CC-349 20,378 73,500 NAPS #26 U2 2-CC-722 20,378 73,500 NAPS #27 U2 2-CC-140 19,167 73,500 NAPS #27 U2 2-CC-717 19,167 73,500 NAPS#27 U2 2-CC-718 19,167 73,500 NAPS #46 U2 2-CH-FCV-2160 39,578 72,000 NAPS#46 U2 2-CH-259 17,012 72,000 NAPS #46 U2 2-CH-332 31,522 72,000 NAPS#46 U2 2-CH-333 18,507 72,000 NAPS #46 U2 2-CH-334 18,507 . 72,000 NAPS#46 U2 2-CH-380 18,507 72,000 NAPS#46 U2 2-CH-406 19,078 72,000 NAPS#46 U2 2-CH-385 18,507 72,000 NAPS #46 U2 2-RC-HCV-2556A 39,576 72,000 NAPS #46 U2 2-RC-HCV-25568 39,576 72,000 NAPS #46 U2 2-RC-HCV-2556C 39,576 72,000 NAPS #106 U2 2-SI-TV-2842 19,022 72,000 NAPS#106 U2 2-Sl-248 6,138 72,000 NAPS#106 U2 2-Sl-231 10,753 72,000 NAPS#106 U2 2-SI-TV-2859 19,023 72,000 NAPS #106 U2 2-Sl-49 24,600 72,000 SPS #20 U1 1-SI-HCV-1851A 29,556 72,000 SPS #20 U1 1-SI-HCV-1851 B 29,556 72,000 SPS #20 U1 1-SI-HCV-1851 C 29,556 72,000 SPS #20 U1 1-Sl-32 15,860 72,000 SPS #20 U1 1-Sl-181 13,796 72,000 SPS #28 U1 1-CH-TV-1204A 17,099
  • 72,000 SPS #28 U1 1-CH-TV-12048 32,628 72,000 SPS #28 U1 1-CH-419 10,598 72,000 SPS #28 U1 1-CH-440 10,598 72,000 SPS #28 U1 1-CH-457 31,704 72,000 SPS #46 U1 1-CH - FCV-1160 30,260 72,000 SPS #46 U1 1-RC-HCV-1556A 30,260 72,000
  • SPS #46 U1 1-RC-HCV-15568 30,260 72,000 SPS #46 U1 1-RC-HCV-1556C 30,260 72,000 SPS #46 U1 1-CH-444 18,486 72,000 Page 7 of 11

SPS #46 SPS#20 U1 U2 e

1-CH-316 2-S1-HCV-2851A 17,250 29,566

  • 72,000 72,000
  • SPS#20 U2 2-SI-HCV-2851 B 29,566 72,000 SPS #20 U2 2-SI-HCV-2851 C 29,566 72,000 SPS#20 U2 2-Sl-32 15,860
  • 72,000 SPS#20 U2 2-Sl-181 13,796 72,000 SPS #28 U2 2-CH-TV-2204A 17,099 72,000 SPS#28 U2 2-CH-TV-2204B 35,529 72,000 SPS #28 U2 2-CH-414 10,599 72,000 SPS #28 U2 2-CH-441 31,551 72,000 SPS #28 U2 2-CH-453 31,704 72,000 SPS #46 U2 2-CH-FCV-2160 30,260 72,000 SPS #46 U2 2-RC-HCV-2556A 30,260 72,000 SPS #46 U2 2-RC-HCV-25568 30,260 72,000 SPS #46 U2 2-RC-HCV-2556C 30,260 72,000 SPS #46 U2 2-CH-444 40,386 72,000 SPS #46 U2 2-CH-316 17,250 72,000 Note:

(1) The valve body structural integrity is verified by comparison of valve section and material properties with that of the connected pipe.

(2) Allowable stresses are from Reference 4.

(3) Adequate margin exists between applied stress and allowable stress.

Page 8 of 11

e

  • Table 5.3 - Valve Stress Summary (applicable to flange end valves) 1 Penetration No. Valve No. Stress on Flange Allowable < > (psi)

Bolts (psi)

NAPS #5 U1 1-CC-MOV-1 ODA 25,535 87,500 NAPS #5 U1 1-CC-TV-103A 25,535 87,500 NAPS #12 U1 1-CC-TV-1008 26,000 87,500 NAPS #12 U1 1-CC-TV-1058 26,000 87,500 NAPS #13 U1 1-CC-TV-1 DOC 32,070 87,500 NAPS #13 U1 1-CC-TV-105C 32,070 87,500 NAPS #14 U1 1-CC-TV-1 ODA 30,102 87,500 NAPS #14 U1 1-CC-TV-105A 30,102 87,500 NAPS #25 U1 1-CC-TV-102E 44,646 87,500 NAPS #25 U1 1-CC-TV-102F 44,646 87,500 NAPS#5 U2 2-CC-MOV-200A 21,758 87,500 NAPS #5 U2 2-CC-TV-203A 21,758 87,500 NAPS #12 U2 2-CC-TV-2008 28,001 87,500 NAPS #12 U2 2-CC-lV-2058 28,001 87,500 NAPS #13 U2 2-CC-TV-200C 27,423 87,500 NAPS #13 U2 2-CC-TV-205C 27,423 87,500 NAPS #14 U2 2-CC-TV-200A 27,423 87,500 NAPS#14 U2 2-CC-TV-205A 27,423 87,500 NAPS#25 U2 2-CC-TV-202E 47,142 87,500 NAPS#25 U2 2-CC-TV-202F 47,142 87,500 NAPS#26 U2 2-CC-TV-202A 45,110 87,500 NAPS#26 U2 2-CC-TV-2028 45,110 87,500 NAPS#27 U2 2-CC-TV-202C 35,281 87,500 NAPS#27 U2 2-CC-TV-2020 35,281 87,500 Note:

(1) 0. 7Su or Sy (A 193 Gr. 87 Bolts) whichever is lower.

Page 9 of 11

e

6.0 CONCLUSION

S:

[1). The pressure determination of the pipe takes credit for a limited amount of strain in the pipe at equilibrium state.

[2] A detailed evaluation of piping for verification of structural integrity was not considered necessary when the faulted pressure was less than 1.2 times the design pressure.

[3] The results of the analysis and determination of Code compliance were based upon the ASME Boiler and Pressure Vessel Code, Section Ill, Appendix F.

These results demonstrate that the structural integrity of the containment penetration piping and associated valves will be maintained when subjected to thermally induced overpressurization during postulated OBA conditions (i.e., LOCA/MSLB).

[4] An adequate technical basis exists to conclude that there are no safety significant issues that could affect containment integrity and equipment operability during OBA conditions. Thus, the evaluation results provide assurance of continued equipment operability and containment integrity during OBA conditions and addresses the concern raised in NRC Generic Letter 96-06.

Page 10 of 11

e

7.0 REFERENCES

[1]. ANSI B31.7, 1969 w/addenda through 1970 "Nuclear Power Piping Code".

[2] ANSI B31.1, 1967, "Standard Code for Pressure Piping."

[3] ANSI B31.1, 1955, "Code for Pressure piping up to and including Code Case N7."

[4] ASME B & PV Code, Section Ill, Appendix I, 1989. *

[5] ASME B & PV Code, Section Ill, Appendix F, 1989.

[6] The NRC Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions, September 30, 1996."

[7] Virginia Power Calculation ME-0526, "Containment Piping Penetration Overpressurization During a OBA for Containment Penetrations 5, 12, 13, 14, 20, 25, 26, 27, 28 and 46 for North Anna Power Station, Units 1 & 2 and Penetration 106 for North Power Station, Unit 2."

[8] Virginia Power Calculation ME-0527, "Containment Piping Penetration Overpressurization During a OBA for Containment Penetration 20, 28, 46 and 110 for Surry Power Station Units 1 & 2."

[9] Letter from Virginia Power to NRC, Serial No. 96-516C, dated February 25, 1998, "Supplemental Response to Generic Letter (GL) 96-06 Acceptance Criteria for Design Adequacy Evaluation."

[1 OJ Virginia Power Calculation ME-0569, "ASME B & PV Code, Section Ill, Appendix F, Valve Analysis for NRC Generic Letter 96-06."

[11] Letter from Virginia Power to NRC, Serial No. 96-516A, dated January 28, 1997, "NRC Generic Letter 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident."

[12] Virginia Power Technical Report No. CE-0102, Rev. 0, "Evaluation of Structural Integrity of Containment Penetration Piping and Isolation Valves During Postulated* LOCA/MSLB *scenarios *using ..ASME *code *section Ill, Appendix F Criteria."

Page 11 of 11

I I

. e Attachment 2 Draft of Proposed UFSAR Revision to Reflect the Resolution of GL 96-06 Surry Power Station Units 1 and 2 Virginia Electric and Power Company

Attachment 2

  • Draft of Proposed UFSAR Revision to Reflect the Resolution of GL 96-06 Surry Power Station Units 1 and 2
  • Insert 1 Insert in SPS UFSAR Revision 30-09/1/98, Section 15.5, after the 4th paragraph in page 15.5-18:

As a part of the issues identified in NRC GL 96-06, isolated containment penetration piping with confined fluid was reviewed for susceptibility to thermal overpressurization following a OBA. The linear elastic analysis criteria stipulated in the 1989 version of the ASME Boiler and Pressure Vessel Code Section Ill, Appendix F, were used for structural integrity evaluation. The internal pressure in piping penetrations during a design basis accident (LOCA or MSLB) was calculated by taking into account the differences in the expansion of the fluid and the pipe, the temperature increase immediately following the OBA and credit for a limited amount of circumferential strain in the pipe. The analysis established that thermally induced overpressurization of isolated water-filled piping sections in the containment boundary could not jeopardize the ability of the accident mitigating systems to perform their safety functions and could not lead to a breach of containment integrity (Reference XX).

  • Insert 2 Add in SPS UFSAR Revision 30-09/1/98, Section 15.5 References, the following new reference:

XX. Letter Dated March XX, 1999, Serial No.99-134, from Virginia Power to the NRC, Supplemental Response to Generic Letter 96-06.

Revision 30-09/1/98 e SPS~ 15.5-18 The circumferential groove in the attachment plate, between the sleeve and penetration with its outside threaded connection, serves as a test chamber for the testing of the welds joining the attachment plate and penetration.

All penetrations are anchored in the reinforced concrete containment wall. The anchor strength is equal to the full yield strength of the pipe with regard to torsion, bending, and shear, and to the maximum possible pipe jet reaction. All stresses induced in the liner by these combinations of loadings are only those reflected by the resulting distortions in the reinforced concrete containment wall, and are minor in intensity. So, loads will not be imposed on the liner, thereby preserving its integrity.

All highly stressed insert plates at penetrations and equipment supports that are welded into the liner to transfer loads into the concrete have been ultrasonically tested to check for possible laminations. Tests were conducted on all plates where analysis showed a higher than average stress field, although all such plates are stressed well below the allowable limits for the materials.

These tests show that no faults exist in the insert plates.

The pipes anchored to the containment penetrations between containment isolation valves constitute an extension of the containment, and are designed in accordance with the USA Standard Code for Pressure Piping - Power Piping, USAS B31.l.0-1967, with respect to materials and allowable stress. Analyses of stresses due to thermal expansion and shock loadings from earthquake, pipe jet reaction, and other causes were made using established digital computer calculation techniques.

In order to determine the loading combinations that act on a penetration, the pipe line passing through the penetration sleeve was assumed to have failed transversely at several locations along its run. The location at which the reaction of the ensuing jet of fluid flowing from the broken end first causes the pipe to completely yield, in either bending or torsion, was taken as the design case from which all resultant combinations of penetration loading were determined for that particular pipe line. The maximum stress allowed on any individual element of the penetration is 90% of the minimum yield point.

The intent of this criterion is to keep the material assembly components within the elastic range of the material. Under operating conditions of pressure, temperature, and external loads, the stresses in the assembly will be within the limits established in Section ill of the ASME Pressure

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Vessel Code.

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..fvi s&c"'-\' ~11 liner seams were strength-welded. Small steel channels weld_ed continuously along the edges of their flanges to the liner plate cover the plate weld seams, in a manner similar to those installed at the Connecticut Yankee Station. These channels are zoned into test areas by dams welded to the ends of the sections of the channels. Fittings are provided in the channels for periodic testing of the weld seams for leaktightness under pressure. Typical liner details are shown in Figure 15.5-12. Testing of the liner is described in Section 5.5.

Revision 30-09/1/98 e SPS£ 15.5-40 periodic testing for structural purposes could be duplicated if at any time further tests were required. The minimum test level required to verify continued structural integrity would be no less than the 115%, or 52-psig initial test pressure.

Periodic inspection of the steel liner is accomplished by a type A leak rate test in accordance with 10 CPR 50 Appendix J. All welded joints and all penetrations of the liner are designed for periodic halogen gas testing.

In summary, no basis exists for attempting to develop structural performance information from leak rate tests conducted at moderate pressures.

15.5 REFERENCES

1. G. N. Bycroft, Forced Vibrations of a Rigid Circular Plate on a Semi-Infinite Elastic Space and on an Elastic Stratum, Philosophical Transactions, Royal Society, London, Series A, Vol. 248, pp. 327-368.
2. Karl Terzaghi, Evaluation of Coefficients of Subgrade Reaction, Geotechnique, Vol. 5, pp. 297-326, 1955.
3. B. 0. Hardin and W. L. Black, Vibration Modulus of Normally Consolidated Clay, Symposium on Wave Propagation and Dynamic Properties of Soils, University of New Mexico, 1967.
4. Stone & Webster Engineering Corporation, Nuclear Containment Structure Access Openirig.
5. B. Budiansky and P. Radkowski, Numerical Analysis of Unsymmetrical Bending of Shells of Revolution, A/AA Journal, August 1963.
6. S. Gere and S. Timoshenko, Theory of Elastic Stability, second edition.
7. J. N. Goodier and S. Timoshenko, Theory of Elasticity, second edition.
8. Stone & Webster Engineering Corporation, Report on Pressure Testing of Reactor Containment for Connecticut Yankee Atomic Power Plant, Connecticut Yankee Atomic Power Company, Haddam, Connecticut, 1967.
9. D. A. Davenport, Penetration of Reactor Containment Shells, Nuclear Safety, Vol. 2, No. 2, December 1960.

~

  • Attachment 3
  • Draft of Proposed UFSAR Revision to Reflect the Resolution of GL 96-06 North Anna Power Station Units 1 and 2 Virginia Electric and Power Company
  • Attachment 3
  • Draft of Proposed UFSAR Revision to Reflect the Resolution of GL 96-06 North Anna Power Station Units 1 and 2
  • Insert 1 Insert in NAPS UFSAR Revision 34-09/1/98, Section 6.2.4.1, before the last paragraph in page 6.2-92:

As a part of the issues identified in NRG GL 96-06, isolated containment penetration piping with confined fluid was reviewed for susceptibility to thermal overpressurization following a OBA. The linear elastic analysis criteria stipulated in the 1989 version of the ASME Boiler and Pressure Vessel Code Section Ill, Appendix F, were used for structural integrity evaluation. The internal pressure in piping penetrations during a

  • design basis accident (LOCA or MSLB) was calculated by taking into account the differences in the expansion of the fluid and the pipe, the temperature increase immediately following the OBA and credit for a limited amount of circumferential strain in the pipe. The analysis established that thermally induced overpressurization of isolated water-filled piping sections in the containment boundary could not jeopardize the ability of the accident mitigating systems to perform their safety functions and could not lead to a breach of containment integrity (Reference XX).
  • Insert 2 Add in NAPS UFSAR Revision 34-09/1/98, Section 6.2 References, the following new reference:

XX. Letter Dated March XX, 1999, Serial No.99-134, from Virginia Power to the NRG, Supplemental Response to Generic Letter 96-06.

Revision 3~9/1/98 *

  • 6.2.4 Containment Isolation System
  • NAPSUFSAR 6.2-92 6.2.4.1
  • Design Bases The containment isolation system has the following desigri bases:
1. For pipe penetrations through the containment, it provides, during accident conditions, at least two barriers between the atmosphere outside the containment structure and
a. The atmosphere inside the containment structure, or
b. The fluid inside the reactor coolant pressure boundary.
2. The design pressure of all piping and connecting component forming the isolation boundary is greater than the 45-psig design pressure of the containment. Piping forming the isolation boundary is designed to Class I or II of the American Standard Code for Pressure Piping -

ANSI B31.7-1969 Nuclear Power Piping.

3. Failure of a single valve or barrier does not prevent isolation.
4. Operation of the containment isolation system is automatic.
5. All isolation valves and equipment are protected from missiles and water jets originating from the reactor coolant system (RCS).
6. All remotely actuated valves and automatically operated isolation valves have their positions indicated in, and can be operated from, the main control room.

All isolation valves located outside the containment in accordance with General Design Criteria 55, 56, and 57 are located as close to the penetration as possible without limiting the service accessibility of the valves or interfering with other valves, piping, or structural members.

Approximately 70% of all outside isolation valves are located within 10 feet of the penetration.

The six valves not within about 20 feet of their penetration are on 3/8-inch lines and are located 50 and 60 feet from their penetration. These six valves are located at this distance to maintain separation of components, as in the leakage monitoring system, or due to the physical size of the isolation valve, such as in the sampling system.

The pressure retaining integrity of the containment pipe penetrations will be maintained under an applicable pressure, temperature, and mechanical load combination, including SSE effects. The intent of Regulatory Guide 1.29 for these penetrations is met by the load .

combinations .and-elastic-stress-limits -specified-in-Table *3;8-8. The-plastic-pipe* loads Mp and T p, which are far greater than the actual calculated pipe seismic loads, plus pipe design pressure and temperature effects, are each sufficient to fully yield the loaded pipe across its entire cross section at the penetration. The resulting penetration assembly stresses for these loads are limited to elastic such as 3S.

-.- e.rk' i ..* .

_!.J'l-:::. ontainment pipe penetrations are designed, built, inspected, and tested to the requirements of B31.7-1969, Class I or IL In 1971, these requirements were incorporated into

Revision 34-09/1/98 *

  • NAPSUFSAR
48. USNRC, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-*

6.2-120 Coolant-Accident, VSNRC Regulatory Guideline 1.82, November 1985.

49. A. W. Serkiz, Containment Emergency Sump Performance, VSNRC, NUREG-0897, October 1985.
50. Westinghouse LOCA Mass and Energy Release Model for Containment Design -March 1979 Version, WCAP-10325-A, April 1979. (Proprietary)
51. Westinghouse ECCS Evaluation Model - 1981 Version, WCAP-9220-P-A, Rev. 1 (Proprietary), WCAP-9211-A, Rev.. J, February 1982.
52. Mixing of Emergency Core Cooling Water with Steam: 1/3 Scale Test and Summary, (WCAP-8423), EPRI 294-2, Final Report, June 1975.
53. Topical Report Westinghouse Mass and Energy Release Data for Containment Design, WCAP-8264-P-A, Rev. 1, August 1975. (Proprietary)
54. American National Standard for Decay Heat Power in Light Water Reactors, ANSI/ANS-5.1-1979, August 1979.
55. Amendment No. 126 to Facility Operating Licence No. DPR-58 (TAC No. 7106), for D. C.

Cook Nuclear Plant Unit 1, Docket No. 50-315, June 9, 1989.

56. J. Wysocki and R. Kolbe, Methodology for Evaluation of Insulation Debris Effects, Burns and Roe, Inc., and Sandia National Laboratories, NUREG/CR-2791 and SAND82-7067, September 1982.
57. Letter Dated January 28, 1997, Serial No. 96-516A, From Virginia Power to the NRC, Generic Letter 96-06.
58. T.G. Carson, Critical Calculation Review, Systems Design Basis I)ocuments, North Anna Power Station, Stone & Webster Engineering Corporation, to D.L. Benson, Virginia Power, November 20, 1989.
59. J. G. Knudsen and R. K. Hilliard, 1969, Fission Product Transport by Natural Processes in Containment Vessels, Battle-Northwest, Richland, Washington, BNWL-943.
60. CORHYD - A Computer Program to Calculate Hydrogen Concentrations After a Design Basis Accident, NU-111, User's Manual Reissued 1983, Stone & Webster Engineering Corporation. (Proprietary)
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