NRC Generic Letter 1986-01
JAN JN3 IMTO ALL BWR APPLICANTS AND LICENSEESGentlemen:
SUBJECT: SAFETY CONCERNS ASSOCIATED WITH PIPE BREAKS INTHE BWR SCRAM SYSTEM (GENERIC LETTER 86-01)On April 10, 1981, the NRC staff sent a generic letter to all BWRapplicants and licensees requesting them to provide their plant specificresponses addressing the concerns identified in Draft NUREG-0785, "SafetyConcerns Associated with Pipe Breaks in the BWR Scram System." OnAugust 31, 1981 the staff sent Generic Letters 81-34 to BWR licensees and81-35 to BWR license applicants, wherein it was stated that plant specificresponses conforming to the guidance contained in NUREG-0803, "Generic SERRegarding Integrity of BWR Scram System Piping" would satisfy the requestfor information in the April 10, 1981 letter. In Generic Letter 81-35, thestaff further stated that pipe failure in the BWR scram system is not asafety issue for the Mark III containment designs.The NUREG-0803 guidelines essentially addressed the need for improvement inprocedures, periodic inservice inspection and surveillance for the scramdischarge volume (SDV) system, and environmental qualification for essentialequipment needed for mitigation of the consequences of staff-postulated pipefailures in the SDV piping system. These guidelines were developed toaddress the consequences of a postulated leakage crack in the SDV pipingand resulting large leakage (up to 550 gpm) downstream of the systemisolation valves. Such a leak would have the potential to causedegradation of the needed mitigation equipment. At the time they weredeveloped, these conservative assumptions and guidelines were based on 1)lack of generically identifiable failure mechanisms for the SOV pipingsystem, 2) scarcity of available data for the system including uncertaintyregarding the operability of mitigation equipment in a possibly harshenvironment, and 3) lack of adequate guidance in the BWR Owners Group(BWROG) Emergency Procedure Guidelines (EPGs) for handling reactor buildingand environmental problems that may arise as a consequence of such anaccident.Based on its review of BWROG and General Electric Company (GE) suppliedgeneric information (NEDO-22209, BWROG-8325 and BWROG-8420) and staffgeneric analyses of the SDV piping system integrity, the staff hasconcluded that in accordance with Branch Technical Position (BTP) MEB 3-1,Position B.2.e in Standard Review Plan 3.6.2, through-wall leakage cracksinstead of breaks may be postulated in the piping of those fluid systemsthat qualify as high-energy fluid systems (temperature greater than 200degrees F or pressure greater than 275 psig) only for short operationalperiods (about 2 percent of the time) but qualify as moderate energy fluidsystems (temperature less than or equal to 200 degrees F and pressure lessthan or equal to 275 psig) for the major operational period. Furthermore,8009029
-2 --2- JAN1 e 3 198the staff has concluded that, based on its classification and low stressthreshold, the SDV piping system satisfies BTP MEB 3-1, Position B.2.c(1)in that a through-wall leakage crack need not be postulated.Since the SDV piping system fulfills the above criteria, breaks andthrough-wall cracks in the SDV piping need not be postulated. In addition,the staff has concluded that, even if a staff-postulated through-wall flawis initially present in the SDV system, it will grow negligibly and will notpropagate into a break under the staff defined piping loads. Further,leakage from such a flaw will be small (less than or equal to about 5 gpm)and, therefore, a harsh environment over large areas of the reactor buildingwhich could affect redundant safety-related mitigating equipment will notresult. Thus, the potentially exposed safety-related equipment need not bequalified for operation in a harsh environment associated with an SDV break.The staff has also concluded that the revised BWROG Emergency ProcedureGuidelines for secondary containment control (NEDO-24934), together withnormal plant procedures and the proposed periodic visual verification of thescram system piping integrity (BWROG-8420), provide sufficient measures fordetecting and mitigating the consequences of leakage which may occur in theSDV piping system. The design basis of the SDV piping system hasconsidered transient forces resulting from the worst case control rod drive(CRD) system actuation. Although water hammer has been analyticallypostulated and hydraulic instabilities have been experienced in the CRDsystem, no events have been experienced of a severity significant enough toconstitute a water hammer. Therefore, water hammer is not considered acontributing factor in potential SDV pipe breaks.Accordingly, this completes our review of the safety concerns associatedwith pipe breaks in the BWR scram system. No OMB clearance is requiredsince no information is requested.This information is being provided to BWR applicants and licensees withMark III Containments for informational purposes only.xs1 isM 1'A. ,:, a .godtrRobert M. Bernero, DirectorDivision of BWR Licensing
Enclosure:
Staff Safety EvaluationDISTRIBUTIONGeneric LetterDBL:PD#2/ D: t DBL:DIR M6MGrotenhu :ajs D1ie \\ RBernero12/)( /85 12t 1/85 1t/ 3 /85 EnclosureGENERIC SAFETY EVALUATION REPORT REGARDING INTEGRITYOF BWR SCRAM DISCHARGE PIPING SYSTEM1.0 Introduction1.1 NUREG-0803, Bases and AssumptionsDuring the investigation of the Browns Ferry Unit 3 control rod partialinsertion event on June 28, 1980, the NRC staff identified potentialsafety concerns associated with postulated pipe failures in BWR scramdischarge piping systems. These concerns were documented in a draftreport entitled, Stfety Concerns Associated W1th Pipe Breaks in the BWRScram System" (NUREG-0785) published on April 3, 1981. Subsequently,the NRC sent a Generic Letter (letter, D. G. Elsenhut, April 10, 1981)to all BWR licensees requesting them to address the concerns identifiedin NUREG-0785. In response to this letter, and after a conference withthe NRC staff, the General Electric Company (GE) provided generic topicalreport NEDO-24342, "GE Evaluation in Response to NRC Request RegardingBWR Scram System Pipe Breaks" by letter dated April 36, 1981. Thisgeneric submittal was reviewed by a multidisciplinary staff group.-During the course of the review, the staff issued a report entitled*Generic Safety Evaluation Report Regarding Integrity of BWR Scram SystemPiping" (NUREG-0803) for resolving this matter. In this report, thestaff provided specific guidelines and criteria to BWR licensees whichwas meant to ensure (1) the integrity of the scram discharge volume(SDV) piping system, (2) the leak detection and mitigation capabilitiesfor the staff postulated piping failure in the system and (3) thequalification of essential equipment (needed for detection and mitigation)exposed to the expected environment resulting from the postulated failure.Following the publication of NUREG-0803, the NRC sent Generic Letters81-34 aid 81-35 dated August 31, 1981 (NRC letters, D. G. E£senhut)to all the BWR licensees and applicants requesting them to provide theirpant specific responses to the NUREG-0803 guidelines within a stipulatedperiod. The bases for these guidelines are summarized below:
2At the tfiue the NUREG-0803 guidelines were formalized, the staff found1. lack of generically identifiable failure mechanisms for the SDVpiping system and scarcity of available data for the system;2. lack of adequate information on the capability to detect a postulatedfailure in the system piping in a timely manner to permit correctiveoperator actions to mitigate the consequences of such a failure;3. lack of adequate guidance in the BWR Owners' Group (BWRDG) emergencyprocedure guidelines (EPG) for handling reactor building and environ-mental problems that may arise as a consequence of an accident; and4. uncertainty regarding the operability of essential equipment neededfor mitigation since it may not be environmentally qualified for thelocal conditions that may exist under the postulated SDY pipingfailure.-NUREG-0803 recognized that the safety concerns that stem from possible con-sequences in the reactor building resulting from a postulated leakage crackin the SDV header piping will not be applicable to BWRs with Mark III con-tainment designs, because the safety-related mitigation equipment is notlocated in the reactor building, but will be applicable to Mark I and IIcontainment designs. The staff, however, concluded that the leak rate fromthe postulated pipe failure in the SDY piping system assumed by GE inNED0-24342 was reasonable. Specific guidelines were developed which con-servatively assumed that a leakage crack (equivalent break area of about0.007 ft2) occurs in the SDV header piping system following a scram butbefore the scram is reset and that it results in fluid leakage at a rateof up to 550 gpm downstream of the system isolation valves. In developingthese guidelines, the staff additionally considered the potential forthis large leakage to cause degradation of the mitigation equipment,particularly if the piping failure can not be isolate In NUREG-080OsJthe staff used an estimated SDV piping failure frequency of 104per plant year in the generic risk assessment analysis.- Given the lack ofgenerically Identifiable failure mechanisms for the SDV piping failure at thattime when NUREG-0803 was prepared, the staff regarded the estimated frequencyto be extremely conservative. However, the staff acknowledged the uncertaintyof the estimate because of the assumptions that were made and the scarcityof the available data at that time.Using the above value and assuming that the required mitigation equipment willbe operable, a core melt frequency of 166 per plant year was estimated. It wasconcluded that the sequence of events following the postulated SDV pipingfailure analyzed in the report will not be a dominant contributor to core melt,provided the required mitigation equipment is not degraded by the adverse SDVpipe failure environment.1.2 BWROG Response To NUREG-0803 Concerns, NEDO-22209In response to NRC generic letters 81-34 and 81-35 relating to theconcerns identified in MUREG-0803, the BWROG submitted a GE topical reportentitled OAnalysis of Scram Discharge Volume System Piping Integrity(NEDO-22209) by letter dated August 23, 1982 and a correction letter datedOctober 26, 1982 (BWROG-B259). In this report, based on their analysis ofSDV piping data for 15 BWRs, the BWROG calculated a probability of 3 x 10per plant year for an unisolatable loss of SDV piping integrity and aprobability of 4 x On events per reactor year for core damage resultingfrom such a loss of integrity. The 8WROG contended that, therefore, therevas no need for qualifying the equipment required to detect and/ormitigate the consequences of such a low probability event. Based on areview of the above submittals (NEDO-22209; BWROG-8259), the staffrequested additional information from the BWROG in a number of areas(NRC, notes of meeting held on February 2, 1983). Included in thisrequest was a concern regarding the effect of a seismic event on SDYpiping integrity and its effect on the SDV pipe failure probability
- .4indicated in NEDO-22209. By letters dated January 28, 1983and June 29, 1983 (BWROG-8303; BWROG-8325), the BWROG provided therequested responses. Specifically, in their June 29, 1983 response, theBWROG determined that the probability of SDV pipe failure from seismicloads was of approximately the same magnitude as the probability of pipefailure from other sources, they had calculated earlier. Consequently,the BWROG concluded that the statements given in NEDO-22209 were stillvalid.1.3 Deterministic Fracture Mechanics EvaluationBased on the staff's review of the BWRDG response (NEDO-22209; BWROG-8259;_ BWRO G-8303; BWROG-8325), it was concluded that an expeditious resolutionof the concerns relating to the SDV piping system integrity required amore detailed consideration of the applicable break mechanisms than thatwhich could be obtained from a probabilistic analysis. Therefore, byletter dated July 25, 1983 (NRC letter, D. 6. Eisenhut to T. J. Dente),the staff requested the BWROG to provide information in connection with adeterministic fracture mechanics evaluation of the SDV piping integrityalong with a discussion of the associated realistic leak rate, leakdetection and mitigation capabilities. Specifically, the staff requestedthe BWROG to provide the following information:1. Perform a fracture mechanics evaluation which is bounding in termsof leak rate, loading conditions, and material properties for thescram system piping. As a minimum, the following specific con-ditions should be met:a. the postulated through-wall flaw size (length) shall be equalto or greater than twice the pipe wall thickness;b. the piping loads applied for flaw stability analysis shall bethose associated with nonral plant conditions in combination*with the Operating Basis Earthquake (OBE); however, for flawleakage calculations, only piping loads associated with normalplant conditions shall be used;_
5c. the flaw should be postulated to be located at the highest stressedl'"gion in the material with the most limiting properties, i.e.,base materials or weld material as applicable;d. valid material test data should be used in the evaluation; ande. the leak rate should be calculated for the above postulatedthrough-wall flaw. In addition, a comparison of the calculatedleak rate with experimental results and/or operational experienceshould be provided.2. Leakage detection capability should be demonstrated to be sufficientto provide adequate margin for detecting leakage from the postulatedcircumferential through-wall flaw. A discussion should be providedregarding the capability to detect the leak In the scram system pipingand take the necessary action to terminate the leakage prior toexceeding the environmental qualification envelope of any affectedessential equipment, i.e., necessary for prompt depressurizationand long-term cooling. This should include consideration ofpotentially harsh environments which may limit access for localmanual actions. Consideration should be given to making appropriatemodifications to existing procedures for prcmpt depressurization inorder to assure that they would cover the above postulated leakagecondition and the available plant-specific leak detection devices.3. Provide a discussion regarding the expected radiation field andrctact exposure level at the scram system piping as it may affectroutine tests and inspection.It response to the above letter (NRC letter, D. 6. Eisenhut toT. J. Dente), the BIROG provided submittal BWROG-8335 by letterdated November 18, 1983. The submittal, included a deterministicfracture mechanics evaluation of the BWR SDV piping system assumingthe staff postulated circumferential through-wall flaw, and pipingloads. On the basis of the analysis, the BWROG concluded that the.'.' ----.-.. -.. .. -.- -- ... .
fiflaw will remain stable and thus will not propagate into a break.The BWROG further concluded that the leak from the flaw will besufficiently low (' 3.5 gpm) that it will not pose a threat to theenvironmental qualification envelope for equipment in the reactorbuildin needed for detection of the leak and mitigation of itsconsequences.1.4 Subsequent DevelopmentsBased on the staff's review of BWROG-8335 and other submittals (NEDO-22209;BWROG-8259; BWROG-8303; BWROG-8325), it was concluded that sufficientgeneric information on the SDV piping system had been provided to warranta new approach to the SDV pipe break concern rather than applying thecriteria of HUREG-0803 for resolving this generic issue. Specifically.the staff concluded that the NUREG-0803 postulated through-wall leakagecrack in the SDV piping system should be re-evaluated against the currentlicensing pipe break criteria of SRP Sections 3.6.1 and 3.6.2 (NUREG-0800;see particularly BTP MEB 3-1, Position B.2.c.(1)J. The staff furtherconcluded that if such a re-evaluation confirms that the SDV piping stresslevels are low enough to preclude the need for postulating the through-wallleakage crack, then the SDV pipe crack and its resulting consequences asidentified in NUREG-0803 need not be addressed. Additional informationwas therefore requested on certain aspects of the BWROG November 18, 1983submittal in a conference call on January 19, 1984 and during a sub-sequent meeting with GE and BWROG representatives on February 23, 1984(NRC, notes of meeting). In the meeting, GE provided responses to staffquestions raised in the conference call. Additionally, the BWROGsubmitted a revision dated May 10, 1984 (BWROG-8420) to their originalsubmittal of November 18, 1983 (BWROG-8335), wherein they included therequested information. Specifically, in the May 10, 1984 submittal(BWROG-8420), the WROG (1) outlined their generic prograr for periodicvisual observation of the SDV piping to check for leaks and (2) outlinedthe generic emergency procedure guidelines provisions given in the GETopical Report NEDO-24934 for handling problems arising from pipe breaksoutside the containment such as SDV pipe break, and (3) provided acalculation of leak rates associated with the staff postulated flaw for 7the maximum and minimum SDV header pipe sizes. Furthermore, in the submittalthe BWROG reiterated their earlier conclusions regarding the unlikelihoodof a leek-age crack in the SDV piping system, the stability of the staffpostulated flaw should It be initially present, and its minimal environ-mental consequences on essential equipment exposed to the leakenvironment should the leak occur.Based on a review of the May 10, 1984 submittal (BWROG-8420), and othersubmittals mentioned above and their own analysis of the SDV piping systemintegrity, the staff has determined that the earlier bases for the con-servative assumptions identified in NUREG-0803 are no longer applicable.It was also determined that a leakage crack in the SDV piping systemneed not be postulated since it has confirmed that the bounding stressvalues ident1fledaby the BWROG satisfy the low stress criteria formoderate energy fluid piping systems given in SRP Section 3.6.2(NUREG-0800).The staff's evaluation of the integrity of the system is given below.Adopting a Odefense-in-depthm strategy, this evaluation addresses(1) the acceptability of SDV piping stress levels against theSRP criteria for postulated crack locations and the conclusion regardingno break or no through-wall leakage crack determination for the SDV pipe,(2) the acceptability of the BWROG deterministic fracture mechanics evalua-tion of the SDV piping system integrity along with the probabilisticfracture mechanics analysis and the conclusion regarding the stability ofthe staff postulated flaw should it be initially present, and (3) theacceptability of the detection and mitigation capability should the leakoccur.-, _ _ -- -. , .-_ -- -_ -._ ._ _
82.0 Evaluation Of The SDV Piping System Integrity2.1 Evaluatfiin Of The Applicability Of Plping Failure Postulation In TheSDV Piping SystemThis section deals with the staff's evaluation of the applicability of apostulated piping failure in the SDV piping system against current SRPlicensing criteria for postulating pipe failures as stated in SRP Section3.6.2. specifically BTP MEB 3-1 Positions B.2.e and B.2.c.(l) andreiterated in SRP Section 3.6.1 (NUREG-0800).General Design Criterion (GDC) 4 of 10 CFR 50, Appendix A. requires thatstructures, components and systems important to safety be designed toaccommodate the effects of postulated accidents, including appropriateprotection against the dynamic effects of postulated pipe rupture.Specific guidance for meeting GDC 4 with regard to pipe rupture is pro-vided in SRP Sections 3.6.1 and 3.6.2 (NUREG-0800)and the attached BranchTechnical Positions BTP ASB 3-1 and BTP MEB 3-1. This guidance includesamong other things, the conditions under which a piping failure need notbe postulated in a piping system.BTP MEB 3-1, Position 8.2.e permits that through-wall leakage cracksinstead of breaks may be postulated in the piping of those fluid systemsthat qualify as high-energy fluid systems (temperature'200°F or pressure> 275 psig) only for short operational periods (about two percent of thetime) but qualify as moderate energy fluid systems (temperature s 200"Fand pressure c275 psig) for the aJor operational period. Since the SDVpiping system fulfills the above criteria, breaks in the SDV piping neednot be postulated.The SDV piping is normally classified as Class 2. However, some plantshave Class I SDV piping. BTP RES 3-1, Position B.2.c.(1) permits that thethrough-wall leakage cracks in piping systems need not be postulated wherethe maximum stres intensity is less than 1.2 Sm for Class I piping and themaximum stress is less than 0.4 (1.2 Sh + SA) for Class 2 piping. Sm is- the design stress intensity for Class 1 piping material as defined inNB-3600 of the ASME Code,Section III and the values for S. are tabulatedin ASME Code, Section 111, Division 1, Appendix I Tables. Sh and S. arethe allowable stress at maximum (hot) temperature, and allowable stressrange fMr thermal expansion, respectively, as defined in Article NC-3600of the ASME Code, Section 111.The piping materials used in the SDV piping are SA 106 6r. B and C carbonsteel and SA 358 304L stainless steel (BWROG-8420). The maximum stressintensity (including SSE loads) for Class 1 design piping is 15 ksi(BWROG-8420). The design stress intensities (S.) for Class I piping are20 Wsi, 22.9 ksi and 15.8 ksi for SA 106 Gr. B. SA 106 6r. C and SA 358304L at 400OF respectively (ASME Code Section III, Division 1, Appendix I,Tables 1-1.1 and 1-1.2). Thus it can be seen that Class 1 piping in theSDV piping system satisfies the criteria for not postulating athrough-wall leakage crack as stated in BTP MEB 3-l, Position B.2.c.(1).A study of different SDV Class 2 pipe sizes found that the maximumstresses (including SSE loads) were 5.14 ksi, 10.34 ksi and 7.3 ksi for 4 inch,12 inch and 24 inch pipes respectively (BWROG-8420). The mechanicalproperties of the Class 2 SDV piping materials are as follows:Material Sh(6D0OF) SA(1DOO cycles) 0.4(1.2 Sh* SAL--- ksi ksi ksiSA 106 6r. B. 15 75 37.2SA 106 Sr. CCarbon steelSA 358 304L 14 105 48.7Stainless steelThus, it follows that Class 2 piping in the SDV piping system alsosatisfies the criteria for not postulating a through-wall leakagecrack as stated in BTP MEB 3-1, Position B.2.c(1). The staff,therefore, concludes that no break or through-wall leakage crack need bepostulated in the SDV pipin .2 Fracture Mechanics Evaluation Of SDV Piping IntegrityThis section deals with the staff's evaluation of the probabilistic anddeterministic fracture mechanics evaluation of the SDY piping integrityprovided; by GE in NEDO-22209 and the BWROG-8420 to support the conclusionthat breaks in the SDY piping system are very unlikely.In NEDO-22209, the loss of piping integrity in the SDV piping system wascalculated based on a consideration of pipe length, scram frequency, andvent and drain valve reliability. Conservative values for the key inputswere selected based on SWR plant data and on generic reliability data.Pipe break probabilities were estimated based on the experience data usedin the NUREG-75/014 (formerly WASH-1400), and on a fracture mechanicsanalysis of the piping system.The results of the above analyses indicated that the probability of anunisolatable loss of scram system piping integrity for an average plant is3 x 1 per plant year. The probability of core damage resulting from aloss of SDV pipe integrity was determined to be approximately 4 x 1011events per reactor year. This is significantly below the proposed NRCsafety goal for core melt events of Ab4 per plant year. Consequently, GEconcluded that the probability of a loss of scram system piping integrityleading to core damage is sufficiently low to preclude the necessity ofqualification or design modifications of equipment required to detectand/or mitigate the consequences of such an event.The staff verified the fracture mechanics equation utilized by GE for theanalysis in KEDO-22209 by comparing the influence function parameters inthe equation with those used by the staff for pressurized thermal shockanalyses of reactor vessels. It was concluded that there was reasonablygood agreement. The staff analyses results had been previously found toagree with analyses performed by ORML. The results reported in NEDO-22209are also consistent with those obtained by Lawrence Livermore Laboratoryin RUREG/CR-3660 for reactor coolant piping. Both reports concludedthat on the basis of probabilistic fracture mechanics, the likelihood ofpipe breaks in the SDV piping system is lo As stated earlier, the staff requested the BWROG to perform adeterministic fracture mechanics analysis for the SDV piping system (NRC,letter from D. G. Eisenhut to T. J. Dente), especially for the largerdiameter pipes (NRC, notes of meeting held February 23, 1984) to assurethat detectable leaks would occur prior to pipe rupture and also tosupport the probabilistic fracture mechanics conclusions given inNEDO-22209 (refer to Section 1.3 of this report for more specificdiscussion regarding the request).In response to the above request, the BWROG performed a genericdeterministic fracture mechanics evaluation of the SDV piping (BWROG-8420)for the staff postulated through-wall flaw in the piping to bound theloading conditions, material properties for the system, and associatedleak rates using the SDV piping system data provided by the participatingutilities. BWROG-8420 indicated that the response to the staff's requestrelating to the expected radiation field and contact exposure level at thescram system piping should be provided by each of the participatingutilities.The deterministic fracture mechanics procedures used by the BWROG(BWROG-8420) to respond to the staff request are based on linear elasticmethods and limit load analyses. This methodology is well documented inthe literature and has been benchmarked against flawed pipe experiments.Three areas of uncertainty exist in evaluating the results of theanalyses:_.._...
121. piping loads,2. material properties, and3. accuracy of evaluation methods.By selecting stresses which bound specific data for all the facilitiesincluded in the study (BWROG-8420), the BWROG has addressed the uncer-tainty of loads. Material property data were also conservatively selectedin that the BSWROG assumed all materials to be SA 106 Grade B and C whichhave limiting toughness properties. Actual materials used in scram dis-charge systems also include stainless steel which is a tougher material.Limit-load analyses have been used routinely by the staff. A limit-loadanalysis is based on equating the forces and moments on the area of thepipe containing a real or a postulated flaw to those applied at the end ofthe pipe. A condition regarding the application of a limit-load analysisto a pipe is that the piping materials must be tough; that is, it willresist crack propagation until the limiting condition -the limiting loadis reached and then the pipe is assumed to fail. Most piping materialsused in a nuclear facility meet this condition and, in general the appliedloads are well below the limit load. This is a simple procedure, comparedto more elaborate and sophisticated fracture mechanics methods also beingused, and therefore its limitations must be recognized. The ratio ofcritical through-wall crack-length to the postulated crack lengthsuggested by the staff was found by the BWROG to range from 2.45 for a 3/4inch diameter pipe up to greater than 8 for pipes 8 inches in diameter andlarger. Thus, the staff concludes that there is adequate margin to accountfor analytical uncertainties in the procedures used.For the reasons stated above, the staff concludes that the deterministicfracture mechanics analyses performed by the BiRDG (BWRDG-8420) were donein a conservative manner and support the earlier conclusions reached by GEon scram discharge piping system integrity by probabilistic fracturemechanics analysis in REDO-22209. The staff concludes therefore, thatthe bouiding values for loading conditions and material properties givenIn the report (BWROG-8420) are reasonable. The staff further concludes 13that there is reasonable assurance that the staff postulated flaw, shouldit be initially present in the SDV piping system, will grow negligiblyand will not propagate into a break under the staff defined piping loads(refer to Section 1.3 of this SER).2.3 Leak Detection And Mitigation Capability 'This section deals with the staff's evaluation of the leak detection andmitigation capabilities for the SDV piping system discussed in thesubmittal BWROG-8420.As stated in Section2.1 of this report, the staff has concluded thateither a break or a through-wall leakage crack in the SDV piping systemneed not be postulated. Furthermore, the staff has also concluded thateven if the staff postulated through-wall flaw is initially present in theSDV piping system, it will not propagate into a break under the staff -defined piping loads, and that the bounding values for loading conditionsand material properties for such a postulated flaw given in BWROG-8420 arereasonable (see Section 2.2 of this report). The resulting leak rates arelow e 3.5 pgm; '5.3 gpm even conservatively assuming 400 scram cycles for-the 40-year plant life time). Therefore, the staff agrees with the BWROGthat this small leakage will not pose a threat to the environmentalqualification envelope for essential equipment needed-for detection of theleak and mitigation of its consequences. The staff's evaluation of theleak detection and mitigation capabilities for the SDV piping system givenin the submittal (BWROG-8420) are, therefore, based on the aboveconclusions.In its May 10, 1984 submittal (BrROG-8420), the SWROG states that theirrecommendations relating to leak detection will be adequate for detectingleaks resulting from the postulated flaw in the SDV piping system. They 14further state that the operator actions specified in normal plantprocedure and/or the generic emergency procedure guidelines (EPGs) forBWRs givh in NEDO-24934 will be adequate for mitigating adverse conse-quences that may result from a flaw or even a break in the SDV pipingsystem.- -Specifically, the BWROG in their submittal (BWROG-8420) recommend thefollowing relating to leak detection for the SDV piping system:1. For plants which employ the leak test criteria for Class 1 piping,leak tests and inspections shall be performed once every refuelingoutage per the criterion for such pipes contained in Section XI ofthe ASME Code.2. For plants which employ the leak test criteria for Class 2 piping,leak tests and inspections shall be performed periodically atrepeating intervals of 3, 4 and 3 years per the criterion for suchpipes contained in the ASME Code,Section XI. Additionally, a post-scram reset walkdown of the SDV piping shall be performed once perrefueling cycle as soon as possible but not more than 30 minutesfollowing the scram reset. This walkdown shall be performedspecifically to investigate evidence of leakage below the SDV headerand instrument volume by visual observation.The BWROG indicates that the walkdown suggested above would be sufficientto detect appreciable leakage from the system should 'it occur and wouldalso enhance the leak detection capability for systems that utilizeClass 2 piping.With regard to the mitigation capability, NEDO-24934 includes the EPGs forsecondary containment control, and lists the entry conditions. Includedamong these is secondary containment sump water level being above thenormal operating level. Among the objectives of these procedures, is.. .1. ... .. .. _. ._ .___ _
15protection of essential equipment needed for shutdown or mitigation orprevention of an accident. The conditions under which these emergencyprocedures will be utilized are symptomatic of potential accidents. TheEPGs encompass instruction for operators for handling problems arisingfrom pipe breaks outside the containment such as an SDV pipe break.The BIROG submittal (BWROG-8420) addresses mitigation by stating that inthe event of a break in the scram discharge piping system, the normalplant procedures will call for a reset of the scram. If, however, theaffected system cannot be isolated or the isolation proves ineffective inmitigating temperature or radiation increases, the generic emergencyoperating procedures will call for rapid depressurization of the reactorto be accomplished by safety valve discharge into the suppression pool.This will reduce the leakage into the reactor building. The submittalalso states that the generic emergency operating procedures guidelines(REDO-24934) specify the operator actions for achieving isolation andreactor depressurization should they be needed to mitigate the leak con-sequences resulting from the postulated flaw.Based on review of the above information, the staff concludes that theperiodic leak tests, inspection and post-scram reset walkdown of SDVpiping recommended by the Ek1.OG (BUROS-8420) provide adequate capabilityfor detection of leakages resulting from the staff postulated through-wallflaw in the SDV piping system. The staff also concludes that normal BWRoperating procedures and the applicable EPGs for secondary containnetcontrol and dealing with problems that may arise in the secondary cortain-sent due to leakage from systems such as the SDV piping system, given inREDO-24934 are adequate to mitigate the consequences of leakage resultingfrom the staff postulated through-wall flaw in the SDV piping syste *63.0 ConclusionBased on review of the GE and BWROG submittals regarding SDV piping systemintegrity for BWR Mark I and Mark 2 containment designs and independentanalyses of the system integrity, the staff concludes the following:1. The SDV piping satisfies the criteria for not postulatingeither a break or a through-wall leakage crack as stated in SRPSection 3.6.2 [NUREG-08000 BTP MEB 3-1, Position 8.2.e andB.2.c.(l)] and reiterated in SRP Section 3.6.1 (NUREG-08Oe,BTP ASS 3-1, Appendix A). Therefore, a break or a through-wallleakage crack need not be postulated in the SDV piping system.2. Probabilistic fracture mechanics evaluation of the SDV piping systemintegrity performed conservatively using the BWR plant data andgeneric reliability data confirms the above conclusion, i.e., theprobability for an unisolatable loss of scram system piping integrityis very low (about 3 x 10 per plant year). The analysis alsoconfirms that the probability of core damage resulting from such aloss of integrity is very low (about 4 x 1o01 events per reactoryear). A deterministic fracture mechanics evaluation of the staffpostulated through-wall flaw in the system performed conservativelysupports the conclusion, that is, even if the staff postulatedthrough-wall flaw Is initially present in the SDV piping system, itwill grow negligibly and will not propagate into a break under thestaff defined piping loads. further, the bounding values for loadingconditions, uaterial properties, and leak rates for the SDVpiping system, given In the BWRDG submittal dealing with thedeterministic fracture mchanics evaluation (BWROG-8420) arereasonable.3. The leakage from the staff postulated through-wall flaw in the SOVpiping system will be low enough (! 3.5 gpm, ' 5.3 gpm even conser-vatively assuming 400 scram cycles for the 40-year plant life time).--
17that a harsh environment will not occur, thus precluding the need forenvironmentally qualifying leak detection and mitigation equipmentexpoSd to the leak environment. Thus, the design of the SDV pipingsystem as described in BWROG-8420 satisfies the intent andpurpose of the applicable guidelines provided in STP ASB 3-1(NUREG-0800).4. Periodic leak testing, inspections and post-scram reset walkdownsrecomnended by the BWROG (BWROG-8420), nonmal BWR operatingprocedures, and the applicable generic secondary containment controlemergency procedure guidelines (EPGs) given in HEDO-24934 areadequate to ensure the detection of a leak resulting from the staffpostulated flaw in the SDV piping system and mitigate itsconsequences. Thus, the leak detection and mitigation capabilitiesfor the SDV piping system meet the intent and purpose of theapplicable guidelines provided in BTP ASB 3-1 (NUREG-08QO).In view of the above considerations, the staff concludes that the designof the SDV piping system described in BWROG-8420 regarding its integrity,the method of verification of the integrity of the system, and the leakdetection and mitigation capabilities for the system provided inBWROG-8420 and NEDO-24934 satisfy the applicable criteria of SRP Sections3.6.1 and 3.6.2 (NVREG-0800) and, are, therefore, acceptable. The stafffurther concludes that the above approach provides sufficient defense indepth to satisfy the concerns identified in NUREG-0803 regarding thepostulated SDV pipe break.-
184.0 ReferencesBWR-Owners Group, BWROG-8259, letter from T. J. Dente to D. G.Eisenhut (NRC) dated October 26, 1982, Subject: NEDO-22209, "Analysisof Scram Discharge Volume System Piping Integrity.8--, BWROG-8303, letter from T. J. Dente to K. Eccleston (NRC) datedJanuary 28, 1983, Subject: Transmittal of Supporting Information onApplication of Scram Time Fraction to Scram Discharge Volume PipIBreak Probability as Used in KEDO-22209.--, BWROG-8325, letter from T. J. Dente to D. S. Efsenhut (NRC) datedJune 29, 1983, Subject: Transmittal of BWR Owners Group Responses toNRC Request for Additional Information on Scram Discharge PipingIntegrity.-9, BWORG-8335, letter from D. R. Helvig to D. 6. Eisenhut (NRC) datedNovember 18, 1983, Subject: Response to NRC Request for AdditionalInformation on Scram Discharge Piping..--, BWROG-8420, letter from D. R. Helwig to D. G. E£senhut (NRC) datedMay 10, 1984, Subject: Transmittal of Additional Information on ScramDischarge Pipe Breaks Requested by NRC Staff at February 23, 1984,Meeting with BWROG.General Electric Company, GE Topical Report NEDO-24342, *GE Evaluationin Response to NRC Request Regarding BWR Scram System Pipe Breaks,"April 1981.--, GE Topical Report NEDO-24934, iEmergency Procedure Guidelines,Revision 3,6 December 1982 (BMR0G-8262, dated December 22, 1982).V-a _ _ __- ---
19--, GE Topical Report NEDO-22209, "Analysis of Scram Discharge VolumePiping Integrity,. August 1982 (BWROG-8254 dated August 23, 1982).U. S. Nuclear Regulatory Commission, Generic Letters 81-34 and 8-35,wSaTety Concerns Associated With Pipe Breaks in the BWR Scram System,"August 31, 1981.-, Letter from D. G. Eisenhut to all BWR licensees, dated April 10,1981, SubJect: Safety Concerns Associated with Pipe Breaks in theBWR Scram System., Letter from D. G. Eisenhut to T. J. Dente (BSWROG) dated July 25,1983, SubJect: Safety Concerns Associated with Pipe Breaks in theBWR Scram Syst&n.--, K. Eccleston Notes of meeting held on February 2, 1983 withGeneral Electric Company and BWROG representatives, dated May 12,1983, SubJect: NUREG-0803 and BWROG Response on Integrity of BWRScram System Piping.--, K. Eccleston Notes of meeting held on February 23, 1984, by NRCwith General Electric Company and BWROG representatives, datedMarch 19, 1984, SubJect: NUREG-0803 and SWROG Response on Integrityof BWR Scram System Piping.--, NIREG-75/014 (formerly MASH-1400)o "Reactor Safety Study: ANAssessment of Risks in V. S. Cmonercial nuclear Power Plants,"December 1976.--, XUREG-078S (draft), 'Safety Concerns Associated with Pipe BreaksIn the BWR Scram System.' April 1981.--, NUREG-0800 (formerly NUREG-75/087), "Standard Review Plan forReview of Safety Analysis Reports for Nuclear Power Plants--LWREdition," July 1981 (includes Branch Technical Positions).--, NUREG-0803, "Generic Safety Evaluation Report Regarding Integrityof BWR Scram System Piping, August 1981.