NLS2018029, Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RR5-03
| ML18143B464 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/16/2018 |
| From: | Dent J Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2018029 | |
| Download: ML18143B464 (82) | |
Text
{{#Wiki_filter:H Nebraska Public Power District Always there when you need us 50.55a NLS2018029 May 16, 2018 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RR5-03
.Cooper Nuclear Station, Docket No. 50-298, DPR-46
References:
- 1. Email from Thomas Wengert, U.S. Nuclear Regulatory Commission, to Jim Shaw, Nebraska Public Power District, dated February 06, 2018, "Cooper Nuclear Station - Final RAI RE: Relief Requests RR-02 and RR-03 (EPIDs L-2017-LRR-065 and -066)"
- 2. Letter from John Dent, Jr., Nebraska Public Power District, to the U.S.
Nuclear Regulatory Commission, dated August 17, 2017, "10 CFR 50.55a Relief Requests for Fifth Ten-Year Inservice Inspection Interval" (MLI 724IA048)
- 3. Letter from John Dent, Jr., Nebraska Public Power District, to the U.S.
Nuclear Regulatory Commission, dated March 8, 2018, "Response to Nuclear Regulatory Commission Request for Additional Information for Relief Requests RR5-02 and RR5-03" (MLI8078A264)
Dear Sir or Madam:
The purpose of this letter is for the Nebraska Public Power District to respond to the Nuclear Regulatory Commission's Request for Additional Information (RAI) (Reference 1) related to the Cooper Nuclear Station (CNS) Relief Request RR5-03 (Reference 2). RAis for Relief Request RR5-02 were previously provided in Reference 3. The responses to the specific RAI questions for Relief Request RR5-03 are provided in the Attachment I to this letter. Attachment 2 contains a revised Relief Request RR5-03. Enclosure I provides the requested drawings. Enclosure 2 contains revised support information for RR5-03 .
. COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com I
NLS2018029 Page 2 of2 The commitments made in Reference 2 are being withdrawn by this letter due to the revision of the reliefrequest. A new commitment is proposed and included immediately following this cover letter. If you have any questions concerning this response, please contact Jim Shaw, Licensing Manager, at (402) 825-2788. Sincerely, l!::2:/!fi_J f('IOi'v- ~ (JOhi\ ~ Vice President - Nuclear and Chief Nuclear Officer /dv Attachments: 1. Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RRS-03
- 2. Revised Relief Request RRS-03
Enclosures:
- 1. Requested Drawings
- 2. RRS-03, Cooper Nuclear Station, Structural Integrity Associates, Inc. (with revised Attachment 3) cc: Regional Administrator w/ attachments and enclosures USNRC - Region IV Cooper Project Manager w/ attachments and enclosures USNRC - NRR Plant Licensing Branch IV Senior Resident Inspector w/ attachments and enclosures USNRC-CNS NPG Distribution w/o attachments and enclosures CNS Records w/ attachments and enclosures
- ~Entergy NUCLEAR MANAGEMENT MANUAL QUALITY RELATED INFORMATIONAL USE Regulatory Correspondence O-EN-Ll-106 REV.13C5 PAGE 29 OF 47 ATTACHMENT 9.3 LICENSEE - IDENTIFIED COMMITMENTS TABLE Sheet 1 of 1 This table identifies actions discussed in this letter which Nebraska Public Power District commits to perform. Any other actions discussed in this submittal are described for the NRC's information and are not regulatory commitments.
TYPE (Check one) SCHEDULED ONE-TIME CONTINUING COMPLETION DATE COMMITMENT/COMMITMENT NO. ACTION COMPLIANCE Commitment# - NLS2018029-01 X Within 60 days of The use of Relief Request RR5-03 is NRC approval of limited to a flaw size resulting in a leak Relief Request rate of no more than 5 gallons per RR5-03 minute.
NLS2018029 Page 1 of 8 Attachment 1 Response to Nuclear Regulatory Commission Request for Additional Information for Relief Request RRS-03 Cooper Nuclear Station, Docket No. 50-298, DPR-46 The Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) regarding Relief Request RR5-03 is shown in italics. The Nebraska Public Power District (NPPD) response to the request is shown in normal font. RAJ RR5-03-l In the "ASME Code Component(s) Affected" Section of the relief request on page 63 of the attachment to the August 17, 2017 submittal, the licensee discusses the RHR and RHRSWB systems. However, in other sections of the relief request, the licensee discusses only the RHRSWB piping. Confirm that the relief request covers only RHRSWB piping, not RHR piping. NPPD Response: RR5-03 only covers the Residual Heat Removal Service Water Booster (RHRSWB) piping. RAJ RR5-03-2 Provide the following: (a) All RHRSWB pipe outside diameters (or nominal pipe sizes) and schedules (i.e., wall thicknesses), (b) Specific maximum operating pressures in the various RHRSWB pipe segments with associated pipe outside diameters and thicknesses, and (c) RHRSWB piping isometric drawings showing the pipe support locations and piping routing. NPPD Response: (a) Nominal Pipe Size Outside Diameter Pipe Schedule Wall Thickness 3/4 1.050 160 0.219 1 1.315 160 0.250 1-1/4 1.660 160 0.250 1-1/2 1.900 160 0.281 2 2.375 80 0.218 3 3.500 80 0.300
NLS2018029 Page 2 of 8 4 4.500 40 0.237 6 6.625 40 0.280 8 8.625 40 0.322 10 10.750 40 0.365 14 14.000 STD 0.375 14 14.000 40 0.438 18 18.000 30 0.438 20 20.000 XS 0.500 (b) The design pressure and temperature of the RHRSWB system is 490 pounds per square inch gauge (psig) and 145°F. Each RHRSWB pump is electrically interlocked with its respective Residual Heat Removal (RHR) heat exchanger outlet valve (SW-M089A or SW-M089B). The RHR heat exchanger outlet valve is throttled to obtain the desired flow of 4000 gallons per minute (gpm). The RHRSWB pump starts when the SW-MOV-M089A(B) valve reaches a throttled position corresponding to 2500 gpm. At this flow rate, a discharge pressure of up to 465 psig has been observed. Once the pump is started, it will continue to run at 2500 gpm until Operations throttles the valve to obtain 4000 gpm. A newly rebuilt RHRSWB pump can operate with a normal operating discharge pressure near 420 psig to provide Service Water (SW) cooling through the RHR Heat exchanger. The pump discharge pressure can vary from approximately 370 psig to 420 psig depending on the condition of the pump. The maximum operating pressure in the various RHRSWB piping segments will vary according to pump discharge pressure (pumps are at the lowest elevation of the system), piping elevation, and friction losses. Paragraph 3 on page 66 of the original request was revised to clarify design and maximum operating pressures. Revised relief request included as Attachment 2. (c) Cooper Nuclear Station (CNS) flow drawings 2006 sheet 4 and 2036 sheet 1, along with their associated isometric key drawings and the isometric drawings for the majority of the piping system, are included in Enclosure 1 as listed in the table below. The pipe support locations are identified on the isometric drawings. Drawing Description Type 2006 Sheet 4 CNS Flow Diagram Control Building Flow Service Water System 2006 Sheet 4 - Iso Key CNS Isometric Key Control Building Isometric Key Service Water System 2036 Sheet 1 CNS Flow Diagram Reactor Building Flow Service Water System 2036 Sheet 1 - Iso Key CNS Isometric Key Reactor Building Isometric Key Service Water System 2851-1 18" SW-1 Class IVP - Reactor Bldg Isometric 2851-2 18" SW-1 Class IVP - Reactor Bldg Isometric 2851-3 18" SW-1 Class IVP - Reactor Bldg Isometric 2851-4 SW-1 Class IV-P, Class-I-S Seismic Service Isometric Water - Reactor Bldg
NLS2018029 Attachment 1 Page 3 of 8 2851-5 14" SW-1 Class IV P Reactor Bldg Isometric 2851-6 18" SW-1, Class IV P, Class IS Seismic, Isometric Control Building 2851-7 SW-1 Service Water, Control Building, Isometric Class IV P 2851-8 14" SW-1 Service Water, Control Building, Isometric Class IVP X2851-200 SW-1 Service Water, Building Reactor, Isometric Class IV P Piping X2851-201 2" SW-1 Drain, Reactor Bldg Isometric X2851-202 SW-1 Service (River) Water, Reactor Bldg Isometric X2851-203 Sheet 1 l" SW-1 Vent, Reactor Bldg Isometric X2851-203 Sheet 2 l" SW-1 Vent, Reactor Bldg Isometric X2851-204 SW-1 Service Water, Control - Basement, Isometric IV-P X2851-205 SW-1 Service (River) Water, Reactor Bldg Isometric X2851-206 2" SW-1 Drain, Reactor Bldg Isometric X285 l-207 Sheet 1 l" SW-1 Vent, Reactor Bldg Isometric X285 l-207 Sheet 2 l" SW-1 Vent, Reactor Bldg Isometric X2851-208 2" SW-1 Drain, Reactor Bldg Isometric X2851-209 SW-1 Service Water, Building Reactor, Isometric Class IV P Piping X2851-212 SW-1 Service Water, Building Reactor, Isometric Class IV P Piping X2851-350 3/4" SW-1 Drain Isometric X2851-351 Valve By-Pass Caps - SW-1, Control Bldg, Isometric Class IV P X2851-355 SW-1 Service Water, Building Control Isometric (Bsmt), Class IV P X2852-388 Service Water Booster Pumps B & D Gland Isometric Water System X2852-389 Service Water Booster Pumps A & C Gland Isometric Water System X2852-392 SW to RHR Corrosion Mon. Control Isometric Building RAJ RRS-03-3 Code Case N-513-4, Paragraph 5(a) requires, in part, that "A sample size of at least five of the most susceptible and accessible locations, or, iffewer than five, all susceptible and accessible locations shall be examined within 30 days ofdetecting the flaw ... " Paragraph 5(b) requires that "... When a flaw is detected, an additional sample of the same size as defined in [paragraph 5}(a) shall be examined ... " j
NLS2018029 Attachment 1 Page 4 of 8 On Page 67 of the attachment to the August 17, 2017 submittal, the licensee proposes the following two deviations from Code Case N-513-4: Deviation No. 1: " ... If a single flaw is identified during scheduled volumetric examinations, CNS may take credit for any previous examination ifperformed within 3 0 days prior to identification of the flaw, as a part of the same set of inspections. Credit will be taken for these examinations to meet the requirements of Code Case N-513-4 paragraph 5(a) ... ". This exception appears to convey that if a single flaw is identified, the licensee will examine a fewer number ofpipe locations than required by paragraphs 5(a) and 5(b) of Code Case N-513-4, because the licensee proposes to take credit for examinations performed within 30 days prior to identification of the flaw. (a) Clarify whether the proposed deviation will require the licensee to examine the same number ofpipe locations as required by paragraph 5(a). (i.e., if a flaw is detected, a sample size ofat least five of the most susceptible and accessible locations, or, iffewer than five, all susceptible and accessible locations shall be examined within 30 days of detecting the flaw.) If not, provide justification. (b) Clarify whether the above proposed deviation will meet the requirement ofparagraph 5(b). If not, provide justification. Deviation No. 2. The licensee proposes to increase the number of augmented examinations from 5 to IO pipe locations. (c) This is an enhancement and improvement to the sample size as specified in paragraph 5(a) of Code Case N-513-4, however, this may introduce unintended requirements. For example, if a flaw is detected, paragraph 5 (b) requires the licensee to increase the number of additional samples of the same size, i.e.,from a sample of JO pipe locations to 20 locations. Clarify if this is the intent ofDeviation No. 2. If not, explain the intent ofDeviation No. 2 with respect to the requirements ofparagraphs 5(a) and 5(b). NPPD Response: (a) NPPD will examine the same number of pipe locations as required by 5(a). NPPD withdraws our request to credit examinations conducted within the 30 days prior to finding a flaw as augmented examination. Therefore, most of paragraph 2 on page 67 of the original request was deleted. Revised relief request included as Attachment 2. (b) NPPD will examine additional locations as specified in the requirements of 5(b) as it applies to 5(a). (c) NPPD will examine the same number of pipe locations as required by 5(a). NPPD withdraws our request to increase the number of augmented examinations from 5 to 10 pipe
NLS2018029 Page 5 of 8 locations. Therefore, portions of paragraph 1 and 3 on page 67 of the original request were deleted. Revised relief request included as Attachment 2. RAJ RR5-03-4 Page 67 of the attachment to the August 17, 2017 submittal states that, if the Code Case is applied to a leaking flaw in the RHRSWB system, for leakage greater than 5 gallons per minute (gpm), the leakage will be stopped by use of a mechanical clamping device. (a) A leak rate of 5 gpm is not insignificant because it may cause flooding in a room if the sump drainage system does not have adequate capacity to drain the fluid, may affect safety-related components, and may impact personnel safety. Discuss the actions that the licensee will take before the leakage reaches 5 gpm. (b) Provide the allowable leakage rate beyond which the relief request will no longer be valid. (c) Discuss the minimum leakage rate that would prevent the RHRSWB from performing its intended fimction, the flow rate in the RHRSWB pipe that is needed to satisfy its designed function, and the average normal flow rate in the RHRSWB piping. (d) Discuss how the leakage rate in the RHRSWB piping can be detected and how the operators in the control room are notified. Discuss the minimum leakage rate that can be detected. NPPD Response: (a) Identified leaks will be quantified and entered into the CNS Corrective Action Program and assessed for operability in accordance with CNS Procedure 0.5.0PS, Operations Review of Condition Reports/Operability Determination. The evaluation process will consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgement. CNS will address all postulated negative effects ofleakage regardless of pressure and conclude there are no unacceptable resultant conditions; otherwise, the leakage would be deemed unacceptable. CNS will evaluate the impact of the leakage on other safety-related or nonsafety-related components with respect to the flooding analysis. In addition, CNS will implement a leak rate limit of 5 gpm, which is conservative in comparison to the leak rate that would be permitted based on the structural and flooding analysis. Therefore, in regards to the original submittal, the last paragraph on page 65 was revised, paragraph 3 on page 67 was deleted, and paragraph 4 on page 67 was revised to remove the safety factor of 4 and the use of an engineered mechanical clamp. Also, Reference 2 for Code Case N-705 was deleted. Revised relief request included as Attachment 2. During normal operation, sump pumps in the four (4) reactor building quad rooms are sized to handle leakage from the reactor building structure that includes the torus area. Two 50
NLS2018029 Page 6 of 8 gpm sump pumps in each quad room have the capacity to pump a total of 100 gpm of leakage from each quad room of the reactor building structure. The sump pumps in the Emergency Condensate Storage Room in the control building basement have a total pumping capacity of 100 gpm. Consequently, during normal operation, a 5 gpm leak would be mitigated due to the capacity of the installed sump pumps. Review of the CNS flooding analysis for impact of the 5 gpm leak in the building areas where the RHRSWB piping passes through or could leak into are summarized below. The table identifies the corresponding break flow in each of these areas that is based on system pressure (adjusting for head loss and elevation difference), break area, and discharge coefficient: Elevation (feet above sea level)/Building Area Break flow analyzed 859' Torus (18" SW-1) 967.5 gpm 881' NW Quad (18" SW-1) 954 gpm 881' SW Quad (18" SW-1) 954 gpm 903' RHR HX A (20" SW-1) 1413 gpm 903' RHR HX B (20" SW-1) 1413 gpm 882' SWBP Room (18" SW-1) 954 gpm All analyzed breaks are acceptable based on equipment available to safely shutdown and maintain the reactor in a safe shutdown condition. The break flows evaluated in the stations internal flooding analysis are significantly greater than a 5 gpm leak proposed by RR5-03. Therefore, from an internal flooding perspective, the 5 gpm leak proposed is very conservative and acceptable. (b) This relief request will be limited to a 5 gpm maximum leak in order to add additional margin in application of this relief request. (c) Each RHRSWB pump is verified to produce up to 4000 gpm as part oflnservice Testing quarterly and biennial performance testing. The RHRSWB System flow assumed in the analyses is 4000 gpm per pump with one pump operating in one loop. In this case, the maximum suppression chamber water temperature and pressure are 196°F and 12.4 psig, respectively, well below the design temperature of281 °F and 56 psig. CNS has evaluated the impact to RHR Heat Exchanger thermal performance based on reduced tube side flow of 3500 gpm and the impact to thermal capability is minimal. The thermal capability is driven largely by the terminal difference between the shell and tube side inlet temperatures. Therefore, since 5 gpm is two orders of magnitude less than 500 gpm, the impact to RHR Heat Exchanger thermal performance is less than minimal. The use of this relief request is limited to RHRSWB piping leaks up to 5 gpm which is well within the uncertainty of the flow instrumentation and within the capacity ofRHRSWB system to provide the required flow. (d) The discharge piping of the RHRSWB system is routed from the RHRSW booster pumps located in the 882' elevation of the CNS control building basement and into the reactor
NLS2018029 Page 7 of 8 building torus area (top of torus) at the 898' elevation, then up into the respective RHR heat exchanger rooms located on the 903' elevation of the reactor building. There are multiple ways a leak in the RHRSWB piping (i.e., both trains) can be detected and communicated to the control room. The CNS Operations department conducts operator tours that vary from once per week (full torus floor area) to twice daily in the reactor building RHR heat exchanger room and control building basement. In addition to operator tours, plant walk-downs conducted by various departments such as Radiation Protection, Maintenance, System engineering, management observations, etc. provide additional opportunities for a water leak to be detected and reported. Generally, leaks from the RHRSWB piping would be expected to be observed at a low point below the leak such as on the floor in these locations. Engagement by the CNS workforce provides reasonable assurance that a leak would be visually identified and reported to the control room in a timely manner. The minimum leakage that can be identified can vary from a small weep to a steady stream depending on the size of the raw water leak opening. More than likely a leak will not be detected until it manifests itself to the point that the leaking raw water would pool onto the floor below the active leak. The RHRSWB system is operated on an intermittent basis (minimum quarterly) so the probability of detecting a leak would be greatest when the RHRSWB system is in operation. Other times, the piping is continuously pressurized at the Service Water operating pressure of approximately 30 to 50 psig. Larger leaks could be detected by the Radwaste operators by observing changes in the floor drain and/or equipment drain collector tank levels. RAJ RR5-03-5 The "Duration ofProposed Alternative" Section of the relief request on Page 68 of the attachment to the August 17, 2017 submittal states, in part, that "A Section XI compliant repair/replacement will be completed prior to exceeding the next scheduled refueling outage or allowable flaw size, whichever comes first." The NRC staff notes that Table 4, "Allowable Axial and Circumferential Flaw Size Comparison, " in Enclosure 2, Attachment 3 of the submittal provides allowable flaw sizes. However, it appears that the allowable flaw size calculations are generic and not specific to CNS. (a) Provide the plant-specific allowable flaw sizes and associated calculations for the RHRSWB piping. (b) Discuss why the allowable leakage rate is not identified as a criterion to determine repair/replacement in the above statement as a condition on the use of the relief request. NPPD Response: (a) Table 4 of Enclosure 2, Attachment 3, is meant to illustrate the effect of pressure on the allowable through-wall flaw size. The "plant-specific" allowable flaw size is a function of
NLS2018029 Page 8 of 8 the applied loading at the specific location for which this Code Case will be applied. The exact location (if any) for which the Case will be applied cannot be known at the time of submittal. The size of a flaw will be limited to a size that does not exceed the acceptance criteria of the evaluation or a flaw size resulting in a leakage rate of greater than 5 gpm, whichever is less. (b) As required by N-513-4, l(f), NPPD will consider the effects ofleakage. The allowable leakage rate is, therefore, used as a criterion to determine repair/replacement. Paragraph 2 on page 68 of the original request was revised to include the limitation of leakage limit, which is the maximum leakage rate allowed by the evaluation or 5 gpm, whichever is lower. Revised relief request included as Attachment 2.
NLS2018029 Page 1 of 6 Attachment 2 Revised Relief Request RRS-03 Cooper Nuclear Station, Docket No. 50-298, DPR-46
NLS2018029 Attachment 2 Page 2 of6 10CFR50.55a Request.No. RRS-03 Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Hardship Without a Compensating Increase in Quality and Safety ASME Code Component(s) Affected All ASME, Section XI, Class 3 RHRSWB system piping with a maximum operating pressure less than or equal to 490 psig and a maximum operating temperature less than 200°F at the Cooper Nuclear Station (CNS). The RHR and RHRSWB Systems are designed such that RHRSWB operates at a higher pressure than RHR. The RHR and RHRSWB Systems are standby systems that typically operate during testing or plant shutdown. Under this design, if there is an internal leak within a RHR heat exchanger, RHRSWB water, which is raw water from the Missouri River, will leak into the RHR System.
- The safety objective of the RHRSWB System is to provide cooling to the RHR System without an uncontrolled release of radioactive material to the environment. The RHRSWB System is designed to provide an adequate supply of cooling water to the RHR heat exchangers during postulated accident and transient conditions to remove the design RHR System heat load and at adequate pressure to prevent uncontrolled release of fission products to the environment due to a RHR heat exchanger tube failure. .
RHRSWB System at CNS has exhibited a history of degradation similar to raw fresh water systems throughout the nuclear industry. Degradation requiring immediate action to address leakage or observed thinning in the system is generally due to localized corrosion mechanisms.
Applicable Code Edition and Addenda
CNS' applicable Code for the fifth IO-year ISi interval and the ISi program is the 2007 Edition of Section XI with the 2008 Addenda. CNS' fifth interval start~.d April 1, 2016 and ends February 28, 2026. Applicable ASME Code Requirements ASME Code, Section XI, IWD-3120(b) requires that components exceeding the acceptance standards ofIWD-3400 be subject to supplemental examination, or to a repair/replacement activity.
Reason for Request
In accordance with 10 CFR 50.55a(z)(2), NPPD is requesting a proposed alternative from the requirement to perform repair/replacement activities for degraded RHRSWB piping which has a
NLS2018029 Attachment 2 Page 3 of 6 10CFR50.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure maximum operating pressure in excess of 275 psig. Moderately degraded piping could require a plant shutdown within the required action statement timefrarnes to repair observed degradation. Plant shutdown activities result in additional dose and plant risk that would be inappropriate when a degraded condition is demonstrated to retain adequate margin to complete the component's function. The use of an acceptable alternative analysis method in lieu of immediate action for a degraded condition will allow NPPD to perform additional extent of condition examinations on the affected systems while allowing time for safe and orderly long term repair actions if necessary. Actions to remove degraded piping from service could have a detrimental overall risk impact by requiring a plant shutdown, thus requiring use of a system that is in standby during normal operation. Accordingly, compliance with the current code requirements results in a hardship without a compensating increase in the level of quality and safety. ASME Code Case N-513-3 does not allow evaluation of flaws located away from attaching circumferential piping welds that are in elbows, bent pipe, reducers, expanders, and branch tees. ASME Code Case N-513-3 also does not allow evaluation of flaws located in heat exchanger external tubing or piping. ASME Code Case N-513-4 provides guidance for evaluation of flaws in these locations. Proposed Alternative and Basis for Use NPPD is requesting approval to apply ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division 1," to the RHRSWB System piping having a maximum operating pressure of 490 psig. The operational and configuration limitations of Code Case N-513-4~ paragraphs l(a), l(b), and l(d), shall apply. The maximum operating temperature of200°F in paragraph l(c) shall also apply. Application of the Code Case will avoid accruing additional personnel radiation exposure and increased plant risk associated with a plant shutdown to comply with the cited Code requirements. The NRC issued Generic Letter 90-05 (Reference 1), "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," to address the acceptability of limited degradation in moderate energy piping. The generic letter defines conditions that would be acceptable to utilize temporary non-code repairs with NRC approval. The ASME recognized that relatively small flaws could remain in service without risk to the structural integrity of a piping system and developed Code Case N-513. NRC approval of Code Case N-513 versions in Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1 (Reference 3)," allows temporary acceptance of partial through-wall or through-wall flaws provided all conditions of the Code Case and NRC conditions are met. The temporary acceptance period has historically been the time to the next scheduled refueling outage. The Code Case also requires the Owner to demonstrate system operability due to leakage.
NLS2018029 Attachment 2 Page 4 of6 10CFR50.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure The ASME recognized that the limitations in Code Case N-513-3 were preventing needed use in piping components such as elbows, bent pipe, reducers, expanders, and branch tees and external tubing or piping attached to heat exchangers. Code Case N-513-4 was approved by the ASME to expand use on these locations and to revise several other areas of the Code Case. Enclosure 2, Attachment 1 provides a marked-up N-513-3 version of the Code Case to highlight the changes compared to the NRC approved N-513-3 version. *The following provides a high level overview of the Code Case N-513-4 changes:
- Revised the maximum allowed time of use from no longer than 26 months to the next scheduled refueling outage.
- Added applicability to piping elbows, bent pipe, reducers, expanders, and branch tees where the flaw is located more than (Rot) 112 from the centerline of the attaching circumferential piping weld.
- Expanded use to external tubing or piping attached to heat. exchangers.
- Revised to limit the use to liquid systems.
- Revised to clarify treatment of Service Level load combinations.
- Revised to address treatment of flaws in austenitic pipe flux welds.
- Revised to require minimum wall thickness acceptance criteria to consider longitudinal stress in addition to hoop stress.
- Other minor editorial changes to improve the clarity of the Code Case.
The technical basis for changes in Code Case N-513-4 when compared to NRC approved Code Case N-513-3 is provided in Enclosure 2, Attachment 2. The design basis is considered for each leak and evaluated using the NPPD Operability Evaluation process. The evaluation process must consider requirements or commitments established for the system, continued degradation and potential consequences, operating experience, and engineering judgment. As required by the Code Case, the evaluation process considers but is not limited to system make-up capacity, containment integrity with the leak not isolated, effects on adjacent equipment, and the potential for room flooding. Leakage rate is not typically a good indicator of overall structuml stability, where the allowable through-wall flaw sizes are often on the order of inches. The periodic inspection interval defined using paragraph 2(e) of Code Case N-513-4 provides evidence that a leaking flaw continues to meet the flaw acceptance criteria and that the flaw growth rate is such that the flaw will not grow to an unacceptable size. The effects ofleakage may impact the operability determination or the plant flooding analyses specified in paragraph l(f). For a leaking flaw, the allowable leakage rate will be limited to 5 gpm to limit effects of jet thrust force even though a structural evaluation of the subject piping
NLS2018029 Attachment 2 Page 5 of 6 10CFR50.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure and leakage effects would allow a much higher leakage rate than 5 gpm. Any leakage, if present, will be limited to the leakage allowed by the evaluation or 5 gpm, whichever is lower. During the temporary acceptance period, leaking flaws will be monitored daily as required by paragraph 2(f) of Code Case N-513-4 to confirm the analysis conditions used in the evaluation remain valid. Significant change in the leakage rate is reason to question that the analysis conditions remain valid, and would require re-inspection per paragraph 2(f) of the Code Case. Any re-inspection must be performed in accordance with paragraph 2(a) of the Code Case. The design pressure of the Class 3 RHRSWB system is 490 psig with a maximum operating pressure of approximately 420 psig. The background, history, and effects of using Code Case N-513-4 at a conservative pressure value of 490 psig in lieu of the current 275 psig limitation provided in the Code Case are contained in Enclosure 2, Attachment 3. A review of previous NRC submittals identified that the NRC has previously granted specific relief for leaks on high energy systems (see Enclosure 2, Attachment 3). NPPD is seeking relief for general application for limited degradation in the RHRSWB System raw water piping for a maximum operating pressure of 490 psig (conservative). Raw water piping degradation is a well understood phenomenon and the evaluation methods in Code Case N-513-4 are widely applied by the industry in raw water piping systems that operate at a pressure less than or equal to 275 psig without incident. The structural aspects of raising the allowable operating pressure to 490 psig were evaluated as discussed in Enclosure 2, Attachment 3. It was determined that Code Case allowable flaw sizes by both the Linear Elastic Fracture Mechanics and branch reinforcement methods used in Code Case N-513-4 were smaller as would be expected. The effects of jet thrust force were evaluated and it was determined there was little difference in force for a 0.50" diameter flaw size at 275 psig versus 490 psig. The study also determined that jet thrust force increases with increasing leakage rate and that it is appropriate to limit the application of this relief request to 490 psig. , Attachment 3 provides:
- 1) A review of relevant NRC approved relief requests
- 2) A structural integrity evaluation that includes:
- Design minimum wall thickness comparison
- Code Case N-513-4 allowable flaw size comparison
- Code Case N-513-4 cover thickness requirement comparison
- 3) A jet thrust force evaluation CNS will follow all requirements of Code Case N-513-4. With regard to augmented examination process as described in Section 5 of the Code Case, a sample size of at least five of the most susceptible and accessible locations will be examined within 30 days of detecting the original flaw.
NLS2018029 Page 6 of 6 10CFRS0.55a Request No. RRS-03 (continued) Cooper Nuclear Station Request to Use Code Case N-513-4 at a Higher System Operating Pressure In summary, NPPD will apply ASME Code Case N-513-4 and RG 1.147, Revision 17 (or later NRC defined revision as applicable) for evaluation ofRHRSWB piping flaws at CNS if Code repairs cannot reasonably be completed within the Technical Specifications required time limit. NPPD will apply a 490 psig maximum operating pressure in lieu of the 275 psig maximum operating pressure defined in paragraph l(c) of the Code Case. Code Case N-513-4 utilizes technical evaluation approaches that are based on principals that are accepted in other Code documents already acceptable to the NRC. In order to bolster defense-in-depth and avoid adverse consequences as a result of increasing the maximum operating pressure to 490 psig, NPPD is making one additional commitment to apply a 5 gpm leakage limit or lower as allowed by the engineering evaluation. This Relief Request is no longer valid for leakages greater than 5 gpm. The application of this Code Case, along with the additional commitment above, will maintain acceptable structural and leakage integrity while minimizing plant risk and personnel exposure by minimizing the number of plant transients that could be incurred if degradation is required to be repaired based on ASME Section XI acceptance criteria only. Duration Of Proposed Alternative The proposed alternative is for use of Code Case N-513-4 for Class 3 RHRSWB piping and components within the scope of the Code Case and the request herein. A Section XI compliant repair/replacement will be completed prior to exceeding the next scheduled refueling outage, or allowable flaw size, or leakage in excess of 5 gpm, whichever comes first. This relief request will be applied for the duration of the fifth 10-year ISI interval. If a flaw is evaluated near the end of the interval and the next refueling outage is in the subsequent interval, the flaw may remain in service under this relief request until the next refueling outage. Precedents US NRC letter to Exelon Generation Company Nuclear Fleet - "... Proposed Alternative to Use ASME Code Case N-513-4," dated September 6, 2016 (ML16230A237). US NRC letter to "Peach Bottom Atomic Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 - Relief from the Requirements of the ASME Code," dated March 19, 2015 (ML15043A496). References
- 1. NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping (Generic Letter 90-05)," dated June 15, 1990.
- 2. NRC Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17," dated August 2014.
NLS2018029 Page 1 of 32 Enclosure 1 Requested Drawings Cooper Nuclear Station, Docket No. 50-298, DPR-46 Note: Enclosed contains 31 drawings.
I 4 5 6 2 - - - - ~ - - - - - - - - - ' - - - - - - - - - - ~ - - - - - - - - - - - ' - - - - - - - - - - - " ' - - ~ - - - - - - - ' - - - - - - - - ~ - - - - - - - ' - - - - o_ _ _ _--'------1_1_ _ _ _ ~_ _ _ _1_2_ _ _~~ SEE DETAIL F OWG. 2036 SHT. 2 ,----0 SEE OETA!L R DWG. 20.36 SHT. 2 NOTES:
- t. AT SAMPLING POINTS CONNECTIONS, PROBES AND ROOT VALVES SHALL BE FURNJSHED AND INSTALLED UNDER CONTRACT E-69-4, PROBES SHALL BE FURNISHED AND INSTALLED lN ACCORDMNCE WITH CONTRACT PARA. 3.4.2 AND DWG. SP-07211 t!G. NO. AS
~LpL __ _l __ 2.
NOTED. CONTINUATION OF SAMPLING PIPING BY E-70-3. LINES DOWNSTREAM OF DRAIN VALVES sw-100, 86, 93, 79, 1179, 1180 AND 120 CONTINUE TO CONTROL BUILDING A BASEMENT SUMP 1L A
- 3. ALL VALVES ARE IN THE SERVICE WATER (SW) SYSTEM 14"SW-2 UNLESS OTHERWISE NOTED.
- 4. RHR SERVICE WATER BOOSTER PUMPS CLASS IS INJEcnON NOTE 4 WATER SYSTEM !S INTENDED FOR BACKUP SERVICE ONLY.
(TYP.) 5. REF. CED 6011141, CCN#l FOR CONNECTION DETAIL SEE NOTE 2
- 6. STRAINER BASKET REMOVED PER MWR 88-2966, IEJN 85-96 AND IR 93-06.
4"(B04) f B
"'J; "
RWD-1 lsw-1 BLIND flANGE B SEE DETAIL D OWG. 2036 SHT. 2 ,-----0 , I
~LpL __ J. __
DETAIL 1 {NOT USED) i
;L,
~
0 _JL 14"SW-2 s.o.
--'=~----,14"(sr4J----+------1 14"SW-2 J."
1-SW-105-14" 14" 85) s.o. SEE NOTE 1
~ -?;-
1426 116!'~::: SEE NOTE 2 C C SEE NOTE 5_) O ~ ~ E E OETAIL G SEE DETAIL T DWG. 2036 SHT. 2 CLASS IISICLASS IS DETAIL DWG. 2036 SHT. 2 2 RJVERWELL ZONE H-12 SUPPLY SEE OWG. 2006 SHT. 5
~
664 ZONE H-7 384B SEE DETAIL E DWG 20.36 SHT. 2 3 D r.* ~ 24" SW-2 FROM PUMP HOUSE
,_<10-,4-l-'r'--~~-o-ll
. N.I.C. [Gg-4 1 ..SW-Z S.O.
14"SW-2 1,-SW-105-14" 117!'~:::R 5
~
1 SEE OETAIL H D FOR CONT. SEE DWG. 2006 - ' 14
="--'~--il/f----f----1 s.o. SEE NOTE l OWG. 2036 SHT. 2 SHT. 3 ZONE G-7 14"(91) DETAIL 3 ZONE H-12 SEE NOTE 2 BLIND FL.ANGE 1~---
C.P. E E CLASS IJS ICL.ASS IS RIVERWELL
~~tP6lc. r-K*---ICl-'-="-'-'--l'*---'-'"--'-'--D-'----'--t~
2006 SHT. 5 ZONE G-7 639 ~40 F F {1090) 18"sw-1 G G 1-SW-124A-14" (1230) 14"SW-2 (1163) SEE llETAIL 3""""\ (1229) THIS DWG. '\- 1V,i" FOR FLUSHING 1)4" FOR FLUSHING (1228) (1164)
~tg:~T LOW [Al{538) ~'tiZrtT LOW [Kl (535)
EMERGENCY SUPPLY H @ SEE DETAIL J, TO CONTROL ROOM AIRCONO. SEE DWG. 2007, ZONE H-3
~rMf,~rr ~[(536) H DWG. 2036, SHT. 2 I!) {232) {1179) (537)[
CNS FLEX STRATEGY 1545 DRAIN HOSE CONNECTION (1180) ~I 10"(1531) , ,rn (541) s.c.
~~~ [ 0 - , - - - - - - - - - - ~ SEE NOTE 2 ,q,20) AS BUILT CONTROL B L D ~ .
454003589 t:. BASEMENT
~---------] ~~~1 STATUS: Release STATUS DATE: 06/02/2017
$ y:~~=~r_9fA OS APPRO\IEO: RDCHAMP VER: Nl. REV: 60 SIZE: F c BY H&V CONTRACTOR
*""2""(1"'°17")----f',,'--J 1 r~T~IR f~T1oN SPB 6-7-93
~ SEE OWG.
ZONE J-8 COOPER NUCLEAR STATION OJF 6-7-9J BURNS & ROE DISCHARGE TO TURBINE 2*sw-2 FLOW DIAGRAM BKW 6~7-9J
~~bW~.G sit~Wf ~i1EsR ---~1-°'sccw-"1cc17~_~,--- - - ~ CONTROL BUILDING SHT. 3 ZONE G-8 SERVICE WATER SYSTEM 2006 SH 4 L 2 3 4 5 6 7 10 11 12 _J
, - - - - - - - - ' - - - - - 2_ _ _----1_ _ _ _3_ _ _ __l__ _ _ _4_ _ _ _L--:------=-5---_J----6=--__:,,it;.-1----....:.7_ _ _ _L _ _ _~8---_JL_ _ _:_9_ _ __.1_ _ ___.'...1o':____ _ _L ___1~1~_ __J_ _ __.21_:2:..__ _~~
SEE DETAIL F DWG. 2036 SHT. 2 ~----0 NOTES:
- 1. AT SAMPLING POINTS CONNECTIONS. P SHALL BE FURNISHED AND INSTALLED PROBES SHA!.L BE FURNISHED ANO AND ROOT VALVES CONTRACT E-69-4, 0 IN ACCOROAANCE WITH CONTRACT PN'.A. 7211 FIG. NO. AS NOTED. CONTINUATION BY E-70-3.
- 2. LINES DOWNSTREAM OF CRAIN VALVES sw-100. 86. 93. 79.
1179, 1180 AND 120 CONTINUE TO CONTROL BUILDING A N.J.C. E69 41 BASEMENT SUMP 1L A 24* SW-2 FROM PUMP HOUSE
~<m. ~0~1N~E~_o,wc.
_____,.-.,__f' _ S.O.
- 3. ALL VALVES ARE IN THE SERVICE WATER (SW) SYSTEM UNLESS OTHERWISE NOTED.
2006
- f----'-14c_"e,SW='--~,4<9...6-)- - - - - - - -
NOTE 4 I - - 26- 5 -1_- 7-
- 4. RHR SERVICE WA1ER BOOSTER PUMPS CLASS JS INJECTION WATER SYSTEM IS INTENDED FOR BACKUP SERVICE ONLY.
(TY!'.) 5. REF. CED 6011141, CCN61 FOR CONNECTION DETAIL
- 6. STRAINER BASKET REMOVED PER MWR 88-2966, IEIN 85-96 AND IR 93-06.
655 ~56 B RWD-1 Jsw-1 B
~
SEE DETAIL D OWG. 2036 SHT. 2 ~-----0 DETAIL 1 (NOT USED) C 21-1--"
~
' /
14
'-'*sw='-,--~i/r*----+-----
2 14.(84) I l-'4c..c*sw"'--i-2r-L_~ I 41aos)_Jr*c. A~-
~;,;-
SEE NOTE 5) r.* 1426 2851-7 s.o. SEE NOTE 1 116
' NUBAR 158
. C SEE DETAIL G
~
OWG. 2036 SHT. 2 SEE DETAIL';; CL.ASS IISICLASS IS DETAIL 2 DWG. 20J6 SHT. 2 RIVERWELl. ZONE H-12 SUPPLY SEE DWG.
@ ;4B ~g~i i~1 5 664 SEE DETAIL E DWG 2036 SHT. 2 D r.* D 24* SW-2 FROM PUMP HOUSE N.I.C. E69-4 1
0 _JlL_ 14*sw-2 s/.0. (256) 1-SW-105-14* 14.(92) 117~,~~~ 5
~
1 SEE DETAIL H FOR CONT. SEE DWG. 2006 > - - - - ' J ' - - - ~ - < ~ . ( g f - i - )- - - , , - - - - , 14"sW-1 s.o. SEE NOTE 1 DETAIL DWG. 2036 SHT. 2 SHT. J ZONE G-7 3 1431 2851-6 ZONE H-12
~
1440 E E CLASS IIS I CLASS IS RM:RWELL
~~:Pfi1c.
2006 SHT. 5 ZONE G-7 639 ~40
~---K~-"""-'.YX1--'-'"--LL-D...L.--..L-':,4oi32~
NC> SEE DETAIL C OWG 2036 SHT. 2 SEE NOTE 1 1-SW-105-14* 14 (19CV)
,1*
1422 2851-6 F F SW-2 SW-1 (1090) (1522) CL.ASS IISIClASS !S I I S.0."-.14*1
~"
(106) CONT. ON CONT. ON DWG. 2262
~5560;~. 5 648 ZONE c-7 J;
~Ly$~LL--K::t---J::::t-'-==-'--t-:r--"'y,'-*-'L--D-1.---1.-'J~
3 RESIDUAL HEAT REMOVAL
~
.1-...
I LRP-RACK-141 SEE DETAIL 5 THIS OWG. I X2852-392 1e*sw-1 2006 SHT. 3 SERVICE WATER BOOSTER PUMPS I I
/~ ZONE G-5
---- ------;;::,c""'------------~*sw~-~1c__~----------------------------JI I G q1532)
Q
~
'CNS FLEX STRATEGY HOSE CONNECTION
{1149) ,,.. SEE DETAIL I OWG. 2036. SHT. 2 SEE DETAIL 4 1-SW-146A-2* (1148) ~- L _- 15*sw-1 DETAIL 4 G 2852-18 2851-7 1-SW-124A-14* ZONE G-5
~
2*sw-2 (1230) 14"SW-2
@ 2851-6 1-Sfl-1168-2" {1163) c~~s~o~cW~ 11-j/ (1227) SEE DETAIL 3 '\._
4*(1533) (1229) THIS OWG. ~ 14 2 1 6 1}4" FOR FLUSHING "SW- \..__,.i (1228) @ {1164) 2*sw-2 2852-16 ~z.r;T LOW ~ (535) H @ -,.~===+a,cc-J VAf.VE SEAT V-J:ING 0C:(536)+'-'Cl..CCSL""'°~,c..\ lF H SEE DETAIL J, OWG. 2036, SHT. 2 {1179) (537)[ I!] (232) DRAIN 1545 1544 {1180) l HLDR- 1;r"" RHRA3 {541) s.c. HLDR- C: SEE NOTE 2 X2851-204 INFORMATION ONLY RHRA2 C.P. 454244547
- J HLDR-RHRA1 t:;
~
CONTROL B L D ~ . BPSEMENT t1~:'i;T~fA ;~ t121) N S.0. STATUS: Relea5e FJO STATUS DATE: 06/02/2017 1546 3/4.SW-1 :J; SEE TELL TALE
- THIS DRAWING GENERATED BY NPPD OS APPRO\IED: RDCHAMP YER: AY REY: 21 SIZE: F
*-~----<~'°'-'
2 7 BY H&V CONTRACTOR _ _ _ _ vi
"(ll ) ~ r~R~1t°~T10N X2852-223 NOTE 2
~ 1:;_2102*
1543 SEE DWG. 2018, ZONE J-8 COOPER NUCLEAR STATION ~~-~-'~1~:-v~o~2 BURNS & ROE STNR-141 ISOMETRIC KEY 8 CONTROL BUILDING f--W_M-r7/_24~~_z..,__ _ _ _~--1~8 DETAIL 5 ZONE G-8 SERVICE WATER SYSTEM 2006 SH 4 3 L 2 3 4 5 6 7 8 10 11 12 _J
________,______2 _ _ _--1._ _ _ _3_ _ _ _. J . . . _ _ ~_ _4_ _ ___JL__ _ _ _s____...L_ _ _~s____:"l~L---.:...7_ _ ___J_ _ _~s::____L _ _ __:9:__ _ _L __~1'.-':o'..___ __J_ _ ____'1._'1__ _ __J__ _ _~12::__ _ _ ~ 1 1-sw-10ss-1s* ,a*-sw-1 1-sw-10SA-1a* ,e*-sw-1 1. UNLESS OTHERWISE INDICATED ALL PRESSURE AND/OR FLOW INSTRUMENT ROOT LINES SHALL BE 3/4*: ROOT VALVES SHALL BE J/4"V253x. LINE UP TO AND INCLUDING ROOT VALVE SHALL BE FURNISHED ANO INSTALJ.£0 UNDER CONTRACT E-69-4. A 2, ALL TEMPERATURE TEST AND PROTECTING A SW-245 WELL PENETRATIONS SHALL BE 1* NPT, VENT ,------,,r--:-:-::----"-'".=-ae,SW!.:-;£2_-"1-:.,SW,c-°'1cs,03e,;A~-1!..SBc.."-.:::.::.J UNLESS OTHERWISE NOTED. SW-396(A} 1-SW-1058-1
- 1* SW:~~~ 3. ALL *sw-1* &: "Sll-2" PIPING SYSTEMS SHOWN
&_ t.:J 2* SW-1294 HEREON SHALL BE ClASS IS SEISMIC ClASSIFICATION, EXCEPT FOR LOW POINT DRAINS.
~RAIN
~ 4. ALL SERVICE WATER LOW POINT DRAINS TO SW-MOV-898 NOT FULLY CLOSED c1-------l y SEE BE ROUTED TO 4 FOOT HIGH COLLECTION STAND PlPE IN SOUTH EAST CORNER OF REACTOR BUILDING TO MINIMIZE LIMIT SWITCH ON VALVE NOTE 13 RHR-121 CONT. ON CONTAMINATION OF RAOWASTE FACILITIES.
r2c'+-----"::J----? DWG. 2040, SEE SH.2 5, "CP" INDICATES LOCATION OF CATHODIC NOTE 13 ZONE G-10 PROTECTION LEADS TO BE INSTALL[O IN ACCORDANCE WITH PARAGRAPH 1,4.14.2, 10 91 A25-59 -------1 ---------1 I CONTRACT E-69-4. B I I
- 6. DELETED B I I
- 7. AT SAMPLING POINTS CONNECTIONS, PROBES AND I
~,~M10 145A ',----t FROM E/S 10 1458',-----t I ROOT VALVES SHALL BE FURNISHED AND INSTALLED UNDER CONTRACT E-69-4. PROBES SHALL BE FURNISHED ANO INSTALLLED IN ACCORDANCE WITH I
f>. HLDR-RECBl I TO
,--ms-10-131 I
I CONTRACT PARA. 3.4.2 AND OWG. SP-07211 FJG. NO. AS NOTED. CONTINUATION OF SAMPLING
~~ ~ I 2030, SH.1 PIPING BY E-70-3.
I
!~'"
~
/!
S.O.
~ S W 104A 2"' ~~
1s* 1 s1 ta* SW-651MV
~ I I
I I I I I ZONE H-9 I I I I I I 8. 9. MANUAL OVERIDE TO CLOSE VALVE ONLY AFTER VALVE HAS OPENED AUTOMATICALLY ON START OF BOOSTER PUMP-TtP. MV-122, 123.124, 125,0N ALL SW BOOSTER PUMPS. 6_ 'S WHERE SHOWN DENOTES WORK LISTED IN PART IIA. PARA. 2.2.3.1 OF SPEC. E-73-58. 15* SW-MOV-89A r;o;;i rdici I
~-~
~-~
(OLD B&R 652MV) 1 t. ALL VALVES ARE SERVICE WATER (SW) UNLESS 10 9-J 10 9-J DENOTED OTHERWISE. C C
~,_,
1-SW-1038-18* 10 9-J 10 9-J 12. REFERENCE GE DRAWING 1 17C3324, SH.2
~,_,
FOR lNSTRUMENT PIPING ANO VALVE INSTAI..LATION AT INSTRUMENT RACK 25-59.
- 13. REFERENCE GE DRAWING 117C3326, SH.2
- J RHWA-PLUG-MOV89A FOR INSTRUMENT PIPING ANO VALVE
@*!'4-lol-----i'C:,- TO EQUIP, DRAIN SEE DWG. 2028, ;l RHWA-PLUG-MOV89B INSTAl..lATION AT INSTRUMENT RACK 25-62.
ZONE E-1 14. FOR LINE DESIGNATION, EXAMPLE 1-SW-1048-18*. THE A OR B DOES NOT NECESSARILY DENOTE LOOP A OR 8. FLOOR FLOOR 15. WHERE LINES ARE INTERCONNECTED ANO CONTINUED ON OTHER DRAWINGS. ZONE NUMBERS ARE APPROXIMATE ONLY. t 6. COUPON HOLDER AND ASSOCIATED COMPRESSION 2" SW-159 GLAND MAY OR MAY NOT BE INSTALLED. IF NOT RESIDUAL HEAT REMOVAL INSTALLED. ISi BOUNDARY/CLASS BREAK REVERTS D HEAT EXCHANGER # 1A 1-SW-130A-2" TO NEAREST CLOSED VALVE (1524). D
,- SW-418 1 }2':..sw-2 1-sw-11a-1W SEE OWG.
2031 SH.2 ZONE E-5 ~ Lµ 2" AlR RELEASE al, SW-1442 SO~
/I
~I"
~sw-to4s-,a*
~o-~
1
---c::J
,a*-sw-2
~~--~~5 1 -SW-413 tO'i T
J E r1~~:-~~:' ________ _f RCC-11 SW-2 ~~s 1-SW-104A-24* E r--------------}-------------------- MO ;'""] I J: 24*-sw-2 CONT. ON 2006 SH.3, ZONE H-9 RMS FROM M0-711 I DJV.1 - - - ..,,:~E ~ol~ SH.2. I TO DISCHARGE FLUME FROM ~0-697 1-SW-104B-24* SEE OWG. FOR CONT. ~ CONT. ON 2031 SH.2, 1* TELL TALE TO NEAREST 24*-sw-2 CONT. ON
~
2031 SH.2 2006 SH.3, ZONE 8-5 2~\\.D~~:1 ZONE H-2 FLR. DRAIN ZONE H-9 SW-184 ZONE E-3 2* SW-251 1* SW-253~ oAAJN [fil vet<T 3" SW-1494 F SPECTACLE F FLANGE 14*-sw-2 ~,,----------------------------- SPECTACLE FLANGE 1----~lf---------------------~ LOW POINT DRAIN HEADER G 2* SW-163 G HOSE CONNECTION CONT. ON 2038, 20NE J-1 , 3*-F:::~lr- {l 2
~- S#-797 1 e7 7 ~
~ ~
~
I
'J.,
2* SW-170 H H CONNECTION FOR PORTABLE PUMP. AS BUILT 454003629 STATUS: Release N'J' VENT SW-489(A} t:
, 11 STATUS DATE: Dl/29/2018 DS i\PPROVED: JAFORBE VER: AM REV: A4 SIZE: F J (EMERGENCY CORE FLOODING - - ;:!ii K.P.B. 12/15/93 14* RHR':;ocv----1
.l, FLOW DIAGRAM OJf 12/15/93 BURNS & ROE
.llilIE; CONT. ON 2040 SH.1. ZONE 0-4 ALL ITTMS MARKED THUS0ARE
*J>S BUILr ITEMS ANO ARE NOT REACTOR BUILDING BKW 12-18-93 SHOWN ON PIPING DRAWINGS.
SERVICE WATER SYSTEM 2036 SH L 2 3 4 5 6 7 8 10 11 12 _J
_ _ _ __:__ _ _-1_ _ _ _.:2_ _ _ _1__ ___:_3:..__ _ __J__ _ _ _4_ _ _ __1___ _ _ _s_____L_ _ _ _6_ _~ ! : . - - - . . 1 - - - - - - - _ _ J _ - - - - s - - - - J _ - - - - - - L - - - - - 1 0_ _ _ _-'-_ _ _ _1_1-::::=---~----12_ _ _-----i~ 1-SW-1058-18* 1-SW-105A-18* 1a*-sw-1
= 1. UNLESS OTHERWISE INDICATED ALL PRESSURE
~~ . - - - - - f 4*-Dw-1s CONT. ON AND/OR FLOW INSTRUMENT ROOT LINES
,..,..._....,,....._r--.,...,_ ~ DW-l4 5
U 2029, ZONE E-.3 SHALL BE 3/4*; ROOT VAI...VES SHALL BE 3/4"253x. LINE UP TO AND INCLUDING ROOT VALVE SHALL BE FURNISHED AND
' ~ INSTALLED UNDER CONTRACT E-69-4.
A 2852-9 f7i ~ 7 ,-.-.,,,,,,,...,_..,,-., 1._1 2851-.3 2, ALL TEMPERATURE 1£ST AND PROTECTING A SW-245
~ ~
WELL PENETRATIONS SHALL BE 1* NPT. VENT .----~---~'~"*~-s~w~-~'~~,-~s~w-~1~0J~A~-'="~*----1 UNLESS OTHERWISE NOTED. sw-.396[:fil 18*-sw-1 1-SW-1058-1
- 1* sw?..~~r: X2852-239 3. ALL *sw-1* & *sw-2* PIPING SYSTEMS SHOWN A __ .:. 2* SW-1294 HEREON SHALL BE CL.ASS IS SEISMIC CLASSIFICATION, EXCEPT FOR LOW POINT DRAINS.
~RAIN 4. ALL SERVICE WATER LOW POINT DRAINS TO BE ROUTED TO 4 FOOT HIGH COLLECTION STAND PIPE IN SOUTH EAST CORNER OF REACTOR BUILDING TO MINIMIZE CONTAMINATION OF RADWASTE FACILITIES.
- 5. *cp* INDICATES LOCATION OF CATHODIC PROTECTION LEADS TO BE INSTALLED IN ACCORDANCE WITH PARAGRAPH 1.4.14.2,
---------1 CONTRACT E-69-4, I
B I 6. DELETED B I
- 7. AT SAMPLING POINTS CONNECTIONS, PROBES AND I ROOT VALVES SHALL BE FURNISHED AND INSTALLED UNDER CONTRACT E-69-4. PROBES SHALL BE lJ<oTE ~~M10 1458',,----t I
FURNISHED AND INSTALLLEO IN ACCORDANCE WITH CONTRACT PARA. J.4.2 AND owe. SP-07211 FIG. HLDR-REC81 f" TO r--TRS-10-131 DWG. 2030, SH.1 TO 1--rns-10-1.31 2030. SH.1 I I NO. AS NOTED. CONTINUATION OF SAMPLING PIPING BY E-70-J. ZONE H-9 ZONE H-9 I
~~i
- 8. MANUAL OVERIDE TO CLOSE VALVE ONLY I AFTER VALVE HAS OPENED AUTOMATICALLY Tt X2851-209 I ON START Of BOOSTER PUMP-TfP. MV-122, 2851-.3 I 123, 124.125,0N ALL SW BOOSTER PUMPS.
1-SW-104A~Tr-- *- I
~=~~ 18* SW-651MV I 9. 6 's WHERE SHOWN DENOTES WORK LISTED IN PART IIA, PARA. 2.2.J.1 OF SPEC. E-7J-5B.
I 2B51-1
" ~-~ 11. ALL VALVES ARE SERVICE WATER {SW) UNLESS DENOTED OTHERWISE.
C '"' ~~ C
~
10 9-J 10 9-:5 12. REFERENCE GE DRAWING 117CJJ24, SH.2
~,_,
FOR INSTRUMENT PIPING AND VALVE 9-J INSTALLATION AT INSTRUMENT RACK 25-59. X2852-201
- 13. REFERENCE GE DRAWING 117CJJ26, SH.2 J RHWA-PLUG-MOV89A FOR INSTRUMENT PIPING AND VALVE INSTALLATION AT INSTRUMENT RACK 25-62.
- l RHWA-PLUG-MOV89B
- 14. FOR LINE DESIGNATION, EXAMPLE 1-SW-1048-18*,
THE A OR B DOES NOT NECESSARILY OENOTE LOOP A OR 8. FLOOR
~ FLOOR 15. WHERE LINES ARE INTERCONNECTED AND CONTINUED
<Xl 2851-4 ON OTHER DRAWINGS, ZONE NUMBERS
~f---~~
~
ARE APPROXIMATE ONLY.
~l
- 16. COUPON HOLDER AND ASSOCIATED COMPRESSION GLAND MAY OR MAY NOT BE INSTALLED. IF NOT RESIDUAL HEAT REMOVAL RESIDUAL HEAT REMOVAL SW-2 sw-1 I INSTALLED, ISi BOUNDARY/CL.ASS BREAK REVERTS D HEAT EXCHANGER # 1A 1-SW-130A-2* HEAT EXCHANGER # l B ~ TO NEAREST CLOSED VALVE (1524). D 2851-2 1 Y.z:.sw-2 1-sw-118-1¥.z* 1-SW-1308-2*
SEE DWG. 20.31 SH.2 ZONE E-5 X2852-209 1-SW-106A-18* 2852-5 E 1-SW-104A-24* E 24"-SW-2 ~g~J* s°tt\, ZONE H-9 TO DISCHARGE FLUME SEE DWG. 20.31 SH.2 ZONE 6-5 3* SW-1494 F SPECTACLE F fl,\NGE 14"-SW-2 ~-------------.-----------------' SPECTACLE fl,\NGE
>------,~--------~-\---------------
LOW POINT DRAIN HEADER 2852-54 AVSPVB55 X2852-206 G :J 2* SW-16.3 G
~
X2852-212 z HOSE CONNECTION g l
~ )3
~ ~ X2852-207
~ C~2a§:s1:1-},::::>-.'. 2851-1 2652-6 2852-9 2" SW-170 H H 2852-8 CONNECTION FOR PORTABLE PUMP.
2624-JA INFORMATION ONLY 454031323 STATUS: Releose FIO STAT\JS DATE: 01/29/2018 OS APPROI/EO: JAFORBE VER: AM REV: 40 SIZE: F
"""9Eft K.P.B. 12/15/93
~~~!;.,!!~~,J!!~~~E..S~;,j::::::e;;;;,:::::+/-::=::c::::+/-IS=OIM:=:ETtR=[IC::::::IK=E=Y:;::::::t::::::;::::::::l:=::;::::=f.~~,~1r,;~7s~;'-"J,, BURNS & ROE CONT. ON 2040 SH.1, ZONE 0-4
=ALL ITEMS MARKED mus 0
*AS BUILr ITEMS ANO ARE NOT ARE REACTOR BU[LD[NG ~
8KW 12-18-93 SHOWN ON PIPING DRAWINGS. SERV!CE WATER SYSTEM 2036 SH L 2 3 4 5 6 7 8 10 11 12 _J
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(STAINLESS STEEL) MATERIAL SPECS. (CARBON STEEL) MATERIAL LIST DC 9D-174A CODE DESCRIPTION SIZE MK. QUA. SIZE CODE NO. SCH. lYP. SPEC. & GR. DESCRIPTION P-1D PIPE 2* &: SMALLER 3/4" SW S.S. GLOBE VALVE 800 lb. F-6A nTI!NGS .-sTM A240 TP 316 RESTRICTING ORIFICE J-1 S.W. JOINTS 304 FLEXITE GASKEr 300 lb. J-6 SCRO. JOINTS 4 3/4" SW S.S. BAU. VALVE Fl-2A-3RF FLANGES 30° 3/4" 80 ASTM A.312 TP:504 PIPE B-2A BOLTS 3/4" SW "5TM A182 f.304 90' ELBOW 3000 PSI G-1 GASKETS SEE NOTE B 1 112* )( 3/4. SW ASTM A105 GR2. REDUCER INSERT 6000 PSI U-4 3/4" SW AS1M A10S CR:2. SOCKOLET 6000 PSI UNIONS 2* & SMALLER END Fl-1A-3RF FlANGES SEE NOTE 7 3/4" SW 1EE 3000 PSI J/4'"' X 1/2* SCRO ASTM A182 F304 REDUCING COUPLING JOOO PSI 3/4" ASTM A182 F304 COUPLING 3000 PSI AIS! 316 S.S. SCRO TUBE & SOCKET 0-100 PSI PRESSURE GAUGE ASME SA 351 3/4" SCRO GRCFBt.4 RELIEF VALVE 1 1;2* SW ASTM A105 GR2. TEE 6000 PSI SCRO .ASTM .'312 TP J04 NIPPLE 3/4" 1 ENO 4 3/4" SCRO UAI..LEABLE lRON CHICAGO QUICK CONNECTS PLATED STEEL LONG SHANK COUPLING & SINGLE BOLT CLAMP 3/4" SW FlANGE 300 lb. BOI..T--ASTMA193G 16 S/8" NUT-ASTM A194 BOLTS & NUTS
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BENDING PROCEDURES DESIGN CONDITIONS PRESSURE RWD-1 SW-1 PROCUREMENT ANALYTICAL 175psig 490psig TEMPERATURE PROCUREMENT ANALYTICAL 100"F 145"F HYDROSTATIC TEST 735pSig (EE 02-033) WELDING PROCEDURES N.D.T. REQUIREMENTS MP-LP X-RAY VISUAL 100% SPECIAL NOTES
- 1. nus DRAWING SUPERSEDES DRAWING 2852-65.
- 2. THE RWD-1 SYSTEM OESJGNATION REFERS TO THE GlANO WAlER SOURCE DURING NORMAL OPERATION. HOWEVER, COMPONENTS ARE Sllll. CONSIDERED PART OF THE SERVICE WAlER SYSTEM.
- 3. REFERENCE DC 90-174A ANO AMENDMENT 1.
- 4. SLOPE IS APPROXIMAlELY 1/8* PER FOOT.
- 5. THREADED CONNECTIONS ARE SEALED WITH RECTORSEAI.. OR AN APPROVED EQUIVALENT.
- 6. CARBON STEEL MATERIAL SPECIFICATIONS PER E-69-4 MAlERlAL SPECIFICATION CODE TABLE.
- 7. THOSE IDENTIFIED WITH AN @ ARE FL-1A-3RF.
OTHERS ARE FL-2A-3RF. B. ALL GASKETS ARE Fl.£XITAWC FOR CARBON STEEL LINES, G-1. Nebraska Public Power District COOPER NUCLEAR STATION JELCO INCORPORATED SCALE, NONE DATIC 5/27/94 DRAWN BY: SPB F1LE NO: CHECKED: DJF APPROVEO, GAS 1/1 B/95 Pl.AN TITLE SERVICE WATER BOOSTER PUMPS B & D AS BUILT GLAND WATER SYSTEM GLAND WATER INJECTION FOR SW BOOSTER PUMPS. CADD DRAWING CONTROL BUILDING/CLASS !VP P.IPING FOR PREVIOUS REVISIONS, SEE SUPERSEDED CARDS. THIS DRAWING GENERATED BY NPPD 453242693 DO NOT REVISE MANUALLY CAOD FILE: CQ014757 B&R DWG. 2006, SHEET 4 REVISIONS BY N.P.P.D. WELDING SYMBOLS FlANGE SYMBOLS GATE GLOBE RELIEF CHECK BUTTERFLY FLOW MTR F1L1ER REV NO REVISIONS OFf CKD APP DATE T T T -=-r=- rfi""" W. E. --L,.- -1/1- -t:=::J-- --u- X-2852-388 PLAN NO. WELD NECK __J SUP-ON J I SCREWED L l VANSTONE SOCKEf WELD FLG'D -J[><]I- -1~1- fl- -lv-11-- -11/11- -1[::::::::JI- -lf-Djl- CLASS !VP-IS SEISMIC N07 L
l (STAINLESS STEEL) MATERIAL SPECS. (CARBON STEEL) MATERIAL LIST DC 9D-174A CODE DESCRIPTION SIZE MK. QUA. SIZE CODE NO. SCH. lYP. SPEC. & GR. DESCRIPTION P-1D PIPE 2" & SMALLER 3/4" SW S.S. GLOBE VALVE 800 lb. F-6A F'JTTINGS 2 >SIM "240 TP 316 RESTRICTING ORIF1CE J-1 S.W. JOINTS 304 FLEXITE GASKET 300 lb. J-6 SCRD. JOINTS 3/4" SW S.S. BALL VAf.VE FL-2A-3RF FLANGES 30' 3/4" 80 ASTMAJ12TPJ04 PIPE RO 8-2A BOLTS 3/4" SW ASTV A182 F304 90" ELBOW 3000 PSI 94 G-1 GASKETS SEE NOTE 7 1 1/2" X 3/4- SW ASTV A105 GR2 REDUCER INSERT 6000 PSI SW SOCKOLEr 6000 PSI U-4 UNIONS 2" & SMALLER 3/4" END ASTM A105 CR2. Fl-1A-3RF FLANGES SEE NOTE 6 3/4" SW ASTM A182 F304 TEE 3000 PSI 3/4~ X 1/2" SCRO ASTM Al82 FJ04 REDUCING COUPLING 3000 PSI 3/4" COUPLING 3000 PSI AlSI 316 S.S. SCRO nJBE a:. SOCKEr 0-160 PSI PRESSURE GAUGE
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'<>* B0I.T--ASTMA193C 16 5/8" NUT--,1,Sll,1 A194 BOLTS & NUTS BENDING PROCEDURES DESIGN CONDITIONS PRESSURE RWD-1 SW-1 PROCUREMENT ANALYTICAL 175psig 490psig TEMPERATURE PROCUREMENT ANALYTICAL 1oo*F 145"F HYDROSTATIC TEST 735esig (EE 02-033)
WELDING PROCEDURES N.D.T. REQUIREMENTS MP-LP X-RAY VISUAL 100% SPECIAL NOTES
- 1. TI-US DRAWING SUPERSEDES DRAWING 2852-67.
- 2. lHE RWD-1 SYSTEM DESIGNATION REFLECTS THE GlAND WATER SOURCE DURING NORMAi.. OPERATION. HOWEVER, COMPONENTS ARE STIU. CONSIDERED PART OF THE SERVICE WATER SYSTEM.
- 3. REFERENCE DC90-174A
- 4. TI-lREADED CONNECTIONS ARE SEALED WITH RECTORSEA!.. OR AN APPROVED *EQUIVALENT.
- 5. CARBON STEEL MA1ERIAL SPECIFICATIONS PER E-69-4 MATERIAL SPECIFICATION CODE TABLES.
- 6. THOSE IDENTIFIED WITH AN @ ARE FL-1A-3RF.
OTHERS ARE Fl-2A-3RF.
- 7. ALL GASKETS FOR CARBON STEEL (C.S.) LINES ARE FLEXITALLIC (G-1).
Nebraska Public Power District COOPER NUCLEAR STATION JELCO INCORPORATED SCALE, NONE DATE, 5/17 /94 DRAWN BY: SPB FILE NO: CHECKED: DJF APPROVED: GAS 1/28/95 PLAN TITLE SERVICE WATER BOOSTER PUMPS A & C AS BUILT GI.A.ND WATER SYSTEM GLAND WA1ER INJECTION FOR SW BOOSTER PUMPS CADD DRAWING CONTROL BUILDING/CLASS lVP FOR PREVIOUS REVISIONS, SEE SUPERSEDED CARDS. THIS DRAWING GENERATED BY NPPD 453242694 DD NOT REVISE MANUALLY COOl 4756 i CADD m.E: B & R DRAWING 2006 SHEET 4 NO NOS DCN 00-0892 . REVISIONS BY N.P.P.D. REVISIONS DFT CKD APP RAC DJF KG 2-9-01 DATE r,N:;;;0';-6r,E~E,!'0~2-;;;0c;3';;3;s0°',Cs;N;;;0,'2;;c-;"04';;9'a'6)'°"_ _ _-+H;;;FY~DJ~F+-:,:KG,..+,1:"07 N07 CED 6008707 DCN 02-0835 RRT OJF KG 2-13-03
-4:;:c-:,:D~2 iwaDING SYMBOLS SHOP WELD SCREWEDFJELD WELD WELD NECK _J J FLANGE SYMBOLS T T T-=r-rf1 I L l SOCKET WELD
- w. E.
FLG'D GATE GLOBE RELIEF
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MATERIAL LIST MK. QUA. SIZE CODE NO. SCH. TYP. SPEC. & GR. DESCRIPTION
,* F6-A JOOO# SW A234 WPB SOCOLEr 6" ,. P1-F SCH80 SMLS Atos GRB PJPE
,. 2200# SW A216 WPB BALL VM.VE
,. F6-B 6000# SW SA105 GR2 SOCOLEr 2*-e* so' ,. P1-J SCH160 SMLS A106 GRB PJPE
,* F6-8 60001/ SCRO SA105 GR2 CAP J1 SW 831.1-1967 JOINT J6 SCRO 82.1 JOINT
,. F6-8 6000# CRO/S SA105 GR2 COUPLING 13 ,. F6-8 60001 SW SA105 GR2 ELBOW
,. U-3 6000# SW SA105 GR2 UNION
,. JOOLB SW A216 WPB STRAINER 1/2" P1-J SCH160 SCRO A106 GRB PIPE NJPPLE 1/2" 2200# SCRO A216 WPB BALL VALVE
,*-,o* ~ FS-8 6000/j SW SA105 GR2 REDUCING INSERT
~ *'"'<> l 2' 3/4" P1-J SCH160 SMLS A106 GRB PIPE
'.? t 3/4" 1000{/ SCRO 316 ss FLOW METER
" 3/4" 3/4" U-3 6000#
800# SW SW SA105 GR2 A105 UNION GLOBE VALVE
,. F6-B 6000# SCRO SA105 GR2 TEE 1* SCRO 316 ss COUPON HOLDER 1* F6-B 6000# SW SA105 GR2 TEE 1* F6-B 6000# SW SA105 GR2 ELBOW, 45 DEG 1/2" F6-B 6000# SCRO SA105 GR2 CAP BENDING PROCEDURES DESIGN CONDITIONS PRESSURE
-5 SW-1 PIPING 490 PSIG 548 SW-2 PIPING 75 PSIG TEMPERATURE SW-1 PIPING 145"F SW-2 PIPING 145"F DETAIL 1 HYDROSTATIC TEST (SEE NOTES & 2)
CORROSION MONITORING RACK LRP-RACK-141 WELDING PROCEDURES N.D.T. REQUIREMENTS MP-LP X-RAY VISUAL
,p sw-s 552 0%
SPECIAL NOTES 0%
- 1. HYDROSTATIC TEST REQUIREMENTS FOR SW-1 PIPING IS 735 PSIG.
100%
- 2. HYDROSTATIC TEST REQUIREMENTS FOR SW-2
. PIPING IS 112 PSIG .
- 3. PIPE SECTION INSTALLED BETWEEN SW-V-1 S22 AND COUPLING IS SCHED 80. CED 6034480 PROVIDES JUSTIFICATION FOR USE OF SCHED
+
I 80 PIPING IN PLACE OF SCHED 160 PIPING (AS SPEC. IN CONTRACT E-69-4 FOR SW-1 PIPING) IN THIS LOCATION. I
~1~
~i~
Nebraska Public Power District COOPER NUCLEAR STATION JELCO INCORPORATED SCALE: NONE DATE, 1/16/15 DRAWN BY: DJB FILE NO: ENGINEER: JDMADDO ORIGINAL ECM: DR-2013-0969 PLAN TITLE 460004103~S_W~T_O~RH_R~C_O_RR_O_S_IO_N~M_O_N~. STATUS: Release CONTROL BUILDING AS BUILT 61A1;'MtJ&, 1]tJ%Jg16 t-::C:cLA=-S--::S=I_V_P_ _ _ _ _ _ _ _ _ _ _ _ _--1 VER: AB REV: DO SIZE: D REF DWG: NO. VERSIONS REVISIONS BY N.P.P.D. DESCRIPTION AB 00 INITIAL ISSUE, AS-BUJLT, DR-2013-0969 OFT DATE ENG DLR 11/16/16 JOMADOO f---t~(FR~O~"~VER=S~IO~N~M~---------t-t----t-----1 iWELDING SYMBOLS SHOP WElD SCREWa;iELD WELD i WELD NECK FLANGE SYMBOLS T T T--=r=-T _j SLIP-ON J I SCREWED L L SOCKET WELD VANSTONE W. E. FLG'O GATE GLOBE
-jt><JI- -j~I-RELIEF CHECK BUTTERFLY FLOW MTR FILTER F. S.: 2006 SH.4 ZONE: F-2, F-9, G-6 PLAN NO.
X2852-392 CLASS IS SEISMIC SOCKET WELD L
NLS2018029 Page 1 of 33 Enclosure 2 RRS-03, Cooper Nuclear Station Structural Integrity Associates, Inc. (with revised Attachment 3) Cooper Nuclear Station, Docket No. 50-298, DPR-46 Final Code Case N-513-3 Markup for Revision 4 Technical Basis for Proposed Fourth Revision to ASME Code Case N-513 Cooper Nuclear Station Technical Basis for Proposed Alternative to Use ASME Code Case N-513-4 and Scope Expansion to a Higher Pressure Limit, Revision 1 Note: Enclosed report contains 32 pages not counting this cover page.
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 1 of 32 Attachment 1 Final Code Case N-513-3 Markup for Revision 4
Attachment 1 - 3 RRS-03 Cooper Nuclear Station 2 of 32 Record # 12-841 CASE CASES OF ASJ\'IE BOILER AND PRESSURE VESSEL CODE N-513-4.J Approval Date: January 26, 2009 Code Cases will remain available for use until anm1lled by the applicable Standards Committee. Case N-513-:!.J- ie0) The piping cbi<>n Code shall be used in Evaluation Criteria for Temporary Acceptance of deLerminina the stress indices B1 and 13°. and stress Flaws in Moderate Energy Class 2 or 3 Piping intensification fa~tor. i, for flaw evaluation follmvlng Section XI, Division 1 Code applicabilitv limits in terms of component ucometrv. sut..:h as D,Jtuom rnlio. [f the piping d.eslgn Inquiry: What requirements may be used for Code does not providt! stres.5 inclices. Section IIL ')004 tern porary acceptance of t1aws, including through-wall Eclition or later Editions and Addenda mm* be used to t1aws, in moderate energy Class 2 or 3 piping including define B: and Be. dbo\vs. bent pipe, reducers. expanders. and branch tc~s. ([,#) The provisions of this Case demonstrate the without performing a repair/replacement activity? integrity of the item and not the consequences of leakage. It is the responsibility of the Owner to Reply: It is the opinion of the Committee that the dam eRstrnta s;*st,m ojl>fabidi~* considerif>§ effects of following requirements may be used to accept flaws, leakage tn demonslrnting svstem onembiliLv and including through-wall flaws, in moderate energy Class perfom1inc* nlant floodino analvscs. 2 or 3 piping inc\udin2 elbows. bent pip,,. reducers. (ej The e.alu,1ticn period, T.;;.,.., is the operntionnl expanders. and hrnnch tees, without performing a time lef whieh tl:e ternrernry aeeejltHnae ariterin are repair/replacement activity for a limited time, not satisfied liur Rel e.cceediflg 26 moml.s frnm the initial exceeding the e1*slusrie.fl Faried ss defined ifl trus diseeYeP: efthe cenditieR. ~time to Lhc nexl scheclulccl refueling OLttage. 2 PROCEDURE SCOPE (a) The flaw geometry shall be characterized by (a) These requirements apply to the ASlvJE Section volumetric inspection methods or by physical III, ANSI B31. I, and ANSI B3 l. 7 piping, classified by measurement. The full pipe circumference at the flaw the Owner as Class 2 or 3 that is accessible for location shall be inspected to characterize the length and inspection. The provisions of this Case do not apply to depth of all flaws in the pipe section. the following: (b} Flaw shall be classified as planar or nonplanar. (1) pumps, valves, expansion joints, and heat (c) When multiple flaws, including irregular exchangers. except as provided in rbi; (compound) shape flaws, are detected, the interaction (2) weld metal of socket welded joints; and combined area loss of flaws in a given pipe section (3) leakage through a flange joint; shall be accounted for in the t1aw evaluation. (4) threaded connections employing (d) A flaw evaluation shall be performed to nonstructural seal welds for leakage protection. determine the conditions for flaw acceptance. Section 3 (I>) This Case mav be applied 10 heat exchanger provides accepted methods for conducting the required external tubing or piping. provided the flaw is analysis. characterized in accordance \Yith 2(a) and leakage is (e) Frequent periodic inspections of no more than monitored. 30 day intervals shall be used to determine if flaws are (;;J) The provisions of this Case apply to Class 2 or growing and to establish the time.-'i'.......,-_at which the 3 piping in liquid svstems whose maximum operating detected flaw will reach the allowable size. temperature does not exceed 200°F (93°C) and whose Alternatively, a flaw growth evaluation may be maximum operating pressure does not exceed 275 psig performed to predict the time;-4~ at which the (1.9 MPa). detected tlaw will grow to the allowable size. The flaw (!!_,') The following flaw evaluation criteria are growth analysis shall consider the relevant growth permitted for pipe and tube includine: elbows bent pine, mechanisms such as general corrosion or wastage, reducers. expanders. and branch tee~. The straight pipe fatigue, or stress corrosion cracking. When a flaw flaw evaluation criteria are permitted for adjoining growth analysis is used to establish the allowable time fittings and flanges to a distance of (R,t)'h from the weld for temporary operation, periodic examinations of no centerline. more than 90 day intervals shall be conducted to verity the flaw growth_ analysis predictions. Draft 15 (05/05/14)
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 3 of32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-:!~ (!) For through-wall leaking flaws, leakage shall be When through-wall axial flaws are evaluated, the eliser,;ea ls,monitored daily walkas",H.: to confirm the allowable flaw length is: analysis conditions used in the evaluation remain valid. (g) If examinations reveal flaw growth rate to be unacceptable. a repair---BF-{replacement activity shall be performed. e., = u8"'RI[((sF~)zcrJ-if (l) (h) Repair----M--,'.replacement activities shall be performed no later than when the predicted flaw size CT1, = pD,/21 ~ (2) from either periodic inspection or by flaw growth a1 =(s, +s,)12 ~ (3) analysis exceeds the acceptance criteria of 4, or durinQ the next scheduled retueling outage, whichever occurs where ~~------ p = pressure for the loading condition Note: Z has been first. Repair~replacement activities shall be in Do = pipe outside diameter added to equation accordance with IWA-4000 er P~.'.\ 'GG0. respe,ti,*et,., iA esitie11s flf!d i\ddeHaa rri lf te tLe 1991 ,\,ldeAea; 0' = How stress (]). aAd, iR tche 199l :\,IElenJa aRd later, iA aeesfaarn,e will, 0 = Code specified yield strength TV'/\ 1000
& = Code specified ultimate tensile strength and (i) Evaluations and examination shall be SFm = structural factor on primary membrane stress as specified in C-2622 documented in accordance with IWA-6300. The Owner Z = load multiplier for ductile tlaw e~1ension shall document the use of this Case on the applicable !equal to 1.0 when using limit load criteria) data report form.
Material properties at the temperature of interest 3 FLAW EVALUATION shall be used. Planar flaws in strni2ht pipe shall be evaluated in FIG.1 THROUGH-WALL FLAWGEOMETRY accordance with the requirements in 3.1. Nonplanar flaws in straight pipe shall be evaluated in accordance with the requirements in 3.2. Through-wall flaws in elbows and bent pipe shall be evaluated in accordance with the requirements in 3.3. ThrouS!h-wall flaws in reducers. expanders. and branch tees shall be evaluated in accordance with the requirements in 3.4 and 3.5. rc,nectivelv. Flaw growth evaluation shall be perl'ormed in accordance with the requirements in 3.§.l-. Nonferrous (c) For planar flaws in ferritic piping, the evaluation materials shall be evaluated in accordance with the procedure of Appendix C shall be used. Flaw depths up requirements in 3.].4. to 100% of wall thickness may be evaluated. Flaw For all Haw evaluations all Service Level load depth a is defined in Figures C-4'.I to-l and C'-43 l 0-'.! combinations shall be evaluated to determine the most When through-wall circumferential flaws are evaluated limitirrn allowable flaw size. in accordance with C-5300 or C-6300, the flaw depth to thickness ratio, alt, shall be set to unity. When applying 3.1 Planar Flaws in Straight Pipe the Appendix C screening criteria for through-wall axial (a) For planar flaws, the flaw shall be bounded by a flaws, alt shall be set to unity, and the reference limit rectangular or circumferential planar area in accordance load hoop stress, o1, shall be defined as o/Ji.,J,. When with the methods described in Appendix C. IW A-3300 through-wall axial flaws are evaluated in accordance shall be used to determine when multiple proximate with C-5400 or C-6400, the allowable length is defined flaws are to be evaluated as a single flaw. The geometry by eqs. (1) tlrrough (3), with the appropriate structural of a through-wall planar flaw is shown in Fig. 1. factors from Appendix C, C-2622. When through-wall (b) For planar flaws in austenitic piping, the flaws are evaluated in accordance with C-7300 or C-evaluation procedure in Appendix C shall be used. Flaw 7400, the formulas for evaluation given in C-4300 may depths up to 1000/o of wall thickness may be evaluated. be used, but with values for Fm, Fb , and F applicable to When through-wall circumferential flaws are evaluated, through-wall flaws. Relations for Fm, F., and F that take the formulas for evaluation given in C-5320 or C-6320, into account flaw shape and pipe geometry (Rlt ratio) as app I irnble, of Appendix C may be used, with the flaw shall be used. The appendix to this Case provides depth to thickness ratio. alt, equal to unity. equations for Fm, Fb, and F for a selected range of Rlt. Geometry of a through-wall crack is shown in Fig. 1. Draft 15 (05/05/14) 2
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 4 of32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513..,4J FIG. 2 SEPARATION REQUIREMENTS FOR ADJACENT THINNED AREAS 3.2 Nonplanar Flaws in Straight PiJ,e defmed in Fig. 3. When the above requirement is not ___fa) The evaluation shall consider the depth and satisfied, (i>ijj) shall be met. extent of the affected area and require that the wall (!£.;t.) When Lm is less than or equal to 2.65 thickness exceed tm;" for a distance that is the greater of (Rotmm)l/2 and tnom is greater than 1.13 tm;" , ta1oc is 2.5 -./Rtru,m or 2Lm,a,g between adjacent thinned regions, determined by satisfying both of the following where R is the mean radius of the piping item based on equations: nominal wall thickness and Lm,ovg is the average of the extent of Lm below tm;" for adjacent areas (see Fig. 2). Alternatively, the adjacent thinned regions shall be (;c, ~ 1.5.Jfiotmm [1-t"~"]+l.O (5) considered a single thinned region in the evaluation. t"'"' L t,,;,, __(b) For nonplanar flaws, the pipe is acceptable when either ( b )Lll.ill.!.d_(µj(l),_Q!: lb)('.:\ and (b )(3) are (6) met. __L.ll_]}he remaining pipe thickness (tp) is greater than or equal to the minimum wall thickness tm;": When the above requirements are not satisfied, ( ~ shall be met. pDo (4) (!.!'..r.) When the requirements of(+{), (.;!!.!.), and 2( S +0.4p) (,?gj) above are not satisfied, ta1oc is determined from Curve 2 of Fig. 4. IA ae<litieu, :...,. .1lrn!! .. atisfj the where fe!le .. itt"' e<juutieH: p = maximum operating pressure at flaw location S = allowable stress at operating temperature 12! The remaining degraded pipe section meets the t,.,, 1.8 longitudinal stress limits in the design Code for the ll..!12.!!l&. ~~-is-fiie-ooHH!Hll-pijae leagirndinal ! , ~ - . , (3i As an altemative to lb)(l'i,>,lto;,arinl;, an re~Q1!tia; frem all Su , iec Le ,el B p11ma11 1,ip, evaluation of the rermrining pipe* thickness (t"! may be leading& performed as given below. The evaluation procedure is (c) When there is through-wall leakage along a a function of the depth and the extent of the affected portion of the thinned wall, as illustrated in Fig. 5, the area as illustrated in Fig. 3. tlaw may be evaluated by the branch reinforcement (!.+) When Wm is less than or equal to 0.5 method The thinned area including the through-wall (Rot) 112 , where Ro is the outside radius and Wm is defmed opening shall be represented by a circular penetration at in Fig. 3, the flaw can be classified as a planar flaw and the flaw location. Only the portion of the flaw lying evaluated in accordance with 3.l(a) through 3.l(c), within ta,1i need be considered as illustrated in Fig. 6. above. When the above requirement is not satisfied, ,¥hen evaluating multiple flaws in accordance with (;ill) shall be met. 3.Z(a), only the portions of the tlaws contained within (!i_J) When Lm,t) is not greater than (Rdmm )112 , to,1i need be considered. ta10, is determined from Curve 1 of Fig. 4, where Lm~) is Draft 15 (05/05/14)
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 5 of 32 CASE (continued) CASES OF AS1VIE BOILER AND PRESSURE VESSEL CODE N-513-43 FIG. 3 ILLUSTRATION OF NONPLANAR FLAW DUE TO WALL THINNING l Lm<n Transverse (clrcumferentiall Axial direction direction Editor's }Tc,te. This figure 3 is t(I be deleted and reolaced \:v1th the figi-1re 3 on the fdlo'i:vtng c,age. Draft 15 (05/05/14) 4
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 6 of32 CASE ( continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4;} FIG. 3 ILLUSTRATION OF NONPLANAR FLAW DUE TO WALL THINNING t tmin
-Lm(a)-
--',--~-
l lml/tJ ITransverse (circumferential) direction FIG. 4 ALLOWABLE WALL THICKNESS AND LENGTH OF LOCALLY THINNED AREA 0 '------'----'-----'----'------'------~--~ 0 5 7 8 Draft 15 (05/05/14)
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 7 of 32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4J. FIG. 5 ILLUSTRATION OF THROUGH-WALL NON PLANAR FLAW DUE TO WALL THINNING Through-wall
/ opening tnom Transverse l..c;rn (circumferential)
Axial direction direction
<~~~~~~~~~~~~~~-, Edit:ir's Note Tlris Figure 5 is to be deleted and replaced with the Figure 5 on the follov.ingpage.
Draft 15 (05/05/14) J
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 8 of32 CASE (continued) CASES OF ASME BOILER Ai'ID PRESSURE VESSEL CODE N-513-4J FIG. 5 ILLUSTRATION OF THROUGH-WALL NONPLANAR FLAW DUE TO WALL THINNING Through-wall I t . / opening t + mrn
+h-~~~~~~~~
-laxial--
----~
i 1 Lcirc I Transverse (circumferential) direction Dratl 15 (05/05/14) 7
Attachment 1 - 3 RRS-03 Cooper Nuclear Station 9 of32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-:!,J The m 1111m wn wall thickness, t.,,,, shall be are adjusted to account fcir the geomctrv differences. as determined by eq. (4). For evaluation purposes, the described below. Alternative methods mav be used to adjusted wall thickness, tad;, is. a postulated thickness as calculaie the stres~s used in evaluation. shown in Fig. 6. The pipe wall thickness is defined as The hoop stress. a,. for elbow and bent pipe the thickness of the pipe in the non-degraded region as evaluation shall be: shown in Fig. 6(a). The diameter of the opening is equal to dad/ as defined by fad; as shown in Fig. 6(a). The postulated value for tad/ shall be greater than t,,n and (]" =(PDa)[ '2R,,,,d+Rasin¢ ]+(l.95)RaM, ,9*, shall not exceed the pipe wall thickness. The tadJ value
' 2t 2(R,".,, + R, sin¢,) h'" I -L?!J.
may be varied between t.. ,,, and the pipe wall thickness to determine whether there is a combination of t,d; and dad1 that satisfies the branch reinforcement requirements. The values of t,dJ and d,dJ of Fig. 6(b) shall satisfy: Ri,,,,,, = elbow or bent pipe bend radius
¢ = ~ircum fercnlial ~mule defined in Fi,,ure 7 h = flexibilitv characteristic Q~)
Al; = resultant primarv bending moment I = moment of inertia based on evaluation wall thickness t The remammg ligament average thickness, t,.ovg, over the degraded area bounded by cl.a; shall satisfy: Equation 9 is onlv applicable for elbows and bent pipe where h > 0.1. (Jl.9-) The axial membrane pressure stress, a~. for elbow t,_..,. ;;, 0.353d.,j~ and bent pipe evaluation shall be: In addition the pipe section including the equivalent hole representation shall m ect the --(J"m = B(2[ l pD,) - - - - - ~ ~ i]i)'1 longitudinal stress limits in the design Code for the piping. __If a flaw growth analysis is performed, the growth where B1 is a primarv stress index as defined in SeGtion in flaw dimensions shall consider the degradation In for the piping item. B, shall be equal lo 0.5 for mechanisms as relevant to the application. The flaw is elbows and bent pipe. acceptable when there is sufficient thickness in the The axial bending stress. r;;** for elbow and bent degraded area to provide the required area pipe evaluation shall be: reinforcement. (d) Alternatively, if there is a through-wall opening along a portion of the thinned wall as illustrated in Fig. 5 the flaw may be evaluated as two independent planar
-~ ' (J"=B(~~)------~*~1=11
' 21 through-wall flaws, one oriented in the axial direction and the other oriented in the circwnferential direction. where B*, is a prirnarv stress index as defined in Section The minimwn wall thickness t.,,,,, shall be determined II[ for the pipinl! item.
by eq. (4). The allowable through-wall lengths in the The thermal expansion stress. r;_. for elbow and axial and circwnferential directions shall be determined bent pipe ev,iluation shall be: by varying fad; shown in Fig. 5 from ln,m to fmm. The allowable through-wall flaw lengths based on lodJ shall be greater than or equal to the corresponding L"""' and _ _ a :::::z 11
*(DJvf,) - - - - - - - ~ =
~
il2'i L"" (see Fig. 5) as determined from 3. l(a) and 3. l(b) or 3.l(c), as appropriate. The remaining ligament average thickness, t,=g, over the degraded area bounded by Lax,al and L,.,,., shall satisfy eq. @).
= stress intensification factor as defined in the 3.3 Through-wall Flaws in Elbows and Bent Pipe de~i2:n Code for the piping item Jf.. ~ resultant tl1ermal expansion moment Through-wall flaws in dhows aml bent pipe mav be evaluated usi.rm the straiuht pipe procedures uiven in 3.1 or .l2(c!'i. provided the stresses used in th, ernluntion Draft 15 (05/05/14)
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 10 of 32 CASE (continued) CASES OF A~ill BOILER AND PRESSURE VESSEL CODE N-513-,1.J 3.4 Through-wall Fhn,*s in Reducers mu) The a'(ial bending strcs.<..: U1,. and themrnl expansion Expund~rs stress. (J,_ for branch tee evaluation shall be detennined Thrmrn:h-wall tlmvs in rt~ducers and expanders mav from eq. (15) and eq. ( 16). respcctivelv. be! evaluated using the ::;tmhi:ht L*ipe procedures ei\*cn in 3.1 or 3.2J d'i provided the stresses usd in the 3.<~ Flaw Growth Evaluation evaluation are adjusted to account for the gcomdrv If-;. flaw growth analysis is perfom1ed, the growth ,lifferences as described below. Altemati,*e methmls analysis shall consider both corrosion and crack-growth mav be: used to calculate the slressl!s rn;ed m cvalUc.'1tion. mechanisms as relevant to the application. fiu. 8 illustrates the reducer and expander zones In perfolilling a flaw growth analysis, the disct1ssed below. Evaluation of flaws in the small end procedures in C-3000 may be used as guidance. transition zone is outside the scone 1..*f this Case. Relevant growth rate mechanisms shall be considered. The hoop stress. u~. and axial membnme presstrrc When stress corrosion cracking (SCC) is active, the stress cr,if, l(x reducer or ~xpander evaluation shall be: following growth rate equation shall be used: _ _ a,,= (U pD,) - - - ~ il3) (11G) where Jaldt is flaw growth rate in incheSihour, Km~ is _ _ _ a .* =B,( ~J-----~(~1~4) the maximum stress intensity factor under long-term steady state conditions in ksi in. 0*', Sr is a temperature where D, is the small end OD for flaws in the small end correction factor, and C and 11 are material constants. and the large end OD for all other flaws. The axial bendino stress. u,-._ and then11al ~xpnnsion for intergranular sec in austenitic steels, where T :S stress. u~, for reducer or expand~r evaluar ion shall be: 200°F (93 °C). C = l.79x 10-* a = R ( DJ1'
" 2I
)-----~(~1 )~**1 Sr n
=I
=2.161 ar = i(D,AI,)
21 ~~~~~~~~~
- 16' For transgranular sec in austenitic steels, where T :S 200°F (93°e).
where I is based on the ck,grndcd secti,,n. C = 1.79 x 10-7 3.5 Through-wall Flaws in .Branch Tees Sr = 3.71 x 10s [lOto.01042r-12.2iJJ Rrnnch reinforcement requirements $hall be met in n =2.161 acc,,niance with the design Code. If the desil!n Code did not requ-ire reinforcement for evaluation purposes. a The temperature, T, is the metal temperature in rninforcernent recrion is dd"ined a~ a renion of radius D. degrees Fahrenheit The flaw growth rate curves for the of the branch pipe from the center of the branch above SCC growth mechanisms are shown in Figs. 2_::; connection. Throu"h-wall flaws in branch kes oursiJe and Jil&. Other growth rate parameters in eq. (124) may of the reinforcement re2:ion may be evaluated usim the be used, provided they are supported by appropriate straight rnpe procedures given in 3.1 or .3.2\d). provided data. the stresScs used in the evaluati<m nre adjust~d. to account for the geom euv di ffercnccs as described 3.14 Nonferrous .Materials below. Alternative methods mav be used ro calculate the For nonferrous materials, nonplanar and planar stresses used in evaluation. Evaluation of flaws in the llaws may be evaluated following the general approach region \)f branch reinforcement is \)Utsidc the scope of of 3.1 through 3.!a_B. For planar flaws in ductile this Case. materials, the approach given iR :.l('ei ans -~.3for Th~ hoop strt:ss, U;,_ and axial membrane pressur~ ,mstenitic pipe may be used; otherwise, the approach stress, U1.;- for branch te~ ~vnluatiL"'J1 shall L~ J~tcrmincJ given iH :. I (e~ aml J.Jfor fC1Titic pipe should be from eg. ( 13\ and eq. ([4\ re,'Dectively. The outside applied. Structural factors provided in 4 shall be used. It dim11e1"r for each of these equations shall be for the. is the responsibility of the evaluator to establish branch or nm pipe, depending on rhe flaw location. conservative estimates of strength and fracture toughness for the piping material. Draft 15 (05/05/14) 9
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 11 of 32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-~ FIG. 6 ILLUSTRATION OF ADJUSTED WALL THICKNESS AND EQUIVALENT HOLE DIAMETERS Through-wall
/ opening T
Pipe
~~'"---___,J~~tall (al 4dJusted Wall Thlckpess Shift figure (bJ to the right so that dadj TItmin
- width lines up with figure(a).
*t;~-i-----------~ ~~~'-~e~d~~~.~----'
l-dadj-1 (bl Equivalent Hole Representation FIG. 7 CIRCUMFERENTIAL ANGLE DEFINED
----~~~~;~
extrados Drail 15 (05/05/14) 10
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 12 of 32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4J. FIG. 8 ZONES OF A REDUCER OR EXPANDER Lorge end transition zone Central conical section Small end transition zone it------------------- GENERAL NOTE: Transitio~ zones extond from the point on the ends where the diameter begins to change to the potnt on the central cone where the cone angle is constant. Draft 15 (05/05/14) 11
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 13 of 32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-.f! FIG.~ FLAW GROWTH RA TE FOR IGSCC IN AUSTEN ITIC PIPING I.CE-02
~ AusJ;nHlc Ftplr,;i r* 21>:f I.CE.03 V
/
a V
~ 1.QE.OS
.c:
"0" /
1l V 0 1..ce.D5
,~
I/
/
/
I.CE.OS 10 100
*:S.EHEtJ.L HDTE: Ii[ CQ.nff!iilll: LJJ Jr..i,r
- 7.M X 1.cl' mm's.!!:.i L.O lc;f in.=-** lJ]Q,~ MP, rnM; 'C * ['F ~ !!2lh.*e->.
Draft 15 (05/05/14) 12
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 14 of 32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4J. FIG . .1!:,lf, FLAW GR QINTH RATE FOR TGSC C IN AUST ENIT IC PIPING UE-02
---1 Amt..-111: Apl~
T~2D:if ,' U.tJ3 f V T.C-'l4 f
;' V
=T*:ilCf
/ /
/ I T* IQ[f' V
/
UE-07
/
1.CEJJB Dr,ft 15 (05/05/l 4) 13
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 15 of 32 CASE (continued) CASES OF AS1"IE BOILER AND PRESSURE VESS.EL CODE N-513-,1.J 4 ACCEPTAJ.'l'CE CRITERIA nondimensional stress intensity factor for through-wall circumferential flaw W1der Piping containing a circumferential planar flaw is membrane stress acceptable for temporary service when flaw evaluation I moment of inertia based on evaluation provides a margin using the structural factors in thickness t Appendix C, C-2621. For axial planar flaws, the mmcimum stress intensity factor W1der structural factors for temporary acceptance are as long term steady state conditions specified in Appendix C, C-2622. Straight pllip,!:iHg L maximwn extent of a local thimied area containing a nonplanar part:-lhrough-wall flaw is with t< l.,m acceptable for temporary service if the remaining pine Laxia! length of idealized through-wall planar section meets the lom!itudinal stress lin1its in the desi,m flaw opening in the axial direction of the Code for the piping and lp 2'. t'1,~ where t'"" is pipe, as illustrated in Fig. 5 detem1ined from 3.2(b). Straight rPiJJ£ffi§c containing a Lc,rc length of idealized through-wall planar nonplanar through-wall flaw is acceptable for temporary flaw opening in the circumferential service when the flaw conditions of 3.2(c) or 3.2(d) are direction of the pipe, as illustrated in Fig. 5 satisfied. An elbow or bent pipe containing a nonplanar L,,, maximum extent of a local thimied area through-wall flaw is acceptable for ternporarv service if with t < tmm the flaw conditions of 3.3 are sntisfied. A reducer or L,,,(,! axial extent of wall thinning below lmm ~xpander containina a nonpbnar thr..._1uoh-wall flaw is Lmr<! circumferential extent of wall thinning acceptable for temporary service if the tlaw conditions below t,,,, of 3.4 are satisfied. A branch tee containine: a nonplanar Lm.avg average of the extent of L,., below through-wall l1aw is acceptable for ternporarv service if lmm for adjacent thinned areas the flaw conditions of 3. 5 are satisfied. maximum extent of thinned area, i bulging factor for axial flaw 5 AUGMENTED EXAMINATION resultant n1irnarv benclin11. moment
!v[, resultant lhennal expansion moment An augmented volwnetric examination or physical R mean pipe radius measurement to assess degradation of the affected R"'""'-'--=el=b=o~v.-'o=.r~l=)e=n=t~p~iP=*e=c=e=n=te=r=li=n=e=b=ei=1d~ra='=li=us system shall be performed as follows: R, outside pipe radius (a) From the engineering evaluation, the most s allowable stress at operating temperature susceptible locations shall be identified A sample size SFm structural factor on primary membrane of at least five of the most susceptible and accessible stress locations, or, if fewer than five, all susceptible and coefficient for temperature dependence in accessible locations shall be examined within 30 days of the crack growth relationship detecting the flaw. S., Code-specified ultimate tensile strength (b) When a flaw is detected, an additional sample s, Code-specified yield strength of the same size as defined in 5(a) shall be examined. T metal temperature (c) This process shall be repeated within 15 days w,. maximW11 extent of a local thinned area for each successive sample, W1til no significant flaw is perpendicular to L,. with t < t,,,,
detected or Wltil I 00% of susceptible and accessible minimum distance between thinned areas i locations have been examined. andj z load multiplier for ductile flaw extension 6 NOMENCLATURE a flaw depth C half crack length Section Ill primarv stress indices da!dt flaw growth rate for stress corrosion C coefficient in the crack growth relationship cracking D. inside pipe diameter diameter equivalent circular hole at t,d1 D, outside pipe diameter diameter of equivalent circular hole at F nondimensional stress intensity factor for lmin through-wall a'<ial flaw W1der hoop stress h flexibilitv characteristic nondimensional stress intensity factor for stress intensification factor through-wall circumferential flaw W1der f. total crack length - 2c pipe bending stress £ ,II allowable axial through-wall flaw length n exponent in the crack growth relationship Draft 15 (05/05/14) 14
Attachment 1 - 3 RRS-03 Cooper Nuclear Station 16 of 32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4-J p maximum operating pressure at flaw location evaluation wall thickness. surrounding the dc2rnded area adjusted wall thickness which is varied for evaluation purposes in the evaluation of a through-wall nonplanar flaw faioc allowable local thickness for a nonplanar flaw
- lc,avg average remaining wall thickness covering degraded area with through-wall leak bounded by d,d; tmm minimum wall thickness required for pressure loading nominal wall thickness minimum remaining wall thickness a maximum cone am!lc: at the center of a reducer nondirnensional half crack length for through-wall axial flaw rp ___. .,cceire,CceUa,m.,_,f,cer'-'e'-'-n"'lie.,a.,_l,,.am,=]e,e..cfccrl,c)lTI"'-e"']"'b"'O\,.,\'-'O"'r'--"'bcee'n=tl flank en. axlal bending str~ss for primarv loading a.. axirrl thennal expansion str~ss l7f material flow stress 17h pipe hoop stress due to pressure and bending moment (for elbows and bent
_.l2filil
.-~ !l.aFHinal lengit>!ili!l.sl b,RdiRg stF@:*s far t*FiFHaFy leaEliflg without skeJs inomsifi eaEier, factor 171 reference limit load hoop stress CY>f! axial prt!ssure stress 17y material yield strength at temperature, as defined in C-4300 ~ time required wr tJae dete2ted flaw to ,,,row to ti'le ello*.,*alale !law size, lrnt Act e,ceeedi,ag JC m ,,,,ths from the initial diseevary e.fthe eoRElitioR
() half crack angle for through-wall circumferential flaw 7 APPLICAfilLITY This Ca,*, i:.* aprlisabl~ frem tha l<)g] Raition with lfw Wimer 19~'5 .\eaaflaa, 1hrilugl, the '.)?07 Eai1ien with the ~oog ;\Jdenaa. Reference to Appendix C in Editor's Note: For Applicabilitv [ndex. this Case shall apply to Appendix C of the 2004 Edition applicabilitv is lrom l 996 Addenda to or later editions or addenda. For editions t!fl6-0r addenda 2013 Edition. prior to the 2004 Edition, Class 1 pipe flaw evaluation procedures may be used for other piping classes. As a matter of defmition, the current tenn "structural factor is equivalent to the tenn "safety factor," which is used in earlier editions and addenda. Draft 15 (05/05/14) 15
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 17 of 32 CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-513-4.J MANDATORY APPENDIX I RELATIONS FOR Fm, Fb, AND FFOR THROUGH-WALL FLAWS 1-1 DEFINITIONS Ab = -3.26543 + 1.52784 (Rlt) - 0.072698 (R/t )2
+ 0.0016011 (Rlt)'
For through-wall flaws, the crack depth_-fa.:, will Bb = ll.36322-3.91412(R/t)+0.18619(R/t) 2 be replaced with half crack length_--fcJ in the stress - 0.004099 (Rlt)' intensity factor equations in C-7300 and C-7400 of Cb =-3.18609+3.84763 (Rlt)-0.18304(R/t)2 Section XI, Appendix C. Also, Q will be set equal to + 0 .00403 (RI t) 3 unity in C-7400. Equations for Fm and A are accurate for Rlt between 5 1-2 CIRCUMFERENTIAL FLAWS and 20 and become increasingly conservative for Rlt greater than 20. Alternative solutions for F,,, and Fb may For a range of Rlt between 5 and 20, the following be used whenR/t is greater than 20. equations for Fm and Fe may be used: 1-3 A.XIAL FLAWS Fm= 1 + A,,, (817r)l.5 + B,,, (B!tr)2 *' + C,,, ((bi/*' Fb = 1 + Ab (8/tr) 1*' + Be (Bltr) 25 + Cb (Bltr.Y'*' For internal pressure loading, the following equation for F may be used: where F = I + 0.072449A. + 0.64856,.2 - 0.2327 )..3 e = half crack angle = c/R + 0.038154 )..4 - 0.0023487 )..' R = mean pipe radius t = ~ w a l l thickness where and A,,, = -2.02917 + 1.67763 (R!t) - 0.07987 (R/t )2 c = half crack length
+ 0.00176 (R!{'J ).. = cl(Rt) 112 B,, = 7.09987 - 4.42394 (R!t) + 0.21036 (R/t ) 2 The equation for F is accurate for A between O and 5.
- 0.00463 (R/t)' Alternative solutions for F may be used when ).. is Cm = 7.79661 + 5.16676 (Rlt)- 0.24577 (R/t) 2 greater than 5.
+ 0.00541 (Rlt)'
Draft 15 (05/05/14) 16
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 18 of 32 Attachment 2 Technical Basis for Proposed Fourth Revision to ASME Code Case N-513
Attachment 1 - 3 RRS-03 Cooper Nuclear Station 19 of 32 Proceedings of the ASME 2014 Pressure Vessels & Piping Conference PVP2014 July20-24, 2014, Anaheim, California, USA PVP2014-28355 TECHNICAL BASIS FOR PROPOSED FOURTH REVISION TO ASME CODE CASE N-513 Robert 0. McGill GuyDeBoo Structural Integrity Associates Exelon Generation Company San Jose, CA Warrenville, IL rmcgil l@structint.com quy.deboo@exeloncorp.com Russell C. Cipolla Eric J. Houston lntertekAIM Structural Integrity Associates Sunnyvale, CA San Jose, CA russell.cipolla@intertek.com ehouston@structint.com ABSTRACT Resultant thennal expansion moment Elbow or bent pipe bend radius Code Case N-513 provides evaluation rules and criteria Outside radius for temporary acceptance of flaws, including through-wall Flexibilitv characteristic flaws. in moderate energy piping. The application of the Code Stress int~nsification factor as defined in the Code of Case is restricted to moderate energy, Class 2 and 3 systems, Record for the piping item so that safety issues regarding short-term, degraded system p Maximum operating pressure at flaw location operation are minimized. The first version of the Code Case t Evaluation thickness was published in 1997. Since then, there have been three
<lb Axial bending stress revisions to augment and clarify the evaluation requirements u, Thennal expansion stress and acceptance criteria of the Code Case that have been Uh Hoop stress published by AS1'IE. The technical bases for the original Um Axial membrane stress version of the Code Case and the three revisions have been
¢ Circumferential angle previously published.
There is currently work underway to incorporate INTRODUCTION additional changes to the Code Case and this paper provides Background the technical basis for the changes proposed in a fourth revision. These changes include addressing the current Code Case N-513-3 [I] (currently approved revision) condition on the Code Case acceptance by the US Nuclear provides evaluation rules and criteria for the temporary Regulatory Commission (NRC), clarification of the Code Case acceptance of flaws, including through-wall flaws, in applicability limits and expansion of Code Case scope to moderate energy piping. The provisions of this Code Case are additional piping components. New flaw evaluation focused on preventing gross failure of the affected pipe for a procedures are given for through-wall flaws in elbows, bent temporary period. However, it also requires the piping system pipe, reducers, expanders and branch tees. These procedures and adjacent equipment functionality be demonstrated for lost evaluate flaws in the piping components as if in straight pipe fluid inventory, spraying and t1ooding caused by the leakage. by adjusting hoop and axial stresses to account for the The Code Case provides rules for the evaluation of degraded geometry differences. These changes and their technical bases pipe and tube for a short operating period, with inspection and are described in this paper. monitoring requirements of the degraded condition as part of the overall integrity assessment. The application of the Code NOMENCLATURE Case is restricted to moderate energy Class 2 and Class 3 systems, so that the safety issues regarding short-tenn system B,, B2 Primary stress index defined in AS1'IE Section III for operation are minimized. Moderate energy piping is defined the piping item as those piping systems where the maximum operating D, Outside diameter pressure and temperature do not exceed 275 psig (1.9 MPa) I Moment of inertia based on t Cdegraded section) and 200°F (93°C), respectively. Mb Resultant primary bending moment Copyright© 2014 by ASME
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 20 of 32 Currently, the scope of the Code Case is limited to to accept flaws ... without performing a repair/replacement straight pipe with some provision for flaw evaluation into activity for a limited time, not exceeding the time to the next fittings for a short distance from the weld attachment to scheduled refueling outage." Note that the word "refueling" straight pipe. There have been many instances where through- has been added for further clarification regarding the nature wall leaks have been observed in pipe components (i.e., and duration of the plant outage. elbows, reducers, expanders or branch tees) or bent pipe outside of the current Code Case scope. As a result, utilities This language is consistent with the introduction ofNRC are forced into taking alternate actions (e.g., repair, Generic Letter 90-05 [3] which the original N-513 was based. replacement, or making requests for NRC relief) that can be The second paragraph of90-05 states: "Temporary non-code significantly more burdensome without any measurable repairs are applicable until the next scheduled outage increase in plant safety. The fourth revision of the Code Case exceeding 30 days, but no later than the next scheduled provides evaluation rules and criteria for the temporary refueling outage." acceptance of flaws for these instances. In addition, other enhancements are included with the fourth revision as Flaw Evaluation Criteria for Piping Components and Bent summarized below. fi.fil. Evaluation and acceptance criteria have been added to N-Summarv of Code Case N-513-4 Changes 513-4 for flaws in elbows, bent pipe, reducers, expanders and The list below summarizes the proposed changes included branch tees. A sun plified approach has been adopted based on in Revision 4 of Code Case N-513. Note that a brief reason the evaluations and results from the Second International for each change is included in parenthesis. Piping Integrity Research Group (IPIRG-2) program reported in Reference [4]. The flaw evaluation for the piping Temporary acceptance period redefined (addresses component is conducted as ifin straight pipe by scaling hoop NRC condition given in Regulatoty Guide 1.147 [2]) and axial stresses using ASME piping design code stress
- Flaw evaluation criteria included for elbows. bent indices and stress intensification factors to account for the pipe. reducers, expanders and branch tees (scope stress variations caused by the geometric differences. In expansion) Reference [4], this approach was determined to be very Allow flaw evaluation of heat exchanger tubing in conservative by comparing the failure moments predicted specific instances (scope expansion) using this approach to the measured failure moments from the Daily walkdown requirement for through-wall leaks elbow tests for through-wall circumferential flaws. Details of provides additional flexibility for user the simplified approach are given in the following sections.
implementation (scope expansion)
- Limit scope to only liquid systems (scope Flaw Evaluation in Elbows and Bent Pipe. Through-clarification) wall flaws in elbows and bent pipe may be evaluated using the Treatment of Service Level load combinations (scope straight pipe procedures given in N-513-4 provided the clarification) stresses used in the evaluation are adjusted to account for Treatment of flaws in austenitic pipe flux welds geometry differences. The hoop stress for elbow and bent pipe evaluation shall be (Equation 9 ofN-513-4):
(scope clarification) Minimum wall thickness acceptance criteria to consider longitudinal stresses in addition to hoop stress (scope clarification)
- a. =(pD, )[ 2R,,,d +R,sin¢ ]+l l;~)R,M,1 (9)
" 2t '.!.(R,,,,,d + R, sin¢) Ji- I In addition, several editorial changes to improve the Equation 9 is only applicable for elbows and bent pipe clarity of the Code Case are included.
where Ii 2'. 0.1. CODE CASE N-513-4 CHANGES AND TECHNICAL The axial membrane pressure stress for elbow and bent BASIS pipe evaluation shall be (Equation 10 ofN-513-4): The following subsections provide details regarding each change and their technical basis. er rn
=B ( pD,)
I 2t (10) Temporarv Acceptance Period N-5 l 3-3 specifies a temporary acceptance period that B 1 shall be equal to 0.5 for elbows and bent pipe. could extent out to 26 months. The NRC did not endorse this maximum period length and in the latest revision ofNRC The axial bending stress for elbow and bent pipe Regulatory Guide 1.147, placed a condition on N-513-3 evaluation shall be (Equation 11 ofN-513-4): stating, "... repair or replacement activity temporarily deferred under the provisions of this Code Case shall be performed during the next scheduled outage." N-513-4 (11) addresses this condition by removing the ma'l:imum duration limit and stating,"... the following requirements may be used Copyright© 2014 by ASlvlE
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 21 of 32 Note that if the piping design Code of Record does not provide a B2, it may be taken from the AS1'1E Section III, 2004 Edition or later Editions and Addenda. (Y
*(D/v/,)
=t - -
2I (16) The thermal expansion stress for elbow and bent pipe evaluation shall be (Equation 12 ofN-513-4): Equations 13-16 are consistent with the piping design by rule approach in AS1'1E Section III. The reducer and expander zones shown in Figure 8 ofN-513-4 are consistent with those a,= { R.~I,) (12) shown in Code Case N-597-2 [8]. As with elbows and bent pipe, N-513-4 allows for alternate methods to be used to calculate stresses used in the flaw evaluation. The first term of Equation 9 accounts for the hoop membrane stress variation around the circumference of the Flaw Evaluation in Branch Tees. Branch reinforcement elbow (higher at the intrados, lower at the extrados assuming a requirements shall be met in accordance with the C-0nstruction uniform wall thickness). The second term conservatively Code. If the Construction Code did not require reinforcement, accounts for the through-wall bending in the hoop direction a reinforcement region is defined as a region of radius D, of that results from elbow ovalization due to in-plane or out-of- the branch pipe from the center of the branch connection for plane bending and was taken from Reference [5]. Equations evaluation purposes. Through-wall flaws in branch tees 10-12 are consistent with the piping design by rule approach outside of the reinforcement region may be evaluated using in AS:ME Section III, NC/ND-3600 [6]. the straight pipe procedures given in N-513-4 provided the stresses used in the evaluation are adjusted as described below It is recognized in Reference [4] that the simplified flaw to account for the geometry differences. Evaluation of flaws evaluation approach may be overly conservative in some in the region of branch reinforcement is outside the scope of instances. Thus, N-513-4 allows for alternate methods to be this Case. used to calculate stresses used in the flaw evaluation. For example, the French have developed a more comprehensive The hoop stress and axial membrane pressure stress for approach in analyzing flaws in cracked elbows. They have branch tee evaluation shall be determined from N-513-4 implemented finite element analysis based correlations to Equation 13 and 14, respectively. The outside diameter for determine location specific stresses for flaw evaluation. each of these equations shall be for either the branch or run Reference [7] provides additional details. pipe depending on the flaw location. The axial bending stress and thermal expansion stress for branch tee evaluation shall be Flaw Evaluation in Reducers and Expanders. Through- determined from N-513-4 Equation 15 and 16, respectively. wall flaws iil reducers and expanders may be evaluated using the straight pipe procedures given in N-513-4 provided the As discussed previously, Equations 13-16 are consistent stresses used in the evaluation are adjusted to account for the with the piping design by rule approach in ASME Section III. geometry differences. Figure 8 ofN-513-4 illustrates the The limitation regarding flaw evaluation within the branch reducer and expander zones discussed below. Evaluation of reinforcement region is consistent with guidance given in flaws in the small end transition zone is outside the scope of Code Case N-597-2. As with elbows and bent pipe, N-513-4 N-513-4. The hoop stress and axial membrane pressure stress allows for alternate methods to be used to calculate stresses for reducer and expander evaluation shall be: used in the flaw evaluation. Flaw Evaluation of Heat Exchanger Tubing (13) N-513-4 allows for flaw evaluation in heat exchanger tubing provided the flaw can be characterized and the leakage monitored. The technical basis is that knowledge of flaw um pD
=B1 ( - -
0 ) (14) geometry is needed to evaluate the structural integrity of the
. 2t tubing containing the flaw. An example application from plant operating experience would be for leaks in nonferrous Note that D0 is either the small end OD for the small end tubing of air cooling coils. The heat exchangers being of or is the large end OD for the large end. For the large end plate-fin design could have tubing accessible for volumetric transition zone and central conical section, D, is the large end inspection.
OD. Also, if the piping design Code of Record does not provide a B 1, it may be taken from the AS1'1E Section III, Flexibilitv in Daily Walkdown Requirement 2004 Edition or later Editions and Addenda. N-513-4 changes the daily walkdown requirement for The axial bending stress and thermal expansion stress for leaking flaws from " ... leakage shall be observed by daily reducer and expander evaluation shall be: walkdowns ... " to ** ... leakage shall be monitored daily ... " This change allows for other techniques to be employed in a, -B
- 2 (DJ\,/'
2I ) (15) meeting the intent of the requirement instead of only physical walkdowns. Such techniques could include remote visual Copyright <1:J 2014 by AS1'1E
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 22 of32 equipment or leakage detection systems to determine if and branch tees. In addition, several other scope expansions leakage rates are changing. and clarifications are included as well as editorial improvements. With these changes, it is believed the Code Scope Limited to Liquid Systems Case will be of even greater use to utilities in avoiding unscheduled plant shutdowns without negatively impacting Provisions ofN-513-4 are now specifically limited to only plant safety. liquid systems. The methods ofN-513-4 should not be applied to the piping of air or other compressible fluid ACKNOWLEDGEMENTS systems. The development of Code Case N-513-4 was through the Treatment of Service Level Load Combinations efforts of the Working Group on Pipe Flaw Evaluation. The authors wish to recognize the efforts of all Working Group N-513-4 now specifically requires all Service Level load members especially the Working Group Chair, Dr. Doug combinations to be considered in flaw evaluations to Scarth. determine the most limiting. N-513-3 was previously silent as to what Service Level loading to consider. However, this REFERENCES requirement was implied in previous Code Case revisions when it required the use of pipe flaw evaluation procedures 1. ASME Code Case N-513-3, "Evaluation Criteria for specified in the referenced, non-mandatory appendices of Temporary Acceptance of Flaws in Moderate Energy AS!vIE Section XI. Appendix C in the 2002 Addenda and Class 2 or 3 Piping Section XI, Division l," Cases of later editions and addenda, and Appendix H for ferritic the AS!vIE Boiler and Pressure Vessel Code, January materials in earlier editions, required all operating conditions 26, 2009. or Service Levels be evaluated. 2. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, AS!vIE Section XI, Division I,,, Treatment of Flaws in Austenitic Pipe Flux Welds Revision 16, October 2010.
- 3. NRC Generic Letter 90-05, "Guidance for A reference to AS!vIE Section XI, Appendix C, C-6320 Performing Temporary Non-Code Repair of ASME has been added to paragraph 3. l(b) to address the instance Code Class I, 2, and 3 Piping," (June 15, 1990).
where a flaw in an austenitic pipe flux weld requires
- 4. Kilinski, T., Mohan, R, Rudlancl D., Fleming, M.,
evaluation. Such a flaw would require the use of Elastic Olson, R., Scott, P., Brust, F., Ghadiali, N., Plastic Fracture Mechanics (EPFM) criteria instead of Limit Witkowski, G., and Hopper, A, "Fracture Behavior Load criteria. Equation 1 ofN-513-4 now includes a Z factor of Circumferentially Surface-Cracked Elbows," to account for this specific application.
- NPREG/CR-6444, BMI-2192, December 1996.
- 5. Moore, S.E., and Rodabaugh, E.C., "Background for When the original version ofN-513 was developed, Changes in the 1981 Edition of the AS!vIE Nuclear Appendix C of AS!vIE Section XI only included flaw Power Plant Components Code for Controlling evaluation criteria for austenitic piping. The change from N-Primary Loads in Piping Systems," Journal of 513-1 to -2 included the reference changes to the new Pressure Vessel Technology, Volume 104, pp. 351 -
Appendix C that combined austenitic and ferritic flaw 361, November 1982. evaluation criteria. The need to reference EPFM criteria for
- 6. AS!vIE Boiler and Pressure Vessel Code, Section III, an austenitic flux weld flaw evaluation was overlooked.
NC/ND-3600, 2004 Edition.
- 7. Marie, S., Chapuliot, S., Kayser, Y., Lacire, M.H.,
Minimum Wall Thickness Acceptance Criteria Drubay, B., Barthelet, B., Le-Delliou, P., Rougier, In establishing a minimum wall thickness acceptance V., Naudin, C., Gilles, P., and Triay, M., "French criteria for paragraphs 3.2(b) and 3.2 (c) ofN-513-4, the RSE-M and RCC-l'v!R Code Appendices for Flaw requirement to consider longitudinal stresses in addition to lillalysis: Presentation of the Fracture Parameters hoop stress is added. While it is unlikely that a longitudinal Calculation - Part IV: Cracked Elbow," International stress based minimum wall thickness would be limiting Journal of Pressure Vessels and Piping 84, pp. 659-compared to the hoop stress based minimum wall thickness, 686, May 2007. the user should be aware of this possibility and confirm an 8. ASME Code Case N-597-2, "Requirements for appropriate tmtn as an acceptance criterion. Analytical Evaluation of Pipe Wall Thinning, Section XI, Division l," Cases of the ASJ\,ffi Boiler and CONCLUSIONS Pressure Vessel Code, November 18, 2003. This document serves as the technical basis for the fourth revision of Code Case N-513. This Code Case provides evaluation rules and criteria for temporary acceptance of flaws, including through-wall flaws, in moderate energy Class 2 or Class 3 piping. Currently, the scope of the Code Case is limited to straight pipe with some provision for flaw evaluation into fittings for a short distance from the weld attachment to straight pipe. N-513-4 provides flaw evaluation rules and criteria for elbows, bent pipe, reducers, expanders Copyright © 2014 by ASlvIE
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 23 of 32 Attachment 3 Cooper Nuclear Station Technical Basis for Proposed Alternative to Use ASME Code Case N-513-4 and Scope Expansion to a Higher Pressure Limit, Revision I
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 24 of 32 INTRODUCTION Background of Code Case N-513-4 Code Case N-513-4 [1] provides evaluation rules and criteria for the temporary acceptance of flaws, including through-wall flaws, in moderate energy piping. Moderate energy piping is defined as those piping systems where the maximum operating pressure and temperature do not exceed 275 psig and 200°F, respectively. The provisions of the Code Case are focused on preventing gross failure of the affected pipe for a temporary period while permitting leakage within the plant's Technical Specification. The Code Case provides rules for the evaluation of degraded pipe and tube for a short operating period, with inspection and monitoring requirements of the degraded condition as part of the overall integrity assessment. The previous Code Case revision, N-513-3 [2], is conditionally approved by the NRC in Regulatory Guide 1.147 [3]. The single condition deals with a requirement to perform the repair or replacement activity, temporarily deferred, during the next scheduled outage. Since the introduction of this Code Case, many utilities have used it as a basis for continued operation of degraded piping in moderate energy systems and that has resulted in significantly fewer relief requests to the NRC. Consequently, the industry has benefited from substantial cost savings while maintaining safety. To date, there have been no known instances where the use of the Code Case has resulted in any safety issues in the industry. The scope of Code Case N-513-3 is limited to straight pipe with some provision for flaw evaluation into fittings for a short distance from the weld attachment to straight pipe. There have been many instances where through-wall leaks have been observed in pipe components (i.e., elbows, reducers, expanders or branch tees) or bent pipe outside of the current Code Case scope. As a result, utilities are forced into taking alternate actions (e.g., repair, replacement or making requests for NRC relief) that can be significantly more burdensome without any measurable increase in plant safety. Code Case N-513-4 provides evaluation rules and criteria for the temporary acceptance of flaws for these instances. In addition, other enhancements are included with the fourth revision as described herein. Background of Moderate Energy Definition The genesis of Code Case N-513 was NRC Generic Letter (GL) 90-05 [4]. Prior to Code Case N-513, this GL was the only available guidance for plants regarding operational leakage in moderate energy piping. The definition of moderate energy piping in Code Case N-513 is consistent with GL 90-05. The scope of GL 90-05 is limited to Class 3 piping, but does address moderate and high energy systems. While non-code repairs are allowed by GL 90-05 (with NRC review) for the temporary period of operation prior to Code compliant repair/replacement, an additional requirement for the repair having load-bearing capability is necessary for high energy pipe applications. Both GL 90-05 and the latest approved revision of Code Case N-513 are identified as methods available to evaluate the structural integrity of piping with discovered leakage in the NRC Inspection Manual [5]. The definition of moderate energy (the 200°F and 275 psig limits) was first introduced in a NRC letter from A. Giambusso to licensees in 1972 to address postulated piping breaks in fluid
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 25 of 32 systems outside containment (discussed in the Branch Technical Position 3-3 of the current Standard Review Plan [6]). The importance of the 200°F temperature limit being below the boiling point of water at atmospheric pressure is clear in the definition of moderate energy piping. However, the basis for the pressure limit of 275 psig is less evident. In conversations with the NRC, there appears to be a link between the 275 psig limit and the recommended working pressure limit for Class 150 pipe flanges and flanged fittings given in the ASME B16.5 standard [7]. This basis is not known to be documented. A technical approach and basis are presented herein to justify the scope expansion of Code Case N-513-4 to a higher pressure limit. The Class 3 RHRSW system at Cooper has a design pressure of 490 psig [8]. It is desired to expand the scope ofN-513-4 such that it may be generically applied to this specific system as it has experienced part-wall degradation in the past. TECHNICAL BASIS FOR ALTERNATE USE OF ASME CODE CASE N-513-4 Summary of Code Case N-513-4 Changes The list below summarizes the major differences between the NRC-approved Code Case N-513-3 and Code Case N-513-4. Note that a brief reason for each change is included in parenthesis.
- Temporary acceptance period redefined from no longer than 26 months to the next refueling outage (addresses NRC condition given in Regulatory Guide 1.147)
- Flaw evaluation criteria included for elbows, bent pipe, reducers, expanders and branch tees (scope expansion)
- Allow flaw evaluation of heat exchanger tubing in specific instances (scope expansion)
- Daily walkdown requirement for through-wall leaks provides additional flexibility for user implementation (scope expansion)
- Limit scope to only liquid systems (scope clarification)
- Treatment of Service Level load combinations (scope clarification)
- Treatment of flaws in austenitic pipe flux welds (scope clarification)
- Minimum wall thickness acceptance criteria to consider longitudinal stresses in addition to hoop stress (scope clarification)
In addition, several editorial changes to improve the clarity of the Code Case are included. A detailed markup of Code Case N-513-3 showing all of the changes made for Revision 4 is provided in Attachment 1. Technical Basis The technical basis for the changes listed above is given in an ASME Pressure Vessel and Piping (PVP) conference paper [9] and is provided in Attachment 2. _j
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 26 of 32 Precedents There have been several submittals approved for N-513-4 use in specific applications and for generic use by Exelon. The table below lists several Safety Evaluation Reports as precedents for use of Code Case N-513-4. Table 1: Code Case N-513-4 Precedents SER Accession No. Plant Application Additional Requirements ML16230A237 Exelon Fleet Generic N-513-4 Critical leakage determination ML15070A428 ANO Leaking sweepolet 5 gpm leakage limit ML14316A167 Fort Calhoun Leaking elbow None ML14335A551 Peach Bottom Leaking elbow 5 gpm leakage limit TECHNICAL BASIS FOR CODE CASE N-514-4 SCOPE EXPANSION TO HIGHER PRESSURE LIMIT The technical approach includes several elements that support the following objectives: (i) show NRC precedent in approving relief requests for through-wall leakage in piping or components operating at high energy, (ii) show that structural integrity is not overly impacted when comparing flaw evaluations between piping systems operating at 275 and 490 psig, and (iii) show that possible jet thrust forces resulting from leaking flaws are not of concern. An outline of the technical basis detailing the elements used to support these objectives follows:
- Review of Relevant NRC Approved Relief Requests
- Structural Integrity Evaluation o Design minimum wall thickness comparison o Code Case N-513-4 allowable flaw size comparison o Code Case N-513-4 cover thickness requirement comparison
- Jet Thrust Force Evaluation Note that the evaluation rules and acceptance criteria of Code Case N-513-4 are not limited by pressure and remain valid for the higher limit of 490 psig.
Review of Relevant Relief Requests A detailed search of the NRC ADAMS database was conducted to identify any relief requests for continued operation of degraded high energy piping or components. The search results are summarized in Table 2.
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 27 of 32 Table 2: Summary of Relevant Relief Requests ADAMS Plant / Operating Accession Description Status Condition Number McGuire, Ul / Application for continued ML072140851 Approved 2,500 psig operation of leaking valve San Onofre U2, Generic application for continued ML101440381 U3 / < 275 psig operation of high temperature Approved (275° F) through-wall leaking pipe Exelon Plants / Generic request to use Code Case ML15043A496 Approved* 375 psig N-513-3 at a higher pressure limit
- In addition to the N-513-3 requirements, the NRC imposed an allowable leak rate limit of 5 gpm.
Structural Integrity Evaluation Design minimum wall thickness comparison ASME B3 l. l [ 1O] defines the design minimum wall thickness required for pressure loading, tm, as (without a corrosion allowance): t =_ __cp_D__,oc.___ (1) m 2(S+0.4p) where: p Design pressure Do Outside pipe diameter s Material allowable stress. Substituting maximum operating pressure for design pressure in Equation 1, the minimum required wall thickness increases 77% with an increase in pressure from 275 to 490 psig. While this percentage increase appears significant, the actual change in minimum required wall thickness is relatively low as the hoop stress at these pressures is small. Table 3 shows a comparison of minimum wall thicknesses for various pipe sizes. Table 3: Minimum Required Wall Thickness Comparison Nominal Pipe Size tmin for 275 psig* (in) tmin for 490 psig* (in) 6-inch 0.053 0.094 12-inch 0.102 0.181 18-inch 0.144 0.255
- A material allowable stress of 17.1 ksi is assumed.
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 28 of 32 Code Case N-513-4 allowable flaw size comparison Code Case N-513-4 allows for nonplanar, through-wall flaws to be evaluated as two independent planar through-wall flaws, one orientated in the axial direction and one orientated in the circumferential direction (i.e., a planar flaw characterization approach). The Code Case acceptance criteria require the flaw region be bounded by the area defined by the allowable axial and circumferential flaw sizes. Several example N-513-4 calculations were conducted illustrating the influence of higher pressure on allowable flaw size. Following the Code Case guidance, a linear elastic fracture mechanics (LEFM) evaluation was performed for various carbon steel pipe sizes to determine the maximum allowable flaw sizes at maximum operating pressures of 275 and 490 psig. Table 4 summarizes the results and several notes are provided giving more details regarding the analysis inputs. The influence of the higher pressure is clearly seen and a greater impact is observed in the axial direction as expected since pressure hoop stress is twice the axial membrane stress due to pressure. While the higher pressure does decrease the allowable flaws sizes, the effect does not impact the functionality or validity of the Code Case approach. Table 4: Allowable Axial and Circumferential Flaw Size Comparison Nominal Axial Direction (in) Circumferential Direction (in) Pipe Size 275 psig 490 psig %~ 275 psig 490 psig %~ 6-inch Mb=40 3.1 1.9 39% 2.8 2.1 25% in-kips 12-inch Mb= 170 3.0 1.5 50% 3.3 2.1 36% in-kips 18-inch Mb= 260 2.2 0.9 59% 3.3 1.8 44% in-kips Notes: - Piping material assumed A106 Grade B; standard schedule thickness.
- Applied bending moment for each pipe size results in a stress ratio of about 0.25 at 275 psig.
- Allowable flaw sizes based on Service Level B structural factors.
- Analysis based on a conservative lower shelf toughness value of 45 in-lb/in 2
- Code Case N-513-4 cover thickness requirement comparison Code Case N-513-4 provides a branch reinforcement method to evaluate nonplanar, through-wall flaws. As part of the branch reinforcement approach, an opening is modeled such that its diameter fully bounds the leaking flaw. In practice, there could exist a remaining wall ligament within the modeled opening. Equation 9 of Code Case N-513-4 provides assurance against pressure blowout (i.e., wall ligament failure) by requiring an average cover thickness, tc,avg, within the modeled opening:
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 29 of 32 tc,avg ';?_ 0.353dadjJ[ (2) where: dadj Modeled opening diameter p Maximum operating pressure s Material allowable stress. The average cover thickness requirement increases about 33% with an increase in maximum operating pressure from 275 to 490 psig. Figure 1 illustrates the average cover thickness required as a function of adjusted diameter for 275 and 490 psig. Note that a material allowable stress of 17 .1 ksi is assumed. Typically, modeled openings have a diameter < 1 inch and the change in the required cover thickness is small(< 15 mils). 0.10 0.09 0.08 c§. 0.07 -+-275 psig V1 V1 CJJ _s:i 0.06 ....,.490psig u
..c I--
w 0.05 0 u
-o 0.04
~
af er: 0.03 0.02 0.01 0.00 0 0.25 0.5 0.75 1 Adjusted Diameter (in) Figure 1: Required Cover Thickness vs. Adjusted Diameter for 275 and 490 psig
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 30 of 32 Jet Thrust Force Evaluation Part 3.6.2 of the NRC Standard Review Plan [11] provides a simplified dynamic analysis model to quantify the jet thrust force, T, of water from a pipe break. The following equation is given: T=KpA (3) where: K Thrust coefficient (2.0 for subcooled, nonflashing water) p System pressure prior to pipe break A Pipe break area. Figure 2 shows a comparison of jet thrust force for pressures of 275 and 490 psig over a range of through-wall opening diameters. For small through-wall opening diameters(< 0.5 in), the difference in the jet thrust force is small. It is unlikely larger openings would be tolerated by the plant as this would result in excessive leakage and possibly impact system operability or room flooding abatement. 800 700 600
...._275 psig 500 ..+,,,,490 psig
;;=-
..a
<lJ
~ 400 0
LL
+-' <lJ 300 200 100 0
0 0.25 0.5 0.75 1 Through-Wall Hole Diameter (in) Figure 2: Jet Force vs. Through-wall Hole Diameter for 275 and 490 psig
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 31 of 32 ADDITIONAL REQUIREMENTS Implementation of Code Case N-513-4 requires additional actions be satisfied by the plant including observing leakage daily to confirm analysis conditions used in the evaluation remain valid, frequent periodic inspections to track flaw growth and augmented examinations to assess degradation of the affected system. Furthermore, the following additional requirement is included as part of the Relief Request in order to bolster defense-in-depth and avoid adverse consequences: CNS will limit the leak rate from a through wall flaw, if present, to the leakage allowed by the engineering evaluation or 5 gpm, whichever is lower. CONCLUSIONS This attachment provides a technical basis for proposed alternative use of ASME Section XI Code Case N-513-4 and its scope expansion to a higher pressure limit from 275 to 490 psig for the Cooper Nuclear Station, specific to the RHRSW system. The discussion provided herein demonstrates that the use of Code Case N-513-4 with the higher pressure limit will reduce plant burden without any adverse effect on safety. The technical basis is comprised of four primary elements. First, Attachment 2 provides the technical basis for the new content and evaluation methods in Code Case N-513-4 including those for flaw analysis in piping components such as elbows, reducers and branch tees. Second, relevant NRC relief requests were reviewed. They demonstrate that there is precedent for the temporary acceptance of leaking flaws in Class 2/3 piping and components at pressures higher than 275 psig. Third, a structural integrity evaluation was performed to determine the impact of the increased pressure on design minimum wall thickness, N-513-4 allowable flaw sizes and the N-513-4 cover thickness requirement. While the influence of the higher pressure was observed in the evaluation, structural integrity can still be demonstrated. The functionality and validity of the Code Case methods at higher pressure were confirmed. Finally, jet thrust forces were estimated for a leaking pipe at 275 and 490 psig. A significant change in jet thrust forces was only seen with large opening areas that would result in high leak rates, i.e., rates that would challenge system functionality or local spray and/or compartment flooding requirements. As part of the generic Relief Request, one additional requirement is introduced to bolster defense-in-depth by limiting the leak rate from a through wall flaw, if present, to the leakage allowed by the engineering evaluation or 5 gpm, whichever is lower. REFERENCES
- 1. ASME Code Case N-513-4, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l ," Cases of the ASME Boiler and Pressure Vessel Code, May 7, 2014.
Attachment 1 - 3 RR5-03 Cooper Nuclear Station 32 of 32
- 2. ASME Code Case N-513-3, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping Section XI, Division l," Cases of the ASME Boiler and Pressure Vessel Code, January 26, 2009.
- 3. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 17, August 2014.
- 4. NRC Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping," (June 15, 1990).
- 5. NRC Inspection Manual, Chapter 0326, "Operability Determinations & Functionality Assessments for Conditions Adverse to Quality or Safety," Issue Date: 12/03/15.
- 6. NRC Standard Review Plan, NUREG-0800, Branch Technical Position 3-3, "Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," Revision 3, March 2007.
- 7. American National Standard, "Pipe Flanges and Flanged Fittings," ASME Bl6.5-2003.
- 8. Nebraska Public Power District Design Calculation No. NEDC 02-051, Revision 1, "Design Temperature Determination for SW, CW, WW, & CS Piping; Design Pressure Determination for SW-1 piping," September 17, 2002, SI File No. 1700405.201.
- 9. McGill, R.O., Cipolla, R.C., DeBoo, G. and Houston, E.J., "Technical Basis for Proposed Fourth Revision to ASME Code Case N-513," Proceedings of the ASME 2014 Pressure Vessels & Piping Conference, PVP2014-28355.
- 10. USA Standard, USAS B31.l.O, "Power Piping," 1967 Edition.
- 11. NRC Standard Review Plan, NUREG-0800, Part 3.6.2, "Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping,"
Revision 2, March 2007.}}