NRC-97-0043, Provides Info Re Recently Identified Issued Involving Effects of Postulated HELBs Inside Containment on EECWs & Utils Basis for Conclusion
| ML20137N116 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/04/1997 |
| From: | Romberg W DETROIT EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137N122 | List: |
| References | |
| CON-NRC-97-0043, CON-NRC-97-43 NUDOCS 9704080263 | |
| Download: ML20137N116 (30) | |
Text
.
Wayne S. Rombe,g Assis.tarr ytte Prescent and Manager,Tectncal 00 Norm Dite Hwy ECHSOn ut%"Or""
W ar.au.
l April 4,1997 r
[
U.S. Nuclear Regulatory Commission i
Attn: Document Control Desk Washington, D.C. 20555-0001
References:
Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
Subject:
Effects of Postulated Containment Pipe Ruptures i
On Emergency Equipment Cooling Water System i
The purpose of this letter is to provide information related to a recently identified issue involving the effects of postulated High Energy Line Breaks (HELBs) inside containment on the Emergency Equipment Cooling Water System (EECWS). Postulated HELB impacts on the EECWS inside containment were not included as part of the licensing basis fe. Fermi 2. Consequently, it is Detroit Edison's position that this issue does not represent a non-conformance with tl.e licensing basis. This position is based on an i
extensive review oflicensing documents which indicate that the effects of postulated IIELBs were considered only for specific systems. The specific systems are identified in the Final Safety Analysis Report (FSAR) and NRC Safety Evaluation Report (SER).
Notwithstanding Detroit Edison's position that impacts of HELBs on the EECWS inside
/
containment are not part of the plant licensing basis, Detroit Edison has performed a comprehensive engineering evaluation of this issue from the safety standpoint and has
/
concluded that safe operation of the facility is not compromised. This letter provides Detroit Edison's basis for thi. conclusion.
Licensing Backgroand 0l i
Initial discussion regarding the impacts of HELBs on safety systems dates back to the Preliminary Safety Analysis Report (PSAR). The plant design did not originally include the EECWS, The Reactor Building Closed Cooling Water System (RBCCWS) was proposed to provide cooling water during normal and accident conditions. The RBCCWS OR0003 9704000263 970404 PDR ADOCK 05000341;
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April 4,1997 NRC-97-0043 Page 2 design basis as described in the PSAR did not include dynamic effects from accidents (i.e.,
IIELBs):
"The reactor building closed cooling water system will be designed and provide adequate cooling water to various equipment during all modes of operation. The system will provide a high degree of reliability and will be capable of being tested during normal operation."(PSAR p.10-18)
This is in contrast with the descriptions of other systems such as the Emergency Core Cooling System (ECCS) where the design bases specifically included dynade effects:
"The equipment of the core standby cooling systems shall withstano the physical effects of a loss of coolant accident so that the core can be effectively cooled.
These effects are missiles, fluid jets, high temperature, pressure, and humidity."
(PS AR p. 6-2)
This indicates that from the first stages of facility licensing, dynamic effects oflIELBs were specifically considered only for certain systems.
This theme is apparent in Detroit Edison to Atomic Energy Commission (AEC) correspondence in the 1973 timeframe. During the summer of 1973, Detroit Edison transmitted a series ofletters to the AEC which documented specific considerations regarding selection ofIIELBs inside containment, systems impacted, and design of protective features, such as whip restraints. These considerations were based on General Electric and other vendor specifications for Fermi 2 and other BWRs of the Fermi 2 vintage. In addition to defining systems considered for IIELB impacts, these letters also defined other considerations regarding single failures to be considered in conjunction with IIELBs, and the proce.s used for dispositioning the consequences ofiiELBs on affected systems. The consideration of systems affected by IIELBs was focused on maintenance of core cooling integrity (i.e., ECCS), reactivity control, and containment integrity.
These considerations were also carried through to the FS AR which was originally submitted in 1974. The focus for IIELBs inside containment directly focused on accident mitigation (ECCS, containment, ability to scram, etc.).11ELB impacts inside containment on the EECWS were not discussed in the FSAR The systems that were specifically considered for impacts from IIELBs inside containment are identified in the FSAR Section 3.6 and FSAR Section 6.2. The FSAR relies on the general principle of physical separation of redundant components to provide functional protection from HELBs inside containment. The NRC evaluation ofinside containment IIELB impacts, including the specific systems affected, is documented in SER Section 3.6.1.1. Specific evaluation of the EECWS for the effects of pipe whip and jet impingement inside primary containment was not included:
April 4,1997 NRC-97-0043 Page 3
" Protected cooling systems inside containment include: low pressure coolant injection, core spray, and automatic depressurization Control rod withdrawal lines are protected to the extent that no more than one line in a nine-rod array could be crimped closed. Protection against dynamic effects is provided by pipe whip restraints, equ!pment shields, and physical separation of piping, equipment, and instmmentation. The pipe whip restraint design description, design bases, load combinations, allowable stresses, and design analysis methods and procedur'es are.
summarized in Section 3.6 of the FSAR."
"We have reviewed applicant's design for protection of dynamic effects of postulated pipe ruptures inside containment and conclude they are acceptable. The analytical methods and procedures that were used to establish restraint locations and the pipe restraint dynamic interaction are based on acceptable methods of engineering analyses. The pipe whip restraints are designed to withstand the resultant loadings and to assure the protection of safety-significant structures, systems, and components. The basis for acceptance in our review has been Regulatory Guide L46, " Protection Against Pipe Whip Inside Containment."
(SER 3.6.1.1 P 3-10)
Evaluation OfIIELB Impacts On EECWS I
Notwithstanding Detroit Edison's position that impacts of HELBs on the EECWS inside containment are not part of the plant licensing basis, Detroit Edison has performed a j
comprehensive engineering evaluation of this issue from the safety standpoint and has 1
concluded that safe operation of the facility is not compromised. Detroit Edison's basis for this conclusion is provided in this letter.
This letter addresses evaluations related to EECWS piping impacts. Potential impacts to containment isolation power and control conduit are being evaluated separately. Thirty-six postulated IIELBs inside containment were identified as potentially impacting EECWS piping. These breaks were postulated in accordance with the criteria stated in the Fermi 2 UFSAR Section 3.6.1.2, Design Basis Pipe Break Criteria. The thiny-six pipe breaks were reviewed to determine which of the breaks could be eliminated from consideration based on the following criteria:
i i
- 1. Guidelines in NRC GL 87-11," Relaxation in Arbitrary Intermediate Pipe Rupture Requirements," are used to eliminate the requirement for postulation ofintermediate pipe breaks. If the maximum stress range as calculated by Code equation 10 (ASME Boiler & Pressure Vessel Code,Section III Anicle 3600) exceeds 2.4S, intermediate pipe breaks are not postulated provided that Code equation 12 or 13 stresses are less than 2.4S. and the cumulative usage j
factor (CUF) does not exceed O.L 1
i
April 4,1997 NRC-97-0043 Page 4
- 2. Based on the definition of a terminal end in ANSI /ANS 58.2, " Design Basis For Protection Of Light Water Nuclear Power Plants Against The Effects Of Postulated Pipe Rupture," in-line fittings, such as valves, not assumed to be anchors in the piping code stress analysis, are not terminal ends. Therefore, terminal ends are not postulated at valves separating high and moderate energy piping.
Postulated pipe break identifications are shown in Attachment 1. Pipe stresses and usage factors at analytical node points corresponding to the break locations were obtained from design basis piping calculations. Drawings showing the postulated break locations and corresponding break numbers are included as Attachment 2. Postulated pipe break locations shown on Figure B-9 for Recirculation Loop A piping (RAXX break location designation) are also applicable and correspond to Recirculation Loop B piping (RBXX break location designation).
Results of the evaluation are provided in Attachment I and exceptions to the GL 87-11 crite.ia are summarized below:
- 1. At postulated break locations CB5C (Core Spray Line B), WAl1 A, WAl2C (Feedwater Line A), WB11 A, WB12C (Feedwater Line B) and RPl (RCIC Line), the Code equation 10 stresses are greater than 2.45.. However, at all thirty-seven postulated break locations Code equations 12 and 13 stresses are less than 2.4S.
- 2. Postulated break locations CB6 (Core Spray Line B), RRA3C (RHR Return Loop A), RRB3C (RHR Return Loop B), RS4 and RSS (RHR Supply) are at valves separating high and moderate energy piping and are eliminated because they do not meet the ANSI /ANS 58.2 definition of terminal ends and are considered arbitrary intermediate breaks. This deviates from the definition of terminal ends provided by footnote 3 on page 3.6.2-13 of MEB 3-1 provided in GL 87-11. According to this definition, for piping runs which are maintained pressurized during normal plant conditions for only a portion of the run (i.e., up to the first normally closed valve) a terminal end of such runs is the piping connection to this closed valve. Based on the definition of a terminal end in ANSI /ANS 58 2, in-line fittings such as valves, not assumed to be anchors in the piping code stress analysis, are not terminal ends. Therefore, terminal ends are not postulated at valves separating high and moderate energy piping. The ANSI /ANS 58.2 terminal end definition was used to eliminate these four postulated breaks at intermediate valve locations. The low stresses and CUFs at these locations allow the breaks to be eliminated from consideration.
Based on the design basis stresses and cumulative usage factors (all less than 0.1) for the applicable piping, as summarized in Attachment 1, and application of criteria for i
L i
April 4,1997 NRC-97-0043 Page 5 eliminating postulated pipe breaks as outlined above, it is concluded that thirty-three of the thirty-six postulated pipe break locations may be eliminated from further consideration for impact to EECWS piping. Three of the eliminated pipe break locations, RSS (RHR Supply), RRA3C (RHR Return Loop A), and RRB3C (RHR Return Loop B), potentially impact containment isolation power and control conduit and are being addressed separately.
The remaining three postulated pipe breaks are: RA20A, RB1 A, and RB28A, which are connections of the Reactor Recirculation Loop A and B piping to the RPV nozzles. These locations are classified as terminal ends and can only impact EECWS Division I.
The remaining three break locations are each comprised of three welds in Type 304 stainless steel piping. These breaks were reviewed using mechanistic fracture mechanics modeling techniques to determine if a Leak Before Break (LBB) approach could be used to eliminate dynamic effects associated with postulated breaks at these locations. This evaluation is documented in Structural Integrity Associates (SIA) report, SIR-97-029, dated h1 arch 25,1997 (Attachment 3). LBB can be applied to these three break locations because there is no active mechanism to lead to fatigue crack growth, the piping is not subject to water hammer, is not susceptible to erosion / corrosion (stainless steel), and it is resistant to Intergranular Stress Corrosion Cracking (IGSCC). While Type 304 stainless steel is susceptible to IGSCC, a combination ofin-place mechanical stress improvement and induction heating stress improvement processes applied to the field welds and solution annealing of shop welds have been applied to mitigate IGSCC for these locations.
Fermi 2 is a unique BWR in the area ofIGSCC mitigation and prevention in that IGSCC protective measures were implemented either during or before the first refueling outage.
These protective measures conform with the NRC guidance provided in NUREG-0313, Rev. 2, Technical Report On hiaterial Selection And Processing Guidelines For BWR Coolant Pressure Boundary Piping. During construction, shop fabricated sections and included welds were solution annealed. In the early 1980's, prior to plant operation, Induction Heating Stress Improvement (IHSI) was performed on the 79 welds that could be treated successfully by this process. Fermi implemented the Mechanical Stress improvement Process (hiSIP) in 1989 during Refuel Outage One (RF01) on the last 27 welds that were defined by the NRC as being potentially susceptible to IGSCC. All IGSCC susceptible welds on Recirculation Loop piping have had some form ofIGSCC mitigating treatment.
Welds associated with the three remaining break locations are within the scope of the Fermi 2 Insarvice Inspection (ISI) program. During the later stages of construction, presersice UT examination was performed on all Class I welds in accordance with ash 1E Section XI, IWB 2200. The welds selected for insersice examination included those welds believed to be most susceptible to cracking, i.e., high stress, moderate stress, dissimilar metal, high cumulative usage and additional random selections to achieve a 25%
selection rate. The ISI program was revised to include the augmented inspections per
April 4,1997 NRC-97-0043 l
Page 6 j
NUREG-0313, Rev. 2. Additional welds were included per GL 88-01,"NRC Position On IGSCC In BWR Austenitic Stainless Steel Piping."
Inservice examination of piping welds has been performed in accordance with ASME Section XI requirements. Additionally, examination procedures and personnel have been qualified as required by NUREG-0313 Rev. 2. From RF01 through RF04 examiner and procedure capabilities to detect IGSCC were demonstrated in accordance with the three party agreement at EPRL For RF05, personnel performing ultrasonic examinations of IGSCC susceptible welds were additionally qualified through the Utility Performance Demonstration Initiative (PDI). Examinations were performed using the PDI generic procedure, PDI-UT-2.
No service related defects have been detected during nondestmetive examinations of piping welds at Fermi 2 Nuclear Power Plant. Five of the nine welds comprising the three remaining break locations have been examined on one or more occasions since plant startup. No IGSCC has been detected to date in any piping welds. Additionally, no evidence of fatigue cracking has been detected in any piping system welds. Further, Fermi 2 intends to implement Hydrogen Water Chemistry during the upcoming operating cycle.
Results of the SIA fracture mechanics evaluation (Attachment 1) demonstrate that leakage in excess of one gpm would occur in a leakage crack at any of the remaining break locations when the circumferential crack approaches half of the critical crack length (Table 5-2 of Attachment 3). The Fermi 2 Technical Specification required leakage detection systems are consistent with the recommendations of Regulatory Guide 1.45, Reactor I
Coolant Pressure Boundary Leakage Detection Systems, and include equipment capable of detecting leakage of one gpm over a one hour interval.
i The results of the SIA fracture mechanics evaluation further demonstrate that the time for flaw propagation from one-half to three-quarters critical flaw size is over 18 months (a nominal operating cycle). This allows ample opportunity to detect and take action to correct unidentified leaks both during operation and refueling outages. In addition, plant Technical Specification (TS) 3.4.3.2 requires an orderly plant shutdown be promptly performed based on indications of abnormal Reactor Coolant System (RCS) leakage or changes in leakage. This provides a significant measure of protection from RCS piping failure by limiting the time between initial leak detection and depressurization of the RCS, supporting the conclusion that three remaining postulated HELBs are not credible. The predicted leakage at the three-quarter critical flaw size for eight of the nine welds associated with the remaining break locations exceeds the five gpm Technical Specification limit on unidentified leakage where plant shutdawn is required. The predicted leakage at the ninth weld is slightly less than the five gpm TS limit and significantly exceeds the normal range of unidentified leakage experienced during plant operation. Given the normal range of unidentified leakage, it is unlikely that plant operation would be continued without investigation ifleakage were to progress to the level of the predicted leakage associated with one-half the critical flaw size for the ninth weld.
A
April 4,1997 NRC-97-0043 Page 7 Based on the preceding discussion, Detroit Edison has reached the following conclusions:
Impact of HELBs inside containment on the EECWS are not part of the licensing basis for Fermi 2. Evaluation of the postulated breaks based on current NRC guidance contained in.
GL 87-11 permit elimination of twenty-eight of thirty-six postulated breaks from further consideration. The application of the definition of terminal end provided in ANSI /ANS 58.2 eliminates five additional postulated breaks. The remaining three postulated breaks were determined to be incredible based on fracture mechanics analysis, with leakage capable of being detected well in advance of flaw propagation to the critical flaw size.
If you have any questions, please contact Peter W. Smith, Director, Nuclear Licensing, at (313) 586-4097.
Sincerely, D
/ der cc: A. B. Beach G. A. Harris M. J. Jordan A. J. Kugler
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April 4,1997 L NRC-97-0043
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ATTACHMENT 1 PIPE BREAK ID TABLE
TaWe1 Es6eation of Postulated Pipe Ereak Locations Inside Containsment i
Systems With Break Analysis Allowables Pipe Stress (PSI)-
Resneve Break?
Postmissed Break ID Nede No Ref. No.
2.4Se 3.8Sm EQ.18 <2.4Se <3.SSm EQ.12 <2.4 San EQ.13 <2.4Se CUF
<S.1 Y/N Reesee Hi h/ Moderate Ener8y Core Spray Line B CB6 55 1
42428 53100 34689 Y
Y N/A Y
N/A Y
0.005 Y Y
8 Core Spray Lme B CB5C 60A 1
42428 53100 48680 N
Y N/A Y
N/A Y
0.011 Y Y Intermediate Break
. Feedwater Line A WA15C 15A 2
46680 58350 37725 Y
Y N/A Y
N/A Y
0.035 Y Y Intermediate Break Feedwater Line A walla 205A 2
46680 58350 4 % 97 N
Y 35117 Y
18578 Y
0.01 Y Y Intennediate Break i
Feedwater Line A WAl2A 215A 2
46680 $8350 40752 Y
Y N/A Y
N/A Y
0.004 Y Y Intermediate Break Feedwater Line A WAl2C 215B 2
46680 58350 48227 N
Y 30291 Y
21790 Y
0.007 Y Y Intermediate Break Feedwater Line B WB15C 15A 3
46680 58350 38548 Y
Y N/A Y
N/A Y
0.034 Y Y Intermediate Break Feedwater Line B WB15A ISB 3
46680 58350 24587 Y
Y N/A Y
N/A Y
0.003 Y Y Intermediate Break Feedwater Line B WB11A 205A 3
46680 58350 49313 N
Y 34351 Y
18809 Y
0.012 Y Y Intermediate Break Feedwater Line B WB1IC 205B 3
46680 58350 42929 Y
Y N/A Y
N/A Y
0.007 Y Y Intermediate Break Feedwater Line B WB12C 215B 3
46680 58350 46882 N
Y 2 % 29 Y
19891 Y
0.008 Y Y Intermediate Break Main Steam Line A SA7C 041F 4
45960 57450 40526 Y
Y I4507 Y
26847 Y
0.01 Y Y Intermediate Break M:in Steam Line B SB6A 050N 5
45960 57450 35175 Y
Y 12979 Y
26898 Y
0.01 Y Y Intermediate Break i
M in Steam Line B SB6C 050F 5
45960 57450 36374 Y
Y I4701 Y
28074 Y
0.01 Y Y Intermediate Break i
MIin Steam Line C SC6C 50F 6
45960 57450 36586 Y
Y 14148 Y
27929 Y
0.01 Y Y Intennediate Break M:in Steam Line D SD70 031F 7
45960 57450 37253 Y
Y 12850 Y
27573 Y
0.01 Y Y Intermediate Break j
MS Drains MDBS 495 11 42552 53190 9880 Y
Y 24 Y
3016 Y
0 Y Y Intermediate Break
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IIPCI
'rdC 418F 4
42480 53100 21550 Y
Y 3010 Y
15125 Y
0 Y Y Intermediate Break HPCI HP2A OBN 4
42480 53100 21589 Y
Y 2461 Y
14937 Y
0 Y Y Intermediate Break
[
HPCI HP4 440 4
42480 53100 18189 Y
Y 1578 Y
12423 Y
0 Y Y Intermediate Break HPCI HP3A 426N 4
42480 53100 23470 Y
Y 6846 Y
I4630 Y
0 Y Y Intermediate Break i
Rs.1C RPI 39 5
35360 44200 51%8 N
N 22735 Y
30112 Y
0.05 Y Y Intermediate Break RCIC RP3A 649N 3
35360 44200 31150 Y
Y 12649 Y
13701 Y
0.01 Y Y Intermediate Break RR Line A RA20A 378 8
40020 50025 22222 Y
Y 8081 Y
17800 Y
0 Y N Terminal End RR Line B RBIA 001 10 40020 50025 20701 Y
Y 3134 Y
21139 Y
0 Y N Terminal End 1 of 2
- Table 1 Evaluation of Postulated Pipe Break Locations Inside Contaiemment Systeem Wkh Break Analysis AllowsMes Pipe Stress (PSI)
Neuseve Breek?
- Postelmeed Break ID Nede No Ref. No.
2.4Se 3.0 Sea EQ. It <2.4Sm <3.8Se EQ.12 <2.4 San EQ.13 <2.4Se CUF
<t.1 Y/N Reesee RR Lbe B RB3A 017 10 40020 50025 20069 Y
Y 981 Y-201%
Y 0 Y Y intermediate Break RR Lee E RB3C 019 10 40020 '50025 20069 Y
Y 837 Y
20163 Y
0 Y Y Intermediate Break RR Line B RB28A 318 10 40020 50025 20552 Y
Y 6411 Y
17836 Y
0 Y N Termlmel End RR Line B RB3L 502 10 40020 50025 32181 Y
Y
'2849 Y
29868 Y
0.031 Y Y latermediate Break ~
RHR Retwn (A)
RRA3C 514 8
42480 53100 20587 Y
Y 3569 Y
20432 Y
' O.01 Y.
Y High/ModersteEnergy RHR Retum (B)
.RRB3C 614 10 42480 53100 19468 Y
Y 2853 Y
18346
~Y 0.01 Y Y liigh/ModerateLier8y RHR Supply RSI 502 10 42480 53100 32181 Y
Y 2849 Y
~29868 Y
0.01 Y
.Y Intermediate Break RHRSqly RS4 520 10 42480 53100 18155 Y
Y 2453 Y
17230 Y
0.01 Y Y Egh/ Moderate Energy RHR Supply -
RS5 564 10 42480 53100 16699 Y
Y 925 Y
17189 Y
0.008 Y Y High/ModersteEnerEy RWCU CU13 041 9
42480 53100 40597 Y
Y 13990 Y
182 %
Y-0.016 Y-Y Intermediate Break '
.RWCU
- CUl2C 099F 9
42480 53100 30648 Y
Y 18209 Y
15050 Y
0.001 Y Y Intermediate Break m u er7:
3 Total No, of Breaks 36 No.of Postulated Breaks %hich May Be Removed 33 No.of Postulated Breaks Remaining 3 (Terminal End) 2 of 2
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April 4,1997 NRC-97-0043 6
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