ML20196E010
| ML20196E010 | |
| Person / Time | |
|---|---|
| Issue date: | 01/20/1988 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Jordan E Committee To Review Generic Requirements |
| Shared Package | |
| ML20151L073 | List: |
| References | |
| NUDOCS 8802250336 | |
| Download: ML20196E010 (17) | |
Text
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!@ ctog,\ UNITED STATES
. I g NUCLEAR REGULATORY COMMISSION 3 l WASHINGTON, D. C 205%
\...../ JAN 2 01988 i MEMORANDUM FOR: Edward L. Jordan, Chairman Connittee to Review Generic Requirements FROM: Thomas E. Murley, Director Office of Nuclear Peactor Regulation
SUBJECT:
PROPOSED BULLETIN REGARDING RAPIDLY PROPAGATING FATIGUE CRACKS IN STEAM GENERATOR TURES < NRR requests that the Committee to Review Generic Requirements (CRGR) consider the enclosed subject proposed bulletin. We have determined W t some pressurized-water reactor plants may fail to meet their licensing basis for the reactor coolant pressure boundary as defined by General Design Criterion 14 relative to ensuring an extremely low probability of abnorus' leakage, of rapidly propagating failure, and of gross rupture. The bulletin requests licensees to take the necessary action to determine if any of their plants are susceptible to the rapidly propagating fatigue mecha-nism that led to the steam generator tube rupture event of July 15, 1987, at North Anna Unit 1. Corrective and/or compensatory measures would be requested as necessary. The actions requested by the bulletin will ensure that the reactor coolant pressure boundary will continue to comply with General Desiqn Criterion 14 The bulletin will also ensure compliance with Appendix B of L0 CFR 50, which requires planned and systematic actions, as necessary, to provide adequate assurance that safety-related components will perform satisfactorily in service. Enclosed is the proposed bulletin and the information required by the CRGR charter to support issuance of the bulletin. Larry Shao, Director, Division of Engineering and Systems Technology, is the sponsoring Division Director. 9.~ (tML hd ' T omas E. Murley, Director ffice of Nuclear Reactor Regulation
Enclosure:
As stated CONTACT: Emmett Murphy, NRR 492-7632 , WA&l5083h N jyp.
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1 CRGR Iteer IV.B. Contents of Packages Submitted to CRGR (Rev 4 Stello to List 042387, des 41860 342 ff) The following requirements apply for proposals to reduce existing requirements or (regulatory) positions as well as proposals to increase requirements or (regulatory) positicns. Each package submitted to the CRGR for review shall include fifteen (15) copies of the following infomation: SUPJECT: RAPIDLY PROPAGATING FATIGUE CRACKS IN STEAM GENERATOR TUBES I. The proposed generic requirement as it is proposed to be sent out to licensees. The proposed generic requirements are spelled out in the proposed bulletin enclosed with this review package. The recomended actions apply to plants with Westinghouse steam generators employing carbon I steel support plates (i.e., Models Nos. 13, 27, 44, 51, D1, D2, D3, D4 and E*). A brief sumary of these actions is provided below. A. Review inspection data records for evidence of denting corrosion ' at the uppemost tube support plate. B. For plants without denting, the continued absence of denting should be confirmed periodically } C. For plants with denting, the following actions should be taken:
- 1. Pending staff review and approval of the program in Item I.C.2, below an enhanced primary to secondary leak rate monitoring program should be implemented as ar. interim com-pensatory measure within 45 days of receipt of the bulletin.
- 2. A program should be implemented to minimize the probability -
of a rapidly propagating fatigue failure such as occurred at North Anna Unit 1. The need for corrective actions (e.g., preventive plugging of potentially susceptible tubes, hardware, and/or operational changes to reduce stability ratios) and/or long term compensatory mear fres (e.g., an enhanced leak rate monitoring program) shoul be assessed and implemented as necessary. A description of this program and the results should be submitted for staff review and approval according to the schedule specified in the bulletin. The bulletin will also requests submittal of a rep 3rt within 45 days of receipt of the bulletin describing the status of compliance with bulletin requirements. II. Draft staff papers or other underlying staff documents supporting the requirements or staff iregulatory) positions. (A copy of all materi-als referenced in the locument shalf t,e made available updn request { to the CRGR staff. An/ comittee member may reowst CRGR staff to I obtain a copy of any referenced material for his or her use.)
"South Texas Unit I only 1
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' A. The proposed bulletin represents the staff's followup to the July 15, 1987, steam generator tube rupture (SGTR) event at North Anna Unit 1.
B. The staff's safety evaluation supporting restart of North Anna Unit 1 from the post-event cutage centains cetailed information concerning the circtanstances that coulti render a plant vulnera-ble to a simila Lype failure in the future. Each propused requirement or staff (regulatory) position shall con-tain the sponsoring office's position as to whether the p.oposal would increase requiments or staff (regulatory) positions, imple-ment existing requirements or staff (regulatory ! positions,orwould , relax or reduce existing iequirements or staff tregulatory) positions. A. The reconnended actions of the proposed bulletin are necessary to ensure continued compliance with 10 CFR 50 Appendices A and 8. Specifically, Ger.eral !$rsign Criterion (GDC) 14 of 10 cm 50 - Aopendix A states that the reactor coolant system boundary shall "have an extremely low pmbability of abnormal leakage; of r Ap-l idly propagating failure, and of In addition, 1 10 CFR 50 Appendix B establishes, gross rupture."in part, quality assura rvrsirements for the operation of safety r*1sted components. As used in Appendix B, quality assurance comprises all those ' planned and systematic acticas necessary to provide adequate confidence that a structure, system, or component will perfona satisfactorily in service. The requireseats of Appendix B apply i to all activities affecting the safety-related functio ts of I these components; these include, in part, inspecting, testing, operating, and maintaining. B. It has long been recoqnized that steem generator tubes tre sub-ject to corrosive and mechanic,rily induced degradation. lo en-l sure complience with GDC 14, licenseas are reqv!wd by the technical specifications (TS) to periodically inspect the tubes to ensure that they continue to meet minimum v311 thickress re-quirements. In addition, TS 11ahs on alloweble primary to sec-l ondary leakage ensure that if r, fit.w escaws detection during inservice inspection and penetrates entirely through the tube l wall, causing a leak, that the plant will be shut down before , rupture occurs. '
, C. Inservice tube inspections and primary to secondary leak rate senitorir.g programs have proven effcetive through the years in minie.izing the occurrence of steam generator tube rv2tures
($GTRs). The frequsuy of SG1Rs to cate is about 10 8/ reactor [ year. The staff has M storically considered such a frequency to ! be consistent with th+ intent of GDC 14 and not to be a signiff-cant contributor to overall risk a'isociated with thd operation i of nuclear power plants fhee probabilistic risk assessments per. formed as parts of Unresolved Safety Issues A-3 A-4, and A-5 The North Anna Unit 1 SGTR event on July 15, 1987, however, re)p-resents a failure mode not anticipated at the time the steam generators were designed or when the TS were formulated. The ' t
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4 North Anna failure mode is unique among all other known failure modes in that inservice inspection programs and existing leak ' monitorino programs (in many if not most instances) may be inef-fective in preventing this type of failure. This is due to the fact that the North Anna failure mechanism involves a rapidly prcpagating circumferential fatigue crack. The time required to propagate the crack from the poiht of crack f attiation to a com-plete double-ended tube rupture is estimated by the staff to be between several hours and several days. The final stage of crack propacation when leakage progresses from a small amount (say less than 0.1 gpm) to a very large amount (say greater than 500 gas) may be just a few hours, t D. Based on available infonnation, the staff concludes that the ! presence of all the following requisite conditions could lead to g a rapidly propagating fatigue failure such as occurred at North Anna: a) Denting at the upper support plate, and . b) Fluid-elastic stability ratio approaching that for the tube that ruptured at No'.th Anna, and L c) Absence of effective antistbration support. E. Preliminary scoping analyses by Westinghouse identified 5 plants f that might exhibit all of the reouisite conditions for a fatigue l failure. This list is considered to be highly tentative. per- , haps including plants that do not truly exhibit all of the req-utsite conditions and excluding otner plants that do. ; r F. Given that existing steam generator inspection programs and leak rate monitoring programs may be ineffective in preventing a rap-idly propagating fatigue crack from causing a rupture, then ,' plants containing the requisite conditions for such failures may ! need to take additional actions to ensure continued compliance l with 10 CFR 50 Appendix A (GDC 14) ud Appendix B. The bulletin I requests that licensees and applicants confirm that their plants ! do not contain the requisite conditions. Any licensees with l I plants satisfying all the requisite conditions are requested by the bulletin to take appropriate corrective and/or compensa-
- tory measures.
IV. The proposed method of implementation along with the concurrence (and any comments) of OGC on the method proposed. A. Methods of implementation are described helaw where they ere not ! already implicit in the requirement sumaries provided in re-sponse to Item I.
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- 8. Recommended action 1.A - The presence of denting is to be determined from a review of eddy current test data. Inspection records may be considered adequate for this purpose if at least l 3 percent of the total population of steam generator tubes was
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inspected at the uppemost support plate elevation during the last 40 months. Where inspection records ars nnt adequate for this purpose, such inspections should be perfonned at the nest refueling outage. C. Reconnended action I.C.1 - Prescriptive interim criteria for an enhanced primary-to-secondary leak rate monitoring program are not given in the bulletin although it would most likely invnive more frequent monitoring and trending of rrimary-to-secondary leakage and reduced administrative limits on allowable leakage before plant shutdown must be connenced. In iteu of prescriptive criteria, the bulletin ide.ttifies performance criteria that must be satisfied before a licensee's program may be considered ude-quately effective. These criteria are provivd in sufficient detail in the bulletin to pemit the licensee to assess whether its program is adequately effective.
- 1. Detailed 3-dimensional thermal-hy.1raulic analysis to accu-rately assess flow parameters and fluid-elastic stability .
ratios in accordance with specific guidance provided in the
- bulletin for tube locations of interest. ,
- 2. Assess depth of penetration of each antivibration bar (AVB) into the tube bundle to (11 establish which tubes are not .'
effectivelysupportedbyAYBsand(2)assessflowpeaking factors. This assessment would most likely be accomplished through an assessment of eddy current test data for all tubes of interest such as was done at North Anna Unit 1 Indian Point Unit 2, and Point Beach linit 2. In general, this assessment can be performed on the basis of data from previous baseline eddy cur ent test inspections and/or pre-vious inservice eddy current test inspect *'4 In some cases, however, it may be necessary to per'e , additional eddy current test inspections at - he next scheduled inservice inspection in order to A ve sufficient data.
- 3. Alternettve approaches and/or compensetory measures implemented in lieu of the actions in !Y.C.1 or 2 above may be implemented where ,iustified by the licensee.
D. Reconnended action I.C.2 - An acceptable method for implementing this action includes the followir.g: E. 0GC has no legal objection. ; V. Rcgulatory arilyses generally confoming to the directives and guld- , once of NUREG/2R-0058 and NURE0/CR 3568 (Make suffletent to address the Paperwork Reduction Act, the Regulatory Flexibility Act and Executive Order 12201.) ; A. This *equest for information was approved by the Offfee of Man-agemera and Budget under blanket clearance number 31500011 as , meeting requiraaents of the Paperwork reduction Act and Executive Order 12291. Sufficient hours are included in the NRC t budget for this request. t 4 l
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B. Since this is not a rulemaking action, the Regulatory Flexibility Act does not apply. Yl. Identification of the category of reactor plants to which the Eneric requirement or staff position is to apply (that is, whether it is to apply to new plants only, new OLs only, OLs after a certain date, OLs before a certain date, all Ols, all plants under constraction, all plants, all water reactors, all PWRs only, some vendor types, some 4 vinta etc.)ge types such as BWR 6 and 4. .iet pump and non,iet pump plants, A. As oescribed in Iteo I above, this proposed bulletin is ad-dressed to all pressurized water reactor Ifeensees or pemit holders with Westinghouse steam generators with carbon steel plates -- Models Nos. 13, 27, 44, 51, D1, 02, 03, and D4 and the model E steam generators at South Texas Unit 1. B. The recommended actions vary according to whether a plant finds denting of the tubes at the tipper support plate. This is .. __ discursed in itse I above. VII. For each such category of reactor plants, an evaluation which demon-i strates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The evaluation shall docu-ment for consideration information available concerning any of the , following factors as may be appropriate and any other infomation relevant and material to the proposed action: Response to this ites is not required pursuant to Revision 4 of the CRGR Charter, Section !!!.D. Since the recomended actions of the proposed bulletin are intended to bring facilities into confomance with the rules of the Comission. VIII. For each evaluation conducted pursuant to 10 CFR 50.109, the propos-ing office director's determination, tocether with the rationale for the deternir.stion based on the considerations of all the above, that A. there is a subst ential increte in the overall protection of public health and safety or the comon 0?fanse and security to b derived from the rrcpotal; cnd
- 1. SGTRs constitute a meior degradation of essential safety-related equipt nt; namely the primary coolant bound-a ry. Although such events, like LOCAs in general, are within the design basis, such events constitute a ma,1or reduction in the degree of protection of public health and ,
safety. Such events have been considered reportable to ' Congress as abnormal occurrences pursuant to an NRC policy statement published in the federal Register on February 24, 1977 (Volume 42 No. 37, 7p.10950-10952). .. 5
- 2. The proposed bulletin is intended to identify and remedy conditions that may cause a relatively smal! number of plants to be highly susceptible to a rapidly propagating fatigue crack mechanism. Thus, the staff concludes that the bulletin may result in a substantive increase in the overall protection of public health and safety.
B. The direct and indirect costs of implementation, for the facili-ties affected, are justified in view of this increased protection. '
- 1. Direct and indirect costs associated with the actions re-ouired by the bulletin involve primarily the conduct of analyses to detemine the degree of susceptibility to the fatigue crack mechanism and enhanced surveillance of prima-ry to secondary leakage pending coupletion of the analysis. ;
A relatively small nus6er of plants (i.e., fewer than 10) may find it necessary to plug a small neber nf tubes. ' Preliminary estimatas of potential costs are provided in .
., Table 1. .
- 2. Although the costs
- implementing the actions required by the bolle'.in have 5.. . been estimated in detail by the i
staff, the costs are expected to be very modest (see Table 1) and justifiable in view of the increased protection to public health and safety. Furthemore, these costs, industry wide, should be largely, if not totally, offset by avoided costs that could be expected to accompany any recovery activity associated with an SGTR tvent. Such avoided costs would include costs of an extended plant shutdown that would be inevitably associated with an SGTR event.
- 3. A preliminary estimate indicates that occupational radia-tion exposures associated with iglesvntation of this bul-letin are negitgible. Table 2 summarizes the basis for this estimate.
IX. For each evaluation conducted for proposed relaxations or decreases in current requirements or staff (regulatory) positions, the propos-ing office director's detemination, together with the rationale for the detemination based on the considerations of all the above, that the pubite health and safety and the consnon defense and security would be adequately protected if the proposed reduction in require-ments or (regulatory) positions were implemented, and the cost say-ings attributed to the action would be substantial enough to .iustify taking the action. This item 1s not appitcable to the proposed bulletin since no i relaxation or decrease in current requirements is being proposed. 6 I
l Table 1 Cost impact One Time Recurring Costs Costs item Per Plant Per Plant Remarks A. $5000 Negligible Conservatively assumed tnat licensee will employ contractor at rats of
$500/ day for 10 days to review inspec-tion data and document results. In most instances, it is expected that licensees can confirm the absence of denting without a detailed review of eddy current tapes and at less expense -
than has been assumed here. B. $5000 Negligible No more than 80 person hours should be necessary to make the procedural char;g-es which are relatively minor. Fur-therwere, the procedural changes should not increase the overall time required to implewent inspection programs in accordancE with existing requirements. Thus, there should be no net impact on recurring costs. C.1.a. $83,000 None This item is estimated to be applicable to ten or fewer plants. The - estimated costs are based on informa-tion provided by one utility who has already perfomed these analyses. C.F.b. $30,000 hone This item is estimated to be applicable to ten or fewer plants. The estimated costs are based on informa-tion provided by one utility who has already performed these analyses, t i 7
Table 1 (continued) One Time Recurring Costs Costs Item Per Plant Per Plant Remarks C.2 $20,000 $33,000/yr. This item is expected to be applicable to 10 or fewer plants for no more than ' 18 months. The one-time cost estimates are based on an assumed 40 person-days needed to develop and document enhanced leakage monitoring procedures (assumed
$500/ day). The recurring cost is also expected to be small since adequate instrumentation and staff are already expected to be present. The mater im-pact of this requirement will involve ,
- more frequent recording ond tracking of leakage rates. The staff has censerva-tively assumed that this may involve approximately three additional person-months per year. This item is not expected to significantly increase the probability of an unschecJ1ed out-age to repalr a primary to secondary leak; rather it will most likely neces-sitate an earlier shutdown than would otherwise be the case, given a primary to secondary leak.*
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'The costs of performing an unscheduled outage to repair a leaking tube is dominated by replacement power costs. A staff study performed as part of the USI A-3, 4, 5 studies in 1983 indicated that net replacement power costs were approximately $0.5M/ day. Outage durations would typically range between 2 and 14 days, resulting in $1M to $7M in replaces?nt power costs.
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P Table 2 Occupational Radiation EFD0sure (ORE) Item ORE Impact Remarks A Negligible it is expected that data from previous inspections will generally be adequate to deter-mine presence of denting B Negligible involves administrative procedures only C.1 Negligible This item is expected to affect ten or fewer i plants. AYB locations can generally be deter-mined from data from previous inspections. How-ever, if past records are inadequate and - F additional tube inspecticns are necessary at a given slant, the resulting ORE impact is estimat-ed to w 3 to 10 person-ree on an average per plant. Corrective actions, if found to be neces-sary as a result of the analyses, will Itkely involve the plugging of a small number of tuks with minimal ORE impact (i.e., no more than 1 or 2 person rems) C.2 Negligible This item is expected to affect ten or fewer plants. Yalue-impact studies perfonned as part of U5!s A-3, 4, and 5 have shown that primay to
- secondary leak rate monitoring programs involve U negligible ORE. ~
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' Enciesuro 1 OMR No.: 31500011
,- NRC8 88-rx UNITED STATES NtfCLEAR REGULATORY COWISSION ,
OFFICE OF NUCLEAR REACTOR REGULATION l WASHINGTON, D.C. 20555 January xx, 1988 NRC BULLETIN NO. 88-XX: RAPIDLY PROPAriATING FATIGUE CRACKS IN STEAM GENERATOR TUBES 1 Addressees: For Action - All holders of operating licenses or construction permits for ! Westinghouse (W)-desioned nuclear power reactors with steam generators having carbon steel support plates. Steam generators in this category include Westinghouse models 13, 27, 44, 51, 01, D2 D3 and D4, and the
- Westinghouse model E steam generators at South Texa,s Unit 1. "
i For Information - All other holders of operating licenses or construction i permits for utstinghouse (W) and Combustion Engineering (CE) designed nuclear power reactors.
Purpose:
The purpose of this bulletin is to request that holders of operating licenses ' or construction pemits for Westinghouse (W)-designed nuclear power reactors with steam generators having carbon steel support plates implement actions ' specified herein to minimize the potenth1 for a steam generator tube rupture ! event caused by a rapidly propagating fatigue crack such as occurred at North Anna Unit 1 on July 15, 1987. _ Description of Circumstances: i on July 15, 1987, a steam generator tube rupture event occurred at Ncrth Anna l Unit 1 shortly after the unit reached 1001 power. For several days prior to l the event, operators had observed erratic air ejeetor radiation monitor read- ' l ings. Grab samples were taken prior to the tube rupture for purposes of ; i performing environmental release calculations. Subsequent analysis of this i i data indicated that increasing primary to secondary leakage had occurred over a 24- to 36-hour period before the tube rupture event. This leakage had been below the limit given in the Technical Specifications. The ruptured tube was j ! located in Row 9 Colusri 51 in steam generator "C." The rupture location in ! d this model 51 steam generator was at the top support plate on the cold leg side d j of the tube. The rupture extended circumferential1y 360* around the tube. f, U The cause of the tube rupture has been determined to be high cycle fatigue. ! l The source in the of the tube and loads is believed a superimposed to be a combination alterrating stress. (Theofmean a mean stress level ; ' stress is pro- j' deced by denting of the tube at the uppermost tube support plate, and the ;! alternating stress is the result of out-of-plane deflection of the U bend ! 4 i
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MpCB 88-XX January xx.1988 Page 2 of 5 portion of the tube above the uppermost support plate, caused by flow-induced vibration.) Denting also shifts the maximum tube bending stress to the vici-nity of the uppermost tube support plate. These loads are sufficient to produce fatigue in an all volatile treatment (AVT) water chemistry environtrent. The specific mechanism for the flow-induced vibretion has been determined to be a fluid-elastic instability. The fluid-elastic mechanism has a significant effect on tube response in cases where the fluid-elastic stability ratio equals or exceeds 1.0 The stability ratio SR, is defined as: SR = Y,9f/Ve where V,ff is the effective crossflow velocity ande V is the critical vclocity beyond which the displacement response to the tube increases rapidly. The most significant contributor to the occurrence of a high fluid-elastic .. stability ratto is believed to nave been either (1) a reduction in darping at the tube-to-tube support plate intersection cr*Jsed by denting and/or (?) locally high flow velocities caused by non-uniform antivibration bar (AVB) penetrations into the tube bundle. In either case, the presence of an AVB support will restrict tube motion and thus preclude the deflection amplitude required for fatigue. The cricinal design configuration required AVRs to be inserted to Rcw 11. However, inspections have shown that some AYBs in the North Anna Unit I steam generators penetrate to Row 8. exceeding the minimum AVB design depth. However, no AVP support was present for the Row 9 Column 51 tube that ruptured. _ Discussion: Based on available information, the staff concludes that the presente of all the following canditions could lead to a rapidly propagating fatigue failure such as occurred at North Anna: (1) denting at the upper support plate (2) a fluid-elastic stability ratio approaching that for the tube that ruptured at North Anna (3) attence of effective AYB support Recomended Actions: {; Within 45 days following receipt of this bulletin, addressees having Westinghouse steam generators with carbon steel support plates are requested to submit a written report detailing the status of their compliance with the , actions specified below for purposes of minimizing the potential for rapidly j, propagating fatigue failure such as occurred at North Anna 1 The weport should include an appropriate schedule for completion of the analyses described under item C below, if applicable. Il
l l NPCP 88-xx i January xx,1988 Page 3 of 5 A. The rest recent steam generator inspection data should be reviewed for evidence of denting at the uppennost tube support plate. Inspection records may be considered adequate for this purpose if at least 37 of the total steam generator tube population was inspected at the uppertmst support plate elevation during the last a0 calendar months, "Dent. g should be considered to include the buildup of magnetite in the tube to-support plate crevices, regardless of whether there is detectable distor-tion of the tubes. The results of this review should be included as part of the 45-dsy report. Where inspection records are not adequate for this purpose, inspections of at least 31 of the total steam generator tube ; population at the uppermost support plate elevation should be performed at the next refueling outage. The schedule for these inspections should be included as part of the 45-day report and the results of the inspections should be submitted within 45 days of their completion. Pending comple- I tion of these inspections, an enhanced primary-to-secondary leak rate t monitoring program should be implemented in accordance with paracraph C.I. . .
.below, ,
p B. For plants where no denting is found at the uppermost support plate, plant procedures should be modified as necessary to ensure that the results of future steam generator tube inspections will be reviewed for evidence of denting at the uppermost support. The NRC staff recommends that plant . procedures should also be modified to state that if denting is found in ' the future, the provisions of item C below will be implemented. Confired tion of implementation of these procedures should be submitted when the results of A above are submit'.ed. C. For plants whe*e denting is found, the NRC staff reco e nds that the following actions be taken:
- 1. Pending completion of the NRC staff review an! approval of the program described in C.? below or completion cf inspections specified in item A above to confir1n that denting does not s vist, the NRC staff i recomnds that an enhanced primary-to-secondary leak rate monitoring program should be implemented as an interim compensatory measure within 45 days of the date of receipt of this bulletin.* Implementa-tion ot' this program should be documented as part of the 45-day report. The enhanced monitoring program is intended to vnsure that !
t if a rapidly pmpagating fatigue crack occurs under flow-induced vibration, the plant power level would be reduced to 50% power or less at least 5 hours before a tube rupture was predicted to occur. The effectiveness of this program should be evaluated soainst an assumed time-dependent leakage curve, sur.h as given in Figure 1.
*While this bulletin war being prepared, licenstes for a few plants comitted to an enhanced primary-to-secondary leak rate mo,titoring program at the staff's [
request. These plants had been identified on a p'eliminary basis by I. t Westinghouse as being potentially susceptible to rapidly propagating fatigue cracks. These enhanced programs should be upgraded as necessary to comply with this paragraph. However, no relaxation from current commitments should be made , I without prior approval by the NRC staff.
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NRCB 88-n January xx, 1988 \- Page 4 of 5 This prograr should consider and provide the necessary leakage reasurement and trending methods, time intervals between measure-wents, alarms and alarm setpoints, intermediate actions based on leak rates or receipt of alams, administrative limits for conrnencing plant shutdown, and tire limitations for (1) reducinq power to less than 50% and (?) shutting down to cold shutdown. Appropriate allow-ances for instrument errors should be considered. Finally, the program should make provision for out of service radiation monitors, including action statements and compensatory measures.
- 2. A program Gould be implemented to minimize the probability of a rapidly propagating fatigue failure such as occurred at North Anna Unit 1. The need for loro-term corrective actions (e.g., preventive plugging of potentially susceptible tubes, hardware, and/or opera-tional changes to reduce stability ratios) and/or long-tem compensa-tory suasures (e.g., enhanced leak rate monitoring program) should be ,*
assessed and implemented as necessary. An appropriate program would include detailed analyses, as described in subparagraphs (a) and (b) below, to assess the potential for such a failure. Alternative approaches and/or compensatory measures implemented in lieu of the actions in subparagraphs (a) or (b) below should be justified. Although the 45-day report should provide a eles indication of actions proposed by Itcensees, including their status and schedule, a detailed description of this program and the results of analyses should be sutwitted subsequently, but early enough to pemit NRC staff review and approval prior to the next scheduled restart from a
- refueling outage. Where the next such restart is scheduled to take l place within 90 days, staff review and approval will not be necessary
, prior to restart from the current refueling outage. An acceptable l schedule for submittal of the above information should be arranged with the NRC plant project manager by all Itcensees to ensure that the staff dill have adequate time and resources to complete its review without adverse impact on the Itcensee's schedule for restart. (a) The analysis would include a 3-dimensional therinal-hydraulic model that is sufficiently refined to accurately assess flow parameters and stability ratios at all tube locations (i.e., rows and colvens) of interest. The performance of this analysis should be similar to that for the tube that ruptured at North Anna. This would allow a direct comparison of the results of the plant-specific analysis fin terms of bundle flow and stabil-ity ratio) with that for North Anna. (Forexample, damping coefficients between ti,e two analyses should be determined on a consistent basis.) (b) The analysis would include an assessment of the depth of pene-tra*, ion of each AVB. The purpose of this assessment'is twofold: (1) to establish which tubes are not effectively supported by AVBs and (2) to permit an assessment of flow peaking factors. l l .
NRCB 88-xx ilanuary xx 1988 Page 5 of 5 (Note: Most steam generators have at least two sets of AYBs. This requirement applies only to the sct that penetrates most deeply into the tube bundle.) The methodology used to determine the depth of penetration of each individual AVB should be described in detail in the written report. The criteria for detemining whether a tube is effectively supported by an AVB should also be identified. (Note: An AVR that penetrates far enouch to produce an eddy current sirjnal in a given tube may not penetrate far enough to provide a fully effective lateral support to that tube.) If addressees cannot perform this suggested approach or meet this suggested schedule, they should ,iustify to the NRC their alternative approaches and schedules. j, Pursuant to Section 182a Atomic Energy Act of 1954, as anended and 10 CFR 50.54(f), the written reports required by this bulletin shall be submitted to . the Eppropriate Regional Administrator under oath or affirmation. In addition, i.' the original of the cover Ictter and a copy of the reports should be transmit- i i ted to the U.S. Nuclear Regulatory Comission. Document Control Desk Washing-ton, D.C. 20555 for reproduction and distribution. This request for information was approved by the Office of Managevent and ' Bubget under blanket clearance number 31500011. Coments on burden and dupli-cation may be directed to the Office of Management and Pudget, Reports Manage-I ment, Room 3208, New Executive Office Building, Washington, D.C. 20503. The NRC intends to review the information collected under this bulletin and determine the adequacy of specific actions proposed by each licensee. The g information will be analyzed and placed in the NRC Public Document Roomt. ; If you have any questions about this matter, please contact the technical contacts listed below or the Regional Administrator of the appropriate NRC regional office. ; l Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contacts: Emett Murphy, NRR {, l (301) 49?-7632 ; Keith Wichman, NRR (301)492-8345 Attachments:
- 1. Figure 1 Leak Rate Versus Thne Chart
- 2. List of Recently Issued NRC Bulletins
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~N OCB 88dx January xx, 1980 Page 5 of 5 (Note: Most steam generators have et least two sets of AVBs.
This requirement applies only to the set that penetrates most deeply into the tube bundle.) The methodology used to determine the depth of penetration of each individual AVB sisould be described in detail in the written report. The criteria for detemining whether a tube is effectively supported by an AVB should also be identified. (Note: An AVR that penetrates far enouch to produce an eddy current signal in a given tube may not penetrate far enough to provide a fully effective lateral support to that tube.) If addressees cannot perform this suggested epproach or meet this suggested schedule, they should justify to the NRC their alternative approaches and schedules. Pursuant to Section 182a, Atomic Energy Act of 1954, as amended and 10 CFR 50.54(f), the written reports required by this bulletin shall be submitted to the appropriate Regional Administrator undet' oath or affirmation. In addition,
- the original of the cover letter and a copy of the reports should be transmit-
- ted t'o the U.S. Nuclear Regulatory Comission, Document Control Desk, Washing-ton, D.C. 20555 for reproduction and distribution.
This request for information was approved by the Office of Manaoement and Budget under blanket clearance number 31500011. Comments on burden and dupli-cation may be directed to the Office of Panagement and Budget. Reports Manage-ment, Room 3208, New Executive Office Building, Washington, D.C. 20503 The NRC intends to review the information collected under this bulletin and detemine the adequacy of specific actions proposed by each licensee. The information will be analyzed and placed in the NRC Public Document Rooms. If you have any questions about this matter, please contact the technical " contacts listed below or the Regional Administrator of the appropriate NRC - regional office. l l Charles E. Rossi, Director Division of Operational Events Assessment l Office of Nuclear Reactor Regulation Technical Contacts: Emett Murphy, NRP (301)492-7632 i Keith Wichman, NRR (301)492-8345 Attachments: ,
- 1. Figure 1 Leak Rate Versus Time Chart '
- 2. List of Recently Issued NRC Bulletins
*SEE PREVIOUS CONCURRENCES D/DOEA:NRR *C/0GCP:00EA:NRR*EMTB: DEST:NRR CERcssi CMerlinger CYCheng 1/ /88 1/05/88 12/08/87
*0GCB:00EA:WPJt *EMTB: DEST:NRR *EMTP: DEST:NRR *EAD/ DEST:NRR *PPMB: ARM SOMacKay EMurphy KWichman JRichardson TechEd 12/06/87 12/08/87 12/08/87 12/08/87 12/9/87
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