View source for B17370, Requests NRC Review & Approval of Leak Before Break Methodology for Portions of Safety Injection & Shutdown Cooling Sys.Rev 0 to SIR-98-048 Rept,Encl

{{Adams
| number = ML20236U627
| issue date = 07/24/1998
| title = Requests NRC Review & Approval of Leak Before Break Methodology for Portions of Safety Injection & Shutdown Cooling Sys.Rev 0 to SIR-98-048 Rept,Encl
| author name = Bowling M
| author affiliation = NORTHEAST NUCLEAR ENERGY CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000336
| license number = 
| contact person = 
| case reference number = RTR-NUREG-1061
| document report number = B17370, NUDOCS 9807300353
| package number = ML20236U629
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 16
}}

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Ned                                                  Hope Ferry Rd. (Route 156), Waterford, CT 06385 Nuclear Energy                                       Maistone Nociear romr sation                        l Northeast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385-0128 (860) 447-1791 Faz (860) 444-4277 The Northeast Utihties System JLL 241998                     j l
Docket No. 50-33G B17370 i
U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Staticn, Unit No. 2 Request For Permission to Apply Leak Before Break Methodology To Portions of Safety iniection and the Shutdown Coolino Systems The purpose of this letter is to request NRC review and approval of Leak Before Break (LBB) methodology for portions of the Safety injection (SI) and the Shutdown Cooling (SDC) systoms. The scope of the proposed LBB application is limited to the ASME Class 1 portion of the Si and the SDC systems extending from their respective connection at the main coolant loop (MCL) piping to the first isolation valve. The portions of the Si and SDC piping systems under consideration are designated as high energy, being subject to the normal operating conditions of the MCL piping. All affected piping considered for this application of LBB is located inside containment.
NNECO also expects, in future correspondence, to request NRC review and approval of Leak Before Break methodology for the Pressurizer Surge Line based upon the expected results of ongoing engineering reviews.
A recent evaluation of the walkdown information on the Si and SDC piping systems, completed in May 1998, concluded that the protection of adjacent closed loop piping systems from potential pipe breaks was not provided as part of the original Millstone                             <
Unit No. 2 design. This concern was reported to the NRC Staff in LER 98-005-00.                                   l NNECO has evaluated various options to address this concern and concluded that the                              )l application of the LBB technology is the most effective and reasonable approach to                                ;
ensure adequate pipe break protection of the closed loop piping systems. Other options considered included the installation of whip restraints and jet shields, relocation                j      ;
j'         of targeted piping, and the installation of added secondary isolation valves in the //
l          closed loop systems.
4 Using the guidance and criteria provided in NUREG 1061, Volunic                              3,   the             l implementation of the LBB methodology will result in the elimination of the requirement                           i 9807300353 980724~                '
PDR     ADOCK 0500o336 P                    PDR Ot0422 5 REV,12 95
* U.S. Nucl:ar Regulitory Commission B17370tPage 2 to consider dynamic effects of postulated ruptuies for the Si and SDC piping. The LBB evaluations for the Si and SDC piping are provided herein for NRC Staff review.
This request for permission for the implementation of the Leak Before Break methodology is organized as follows:
Attachment 1 to this letter provides an executive summary of the NNECO submittal.
Attachments 2 provides the background and introductory remarks and system descriptions related to this request.
Attachments 3 and 4 provide a detailed report on the results of the LBB evaluations for the safety injection and the shutdown cooling piping syitems, respectively. These reports address all applicable NRC LBB evaluation criteria contained in NUREG 1061, Volume 3.
Attachment 5 provides a discussion of the plant specific information and service experience to date with the Si and the SDC piping rystems. Issues, such as NDE inspection results, water hammer experience, and th3 i'npect of any other potential sources of degradation, such as fatigue, vibration, corrwion, thermal stratification, etc.
are addressed.
Attachment 6 contains a discussion of the material properties, welding processes and weld procedures that pertain to the Si and SDC piping systems.
Attachment 7 contains a detailed description of the installed leak detection system inside containment.
Attachment 8 provides a qualitative value impact assessment of this request.
Some of the benefits that will accrue from application of LBB technology include elimination of the need for installing the pipe whip restraints and jet barriers on the Si and SDC piping, elimination of associated occupational man-rem exposure during construction, periodic removal and re-installation of these restraints and barriers, less congestion in the containment, and elimination of potential damage to piping from i         n ' function of rupture restraints.
There are no regulatory commitments contained within this submittal.
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      ~ U.S. Nucl=r Regul tory Commission B17370\Page 3 If you have any additional questions conceming this submittal, please contact Mr. Ravi G. Joshi at (8S0) 440-2080.
I Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
                                                              /~          1%
Martin L. Bowling, Jr. U Recovery Officer - Technical Services Attachments cc:    H. J. Miller, Region 1 Administrator W. D. Travers, Ph.D., Director, Special Projects Office P. F. McKee, Deputy Director, Licensing and Oversight, Special Projects Office W. D. Lanning, Deputy Director, inspections, Special Projects Office J. P. Durr, Branch Chief Inspections, Special Projects Office D. G. Mcdonald, Jr., NRC Senior Project Manager, Millstone Unit 2 D. P. Beaulieu, Senior Resident inspector, Millstone Unit 2                    i I
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Docket No. 50-336 B17370 Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Request For Permission to Apply Leak Before Break Methodology To Portions of Safety injection and the Shutdown Cooling Systems Executive Summary l
July 1998

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U.S. Nucl=r R:gulatory Commission B17370\ Attachment 1\Paga 1 EXECUTIVE 

==SUMMARY==

  . This Millstone Unit 2 request to the NRC for implementation of LBB in accordance with l  the provisions of General Design Criterion (GDC) 4 of Appendix A to 10CFR Part 50 is based on the methodology described in NUREG 1061, volume 3. The subsequent attachments two through eight provide the technical justification for our LBB submittal.
The scope of requested implementation of LBB is limited to the ASME Class 1 portion of the Safety injection (SI) ar.d the Shutdown Cooling (SDC) systems extending from l  their respective connection at the Main Coolant Loop (MCL) to the first isolation valve.
l  The portions of the Sl and SDC piping systems under consideration are designated as high energy, being subject to the normal operating conditions of the MCL piping. All l  affected piping considered for this application of LBB is located inside the containment.
The subject SI and SDC piping is 12 inch Schedule 140 seamless piping fabricated from 316 SS and extend from their respective connection to the MCL to the first isolation valve. Deterministic Elastic-Plastic Fracture Mechenics (EPFM) J-Integral-                                        i Tearing Modulus (J/T) evaluations were performed for all significant weld locations identified in each piping system. In order to demonstrate the required safety margins of
[ on the accident load combination, factor of two on critical crack length, and margin of 10 on leakage detection, several fracture mechanics evaluations were performed:
The results of fracture mechanics evaluation for the Safety injection piping show that a 4.29 inches long crack will leak at the rate of 11 gpm under normal operating conditions
!  and is less than one-half the critical crack length of 8.78 inches calculated for the accident loading condition. This 4.29 inches long crack remains stable under piping loads equal to [ times the accident load condition..
Similarly, the results of fracture mechanics evaluation for the Shutdown cooling piping show that a 7.84 inches long crack will leak at the rate of 12.2 gpm under normal operating conditions and is equal to one-half the critical crack length of 15.68 inches calculated for the accident loading condition. This 7.84 inches long crack remains stable under piping loads equal to 8 times the accident load condition.
The above results were determined based on calculated critical flaw sizes at each weld                                      )'
location uWog generic lower bound base or weld material properties for (a) pressure plus deao weight, plus thermal, plus safe shutdown earthquake (P+DW+TH+SSE) and for (b) 8(P+DW+TH+SSE). Leakage-size-crack (LSC) was then determined as the                                                 ,
smaller of one half of the critical crack length calculated for stress condition (a) or full                                j critical crack length for stress condition (b). Leakage from LSC flaw size established above was then calculated under normal operating load combination of pressure plus                                          ]
dead weight, plus thermal (P+DW+TH). The calculated leakage was then compared to

U.S. Nucl:rr Regulatory Commission B17370\Att: chm::nt 1\Pcgo 2 ten times the installed leak detection capability at Millstone Unit 2 to demonstrate a safety margin of 10 on leakage detection as stipulated in NUREG 1061, volume 3.
Millstone Unit 2 has an installed leak detection capability of 1 gpm unidentified leakage inside containment which is more than u quate to meet the LBB leakage detection requirements calculated above for the di and SDC piping. This capability is in compliance with Regulatory Guide 1.45 requirements.
ASME XI fatigue crack growth analyses of shallow flaws performed for both the SI and the SDC piping systems show an insignificant growth in both flaw depth and flaw length. To account for the reactor water environment, a factor of 2 was applied to the in-air ASME XI crack growth law for austenitic steels. For carbon steel, the ASME Xi fatigue crack growth curve for ferritic steel in water environment was utilized. Deeper flaws grow preferentially through-wall prior to extending in length and becoming unstable.
The susceptibility to degradation mechanisms, such a 3 intergrannular stress corrosion cracking (IGSCC) and        flow accelerated corrosion (FAC) were considered and determined not to be a concern for the affected portions of the systems. Similarly, the potential for loadings, such as vibration, water hammer, and thermal stratification was considered. The Safety injection nozzles and safe ends are protected from thermal fatigue by thermal sleeves. The probability of occurrence of water hammer or significant vibrations were considered low. The layout of the Safety injection and the Shutdown Cooling piping systems, including the close proximity of the RCS isolation valves minimizes their susceptibility to thermal stratification. The impact of thermal aging on the material properties of cast stainless steel was also considered.
In summary, Millstone Unit 2 has satisfied the requirements relative to the leak-before-break analysis, stipulated in NUREG 1061, Volume 3, for the high energy portions of the Safety injection and the Shutdown Cooling piping systems and accordingly requests approval to apply the LRB methodology as allowed per GDC-4. This request will eliminate the need to postulate pipe ruptures and consider their associated dynamic effects from the design basis for the high energy portions of the Si and SDC piping.
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Docket No. 50-336 817370 l
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Attachment 2 l                 Millstone Nuclear Power Station, Unit No. 2
: l. Request For Permission to Apply Leak Sefore Break Methodology l      To Portions of Safety injection and the Shutdown Cooling Systems
!                        Introduction And Background -
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l l                    U.S. Nucirr R:gulatory Commission l                    817370\ Attachment 2\Page 1 INTRODUCTION AND BACKGROUND In the original assessment of high energy line breaks (HELB) inside containment, the protection of closed systems was not addressed. The generic industry oversight in protecting the closed loop systems was recognized in Information Notice 89-55.
Millstone Unit 2 completed a full evaluation of the closed loop systems and supporting walkdowns in May,1998. This review of inside the containment HELB program for its adequacy in protecting closed loop piping systems from a p..stulated impact due to the rupture of an adjacent High Energy Line, identified several deficiencies (Ref.1).
The Safety injection, Shutdown Cooling and Pressurizer Surge Lines were all identified as initiating sources of unacceptabb interaction with closed loop systems. The possible corrective actions for the identified adverse interactions considered options for both targets (moving lines or adding secondary isolation valves) and sources frestraint of ruptured line or elimination of the postulated break).
The best target-related option is the use of a second isolation valve at the containment penetrations to ensure containment integrity under the postulated breaks. While such an approach could provide successful resolution of this issue, it introduces considerable additional maintenance and reliability concerns which would negatively impact plant operation. In total, eight new motor or air operated isolation valves, with associated controls to close on a containment isolation actuation signal (CIAS), would be required to address all identified interactions. The target lines under consideration include four Reactor Building Component Cooling Water (RBCCW) lines (two 6" and two 8" penetrations), two Steam Generator Blowdown (SGBD) lines (two 2" penetrations) and two SGBD Sampling lines (two 1/2" penetrations). The addition of more motor or air operated valves to these systems increases their complexity and decreases their overall efficiency and reliability. Thus, the benefits of this approach are judged to be greatly exceeded by the overall negative impact on plant operation.
Considering the size of the Si and SDC lines (all 12" diameter) and their location within the loop area, where structural attachment locations are minimal, the application of Leak-Before-Break (LBB) methodology to demonstrate low probability of pipe rupture          !
provides the oest promise of success. Given that the protection of closed loop systems      l is considered a prudent enhancement to the existing LB/DB of Millstone Unit 2, and given the previous industry precedent for application of Leak Before Break methodology for large lines attached to the primary loop, this approach is considered the most effective and reasonable implementation approach.
2.1      SYSTEM DESCRIPTION - SAFETY INJECTION SYSTEM Safety injection System (SIS) is part of the Emergency Core Cooling System (ECCS).
The primary system function is to provide both short and long term core cooling by injecting borated water into the core through each of the four main coolant loop cold legs following a loss-of-coolant-accident (LOCA). The safety injection system is divided into two parts: the High Pressure Safety injection (HPSI) system injects borated 1
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U.S. Nucl:ar Regulatory Commission B17370\ Attachment 2\Page 2 water using HPSI pumps at high pressure with flow capability commensurate with RCS inventory loss during a small break LOCA; and the Low Pressure Safety injection (LPSI) system uses two LPSI pumps to provide borated water for long term core cooling and injecting negative reactivity whenever the RCS pressure decreases rapidly as a result of large break LOCA up to and including the break in the largest RCS main coolant loop pipe.
The high energy portion of the HPSI piping is located directly adjacent to the reactor coolant loop cold legs and is part of the RCS pressure boundary. The physical layout of the safety injection piping is shown in Figures 2-1 through 2-4.
2.2    STRESS ANALYSIS & ASME CODE COMPLIANCE -
SAFETY INJECTION SYSTEM:
The stress and fatigue design compliance of the safety injection piping in accordance with the Bechtel Design Specification Number 7604-M-290 and the Class 1 requirements of ASME Ill,1971 was performed by Teledyne Engineering Services.
The requirements of I&E Bulletin 79-14 involving as-built verification were subsequently implemented for the safety injection piping.
The piping analyzed extends from each of the four RCS cold leg loop nozzles to their respective safety injection tank at one end of the branch piping and to the containment
; penetration at the other. Each of the four safety injection piping systems is computer analyzed (ADLPIPE) for pressure, dead weight, thermal expansion, OBE and SSE loads. Seismic inertia analysis is based on the enveloped response spectra method.
Teledyne performed thermal transient temperature distribution analysis using simplified chart methods and detailed analyses with one-dimensional and two-dimensional heat
! transfer programs. Code compliance is demonstrated for normal, upset, emergency, faulted and test conditions. Additionally, cumulative fatigue usage factors for 40 year
! service life are shown to be within Code acceptable limit of 1.0. All pressure boundary attachments to Class 1 piping were designed to ASME Section ill, NB requirements.
Structural Steel supports were qualified in accordance with the AISC requirements.
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U.S. Nucirr Regulatory Commission B17370\ Attachment 2\Page 7                                                                         j 2.3    SYSTEM DESCRIPTION -SHUTDOWN COOLING SYSTEM The primary system function of the shutdown cooling (SDC) system is to reduce the reactor coolant temperature to cold shutdown condition and to maintain this temperature during refueling. The LPSI pumps are used to draw suction from the Main Coolant Loop 2 hot leg piping and discharge it to each of the four cold legs after removal of decay heat in the shutdown cooling heat exchangers. The high energy portion of the SDC suction piping is located directly adjacent to the reactor coolant loop 2 hot leg and is part of the RCS pressure boundary. The physical layout is shown in                 ,
Figure 2-5.
The high energy portion of the SDC discharge piping is located directly adjacent to the reactor coolant cold leg loops and is part of the RCS pressure boundary and is the same piping as the Si piping discussed above in Section 2.1.
2.4    STRESS ANALYSIS & ASME CODE COMPLIANCE - SHUTDOWN COOLING I
The stress and fatigue design compliance of the shutdown cooling piping in accordance with the Bechtel Design Specification Number 7604-M-209 and Class 1 requirements of ASME Ill,1971 was performed by Teledyne Engineering Services.
The requirements of I&E Bulletin 79-14 involving as-built verification were subsequently
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implemented for the shutdown cooling piping.
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Seismic inertia analysis is based en the enveloped response spectra method.
Teledyne performed thermal transient temperature distribution analysis using simplified chart methods and detailed analyses with one-dimensional and two-dimensional heat transfer programs. Compliance with Code is demonstrated for normal, upset, emergency, faulted and test conditions. Additionally, cumulative fatigue usage factor                 i for 40 year service life is shown to be within the Code acceptable limit of 1.0. Ali pressure boundary attachments to Class 1 piping were designed to ASME Section Ill, NB requirements. Structural Steel supports were qualified in accordance with the AISC requirements.

==2.5   REFERENCES==

l  1. 25203-EV-98-0063, " Technical Evaluation for NRC Information Notice No. 89-1 55: " Degradation of Containment isolation Capability By a High-Energy 1.ine Break," issued May 15,1998.

U.S. Nuclxr R:gulatory Commission B17370\ Attachment 2\Page 8                                                                                                                      l
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Docket No. 50-336 B17370 Attachment 3 Millstone Nuclear Power Station, Unit No. 2 Request For Permission to Apply Leak Before Break Methodology To Portions of Safety Injection and the Shutdown Cooling Systems Report Number SIR-98-048, Rev. O.
Prepared By: Structural integrity Associates l
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July 1998
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