Text

License Amendment Request 13-06; Plant-Specific Leak-Before-Break Analysis
	ML13220A073
Person / Time
Site:	Fort Calhoun
Issue date:	08/05/2013
From:	Cortopassi L; Omaha Public Power District
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LIC-13-0100, LAR 13-06
	Download: ML13220A073 (17)
	v • d • e

WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Enclosure, Attachment 3 contains Proprietary information.

Upon removal of Enclosure, Attachment 3, this letter is Decontrolled.

Omaa Piblic Power Dis 444 South 16e Street Mall Omaha, NE 68102-2247 August 5, 2013 LIC-13-0100 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

References:

1. Docket No. 50-285

2. Letter from NRC (Pao-Tsin Kuo) to OPPD (R. T. Ridenoure), "License Renewal Safety Evaluation Report for the Fort Calhoun Station, Unit 1," dated September 5, 2003 (NRC-03-0187) (ML032481209)

SUBJECT:

License Amendment Request 13-06; Plant-Specific Leak-Before-Break Analysis As described in the Safety Evaluation Report (SER) (Reference 2) for Renewed Operating License No. DPR-40, OPPD committed to provide a plant-specific leak-before-break (LBB) analysis before the period of extended operation (PEO), which begins at midnight, August 9, 2013. This commitment is also documented in Item No. 39 of Section 15.4, "License Renewal Commitment Listing" of the Updated Safety Analysis Report (USAR). Therefore, in accordance with our commitment and pursuant to 10 CFR 50.90, the Omaha Public Power District (OPPD) hereby provides the plant-specific LBB analysis performed by Westinghouse.

The enclosed Westinghouse Report (WCAP-17262-P, Revision 1) evaluated the effects of license renewal on the continued applicability of LBB for the Fort Calhoun Station's (FCS) reactor coolant loop piping.

The report demonstrates that the dynamic effects of the pipe rupture resulting from postulated breaks in the FCS reactor coolant primary loop piping need not be considered in the structural design basis for the license renewal period. Based on the report's findings, OPPD proposes to revise USAR Section 4.3.6.

Technical Specification changes are not necessary.

The Enclosure to this letter contains the license amendment request (LAR). Attachments 1 and 2 of the Enclosure contain the USAR page markups and clean pages respectively. Attachment 3 of the Enclosure contains the LBB Analysis, which is proprietary to Westinghouse Electric Company LLC.

As Attachment 3 contains information proprietary to Westinghouse Electric Company LLC, it is supported by an affidavit (Attachment 4) signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Employment with Equal Opportunity

WITHHOLD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390 Enclosure, Attachment 3 contains Proprietary information.

Upon removal of Enclosure, Attachment 3, this letter is Decontrolled.

U. S. Nuclear Regulatory Commission LIC-13-0100 Page 2 Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations. of the Enclosure contains a non-proprietary version of the LBB Analysis that is suitable for public disclosure.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-13-3768 and be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 310, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

This LAR has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c). OPPD has determined that this LAR involves no significant hazards consideration.

The basis for this determination is included in the Enclosure.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of Nebraska official.

The amendment will be implemented within 90 days of NRC approval. No other commitments are contained in this submittal.

If you have any additional questions, or require further information, please contact Mr. Bill R. Hansher at (402) 533-6834.

I declare under penalty of perjury that the foregoing is true and correct; executed on August 5, 2013.

Sincerely, Louis P. Cortopassi Site Vice President and CNO LPC/KGM/mle

Enclosure:

OPPD's Evaluation of the Proposed Change(s) c:S. A. Reynolds, Acting NRC Regional Administrator, Region IV (w/o Enclosure, Attachment 3)

J. M. Sebrosky, NRC Senior Project Manager (w/o Enclosure, Attachment 3)

L. E. Wilkins, NRC Project Manager (w/o Enclosure, Attachment 3)

J. C. Kirkland, NRC Senior Resident Inspector (w/o Enclosure, Attachment 3)

Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska (w/o Enclosure, Attachment 3)

LIC-13-0100 Enclosure Page 1 OPPD's Evaluation of the Proposed Change(s)

License Amendment Request 13-06 Plant-Specific Leak-Before-Break Analysis 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENTS:

1. USAR Pages - Markup

2. USAR Pages - Clean

3. Westinghouse Report "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Fort Calhoun Unit 1," dated July 2013 (WCAP-17262-P, Revision 1) (Proprietary)

4. Westinghouse Electric Company LLC, Affidavit for Withholding WCAP-17262-P from Public Disclosure (CAW-13-3768)

5. Westinghouse Report "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Fort Calhoun Unit 1," dated July 2013 (WCAP-17262-NP, Revision 1) (Non-Proprietary)

LIC-13-0100 Enclosure Page 2 1.0

SUMMARY

DESCRIPTION The Omaha Public Power District (OPPD) hereby proposes to revise the structural design basis for reactor coolant system (RCS) piping described in Section 4.3.6 of the Fort Calhoun Station (FCS), Unit No. 1, Updated Safety Analysis Report (USAR).

This proposal is derived from NRC and industry initiatives based on fracture mechanics technology and material toughness that demonstrate that leak-before-break (LBB) criteria can be applied to RCS piping.

OPPD proposes to eliminate RCS piping rupture from consideration in the FCS structural design basis and requests NRC approval of the attached plant-specific, LBB methodology performed by Westinghouse using the latest LBB criteria. The analysis was performed using the existing RCS leak detection capability and piping stress analysis loads. Thermal aging considerations were performed using approved Westinghouse methodology. The analysis is applicable for the period of extended operation (PEO), which begins at midnight on August 9, 2013.

No Technical Specification (TS) changes are proposed in this license amendment request (LAR).

2.0 DETAILED DESCRIPTION By letter dated July 7, 2003, (Reference 6.1) OPPD committed to complete a plant-specific LBB evaluation of the RCS piping using the latest LBB criteria to include the effects of thermal aging, plant-specific materials, operating temperatures/pressures, loads at welds in the primary loops, and weld fabrication.

OPPD noted that the plant-specific methodology would use the existing plant's RCS leak detection capability and the piping stress analysis loads for the FCS RCS configuration. The analysis was to be applicable for the PEO, and use a methodology from the Westinghouse Electric Company for thermal aging considerations. The letter also noted that Westinghouse had performed numerous plant-specific LBB analyses approved by the NRC, and has addressed thermal aging effects of the cast materials as applicable.

As noted in the License Renewal Safety Evaluation Report (Reference 6.2), the staff reviewed, the information provided by OPPD and determined that it adequately described the methodology to be used for the LBB analysis.

Accordingly, Westinghouse Report, WCAP-17262-P Revision 1 (Attachment 3), titled "Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for FCS Unit 1" is provided to demonstrate FCS compliance with LBB technology for RCS piping. The attached report documents the plant-specific geometry operating parameters, loading, and material properties used in the fracture mechanics evaluation. Mechanical properties were determined at operating temperatures. Since the piping systems include cast stainless steel, fracture toughness considering thermal aging was determined for each heat of material for the fully aged condition.

It should be noted that Revision 1 of WCAP-17262-P is applicable for "current" operating conditions while Revision 0 of WCAP-17262-P is applicable for the Extended Power Uprate (EPU) and Measurement Uncertainty Recapture (MUR) programs.

However, OPPD is not currently pursuing either of those programs and has therefore, attached Revision 1 of WCAP-1 7262-P.

LIC-13-0100 Enclosure Page 3 The evaluation includes analysis of the primary loop pipe of the RCS. This includes the two 38.5" outside diameter (OD) hot legs between the reactor vessel and the steam generator, four 29" OD cold legs between reactor coolant (RC) pumps and the reactor vessel, and four 29" OD cross over legs between the steam generator and RC pumps.

Westinghouse has performed considerable testing and analysis to demonstrate that RCS primary loop pipe breaks can be eliminated from the structural design basis of all Westinghouse plants. The concept of eliminating pipe breaks in the RCS primary loop was first presented to the NRC in 1978 in WCAP-9283 (Reference 6.3). That topical report employed a deterministic fracture mechanics evaluation and a probabilistic analysis to support the elimination of RCS primary loop pipe breaks. That approach was then used as a means of addressing Generic Issue A-2 and Asymmetric LOCA Loads.

As noted in the attached report, the Westinghouse methodology was confirmed by an independent NRC research study through the Lawrence Livermore National Laboratory (LLNL).

The studies applicable to Westinghouse plants east of the Rocky Mountains determined the mean probability of a direct loss-of-coolant accident (LOCA), i.e. RCS primary loop pipe break, to be 4.4 x 10-12 per reactor year and the mean probability of an indirect LOCA to be 107 per reactor year. Although FCS is a Combustion Engineering plant, Westinghouse has confirmed that the methodology is also applicable to FCS. The analyses provided in the Westinghouse report include evaluations of the four failure modes of the primary loop pipe; (1) stress corrosion cracking (SCC) (2) water hammer (3) low cycle and high cycle fatigue (4) wall thinning, creep, and cleavage.

The changes to USAR Section 4.3.6 for which NRC approval is sought are found in Attachments 1 and 2. Attachment 1 contains the markup of USAR Section 4.3.6 showing new text in double underline. contains the revised (i.e., clean) pages showing the text with revision bars in the right margin denoting where changes were made. Attachment 3 contains Westinghouse Topical Report WCAP-17262-P, Revision 1, which contains information proprietary to Westinghouse and should be withheld from public disclosure in accordance with 10 CFR 2.390. Attachment 4 contains an affidavit from Westinghouse Electric Corporation, LLC (WEC) for withholding the proprietary information contained in Attachment 3. The affidavit sets forth the basis for which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). Attachment 5 contains a non-proprietary version of Westinghouse Report WCAP-17262, Revision 1 that may be released to the public.

3.0 TECHNICAL EVALUATION

3.1

System Description

The primary purpose of the RCS is to remove the heat generated in the fuel and to transfer this heat to the secondary plant via the steam generators where this heat is used to produce steam for use as the prime mover in the main turbine-generator.

The secondary purpose of the RCS is to contain fission products that may be released by a fuel element defect and prevent the escape of fission products from the RCS to the environment.

LIC-13-0100 Enclosure Page 4 As stated in USAR Section 4.3.15, a leak detection sensitivity of 1 gallon per minute (GPM) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required as a result of implementing LBB methodology to exempt the reactor coolant loop piping system from consideration of the dynamic effects of a postulated primary pipe break (Reference 6.4).

3.2 Licensing Methodologies USAR Section 4.3.6 states that the RCS is exempt from consideration of the dynamic effects, including transient pressure waves, of a rupture in a hot or cold leg pipe by the application of LBB methodology.

The requirements for the exemption are, 1) the maximum bending moments do not exceed 42,000 in-kips for the highest stressed vessel nozzle/pipe junction and 2) at least one leakage detection system with a sensitivity capable of detecting 1 GPM in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is operable.

The application of the LBB methodology for nuclear power plant piping is provided for in GDC 4 of Appendix A of 10 CFR 50. Guidance for the application of this methodology is provided in NUREG-1061, Volume 3 (Reference 6.5) and in NUREG-0800, Section 3.6.3 (Reference. 6.6).

Generic analyses by Westinghouse to resolve the A-2 (Asymmetric Blowdown Loads on the Reactor Coolant System) issue were approved by the NRC and are documented in Generic Letter 84-04,"Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."

10 CFR 50, Appendix A, GDC 4 - "Environmental and Dynamic Effects Design Bases,"

states:

Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

It should be noted that Fort Calhoun Station was licensed for construction prior to May 21, 1971, and is committed to the draft General Design Criteria (GDC) published for comment in the Federal Register on July 11, 1967 (32 FR 10213) in lieu of 10 CFR 50, Appendix A.

The draft GDC are contained in Appendix G of the FCS USAR and are similar to 10 CFR 50, Appendix A. The draft GDC that are most applicable to LBB are Criterions 9, 16, 33, 34, 35, and 36.

NUREG-1061, Volume 3 provides a methodology that the NRC accepts for LBB submittals. The LBB approach described applies the fracture mechanics technology to demonstrate that high-energy fluid piping is very unlikely to experience double-ended ruptures or their equivalent in longitudinal or diagonal splits. The NUREG also provides a

LIC-13-0100 Enclosure Page 5 step-by-step approach to performing LBB analysis.

Westinghouse has followed the guidance of NUREG-1061, Volume 3 in performing the attached analyses.

NUREG-0800, Revision 1, Section 3.6.3 provides guidance to NRC reviewers on the specific areas to review and acceptance criteria for LBB applications.

The LBB methodology is reviewed for key parameters to ensure that acceptance criteria are satisfied.

Westinghouse Report WCAP-17262-P, Revision 1 justifies the elimination of RCS primary loop pipe breaks from the structural design basis for FCS Unit 1 as follows:

a)

Stress corrosion cracking is precluded by use of fracture resistant materials in the piping system and controls on reactor coolant chemistry, temperature, pressure, and flow during normal operation. Mitigation measures for primary water stress corrosion cracking (PWSCC) sensitive Alloy 82/182 welds will be implemented as required on the FCS reactor vessel nozzles weld locations.

b)

Water hammer should not occur in the RCS piping because of system design, testing, and operational considerations.

c)

The effects of low and high cycle fatigue on the integrity of the primary piping are negligible.

d)

Ample margin exists between the leak rate of small stable flaws and the capability of the FCS reactor coolant system pressure boundary leakage detection system.

e)

Ample margin exists between the small stable flaw sizes of item (d) and larger stable flaws.

f)

Ample margin exists in the material properties used to demonstrate end-of-service life (fully aged) stability of the critical flaws.

For the critical locations, flaws are identified that will be stable because of the ample margins described in (d), (e), and (f) above. The LBB assessment was performed based on deterministic fracture mechanics methods that have been previously approved by the NRC.

Based on the attached report as summarized above, the LBB conditions and margins are satisfied for the FCS Unit 1. Therefore, for the duration of the PEO, RCS loop piping need not be considered in the structural design basis for FCS Unit 1.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.1.1 Regulations 10 CFR 50, Appendix A, General Design Criterion (GDC) 4 allows the use of analyses reviewed and approved by the Commission to eliminate from the design basis, the dynamic effects of the pipe ruptures postulated in Standard Review Plan (SRP) Section

LIC-13-0100 Enclosure Page 6 3.6.2.

In providing the Westinghouse report, OPPD is also fulfilling a regulatory commitment (References 6.1 and 6.2). OPPD initially committed to submit the LAR by December 2006 but as noted in Reference 6.7, the date was changed from December 2006 to prior to the period of extended operation.

4.1.2 Approved Methodologies The application of the LBB methodology for nuclear power plant piping is provided for in GDC 4 of Appendix A of 10 CFR 50. NUREG-1061, Volume 3 provides a methodology that the NRC accepts for LBB submittals. NUREG-0800, Rev. 1 Section 3.6.3 provides guidance to NRC reviewers on the specific areas to review and acceptance criteria for LBB applications.

4.2 Precedent By letter dated February 12, 1993 (Reference 6.8), as supplemented by letters dated August 20, 1993 (Reference 6.9) and June 6, 1994 (Reference 6.10), OPPD submitted a LAR revising TS 2.1.4 "Reactor Coolant System Leakage Limits" LBB methodology to eliminate RCS pipe rupture as the structural design basis.

In Amendment No. 165 (Reference 6.4), the NRC subsequently approved OPPD's application of LBB methodology for RCS piping (primary loop).

By letter dated October 27, 2011, (Reference 6.11) the NRC approved amendments for Prairie Island Units 1 and 2 excluding consideration of the dynamic effects associated with the postulated rupture of the subject RCS piping from their current licensing basis.

4.3 Significant Hazards Consideration The Omaha Public Power District (OPPD) is proposing to revise USAR 4.3.6, by noting that the dynamic effects of the pipe rupture resulting from postulated breaks in the reactor coolant primary loop piping need not be considered in the structural design basis for the period of extended operation (PEO).

OPPD has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

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

Response: No.

The overall performance of protection systems remains within the bounds of the accident analyses.

The design of the reactor protective system (RPS) and engineered safety feature actuation system (ESFAS) are unaffected and these systems will continue to function consistent with their design basis.

Design, material, and construction standards are maintained.

LIC-13-0100 Enclosure Page 7 At Fort Calhoun Station (FCS), the bounding accident for pipe breaks is a large break loss-of-coolant accident (LBLOCA). The consequences of a LBLOCA have been previously evaluated and found acceptable. Since the attached leak-before-break (LBB) methodology verifies the integrity of reactor coolant system (RCS) piping, the probability of a previously evaluated accident is not increased.

The application of the LBB methodology does not change the dose analysis associated with a LBLOCA, and therefore, does not affect the consequences of an accident.

The proposed amendment will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Safety Analysis Report (USAR).

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

No new accident scenarios, failure mechanisms, or single failures are introduced because of the proposed change. All systems, structures, and components (SSCs) required for the mitigation of an event remain capable of performing their design function. The proposed change has no adverse effects on any safety-related SSC and does not challenge the performance or integrity of any safety-related SSC. The methods by which safety-related SSCs perform their safety functions are unchanged. This amendment will not affect the normal method of power operation or change any operating parameters.

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

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change does not involve a significant reduction in a margin of safety because the proposed changes do not reduce the margin of safety described in the FCS Technical Specifications or USAR.

The proposed amendment does not involve a change to any of the fission product barriers (i.e., fuel cladding, reactor coolant system or the containment building). The operability requirements of the Technical Specifications are consistent with the initial condition assumptions of the safety analyses. The proposed change does not affect any Technical Specification limiting conditions for operation (LCO) requirements.

This proposed amendment uses LBB technology combined with leakage monitoring to show that it is acceptable to exclude the dynamic effects of pipe ruptures resulting from postulated breaks in the reactor coolant primary loop piping from consideration in the structural design basis for the period of extended operation.

LIC-13-0100 Enclosure Page 8 The attached Westinghouse report demonstrates that the LBB margins discussed in NUREG-1061, Volume 3 are satisfied.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, OPPD concludes that the proposed amendment presents 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.

4.4 Conclusion 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 Commission'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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) 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 environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1.

Letter from OPPD (R. L. Phelps) to NRC (Document Control Desk), License Renewal Safety Evaluation Report for Fort Calhoun Station, Unit 1-Comments and Responses to Open and Confirmatory Items, dated July 7, 2003 (LIC-03-0089)

(ML031960063) 6.2.

Letter from NRC (Pao-Tsin Kuo) to OPPD (R. T. Ridenoure), License Renewal Safety Evaluation Report for the Fort Calhoun Station, Unit 1, dated September 5, 2003 (NRC-03-0187) (ML032481233) 6.3.

WCAP-9283, "Integrity of the Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," March, 1978 6.4.

Letter from NRC (S. Bloom) to OPPD (T. L. Patterson), "Fort Calhoun Station, Unit No. 1 - Amendment No. 165 to Facility Operating License No. DPR-40 (TAC No.

M85848)," dated August 25, 1994 (NRC-94-0246)

LIC-13-0100 Enclosure Page 9 6.5.

NUREG-1061, Volume 3, "Report of the U. S. Nuclear Regulatory Commission Piping Review Committee," November 1984 6.6.

NUREG-0800, Revision 1, Standard Review Plan: 3.6.3, "Leak-Before-Break Evaluation Procedures," dated March 2007 6.7.

Letter from OPPD (J. A. Reinhart) to NRC (Document Control Desk), Revision to Completion Date for License Renewal Commitment Pertaining to Leak-Before-Break (LBB) Evaluation, dated September 27, 2006 (LIC-06-01 11) (ML062770224) 6.8.

Letter from OPPD (W. G. Gates) to NRC (Document Control Desk), "Application for Amendment of Operating License," dated February 12, 1993 (LIC-93-0074) 6.9.

Letter from OPPD (W. G. Gates) to NRC (Document Control Desk), "Application for Amendment of Operating License," dated August 20, 1993 (LIC-93-0228) 6.10. Letter from OPPD (W. G. Gates) to NRC (Document Control Desk), "Application for Amendment of Operating License (TAC 85848)", dated June 6, 1994 (LIC-94-0133) 6.11. Letter from NRC (T. J. Wengert) to Northern States Power Company-Minnesota (M.

A. Schimel), "Prairie Island Nuclear Generating Plant, Units 1 And 2 - Issuance of Amendments RE: Request to Exclude the Dynamic Effects Associated with Certain Postulated Pipe Ruptures From the Licensing Basis Based Upon Application of Leak-Before-Break Methodology (TAC Nos. ME2976 and ME2977)," dated October 27, 2011 (ML112200856)

LIC-13-0100 Enclosure Page 1 Updated Safety Analysis Report (USAR) Markup

Page 1 of 50 II USAR-4.3 II Reactor Coolant System Component and System Design and Operation Rev 39 Safety Classification:

I Safety Usage Level:

I Information Change No.:

Reason for Change:

Preparer:

Issued:

Fort Calhoun Station

USAR-4.3 Information Use Page 29 of 50 Component and System Design and Operation Rev. 39 4.3.6 Reactor Coolant Piping The reactor coolant piping consists of 32-inch ID hot leg pipes from the reactor vessel outlets to the steam generator inlets and 24-inch ID cold leg pipes between the steam generator outlets to the pump suction nozzles and between the pump discharges and the reactor vessel inlets. The other major piece of reactor coolant piping is the 10-inch, schedule 160 surge line pipe between the pressurizer and the hot leg in loop 1. Design parameters for the reactor coolant piping are given in the piping list Table 4.3-6.

The reactor coolant piping was sized to obtain a coolant velocity which would provide a reasonable balance between erosion, corrosion, pressure drop and system volume. The surge line is sized to limit the frictional pressure loss through it during the maximum insurge so that the pressure differential between the pressurizer and the heat transfer loops is no more than 5 percent of the system design pressure.

The hot and cold leg pipes have no individual supports. The hot and cold legs are supported by connections to the steam generator, reactor vessel and reactor coolant pumps.

The reactor coolant system is exempt from consideration of the dynamic effects, including transient pressure waves, of a rupture in a hot or cold leg pipe by the application of "Leak-Before-Break" methodology (4-45)(4-49).

The requirements for this exemption are that 1) The maximum bending moments do not exceed 42,000 in-kips for the highest stressed vessel nozzle/pipe junction and 2) At least one leakage detection system with a sensitivity capable of detecting 1 gpm in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be operable (4-49).

The dynamic effects of the pipe rupture resulting from postulated breaks in the reactor coolant primary loop piping need not be considered in the structural basis for the period of extended operation. The dynamic effects of ruptures in attached branch lines, however, have not been exempted with LBB and would require NRC approval for such an exemption.

The reactor coolant piping is 316 stainless steel. The 10-inch surge line is also Type 316 stainless steel.

Thermal sleeves are installed in the surge nozzle, charging nozzle and shutdown cooling inlet nozzle to reduce thermal shock effects from auxiliary systems. All nozzles on the reactor coolant piping are constructed of stainless steel.

LIC-13-0100 Enclosure Page 1 Updated Safety Analysis Report (USAR) Clean

Page 1 of 50

-II USAR-4.3 II Reactor Coolant System Component and System Design and Operation Rev 39 Safety Classification:

I Safety Usage Level:

E Information Change No.: