ML20211F102

From kanterella
Jump to navigation Jump to search
Informs That Memo Approves Res/Dsar Taking Appropriate Actions,Within Current Resource Allocations,To Resolve Generic Safety Issue 156.6.1, Pipe Break Effects on Systems & Components Inside Containment
ML20211F102
Person / Time
Issue date: 07/16/1999
From: Thadani A
NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To: Rossi C
NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
REF-GTECI-156, REF-GTECI-NI, TASK-156, TASK-OR NUDOCS 9908300234
Download: ML20211F102 (17)
v • d • e


Text

r July 16,1999 MEM,ORANDUM TO: ChIrles E. Rossi, Dirsctor t

Divisi n of Systsms An: lysis End Regul tory Effectiveness, RES FROM:

Ashok C. Thadani, Director /s/ M. V. Federline for Thadani Office of Nuclear Regulatory Research

SUBJECT:

PRIORITIZATION OF AND TRANSFER OF RESPONSIBILITY FOR GENERIC SAFETY ISSUE 156.6.1, ' PIPE BREAK EFFECTS ON SYSTEMS AND COMPONENTS INSIDE CONTAINMENT" The prioritization of the subject issue showed that it has a HIGH priority. In accordance with Office Letter 7 procedores and the new division of responsibilities that have resulted from the recent RES reorganization, I am assigning lead for the resolution of this issue to your Division. This memorandum approves RES/DSAR taking appropriate actions, within current l

resource allocations, to resolve the issue. The prioritization of the subject issue is provided in l

The prioritization will be incorporated into NUREG-Og33, "A Prioritization of Generic Safety Issues," and is being sent to the regions, other offices, the ACRS, and the PDR, by copy of this memorandum, to allow others the opportunity to comment. Any changes as a result of l

r,omments will be coordinated with you. However, the schedule for the resolution of this issue l

should not be delayed to wait for these comments.

l In accordance with RES Office Letter 7, " Procedure for identification, Prioritization, and Tracking of the Resolution of Generic issues," the resolution of this issue will be monitored by the Generic issue Management Control System (GIMCS). The information needed for this system is indicated on the GIMCS information sheet (Attachment 2). Should you have any questions pertaining to the contents of this memorandum, please contact Ronald Emrit (415-6447). For questions pertaining to the prioritization, please contact Joel Page (415-6784).

l l

Attachments:

l

1. Prioritization Evaluation l
2. Management and Control Indicators for GIMCS cc: B. Sheron, NRR H. Miller, Region I L. Reyes, Region 11 J. Caldwell, Region lli E. Merschoff, Region IV ACRS j

PDR i

DCS 3

9908300234 990716 PDR QTECI GNI156 I

PDR i

Dettbution Reading Fne (JWC) stanch Reading Flie DOCUMENT NAME: G:WEsPAGE\\156CLoSE.699

  • See previous concurrences Te e i een et om de wnent. wimamm in su n.a: c - con weout ers.chmenvencio w. T - copy we ett.chment/enciosw.

w - wo copy l

RES wQL l I

OFFICE MES

  • l ME8
  • l DET
  • V NAME JPage hAAsytend JCreig AThadeni DATE

/

/99

/

/99

/

/99 7 Itud Y OFFICIAL RECORD COf'Y RES F66. Code:: Rty-;AO-3

p u:o l

0 UNITED STATES j

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 20555 4001

\\

p$

l July 16,1999 MEMORANDUM TO: Charles E. Rossi, Director Division of Systems Analysis and Regulatory Effectiveness, RES FROM:

Ashok C. Thadani, Director M D*D Office of Nuclear Regulatory Research

SUBJECT:

PRIORITIZATION OC AND TRANSFER OF RESPONSIBILITY FOR GENERIC SAFETY.SSUE 156.6.1, " PIPE BREAK EFFECTS ON l

SYSTEMS AND COMPONENTS INSIDE CONTAINMENT" The prioritization of the subject issue showed that it has a HIGH priority. In accordance with Office Letter 7 procedures and the new division of responsibilities that have resulted from the recent RES reorganization, I am assigning lead for the resolution of this issue to your Division. This memorandum approves RES/DSAR taking appropriate actions, within current resource allocations, to resolve the issue. The prioritization of the subject issue is provided in.

The prioritization will be incorporated into NUREG-0933, "A Prioritization of Generic Safety issues," and is being sent to the regions, other offices, the ACRS, and the PDR, by copy of this memorandum, to allow others the opportunity to comment. Any changes as a result of comments will be coordinated with you. However, the schedule for the resolution of this issue should not be delayed to wait for these comments.

In accordance with RES Office Letter 7, " Procedure for identification, Prioritization, and Tracking of the Resolution of Generic issues," the resolution of this issue will be monitored by the Generic issue Management Control System (GIMCS). The information needed for this l

system is indicated on the GIMCS information sheet (Attachment 2). Should you have any questions pertaining to the contents of this memorandum, please contact Ronald Emrit (415-6447). For questions pertaining to the prioritization, please contact Joel Page (415-6784).

Attachments:

1. Prioritization Evaluation
2. Management and Control Indicators for GIMCS l

cc: B. Sheron, NRR l

H. Miller, Region i L. Reyes, Region 11 J. Caldwell, Region lil E. Merschoff, Region IV ACRS PDR DCS ht s A D - 3

r.

?

7-7-99 Prioritization of Generic issue 156.6.1, " Pipe Break Effects on Systems and Components" plS.CRIPTION S

Historical Backaround in 1967 the AEC published draft General Design Criteria (GDCs) for comment and interim use.

Until 1972 the staff's implementation of the GDCs required consideration of pipe break effects inside containment. However, due to the lack of documented review criteria, NRC/AEC staff positions were continuatly evolving. Review unifoimity was finally developed in the early 1970s; initiated by a Note from L. Rogers to R. Fraley, " Safety Guides" dated November 9,1972, in which a Draft Safety Guide entitled " Protection Against Pipe Whip Inside Containment" was proposed. This Draft contained some of the first documented deterministic criteria that the staff had been using for several years (to varying degrees) as guidelines for selecting the locations and orientations of postulated pipe breaks inside containment, and for identifying the measures that should be taken to protect safety-related systems and equipment from the dynamic effects of such breaks. Prior to use of these deterministic criteria, the staff used non-deterministic guidelines on a plant-specific basis. This Draft Safety Guide was subsequently revised and 2

issued in May 1973 as Regulatory Guide 1.46 with the same title. The Regulatory Guide was implemented only on a forward-fit basis.

Regarding pipe break effects outside containment: in December 1972 and July 1973, the AEC issued two generic letters to all licensees and CP or OL applicants (References 1 and 2) ;

known as the "Giambusso" and "O' Leary" letters, respectively. These letters extended the pipe break concerns to outside containment, and provided deterministic criteria for break postulation and evaluation of the dynamic effects of postulated breaks. The letters requested that all recipients submit a report to the staff which summarized each plant-specific analysis of this issue. All operating reactor licensees and license applicants submitted the requested analyses in separate correspondence or updated the safety analysis report for the proposed plant to include the analysis. The staff reviewed all of these submitted analyses and prepared safety evaluations for all plants, in November 1975, the staff published SRP Sections 3.6.1 and 3.6.2 that slightly revised the two generic letters discussed above. Thus, after 1975 the specific structural and environmental effects of pipe whip, jet impingement, flooding, etc. on systems and components relied on for safe reactor shutdown were considered.

As stated above, the AEC/NRC has provided requirements to the industry regarding pipe breaks outside of containment through the issuance of the "Giambusso" and "O' Leary" generic letters. Since these requirements are applicable to all the affected plants, pipe breaks outside of containment are considered a compliance issue and have been dropped from this prioritization. By EDO direction, compliance matters are to be dealt with promptly, and not await the generic issue resolution process. Therefore the issue of pipe breaks outside of containment for the 41 affected plants was brought to the attention of NRR by separate correspondence (Reference 3). The remainder of this prioritization discusses only pipe breaks inside containment.

As a part of its plant-specific reviews between 1975 and 1981, the staff used the guidelines in Regulatory Guide 1.46 for postulated pipe breaks inside containment and SRPs 3.6.1 and 3.6.2 for outside containment. In July 1981, SRPs 3.6.1 and 3.6.2 were revised to be applicable to

. both outside and insido containment; thus, eliminating the need for furt Guide 1.46.

Between the period 1983-1987, the NRC Systematic Evaluation Program (S generalissue of pipe breaks inside and outside containment. The objective of th determine to what extent the earliest 10 plants (i.e., SEP-ll) met the licensing crite safety assessments adequate for conversion of provi term operating licenses (FTOLs). As a result of these reviews plants were req engineering evaluations, technical specification or procedural changes, and phys modifications both inside and outside containment. Regarding inside containment modifications: of the two SEP-il plants evaluated for this prioritization (one BWR and PWR), the BWR was required to modify four piping containment penetrations a wide spectrum of implementation associated with the or pipe breaks inside and outside containment.

criteria were also in a state of development. Prior to 19 variations in interpretations of that document resulted i Specifically, true physical separation of wiring to redundant components w accomplished. In 1974, Regulatory Guide 1.75 was published, clarifying the requireme A draft prioritization of this issue resulted in a MEDIUM determination and that be limited to pipe breaks inside containment since the NRC had already provide regarding outside containment pipe breaks to the industry through the issuance of the previously mentioned "Giambusso" and "O' Leary" generic letters.

However, the uncertainty in the analysis was much wider than desired for a ranking. Thus, the issue appeared to warrant additional analysis to enhance the pr In July 1994 a contract was begun with the Idaho National Engineering Laborato 1.

Review of pipe failure rate data, pipe break methodologies, and related publications to determine recommended pipe failure rates (initiating events) applicable to the affected SEP-ill plants.

2.

Review of Updated Final Safety Analysis Reports and related Safety Evaluatio Reports for SEP-il, SEP-lil, and for representative non-SEP plants to iden and prioritize potential safety concerns (i.e., accident sequences). Several pl visits /walkdowns were included as part of this review.

3.

Estimate changes to core damage frequencies for accident sequences that are determined to be of high or medium priority.

4.

Identify potential corrective actions and their estimated costs.

1 :

Based on the results of the INEL research, the enhanced prioritization is presented below.

Safety Sionificance GDC 4 is the primary regulatory requirement of concern. It requires, in part, that structures, systems and components important to safety be appropriately protected against the environmental and dynamic effects that may result from equipment failures, including the effects of pipe whipping and discharging fluids. Several possible scenarios for plants that do not have adequate protection against pipe whip were identified as a result of the research performed in support of the enhanced prioritization.

Related regulatory criteria include common cause failures, protection system independence, and the single failure criterion.

Recommended Solution issue Generic Letters to the affected plants requesting that they perform plant-specific reviews and walkdowns, identify vulnerable pipe break locstions, and inform the NRC of proposed corrective actions.

PRIORITY DETERMINATIONS Numerous scenarios of potential concern were evaluated. The following were considered important enough to be specifically identified for future consideration. All estimated frequencies and probabilities are mean values.

BWRs Case 1 (INEEL BWR Event 1h Failure of Main Steam or Feedwater Pioina Resultina in Pine Whio and Containment Imoact/ Failure. with Resultant Failure of All Safety Iniection Systems This event involves a BWR with a Mark I steel containment; 15 of the 16 affected BWRs are of this design. A DEGB of an unprotected (i.e., no pipe whip restraint or containment liner impact absorber) large reactor coolant recirculation pipe inside containment and near the containment liner might result in puncturing the liner. The resulting unisolable LOCA steam environment would be introduced into the secondary containment building, possibly disabling the ECCS equipment located there. This scenario would greatly increase the probability of core damage and potential offsite doses.

All of the affected BWRs are more than 10 years old, and most use type 304 stainless steel in the primary system piping; a material that is susceptible to IGSCC degradation it should be noted that piping of this material does not qualify for the extremely low rupture probability (Leak-Before-Break) provision of GDC 4. From NUREG-1150, the recirculation loop DEGB frequency for this material is estimated to be 1 E-4/ Reactor-Year (Rx-Yr). The fraction of BWR primary piping inside containment that is either Main Steam (MS) or Feedwater (FW) is estimated to be 4.0 E-1. The fraction of MS or RV piping that can impact the containment metal shell is estimated to be 2.5 E-1.

[

. The research performed indicates that there is considerable variation among the affected plants regarding the amount of pipe whip protection provided and the proximity of high energy lines to potential targets of concern, including redundant trains, (see Other Considerations). It was assumed that the probability of a MS or FW broken pipe rupturing the containment netal shell was 2.5 E-1.

The postulated event may also cause a common mode failure of the ECCS system since much of this equipment is located within the secondary containment and will be exposed to a harsh environment beyond its design basis, or that the ECCS piping will fail due to overpressurization of the containment annulus. In most of the affected plants, the ECCS is located in four different quadrants outside the suppression pool (torus). On the other hand, as stated above, redundant electrical power systems and initiating circuitry may not be physically separated in these older plants. Also, if the ECCS operates initially, the ECCS equipment rooms may not be fully protected from internal flooding as the water from the suppression pool flows out the broken pipe into the secondary containment. Based on these considerations the mean probability of loss of ECCS function was assumed to 8.0 E-1.

l Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 2.0 E-6 From WASH-1400, the nearest scenario to that described above is the large LOCA BWR-3 release category; involving a large LOCA and subsequent containment failure. However, in the WASH-1400 case, the containment failure results from overpressurization; not from pipe whip.

Three of the four specific BWR-3 large LOCA accident sequences have an incidence frequency of 10 E-8/Rx-Yr, and the remaining one is 10 E-7/Rx-Yr; 10 E-8/Rx-Yr was chosen as the base case for this analysis.

Case 2 (INEEL BWR Event 9h Failure of Recirculation Pipino Resultino in Pioe Whio and Containment Imoact/ Failure. with Resultant Failure of All Emeroency Core Coolino Systems This event is similar to Case 1 but involves the Recirculation System piping. From NUREG-1150, the recirculation loop DEGB mean frequency for this material is estimated to be 1 E-4/Rx-Yr. The fraction of BWR primary piping inside containment that is recirculation piping is estimated to be 2.0 E-1. The fraction of recirculation piping that can impact the containment metal shell is estimated to be 5.0 E-1. It was estimated that the mean probability of a recirculation system broken pipe rupturing the containment metal shell was 5.0 E-1. The mean probability of eventual failure of all ECCS by the same modes described for Case 1 is estimated to be 8.0 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 4.0 E-6

. Case 3 (INEEL BWR Event 12): Failure of RHR Pioina Resultina in Pioe Whio and Containment Imoact/ Failure. with Resultant Failure of All Emeraency Core Coolina Systems This event is similar to Cases 1 and 2 but involves the RHR System piping. From NUREG-1150, the RHR DEGB frequency for this material is estimated to be 1 E-4/Rx-Yr. The fraction of BWR primary piping inside containment that is RHR piping is estimated to be 1.0 E-1. The fraction of RHR piping that can impact the containment metal shell is estimated to be 5.0 E-1. The mean probability of a recirculation system broken pipe rupturing the containment metal shell is 1.0 E-1. The mean probability of eventual failure of all ECCS by the same modes described for Cases 1 and 2 is estimated to be 8.0 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 4.0 E-7 Case 4 (INEEL BWR Event 5): Failure of Recirculation Pioina Resultina in Pioe Whio or Jet Imoinaement on Control Rod Drive Bundles. Causino Failure by Crimoina of Enouah Insert / Withdraw Lines to Result in Failure to Scram the Reactor From NUREG-1150, the recirculation loop DEGB frequency for this material is estimated to be

~

1 E-4/Rx-Yr. The fraction of BWR primary piping inside containment that is recirculation piping is estimated to be 2.0 E-1. The fraction of recirculation piping that can impact or impinge on the CRD lines is estimated to be 2.5 E-1. It is estimated that the mean probability of a broken RHR pipe crimping enough CRD lines to prevent a scram (about 5 to 10 adjacent lines) is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 5.0 E-6 Case 5 (INEEL BWR Event 10): Failure of RHR Pioina Resultina in Pioe Whio or Jet Impinaement on Control Rod Drive Bundles. Causina Failure by Crimoina of Enouah Insert / Withdraw Lines to Result in Failure to Scram the Reactor This event is similar to Case 3 but involves the RHR Systerr. piping. The research performed indicates that there is considerable variation among the affected plants regarding the amount of l

pipe whip protection provided and the proximity of high energy lines to potential targets of concern; waldowns showed that in at least one case a large "unisolable from the RCS RHR line was routed directly between the two banks of CRD bundles. An RHR pipe break in this vicinity would surely impinge and/or impact on both banks simultaneously, i

From NUREG-1150, the RHR DEGB frequency for this material is estimated to be 1 E-4/Rx-Yr.

l The fraction of BWR primary piping inside containment that is RHR piping is estimated to be 1.0 j

E-1. The fraction of RHR piping that can impact or impinge on the CRD lines is estiinated to be 2.5 E-1. It is estimated that the mean probability of a broken RHR pipe crimping enough CRD f

[

lines to prevent a scram (about 5 to 10 adjacent lines) is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 2.5 E-6

l* Case 6 (INEEL BWR Event 14): Failure of Hiah Enerav Pioina Resultina in Pioe Whio or Jet imoinaement on Reactor Protection or Instrumentation & Control Electrical. Hydraulic or Pneumatic Lines or Comoonents and Eventually Resultino in Failure of Mitiaation Systems and Core Damaae i

From NUREG-1150, the Large LOCA frequency is 1.0 E-4/Rx-Yr. All high energy piping inside containment is considered. The fraction of high energy piping that can impact or impinge on these lines or components is estimated to be 5.0 E-1. The mean probability of a broken high energy line failing some of these lines or components to the extent that core damage results is estimated as 7.5 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 3.8 E-5 Case 7 (INEEL BWR Event 16): Failure of Hiah Enerav Pioina Resultina in Pioe Whio Imoact on Reactor Buildina Component Coolina Water (RBCCW) System to the Extent That the RBCCW Pressure Boundarv is Broken. Potentially Ooenina a Path to Outside Containment if Containment Isolation Fails to Occur: Also Possible Loss of RBCCW Outside Containment for Mitiaation From NUREG-1150, the Large LOCA frequency is 1.0 E-4/Rx-Yr. All high energy piping inside containment is considered. The fraction of high energy piping that can impact the RBCCW system is estimated as 1.0 E-1. The probability of an HELB broken pipe rupturing the RBCCW system is 5.0 E-1. The probability of failure to close of containment isolation check valve is 1.0 E-3; the probability of failure to close of a containment isolation motor operated valve is 3.0 E-3; this combines for a total of 4.0 E-3. Since the RBCCW surge tank in the secondary containment is vented to atmosphere and has a relatively small volume, it is assumed that its water inventory will drain quickly; for this reason the mean probability of opening a path to atmosphere outside containment is 1.0. Once this scenario proceeds to this point the RBCCW system in secondary containment will become unavailable, including the RHR heat exchanger; therefore, the probability of losing the RBCCW function outside containment to the extent that l

core damage occurs is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 2.0 E-8 The total change in core damage frequency for the above 7 BWR cases is:

l l

dCDF/Rx-Yr = 5.2 E-5 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933)

And, for all 16 affected BWRs:

dCDF/Yr = 8.3 E-4 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933) i

r

]

l l l

BWR Offsite Dose Table 1

GSI-156.6.1 GSI-156.6.1 WASH-1400 WASH-1400 Offsite Dose Event Number dCDF Release Offsite Dose (OSD) per NUREG/CR-(Events /Rx-Yr)

Category (Person-Rem /

(Person-Rem /

6395 Event)

Reactor Year)

BWR Event 1 2.0 E-6 BWR-3 5.1 E+6 10.2 BWR Event 5 5.0 E-6 BWR-4 6.1 E+5 3.1 BWR Event 9 4.0 E-6 BWR-3 5.1 E+6 20.4 BWR Event 10 2.5 E-6 BWR-4 6.1 E+5 1.5 BWR Event 12 4.0 E-7 BWR-3 5.1 E+6 2.0 BWR Event 14 3.8 E-5 BWR-4 6.1 E+5 23.2 BWR Event 16 2.0 E-8 BWR-3 5.1 E+6 0.1 Total 60.5 For the 17 affected BWRs, the estimated change in offsite dose per reactor (d Person-Rem / Reactor)is:

60.5 Person-Rem x 17 Average Remaining Years =

1029 Person-Rem Reactor-Year Reactor (Offsite)

For 20 years of life extension:

60.5 Person-Rem x 37 Average Remaining Years =

2239 Person-Rem Reactor-Year Reactor (Offsite)

And the estimated change in offsite dose for the 16 affected BWRs is:

1029 Person-Rem x 16 Affected BWRs 16,464 Person-Rem *

=

Reactor (Total Offsite, All Affected BWRs)

f For 20 years of life extension:

' 2239 Person-Rem x 16 Affected BWRs 35,824 Person-Rem *

=

Reactor (Total Offsite, All Affected BWRs)

PWRs.

Case 1 (INEEL PWR Event 9): Failure of Non-Leak-Before-Break Reactor Coolant System _,

Feedwater. or Main Steam Pioina Resultina in Pine Whio or Jet Imoinaement on Reactor Protection or Instrumentation & Control Electrical. Hydraulic or Pneumatic Lines or Components and Eventually Resultina in Failure of Mitination Systems and Core Damaae From NUREG-1150, the HELB frequency in the above listed systems is 1.5 E-3/Rx-Yr. All of the listed high energy piping inside containment is considered. The fraction of high energy piping that can impact or impinge on these lines or components is estimated to be 1.0 E-1. The mean probability of a broken high energy line failing some of these lines or components to the

)

extent that core damage results is estimated as 5.0 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 7.5 E-5 Case 2 (INEEL PWR Event 16): Failure of Main Steam or Feedwater Pioina Resultina in Pioe Whio and Containment Imoact/ Failure. with Resultant Failure of All Emeraency Core Coolina Systems I

From NUREG-1150, the DEGB frequency in Feedwater (FW) piping is estimated to be 4 E-4/Rx-Yr; for Main Steam (MS) piping it is estimated as 1 E-4/Rx-Yr. The fraction of FW piping that can impact the containment shell is estimated as 1.0 E-1; the fraction of MS piping is also estimated as 1.0 E-1; this fraction remains 1.0 E-1. The mean probability of a FW or MS system broken pipe rupturing the containment metal shell was 5.0 E-1. The mean probability of additional I&C or ECCS systems failures to the extent that core damage results is estimated as 4.8 E-5 for the case involving FW piping breaks, and 9.8 E-5 for the case involving MS piping breaks.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 1.4 E-9 I

J i

F-

.g.

l Case 3 (INEEL PWR Event 17): Failure of Main Steam or Feedwater Pioina Resultina in Pioe l

Whio Imoact on Component Coolina Water (CCW) Sv= tem to the Extent That the CCW Pressure Boundarv is Broken. Potentially Openina a Path to Outside Containment if Containment Isolation Fails to Occur: Also Possible Loss of CCW Outside Containment for Mitiaation From NUREG-1150, the DEGB frequency in Feedwater (FW) piping is estimated to be 4 E-4/Rx-Yr; for Main Steam (MS) piping it is estimated as 1 E-4/Rx-Yr; this combines for a total of 5.0 E-4. The fraction of FW piping that can impact the CCW system is estimated as 1.0 E i the fraction of MS piping is also estimated as 1.0 E-1; this fraction remains 1.0 E-1. The i

probability of a FW or MS system broken pipe rupturing the CCW system is 5.0 E-1. The l

probability of failure to close of containment isolation check valve is 1.0 E-3; the probability of failure to close of a containment isolation motor operated valve is 3.0 E-3; this combines for a total of 4.0 E-3. Since the CCW surge tank is in the auxiliary building near where mitigation equipment is, is vented to atmosphere and has a relatively small volume, it is assumed that its water inventory will drain quickly; for this reason the mean probability of opening a path to atmosphere outside containment is 1.0. Once this scenario proceeds to this point the CCW system outside containment will become unavailable, including the RHR heat exchanger; therefore, the probability of losing the CCW function outside containment to the extent that core damage occurs is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 1.0 E-7 The total change in core damage frequency for the above 3 PWR cases is:

l dCDF/Rx-Yr = 7.5 E-5 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933)

And, for all 25 affected PWRs:

dCDF/Yr = 1.9 E-3 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933)

PWR Offsite Dose Table GSI-156.6.1 GSI-156.6.1 WASH-1400 WASH-1400 Offsite Dose Event Number dCDF Release Offsite Dose (OSD) per NUREG/CR-(Events /Rx-Yr)

Category (Person-Rem /

(Person-Rem /

6395 Event)

Reactor Year)

PWR Event 9 -

7.5 E-5 PWR-6 1.5 E+5 11.3 PWR Event 16 1.4 E-9 PWR-4 2.7 E+6 0.004 PWR Event 17 1.0 E-7 PWR-4 2.7 E+6 0.3 Total 11.6

O a For the 25 affected PWRs, the estimated change in offsite dose per reactor (d Person-Rem / Reactor)is:

11.6 Person-Rem x 17 Average Remaining Years =

197 Person-Rem Reactor-Year Reactor (Offsite)

  • Ranks MEDIUM / LOW in Figure 2 of NUREG-0933 For 20 years of life extension:

11.6 Person-Rem x 37 Average Remaining Years =

429 Person-Rem Reactor-Year Reactor (Offsite)

  • Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933 And the estimated change in offsite dose for the 25 affected PWRs is:

197 Person-Rem x 25 Affected PWRs 4,925 Person-Rem *

=

Reactor (Total Offsite, All Affected PWRs)

  • Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933 For 20 years of life extension:

429 Person-Rem x 25 Affected PWRs 10,725 Person-Rem *

=

Reactor (TotalOffsite, All Affected PWRs)

  • Ranks MEDIUM in Figure 2 of NUREG-0933 The estimated total offsite dose for the 41 affected plants (BWRs and PWRs) is:

16,464 + 4,925 = 21,389 Person-Rem * (Total Offsite, All Affected Reactors w/o life extension)

  • Ranks MEDIUM in Figure 2 of NUREG-0933 35,824 + 10,725 = 46,549 Person-Rem * (Total Offsite, All Affected BWRs & PWRs w/ life extension)

F' l-I Cost Estimatq Industry Cost: Implementation of the possible solution is assumed to require the performance of engineering analyses inside containment, perform system walkdowns, and provide a report to the NRC. Ultimately, it is expected that operating procedures and/or technical specifications will be modified, inservice inspections will be enhanced, or physical modifications will be done either to piping (probably addition of pipe whip restraints or jet shields) or to the inside containment leakage detection system. It is expected that the cost to each plant will be $1M.

Therefore, for the 41 affected plants (16 BWRs and 25 PWRs) the total implementation cost is estimated to be $41M. This estimate was based on the presumption that the level of effort at the affected plants would be similar to that which resulted for this issue during the SEP program review of the 10 earliest SEP plants.

NRC Cost: Development and implementation of a resolution is estimated to cost $1M; primarily involving review of industry submittals and possible proposed changes to hard..are.

Total Cost: The total industry and NRC cost associated with the possible solution is $42M.

i l

ImpactNatue Assessment l

l S=

Total Cost ($)

Person-Rem (All Reactors)

$42M

=

l 21,389 Person-Rem l

$1960/ Person-Rem

  • w/o Life Extension j

l

=

l

Total Cost ($)

Person-Rem (All Reactors)

$42M

=

46,549 Person-Rem

$900/ Person-Rem

  • w/ 20 Years of Life Extension

=

  • Ranks HIGH in Figure 2 of NUREG-0933 OTHER CONSIDERATIONS 1.

The Updated Safety Analysis Report for an SEP-Ill BWR (i.e., one of the 41 plants potentially affected by this issue) stated that, in the event of a DEGB, the broken pipe would strike the Mark l Containment and deform it significantly. However, another BWR of about the same vintage is known to have been required to add energy absorbing structures to protect the Mark l Containment from pipe whip, prior to receipt of an

. l operating license. Therefore, it appears that there is considerable variation among the l

affected plants regarding the amount of pipe whip protection provided.

2.

Pipe breaks have actually occurred in the industry. Examples include a Surry Feedwater line break, a WNP-2 Fire System valve structural pressure boundary failure, and a Ft. Calhoun 12" Steam line break.

3.

Some suspect configurations were observed in the SEP-Ill walkdown plants; for example, at one BWR a very close proximity exists between a large RHR (unisolable from RCS) pipe and both banks of the Control Rod Drive piping, and at one PWR it appeared that a large volume of piping penetrated the containment near where a large amount of electrical wiring also penetrated the containment. This demonstrates that even through modest efforts (i.e., sampling walkdowns of a sampling of plants) configurations of potential concern have been identified.

4.

Readily available plant documentation provides very little insights regarding actual proximity of high energy piping and potential targets or concern. The potentiallack of adequate separation of redundant system targets (e.g., l&C electrical wiring) is also a concem.

5.

Uncertainty remains a significant factor because of the large scope of this issue. This is because of the large number and types of plants, and significant differences in the specific as-built details applicable to this issue.

l 6.

Many of the affected plants are either currently applying for life extension or are expected to in the near future. Most of the lead life extension applications will be from the affected plants for many years to come.

j 7.

Although there is a large apparent disparity between the BWR and PWR cases evaluated, it must be remembered that much of the background of this issue was based on sampling walkdowns; that is, only selected portions of selected plants were available for these walkdowns. Therefore, it is important to treat the BWR and PWR evaluations equally during the next phase of the evaluation. Also, some of the listed scenarios seem to have low probabilities but potentially high consequences. They should be further evaluated.

CONCLUSION Several potential accident scenarios were identified; 7 for BWRs and 3 for PWRs. Mean values for core damage were estimated for each and the cumulative effect of each group was also estimated. When compared to Figure 2 of NUREG-0933, these values mostly showed that this issue is of HIGH/ MEDIUM safety significance. Further evaluations which included estimates of offsite doses and costs for potential solutions showed that this issue is of HIGH priority.

c l

. REFERENCES 1.

Giambusso, A.,1972, AEC Deputy Director for Reactor Projects, Directorate of Licensing, generic letter to applicable BWR and PWR plants regarding postulated pipe failures outside of the containment structure, issued December 1972.

2.

O' Leary, J.F.,1973, AEC Director, Directorate of Licensing, generic letter to applicable BWR and PWR plants regarding clarification of A. Giambusso letter of 12/72, issued January through July 1973.

3.

Memorandum from E. S. Beckjord (RES) to A. C. Thadani (NRR) dated October 31, 1994.

4.

Memorandum from T. E. Murley (NRR) to E. S. Beckjord (RES) dated February 5,1991.

(User Need Letter).

5.

NUREG/CR-6395, " Enhanced Prioritization of Generic Safety issue 156.6.1, Pipe Break Effects on Systems and Components inside Containment," draft dated June 1999.

6.

WASH-1400 (NUREG-75/014)," Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission, October 1975.

7.

G. S. Holman, C. K. Chou (LLNL), " Assessment of Value-impact Associated with the Elimination of Postulated Pipe Ruptures from the Design Basis for Nuclear Power Plants," March 29,1985.

8.

NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety issue Prioritization Information Development," U.S. Nuclear Regulatory Commission, February 1983, (Supplement 1) May 1983, (Supplement 2) December 1983, (Supplement 3) September 1985, (Supplement 4) July 1986.

9.

NUREG/CR-4627, " Generic Cost Estimates," U.S. Nuclear Regulatory Commission, June 1986.

10.

NUREG/CR-4550," Analysis of Core Damage Frequency: Surry, Unit 1 Internal Events,"

April 1990.

11.

NUREG-0933, "A Prioritization of Generic Safety issues," July 1991.

12.

NUREG/CR-4792, " Probability of Failure in BWR Reactor Coolant Piping,"

Volume 1, March 1989.

j 13.

NUREG-1150," Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," December 1990.

o i

I*

Page 1 of 2 l

Managment and control indicators used in GIMCS are defined as follows:

(.

Generic Issue Number l

1.

Issue No.

Generic Issue Title 2.

Title Date the issue was identified 3.

Identification Date The date that the prioritization evaluation 4.

Prioritization Date was approved by the~RES Director Generic Safety (GSI), Licensing (LI), or 5.

g Regulatory Impact (RI)

High (H) or Medium (M) 6.

Priority Name of assigned individual responsible for 7.

Task Manager resolution The Office, Division, and Branch of the Task 8.

Office /Div/Br Manager who has lead responsibility for resolving the issue.

Technical assistance funds Active 9.

Action Level appropriated for resolution f

and/or Task Manager actively

(

pursuing resolution No technical assistance funds Inactive appropriated for resolution, Task Manager assigned to more important work, or no Task Manager assigned All necessary work has been Resolved completed and no additional resources will expended Coded summary as follows-

10. Status NR - (Nearly-Resolved);

3A - (Resolved with requirements),

i 3B - (Resolved with Ho requirements);

l 5 - (Licensing or Regulatory Impact issued that should be assigned resources for completion)

Task Action Control (TAC) number assigned l

11. TAC Number to the issue Scheduled resolution date for the issue j

1

12. Resolution Date

4.w.

._...m.,_..__m.a_.<

I-Page 2 of_2_

.h[

Who or what authorized work to be done on

13. Work Authori,zation

~

the issue s

Financial identification number assigned to

14. FIN contract (if any) for technical assistance Contractor name
15. Contractor Contract Title (if contract issued)
16. Contract Title

~-

Describes briefly the work necessary to

17. Work Scope technically resolve and complete the generic issue Describes current status of work
18. Status Identifies documents into which the

'19.

Affected Documents technical resolution will be incorporated Identifies problem areas and describes what

20. Problem / Resolution actions are necessary to resolve them i

Selected significant milestones:

21. Milestones (a) the " original" scheduled dates reflect the original Task Action Plan plus addi-tional milestone dates added during task (g-resolution; (b) changes in the the original schedaled

(

dates are listed under " Current";

(c) actual completion dates are listed under

" Actual" I

i i

1 i

1 l

l 1

0 L

F.

l l

7-7-99 Prioritization of Generic issue 156.6.1," Pipe Break Effects on Systems and Components" DESCRIPTION Historical Backcround in 1967 the AEC published draft General Design Criteria (GDCs) for comment and interim use.

)

Until 1972 the staff's implementation of the GDCs required consideration of pipe break effects inside containment. However, due to the lack of documented review criteria, NRC/AEC staff i

positions were continually evolving. Review uniformity was finally developed in the early 1970s; initiated by a Note from L. Rogers to R. Fraley, " Safety Guides" dated November 9,1972, in which a Draft Safety Guide entitled " Protection Against Pipe Whip Inside Containment" was proposed. This Draft contained some of the first documented deterministic criteria that the staff had been using for several years (to varying degrees) as guidelines for selecting the locations and orientations of postulated pipe breaks inside containment, and for identifying the measures that should be taken to protect safety-related systems and equipment from the dynamic effects of such breaks. Prior to use of these deterministic criteria, the staff used non-deterministic guidelines on a plant-specific basis. This Draft Safety Guide was subsequently revised and issued in May 1973 as Regulatory Guide 1,46 with the same title. The Regulatory Guide was implemented only on a forward-fit basis.

Regarding pipe break effects outside containment: in December 1972 and July 1973, the AEC 1

issued two generic letters to all licensees and CP or OL applicants (References 1 and 2) ;

known as the "Giambusso" and "O' Leary" letters, respectively. These letters extended the pipe break concerns to outside containment, and provided deterministic criteria for break postulation and evaluation of the dynamic effects of postulated breaks. The letters requested that all recipients submit a report to the staff which summarized each plant specific analysis of this issue. All operating reactor licensees and license applicants submitted the requested analyses in separate correspondence or updated the safety analysis report for the proposed plant to include the analysis. The staff reviewed all of these submitted analyses and prepared safety evaluations for all plants. In November 1975, the staff published SRP Sections 3.6.1 and 3.6.2 that slightly revised the two generic letters discussed above. Thus, after 1975 the specific structural and environmental effects of pipe whip, jet impingement, flooding, etc. on systems and components relied on for safe reactor shutdown were considered.

As stated above, the AEC/NRC has provided requirements to the industry regarding pipe

~

breaks outside of containment through the issuance of the "Giambusso" and "O' Leary" generic letters. Since these requirements are applicable to all the affected plants, pipe breaks outside of containment are considered a compliance issue and have been dropped from this prioritization. By UDO direction, compliance matters are to be dealt with promptly, and not await the generic issue resolution process. Therefore the issue of pipe breaks outside of containment for the 41 aff ected plants was brought to the attention of NRR by separate correspondence (Reference 3). The remainder of this prioritization discusses only pipe breaks inside containment.

As a part of its plant-specific reviews between 1975 and 1981, the staff used the guidelines in 1-Regulatory Guide 1.46 for postulated pipe breaks inside containment and SRPs 3.6.1 and 3.6.2 for outside containment. In July 1981, SRPs 3.6.1 and 3.6.2 were revised to be applicable to l

, both outside and inside containment; thus, eliminating the need for further use of Regulatory Guide 1.46.

Between the period 1983-1987, the NRC Systematic Evaluation Program (SEP) revisited the generalissue of pipe breaks inside and outside containment. The objective of the SEP was to determine to what extent the earliest 10 plants (i.e., SEP-ll) met the licensing criteria in existence at that time. This objective was later interpreted to ensure that the SEP also provided safety assessments adequate for conversion of provisional operating licenses (POLS) to full-term operating licenses (FTOLs). As a result of these reviews plants were required to perform j

engineering evaluations, technical specification or procedural changes, and physical modifications both inside and outside containment. Regarding inside containment modifications: of the two SEP-Il plants evaluated for this prioritization (one BWR and one PWR), the BWR was required to modify four piping containment penetrations and the PWR 1

was required to modify steam generator blowdown piping supports. This indicates there was a wide spectrum of implementation associated with the original reviews of these early plants for pipe breaks inside and outside containment.

As with the above-described evolution of uniform pipe break criteria, electrical systems design criteria were also in a state of development. Prior to 1974, electrical system designs were generally reviewed in accordance with the guidelines provided in IEEE-279; however, significant variations in interpretations of that document resulted in substantial design differences in plants.

Specifically, true physical separation of wiring to redundant components was not necessarily accomplished. In 1974, Regulatory Guide 1.75 was published, clarifying the requirements.

A draft prioritization of this issue resulted in a MEDIUM determination and that the scope could be limited to pipe breaks inside containment since the NRC had already provided requirements regarding outside containment pipe breaks to the industry through the issuance of the previously mentioned "Giambusso" and "O' Leary" generic letters.

However, the uncertainty in the analysis was much wider than desired for a definitive priority ranking. Thus, the issue appeared to warrant additional analysis to enhance the prioritization, in July 1994 a contract was begun with the Idaho National Engineering Laboratory to:

1.

Review of pipe failure rate data, pipe break methodologies, and related publications to determine recommended pipe failure rates (initiating events) applicable to the affected SEP-lil plants.

2.

Review of Updated Final Safety Analysis Reports and related Safety Evaluation Reports for SEP-ll, SEP-ill, and for representative non-SEP plants to identify and prioritize potential safety concerns (i.e., accident sequences). Several plant visits /walkdowns were included as part of this review.

3.

Estimate changes to core damage frequencies for accident sequences that are determined to be of high or medium priority.

. 4.

Identify potential corrective actions and their estimated costs.

o l

l Based on the results of the INEL research, the enhanced prioritization is presented below.

j Safety Sianificance GDC 4 is the primary regulatory requirement of concern. It requires, in part, that structures, systems and components important to safety be appropriately protected against the environmental and dynamic effects that may result from equipment failures, including the j

effects of pipe whipping and discharging fluids. Several possible scenarios for plants that do j

not have adequate protection against pipe whip ware identified as a result of the research performed in support of the enhanced prioritization.

Related regulatory criteria include common cause failures, protection system independence, i

and the single failure criterion.

Recommended Solution issue Generic Letters to the affected plants requesting that they perform plant-specific reviews and walkdowns, identify vulnerable pipe break locations, and inform the NRC of proposed corrective actions.

PRIORITY DETERMINATIONS l

Numerous scenarios of potential concern were evaluated. The following were considered important enough to be specifically identified for future consideration. All estimated frequencies and probabilities are mean values.

RWRs Case 1 (INEEL BWR Event 1h Failure of Main Steam or Feedwater Pioina Resultina in Pioe Whip and Containment Imoact/ Failure. with Resultant Failure of All Safety iniection Systems This event involves a BWR with a Mark I steel containment; 15 of the 16 affected BWRs are of this design. A DEGB of an unprotected (i.e., no pipe whip restraint or containment liner impact absorber) large reactor coolant recirculation pipe inside containment and near the containment liner might result in puncturing the liner. The resulting unisolable LOCA steam environment would be introduced into the secondary containment building, possibly disabling the ECCS equipment located there. This scenario would greatly increase the probability of core damage and potential offsite doses.

All of the affected BWRs are more than 10 years old, and most use type 304 stainless steelin the primary system piping; a material that is susceptible to IGSCC degradation. It should be noted that piping of this material does not qualify for the extremely low rupture probability (Leak-Before-Break) provision of GDC 4. From NUREG-1150, the recirculation loop DEGB freque~.::y for this material is estimated to be 1 E-4/ Reactor-Year (Rx-Yr). The fraction of BWR primary piping inside containment that is either Main Steam (MS) or Feedwater (FW) is estimated to be 4.0 E-1. The fraction of MS or FW piping that can impact the containment l

metal shell is estimated to be 2.5 E-1.

l 1

I l l The research performed indicates that there is considerable variation among the affected plants regarding the amount of pipe whip protection provided and the proxirnity of high energy lines to potential targets of concern, including redundant trains, (see Other Considerations). It was assumed that the probability of a MS or FW broken pipe rupturing the containment metal shell was 2.5 E 1.

The postulated event may also cause a common mode failure of the ECCS system since much of this equipment is located within the secondary containment and will be exposed to a harsh environment beyond its design basis, or that the ECCS piping will fall due to overpressurization of the containment annulus. In most of the affected plants, the ECCS is located in four different quadrants outside the suppression pool (torus). On the other hand, as stated above, redundant electrical power systems and initiating circuitry may not be physically separated in these older plants. Also, if the ECCS operates initially, the ECCS equipment rooms may not be fully protected from internal flooding as the water from the suppression pool flows out the broken pipe into the secondary containment. Based on these considerations the mean probability of loss of ECCS function was assumed to 8.0 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 2.0 E-6 From WASH-1400, the nearest scenario to that described above is the large LOCA BWR-3 release category; involving a large LOCA and subsequent containment failure. However, in the WASH-1400 case, the containment failure results from overpressurization; not from pipe whip.

Three of the four specific BWR-3 large LOCA accident sequences have an incidence frequency of 10 E-8/Rx-Yr, and the remaining one is 10 E-7/Rx-Yr; 10 E-8/Rx-Yr was chosen as the base case for this analysis.

Case 2 (INEEL BWR Event 9h Failure of Recirculation Pinina Resultina in Pioe Whio and Containment impact / Failure with Resultant Failure of All Emeroency Core Coolina Systems This event is similar to Case 1 but involves the Recirculation System piping. From NUREG-1150, the recirculation loop DEGB mean frequency for this material is estimated to be 1 E-4/Rx-Yr. The fraction of BWR primary piping inside containment that is recirculation piping is estimated to be 2.0 E-1. The fraction of recirculation piping that can impact the containment metal shen is estimated to be 5.0 E-1. It was estimated that the mean probability of a recirculation system broken pipe rupturing the containment metal shell was 5.0 E-1. The mean probability of eventual failure of all ECCS by the same modes described for Case 1 is estimated to be 8.0 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr

' E-6 I

l l

l l

l 5-l Case 3 (INEEL BWR Event 12): Failure of RHR Pipina Resultina in Pioe Whio and Containment imoact/ Failure. with Resultant Failure of All Emeraency Core Coolina Systems This event is similar to Cases 1 and 2 but involves the RHR System piping. From NUREG-1150, the RHR DEGB frequency for this materialis estimated to be 1 E-4/Rx-Yr. The fraction of BWR primary piping inside containment that is RHR piping is estimated to be 1.0 E-1. The fraction of RHR piping that can impact the containment metal shellis estimated to be 5.0 E-1. The mean probability of a recirculation system broken pipe rupturing the containment metal shellis 1.0 E-1. The mean probability of eventual failure of all ECCS by the same modes described for Cases 1 and 2 is estimated to be 8.0 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 4.0 E-7 Case 4 (INEEL BWR Event 5): Failure of Recirculation Pioina Resultina in Pine Whio or Jet imoinaement on Control Rod Drive Bundles. Causino Failure by Crimoina of Enouah insert / Withdraw Lines to Result in Failure to Scram the Reactor From NUREG-1150, the recirculation loop DEGB frequency for this material is estimated to be 1 E-4/Rx-Yr. The fraction of BWR primary piping inside containment that is recirculation piping is estimated to be 2.0 E-1. The fraction of recirculation piping that can impact or impinge on the CRD lines is estimated to be 2.5 E-1. It is estimated that the mean probability of a broken RHR pipe crimping enough CRD lines to prevent a scram (about 5 to 10 adjacent lines) is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 5.0 E-6 Case 5 (INEEL BWR Event 10): Failure of RHR Pioina Resultina in Pioe Whio or Jet Imoinaement on Control Rod Drive Bundles. Causino Failure by Crimoina of Enouah Insert / Withdraw Lines to Result in Failure to Scram the Reactor This event is similar to Case 3 but involves the RHR System piping. The research performed indicates that there is considerable variation among the affected plants regarding the amount of pipe whip protection provided and the proximity of high energy lines to potential targets of concern; waldowns showed that in at least ens case a large "unisolable from the RCS" RHR line was routed directly between ih6 iwo banks of CRD bundles. An RHR pipe break in this vicinity would surely impinge and/or impact on both banks simultaneously.

From NUREG-1150, the RHR DEGB frequency for this material is estimated to be 1 E-4/Rx-Yr.

The fraction of BWR primary piping inside containment that is RHR piping is estimated to be 1.0 E-1. The fraction of RHR piping that can impact or impinge on the CRD lines is estimated to be 2.5 E-1. It is estimated that the mean probability of a broken RHR pipe crimping enough CRD i

l lines to prevent a scram (about 5 to 10 adjacent lines) is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

1 dCDF/Rx Yr = 2.5 E-6 1

I l

l l Case 6 (INEEL BWR Event 14): Failure of Hiah Enerav Pioina Resultina in Pioe Whio or Jet Imoingement on Reactor Protection or Instrumentation & Control Electrical. Hydraulic or Pneumatic Lines or Comoonents and Eventually Resultina in Failure of Mitiaation Systems and Core Damaae From NUREG-1150, the Large LOCA frequency is 1.0 E-4/Rx-Yr. All high energy piping inside containment is considered. The fraction of high energy piping that can impact or impinge on these lines or components is estimated to be 5.0 E-1. The mean probability of a broken high energy line failing some of these lines or components to the extent that core damage results is estimated as 7.5 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 3.8 E-5 Case 7 (INEEL BWR Event 16): Failure of Hiah Enerav Pipina Resultina in Pioe Whio Imoact on Reactor Buildina Component Coolina Water (RBCCW) System to the Extent That the RBCCW Pressure Boundarv is Broken. Potentially Openina a Path to Outside Containment if Containment Isolation Fails to Occur: Also Possible Loss of RBCCW Outside Containment for Mitiaation From NUREG-1150, the Large LOCA frequency is 1.0 E-4/Rx Yr. All high energy piping inside containment is considered. The fraction of high energy piping that can impact the RBCCW system is estimated as 1.0 E-1. The probability cf an HELB broken pipe rupturing the RBCCW system is 5.0 E-1. The ptobability of failure to close of containment isolation check valve is 1.0 E-3; the probability of failure to close of a containment isolation motor operated valve is 3.0 E-3; this combines for a total of 4.0 E-3. Since the RBCCW surge tank in the secondary containment is vented to atmosphere anti has a relatively small volume, it is assumed that its water inventory will drain quickly; for this ceason the mean probability of opening a path to atmosphere outside containment is 1.0. Once this scenario proceeds to this point the RBCCW system in secondary containment will become unavailable, including the RHR heat exchanger, therefore, the probability of losing the RBCCW function outside containment to the extent that core damage occurs is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 2.0 E-8 The total change in core damage frequency for the above 7 BWR cases is:

dCDF/Rx-Yr = 5.2 E-5 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933)

And, for all 16 affected BWRs:

dCDF/Yr = 8.3 E-4 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933)

7-BWR Offsite Dose Table GSI-156.6.1 GSI-156.6.1 WASH-1400 WASH 1400 Offsite Dose Event Number dCDF Release Offsite Dose (OSD) per NUREG/CR-(Events /Rx-Yr)

Category (Person-Rem /

(Person-Rem /

6395 Event)

Reactor Year)

BWR Event 1 2.0 E-6 BWR-3 5.1 E+6 10.2 BWR Event 5 5.0 E-6 BWR-4 6.1 E+5 3.1 BWR Event 9 4.0 E-6 BWR-3 5.1 E+6 20.4 BWR Event 10 2.5 E-6 BWR-4 6.1 E+5 1.5 BWR Event 12 4.0 E-7 BWR-3 5.1 E+6 2.0 BWR Event 14 3.8 E-5 BWR-4 6.1 E+5 23.2 BWR Event 16 2.0 E-8 BWR-3 5.1 E+6 0.1 Total 60.5 For the 17 affected BWRs, the estimated change in offsite dose per reactor (d Person-Rem / Reactor) is:

60.5 Person-Rem x 17 Average Remaining Years =

1029 Person-Rem Reactor Year Reactor (Offsite)

For 20 years of life extension:

60.5 Person-Rem x 37 Average Remaining Years =

2239 Person-Rem Reactor-Year Reactor (Offsite)

  • (Ranks.HIGH/ MEDIUM in Figure 2 of NUREG-0933) i And the estimated change in offsite dose for the 16 affected BWRs is:

l 16,464 Person-Rem

  • 1029 Person-Rem x 16 Affected BWRs

=

Reactor (Total Offsite, All Affected BWRs)

8 For 20 years of life extension:

2239 Person-Rem x 16 Affected BWRs 35,824 Person-Rem *

=

Reactor (Total Offsite, All Affected BWRs)

Feedwater. or Main Steam Pioina Resultina in Pipe Whio or Jet Imoinaement on Reactor l

Protection or Instrumentation & Control Electrical. Hydraulic or Pneurnatic Lines or Components and Eventually Resultina in Failure of Mitiaation Systems and Core Damaae From NUREG-1150, the HELB frequency in the above listed systems is 1.5 E-3/Rx-Yr. All of the listed high energy piping inside containment is considered. The fraction of high energy piping that can impact or impinge on these lines or components is estimated to be 1.0 E-1. The mean probability of a broken high energy line failing some of these lines or components to the extent that core damage results is estimated as 5.0 E-1.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx-Yr = 7.5 E 5 Case 2 (INEEL PWR Event 16): Failure of Main Steam or Feedwater Picina Resultina in Pioe Whio and Containment impact / Failure. with Resultant Failure of All Emeraency Core Coolina Systems From NUREG-1150, the DEGB frequency in Feedwater (FW) piping is estimated to be 4 E-4/Rx-Yr; for Main Steam (MS) piping it is estimated as 1 E-4/Rx-Yr. The fraction of FW piping that can impact the containment shellis estimated as 1.0 E-1; the fraction of MS piping i

is also estimated as 1.0 E-1; this fraction remains 1.0 E-1. The mean probability of a FW or MS system broken pipe rupturing the containment metal shell was 5.0 E-1. The mean probability of i

additional l&C or ECCS systems failures to the extent that core damage results is estimated as 4.8 E-5 for the case involving FW piping breaks, and 9.8 E-5 for the case involving MS piping breaks.

l Based on the above assumptions, the mean value of change in CDF per reactor year is:

1 dCDF/Rx-Yr = 1.4 E-9 1

4 1

I l

-9 Case 3 (INEEL PWR Event 17): Failure of Main Steam or Feedwater Pioina Resultina in Pipe Whio impact on Component Coolina Water (CCW) System to the Extent That the CCW I

Pressure Boundarv is Broken. Potentially Openina a Path to Outside Containment if Containment Isolation Fails to Occur: Also Possible Loss of CCW Outside Containment for Mitiaation From NUREG-1150, the DEGB frequency in Feedwater (FW) piping is estimated to be 4 E-4/Rx-Yr; for Main Steam (MS) piping it is estimated as 1 E-4/Rx-Yr; this combines for a total of 5.0 E-4. The fraction of FW piping that can impact the CCW system is estimated as 1.0 E-1; the fraction of MS piping is also estimated as 1.0 E-1; this fraction remains 1.0 E-1. The probability of a FW or MS system broken pipe rupturing the CCW system is 5.0 E-1. The l

probability of failure to close of containment isolation check valve is 1.0 E-3; the probability of I

failure to close of a containment isolation motor operated valve is 3.0 E-3; this combines for a total of 4.0 E-3. Since the CCW surge tank is in the auxiliary building near where mitigation equipment is, is vented to atmosphere and has a relatively small volume, it is assumed that its water inventory will drain quickly; for this reason the mean probability of opening a path to atmosphere outside containment is 1.0. Once this scenario proceeds to this point the CCW system outside containment will become unavailable, including the RHR heat exchanger; therefore, the probability of losing the CCW function outside containment to the extent that core damage occurs is 1.0.

Based on the above assumptions, the mean value of change in CDF per reactor year is:

dCDF/Rx Yr = 1.0 E-7 The total change in core damage frequency for the above 3 PWR cases is:

dCDF/Rx-Yr = 7.5 E-5 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933)

And, for all 25 affected PWRs:

dCDF/Yr = 1.9 E-3 (Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933)

PWR Offsite Dose Table GSI-156.6.1 GSI-156.6.1 WASH-1400 WASH-1400 Offsite Dose Event Number dCDF Release Offsite Dose (OSD) per NUREG/CR-(Events /Rx-Yr)

Category (Person-Rem /

(Person-Rem /

6395 Event)

Reactor Year)

PWR Event 9 7.5 E-5 PWR-6 1.5 E+5 11.3 PWR Event 16 1.4 E-9 PWR-4 2.7 E46 0.004 PWR Event 17 1.0 E-7 PWR-4 2.7 E+6 0.3 Total 11.6

..-. For the 25 affected PWRs, the estimated change in offsite dose per reactor (d Person-Rem / Reactor) is:

11.6 Person-Rem x 17 Average Remaining Years =

197 Person-Rem

)

Reactor-Year Reactor (Offsite)

  • Ranks MEDIUM / LOW in Figure 2 of NUREG-0933 For 20 years of life extension:

i 11.6 Person-Rem x 37 Average Remaining Years =

429 Person-Rem Reactor-Year Reactor (Offsite)

  • Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933 And the estimated change in offsite dose for the 25 affected PWRs is:

197 Person-Rem x 25 Affected PWRs 4,925 Person-Rem *

=

Reactor (Total Offsite, All Affected PWRs)

  • Ranks HIGH/ MEDIUM in Figure 2 of NUREG-0933 For 20 years of life extension:

429 Person-Rem x 25 Affected PWRs 10,725 Person-Hem *

=

Reactor (Total Offsite, All Affected PWRs)

  • Ranks MEDIUM in Figure 2 of NUREG-0933 The estimated total offsite dose for the 41 affected plants (BWRs and PWRs) is:

16,464 + 4,925 = 21,389 Person-Rem * (Total Offsite, All Affected Reactors w/o life extension)

35,824 + 10,725 = 46,549 Person-Rem * (Total Offsite, All Affected BWRs & PWRs l

w/ life extension)

E

..., Cost Estimate Industry Cost: Implementation of the possible solution is assumed to require the performance of engineering analyses inside containment, pedorm system walkdowns, and provide a report to the NRC.. Ultimately, it is expected that operating procedures and/or technical specifications will be modified, inservice inspections will be enhanced, or physical modifications will be done either to piping (probably addition of pipe whip restraints or jet shields) or to the inside containment leakage detection system. It is expected that the cost to each plant will be $1M.

l Therefore, for the 41 affected plants (16 BWRs and 25 PWRs) the total implementation cost is estimated to be $41M. This estimate was based on the presumption that the level of effort at the affected plants would be similar to that which resulted for this issue during the SEP program review of the 10 earliest SEP plants.

NRC Cost: Development and implementation of a resolution is estimated to cost $1M; primarily involving review of industry submittals and possible proposed changes to hardware.

Total Cost: The totalindustry and NRC cost associated with the possible solution is $42M.

Impact /Value Assessment S=

Total Cost ($)

Person-Rem (All Reactors)

$42M I

=

21,389 Person-Rem

$1960/ Person-Rem

  • w/o Life Extension

=

TotalCost (S)

Person-Rem (All Reactors)

$42M

=

a 46,549 Person-Rem

$900/ Person-Rem

  • w/ 20 Years of Life Extension

=

  • Ranks HIGH in Figure 2 of NUREG-0933 OTHER CONSIDERATIONS l

1.

The Updated Safety Analysis Raport for an SEP-Ill BWR (i.e., one of the 41 plants potentially affected by this issue) stated that, in the event of a DEGB, the broken pipe would strike the Mark l Containment and deform it significantly. However, another BWR of about the same vintage is known to have been required to add energy absorbing structures to protect the Mark l Containment from pipe whip, prior to receipt of an L

12-operating license. Therefore, it appears that there is considerable variation among the affected plants regarding the amount of pipe whip protection provided.

2.

Pipe breaks have actually occurred in the industry. Examples include a Surry Feedwater line break, a WNP-2 Fire System valve structural pressure boundary failure, j

and a Ft. Calhoun 12" Steam line break.

3.

Some suspect configurations were observed in the SEP-Ill walkdown plants; for example, at one BWR a very close proximity exists between a large RHR (unisolable from RCS) pipe and both banks of the Control Rod Drive piping, and at one PWR it i

appeared that a large volume of piping penetrated the containment near where a large amount of electrical wiring also penetrated the containment. This demonstrates that even through modest efforts (i.e., sampling walkdowns of a sampling of plants) configurations of potential concern have been identified.

4.

Readily available plant documentation provides very little insights regarding actual proximity of high energy piping and potential targets or concern. The potential lack of adequate separation of redundant system targets (e.g., l&C electrical wiring) is also a concern.

5.

Uncertainty remains a significant factor because of the large scope of this issue. This is because e,f the large number and types of plants, and significant differences in the l

specific es-built details applicable to this issue.

6.

Many of the affected plants are either currently applying for life extension or are expected to in the near future. Most of the lead life extension applications will be from the affected plants for many years to come.

7.

Although there is a large apparent disparity between the BWR and PWR cases evaluated, it must be remembered that much of the background of this issue was based on sampling walkdowns; that is, only selected portions of selected plants were available for these walkdowns. Therefore, it is important to treat the BWR and PWR evaluations equally during the next phase of the evaluation. Also, some of the listed scenarios seem to have low probabilities but potentially high consequences. They should be further evaluated.

CONCLUSION Several potential accident scenarios were identified; 7 for BWRs and 3 for PWRs. Mean values l

for core damage were estimated for each and the cumulative effect of each group was also estimated. When compared to Figure 2 of NUREG-0933, these values mostly showed that this issue is of HIGH/ MEDIUM safety significance. Further evaluations which included estimates of offsite doses and costs for potential solutions showed that this issue is of HIGH priority.

l

  • .' o REFERENCES 1.

Giambusso, A.,1972, AEC Deputy Director for Reactor Projects, Directorate of Licensing, generic letter to applicable BWR and PWR plants regarding postulated pipe failures outside of the containment structure, issued December 1972.

2.

O' Leary, J.F.,1973, AEC Director, Directorate of Licensing, generic letter to applicable BWR and PWR plante regarding clarification of A. Giambusso letter of 12/72, issued January through July 1973.

3.

Memorandum from E. S. Beckjord (RES) to A. C. Thadani (NRR) dated October 31, 1994.

4.

Memorandum from T. E. Murley (NRR) to E. S. Beckjord (RES) dated February 5,1991.

(User Need Letter).

5.

NUREG/CR-6395," Enhanced Prioritization of Generic Safety Issue 156.6.1, Pipe Break Effects on Systems and Components inside Containment," draft dated June 1999.

6.

WASH-1400 (NUREG-75/014), " Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission, October 1975.

7.

G. S. Holman, C. K. Chou (LLNL), " Assessment of Value-impact Associated with the Elimination of Postulated Pipe Ruptures from the Design Basis for Nuclear Power Plants," March 29,1985.

8.

NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety issue Prioritization Information Development," U.S. Nuclear Regulatory Commission, February 1983, (Supplement 1) May 1983, (Supplement 2) December 1983, (Supplement 3) September 1985, (Supplement 4) July 1986.

9.

NUREG/CR-4627, " Generic Cost Estimates," U.S. Nuclear Regulatory Commission, June 1986.

10.

NUREG/CR-4550," Analysis of Core Damage Frequency: Surry, Unit 1 Internal Events,"

April 1990.

11.

NUREG-0933, "A Prioritization of Generic Safety issues," July 1991.

l 12.

NUREG/CR-4792, " Probability of Failure in BWR Reactor Coolant Piping,"

Volume 1, March 1989.

13.

NUREG 1150," Severe Accident Risks: An Assessment for Five U.S. Nuclear Power l

Plants," December 1990.

1 i

L i

Page 1 of 2 Managment and control indicators used in GIMCS are defined as follows:

(.

Generic Issue Number 1.

Issue No.

Generic Issue Title I

2.

Title Date the issue was identified j

3.

Identification Date The date that the prioritization evaluation 4.

Prioritization Date was approved by the'RES Director Generic Safety (GSI) Licensing (LI), or 5.

Ty[Le Regulatory Impact (RI)

High (H) or Medium (M) 6.

Priority Name of assigned individual responsible for 7.

Task Manager resolution The Office, Division, and Branch of the Task 8.

Office /Div/Br Manager who has lead responsibility for resolving the issue.

Technical assistance funds Active 9.

Action Level appropriated for resolution

(

and/or Task Manager actively

(

pursuing resolution No technical assistance funds Inactive appropriated for resolution, Task Manager assigned to more important work, or no Task Manager assigned All necessary work has been Resolved completed and no additional resources will expended Coded summary as follows:

10. Status NR - (Nearly-Resolved);

3A - (Resolved with requirements);

3B - (Resolved with @ requirements);

5 - (Licensing or Regulatory Impact issued that should be assigned resources for completion)

Task Action Control (TAC) number assigned

11. TAC Number to the issue Scheduled resolution date for the issue 12.

Resolution Date I

L

. A..m.

K.--=-...

....-...w.......

,.a o

  • Page 2 of 1
13. Work Authorization Who or what authorized work to be done on the issue 6

Financial identification number assigned to

14. FIN contract (if any) for technical assistance Contractor name
15. Contractor Contract Title (if contract issued)
16. Contract Title Describes briefly the work necessary to
17. Work Scope technically resolve and complete the generic issue Describes current status of work
18. Status Identifies documents into which the

'19.

Affected Documents technical resolution will be incorporated Identifies problem areas and describes what

20. Problem / Resolution actions are necessary to resolve them Selected r'anificant milestones:
21. Milestones (a) the " original" scheduled dates reflect the original Task Action Plan plus addf-tional silestone dates added during task resolution; (b) changes in the the original scheduled dates are listed under " Current";

(c) actual completion dates are listed under

" Actual"

Retrieved from "https://kanterella.com/w/index.php?title=ML20211F102&oldid=4009725"