ML20154L078
| ML20154L078 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/16/1988 |
| From: | Sneizek J Office of Nuclear Reactor Regulation |
| To: | Jordan E Committee To Review Generic Requirements |
| Shared Package | |
| ML20153B009 | List: |
| References | |
| IEIN-88-001, IEIN-88-1, NUDOCS 8805310146 | |
| Download: ML20154L078 (20) | |
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'o UNITED STATES l
NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D C. 20555 i
\\.....J IMY 161353 tel0RA!EQi IVR: Edward L. Jornian, Chairran Ocemittee to Review Generio Requirecente FRCti:
Jazee H. Sniezek, Deruty Director Office of Ibelear Reactor Regulatico SUNECT:
RBWEST EDR REVID4 OF A DRAFT BULLETIH Ctl 1EEFMAL STRESSES IN PIPI!G OctNECTED 1D REACIVR COOLAlfr SYSTDG Cti January 27,1988, the 100 iseued Inforvatico tbtice 88-01, "Safety Injection Pire Failure," which alerted licensees to a potentially generic problem that had occurt,ed at Farley 2.
he problem was the reeult of leakage of relatively cold water through a valve, which cauced thettal cycling and failure of a section of safety injection piping connected to the reactor coolant system.
Be section of failed piping was downstream free the last check valve and ceuld not be isolated free the reactor coolant system. The piping was repaired after the reactor was chut d;un and the reactor coolant system was partially drained.
he problem at Farley is considered to be generically significent tecause it is difficult to ensure that there will never te leakage acrose seated valvee, ikwever, tecause of the ductility of piping rateriale, catastrtthic failure of fatigued piping ir not considered to te likely. Neverthelece, the exrerience at Farley indicatee thst sece water-cooled reactore ray not cceply entirely with General Design Criterico 14 of 10 CFR 50, Aprendix A, which requitta that the reactor ecolant preesure bcundary te deeigned to have an exttteely small protability of atnorral leakage, of rapidly propagating failure, and of groes rupture,
he encleced draft tulletin would ensure that coepliance is achieved. %t uculd te aMreceed to all holdere of orerating licenses and cenetructico remits for water-cooled rower reactore and wculd requent that they take action to precitde eignificant thenal cycling, which might othervice lead to high-cycle fatigue and failure of unicolable piping connected to the reactor coolant system.
he prorceed tulletin and tackah inforratico required by the CfGR Charter are enciceed.
CEtfrACT: IMxdruff, bT6 x21180 IKuo, tGB x20907 c c g r ? ' n l ' ' 6, nL g
I UAY 10 I300 E. L. Jordan 2
l He request that myiew of this package be scheduled at CfGR's earliest con-t venience. h tulletin is sponsored by Charles E. Rossi, Dimotor, Division of Operational Events Assessment.
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James H.# ni
. Deputy Director g) i Office of thelear Beactor Begulation t
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Enclosures:
- 1. NBC Rillotin Ib. 88-XX. hmal Stresses in
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Piping Connected to Beactor Coolant Systems
- 2. CIUR Item IV.B. Contents of Packages Sutaitted
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I Ehclosure 1 UNITED STATES HUQ2AR REULATORY OttMISSION OETICE OF NUCLEAR REACKS REULATICH WASHI N!CH, DC 20555 May XX, 1966 HRC BULIATIN H0. 68-XX: THERIAL STRESSES IN FIPIM 00tWICIED 10 REACICR 00CtANT SYSTEMS A&irmanaam All holders of operating licenses or omstruction posits for light-water-cooled nuclear power m actors.
Pluv=a t he purpose of this bulletin is to request that licensees (1) reviou their mactor coolant systems (MCSs) to identify any wi.4ted, unisolable piping that oculd be subjected to temperature distritutions which would result in unacceptable themal stresses and (2) take action, where sus piping is identified, to ensure that the piping will not be sabjected to maaoeptable themal stresses.
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On %=mber 9,1967, while Farley 2 was operating at 33 percent power, theh licensee noted inemased moisture and radioactivity within omtainment, The source of leakage unidentified leak rate was detamined to be 0.7 asm.
j was a cimunferential crack extending through the wall of a short, misolable section of emergency oors cooling system (9005) piping that is connected to the cold les of loop B in the BQi. 1his section of piping, consisting of a nosale, two pipe spools, an elbew, and a ched valve, is shoun in Figure 1.
l The creek resulted frem high-cycle themal fatigue that was caused by mia-tively oold water leaking through a ciceed globe valve at a pewasure sufficient to open the ched valve. h leaking globe valve is in the bypass pipe around the bomn injection tank (BIT) as shown in Figure 2.
During normal operation this valve and others isolate the 30Gi piping fn:a the discharge pressure of With a charging pap running and the valve leaking, the charging sumps.
temperature strstification oocurred in the EO:S pipe as indicated in Figurs 1.
In addition, temperature fluctuaticas vers found at the location of the failed weld with peak-to-peak amplitudes as large as 70 desmes F and with periods between 2 and 20 minutes.
Di amma h t At Farley 2 dual-purose puse an used for charging the 305 wita coolant free the chemical and volume contml systee kring nomal operation and i
injecting emergency core coolant at high pmasure during a loes-of-ooolant j
Separate runs of piping from these pamps an connected to accident (14CA).
separate nomslee on the BCS piping for normal charging flew, backup char 1 ring
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l MBCB 86-t May xx, 1968
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Pese 2 of 4 j
f flow, and hot-and cold-les ECQi injection and to a nossle on the pusesuriser for auxiliary pressuriser spray. All of these nns of piping, downstream from the last check valve in each pipe, are susceptible to the kind of failure that occurred in the BCC5 piping wisted to the cold les of loop 3.
j In any light-water-cooled power reactor, themal fatigue of unisolable piping connected to the BCS can coeur when the connected piping is isolated by a I
leaking block valve, the pressure usstream from the block valve is higher than l
4 BCS pressure, and the temperature upstream is significantly cooler than RC8 i
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temperature. Because valves often leak, an unrovisued safety guastion may I
exist for those reactors that can be subjected to these conditions, thder i
these conditions, thornal fatigue of the unisolable piping can result in creek
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d initiation as experienood at Farley 2.
Subjecting flamed piping to emoensive i
stessees induced by a seismic event, waterhammer, or one other cause t
conceivably could result in double-ended failum of the pipe.
l General Design Criterien 14 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations requires that the reactor coolant pressure boundary be designed so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. At Farley 2, the pressure f
l boundary failed well within its desian life, J
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Review systems wested to the BCS to detemine whether unisolable j
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eections of piping wested to the RQi can be subjected to atteseos from i
j temperature stratificaticm or temperatum oscillations that could le j
induced by leaking valves and that were not evaluated in the design l
analysis of the piping. For those addressess who determine that there am no unisolable sections of piping that een be subjected to such strisees,
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j no additional actions are requested except for the report required below.
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For any unisolable sections of safety injection piping that may ha're been f
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subjected to excessive thermal streseos, examine nendestructively 'the l
j welds and heat-affected mones in that piping to provide assurance 'duit t
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there are no existing flaws.
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Plan and implement a program to provide continuing assurance that l
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unisolable sections of all piping oonnected to the BCS will not be l
subjected to ocabined cyclic anri static thornal and other stresses that l
could cause fatigue failure during the remaining life of the unit.
This assurance may be provided by (1) mdesigning and modifying these J
j sections of piping to withstand ocabined stressee caused by various loads including temporal and spatial distributiens of tamperature resulting free leakage across valve seats (2) instamenting this piping to detect adverse tamperature distritutions and establishing appropriate t
MRCD 88-May xx, 1988 Pase 3 of 4 I
limiting conditions for operation on tasseratum distributions, or (3) providing means for ensuring thit presoun upstrose from block valves i
which might leak is monitored aM does not exooed K5 pewssum, For operating planta not in extended outases. Action 1 should be ocupleted 4.
l within 60 days of rooeipt of this W11etin, and Actions 2 and 3, if
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zuguitwd, shculd be ocepleted tefore the and of the next refueling cutase.
l If the next awfueling cutase ords within 90 days after receipt of this tulletin, then Actions 2 and 3 any be ocepleted before the end of the following refueling cutase.
For operating plants in extend rd cutases and for planta under constructico, t
Action 1 should be ocepleted within 60 days of receipt of this bulletin or l
before achieving criticality, Aichever is later, and Actions 2 and 3 nhould bs, ocepleted before achieving criticality, unless criticality is
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scheduled to occur within 90 days of receipt of this bulletin.
In that r
case, Actions 2 and 3 should to ocepleted beform the end of the next l
twfuelire cutase.
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Within 30 days of ocepletion of Action 1 endi addressee shall sutait a I
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I 1etter confirming that the act,ien has been ocupleted and describirw the If the review performed under Action 1 indicates l
resulta of the review.
that a potential problem exista, the confirmatory letter shall include a I
schedule for ocepletina Acticns 2 and 3.
hose addressees who deterairse that there are unisolable sections of 2.
piping that can be subjected to stresses from tasperature stratification or temperature oscillations %at oculd in irh by leaking valves and 3
I that worr not evaluated in the design analysis of the piping shall eutait a letter within 30 days of o:apletion of Actions 2 and 3.
his letter
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should confits that Actions 2 and 3 have been ocepleted and describe the actions taken.
The writte reports, muired abcue, shall to addrwssed to the U. S. Nelear Begrulatory Commission. AT1H: Dcoment Control Desk. Washington, DC 20555, f
under oath or affitantion under the parvisiena of Section 182a Ataic F.nergy Act of 1954, as amended. In esittien, a copy shall be sutaitto:1 to the appstpriate Begi.nal A:tministrator.
This zwquirement for information was appstved by the Office of Management and
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Ibdget un#r clearance nutter 3150-0011.
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NBC8 68-May xx, 1966 Pese 4 of 4 If you have any questions regarding this matter, please contact one of the technical contacts listed below or the Begional Administrator of the
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appropriate NBC Begional offios.
l Charles E. Rossi, Dirwetor Division of Operatione3 Eventa Assesament c1fioe of Ibclear Beactor.Nc4ation I
1 Technical Contacts Boger W. Woodruff, HRR t
l (301) 492-1180 Pao Kuo, NRR (301) 492-0907 l
I Attachente:
1.
Fielre 1 - Farley 2 Temperature Data l
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risurs 2 - Farley 2 EOCS 3.
List of Becently Issue.C NRC R111stins l
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TA T
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G FF B
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PO W
SC E
IP M D
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F O E
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PT A
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T B LRA F
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NORMAL CHARGING TO RCS COLD LEG B
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l PRESSURE SAFETY N8JECTION PURAPS.
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i FAHLEY 2 ECCS I
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-m CBGB Item IV.B.
Contenta of Packages Sutaitted to CBGR (Rev. 4, Stello to List 042387, des 41860 342 ff)
The following requiresnents apply for prx>posals to reduce existing r1!quiresnents or (regulatory) positions as well as proposals to increase requirtienents or (regulatory) positions. Each package sutznitted to the CBGR for review shall include fifteen (15) copies of the following infomation:
SUBJECT:
BULLETIN RB3ARDItU THERMAL STRESSES IN PIPItU CONNECTED 'IO REACTOR COOLAtTT SYSTD!S Chwmtion (1):
The proposed generic requirement or staff position as it is proposed to be sent out to licensees.
Resonnse:
The proposed requirements are set fortir. in the tulletin (Enclosum 1).
Question J114, Draft staff papers or other und:.
c.ng staff documents supporting the require-ments or staff positions.
(A cop, of all materials referenced in the document ehail be made available upor r nuest to the CRGR staff. Any committee member may 1 aquest CRGR staff to obtain a copy of any referenced material for his or her ee.)
Renconee :-
1.
Reportable Event 10919 (50.72 Report), December 9,1987, 2.
Preliminaly tbtification PHO-II-87-80, December 9,1987, 3.
Inspection Reports 50-348/87-36 and 50-384/87-36 describing inspections conducted between December 12 and 16, 1987, 4.
Preliminary Notification PNO-II-87-80A, December 14, 1987, 5.
Memorandum from Reyes to Varga, "HPSI Pipe Crack - Farley Nuclear Plant,"
December 18, 1987, 6.
Memorandum free Lainas to Richardson, "TIA - Review of Farley SI Line Pipe Crack," December 28, 1987, 7.
Sumary (dated February 8,1988) of meeting held on January 15, 1988 between NBC and AFCo representatives to discuss the generic implications of a cracked 6-inch safety injection pipe at Farley 2 (TAC 66773),
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NRC Infomation tbtice No. 88-01, "Safety Injection Pipe Failure,"
January 27, 1988, 9.
Event Followup Report 88-018, "Safety Injection Pipe Crack,"
February 29, 1988.
2 Osaation (iii):
Each proposed requirement or staff position shall contain the sponsoring office's position as to whether the proposal would incrtase staff require-ments or staff positions, would implement existing staff requirments or positions, or would relax or reduce existing requirements or steff positions.
Resonnae:_
Action items in the propcsed Mlletin will implement existing regulatory requirements as follows:
Action 1 10 CFR 50.34(a)(3)(1) requires that principal design criteria meet or exceed requirements established in Appendix A to 10 CFR 50 and that these criteria be identified in the preliminary safety analysis report (PSC). Appendix A, General lesign Criterion (GDC) 14, requires that the reactor coolant pressure boundary be designed so as to have an extremely low probability of abomal leakage, of rapidly propagating failure, and of gross rupture.
10 CFR 50.34(a)(4) requiras that the PSAR include a preliminary analysis of the design of systems 10 CFR with the objective of assessing the risk to public health and safety.
50.34(b)(4) requires that the final safety analysie report (FSAR) include a final ant u.31s of the design of systems with the same objective. Notwithstanding thetz existing regulatory requirements, the reactor coolant pressure boundary did leak abnomally because of a leaking valve in another system. The design of the failed unisolable piping at Farley 2 does not comply entirely with GDC
- 14. It is likely that other licensees have similar problems.
Action 2 10 CFR 55a(g)(6)(ii) allows the Connission to require licensees to follow an augmented inservice inspection prtgram for systems and components for which the Cwmission deems that added assurance of structural reliability is necessary.
Action 3 This action item requires those licensees who are not cceplying with GDC 14 to cceply.
Quegtion (iv1:
The propoced method of implementation with the concurrence (and any cooments) of OGC on the method proposed.
Restence:
The methcd of implecentation will be the proposed tulletin (Encloeure 1 to the request for review). 1here are no issues with the tulletin that require CGC concurrence.
3 Question (v):
Regulatory analyses generally confoming to the directives and guidance of NURED/BR-0058 and NUREG/CR-3568.
Response
This is a compliance issue. No value/ impact analysis was made.
I O witinn (vi):
Identification of the category of reactor plarte to which the generic require-ments or staff position is to apply (that is, whether it is to apply to new plante only, new OLs [ operating licenses) only, OLs after a certain date, all OLs, all plants under construction, all plants, all water reactors, all EWRs
[ pressurized water reactors] only, scoe vendor types, some vintage types such as IER 6 and 4, jet pump and nonjet sump plants, etc).
Resonnse:
he pmposed bulletin would apply to all holders of operating licenses or construction pemits for LWRs.
Question (vii):
For each such category of reactor plante, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The evaluation shall document for consideration infomation available concerning any of the following factors as may be appropriate and any other infomation relevant and material to the proposed action:
(a) Statement of the specific objectives that the proposed action is designed to achieve...
Reconnse:
he event at Farley 2 demonstrates that operating conditions can exist at LWRs that result in noncompliance with GDC 14. The objective of the proposed action is to encure that licensees cceply and remin in compliance.
Continuation of Question (vii):
(b) General descrirtion of the activity that would t,e required by the licensee or applicant in order to cceplete the action...
Renennse:
To cceplete the action, licensees witt. piping that dces not meet GDC 14 would to required to provide assurance that GDC 14 is met by (1) redesigning and redifying unisolable sections of piping connected to the RCS so that these sections of piping will withstand combined stresses caused by various loads
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4 including temporal and spatial distributions of temperature resulting from leakage across valve seats, (2) instmmenting this piping to detect adverse temperature distributions and establishing appmpriate limiting conditions for operation on temperature distritutions, or (3) providing means to ensure that upstream pressure is monitored and does not exceed RCS pressure.
Continuation of Omation (vii):
(c) Potential change in the risk to the piolic from the accidental offsite release of radioactive material...
Resenne:
At present, the risk to the p2blic, from the accidental offsite release of radioactive material due to a IDCA, exceeds that which is implicit in GDC 14.
Compliance with the regulations, as required by the proposed bulletin, would reduce the risk to that intended by promulgation of GDC 14.
Centinuation of Question (vtih (d) Potential impact on radiological exposure of facility employees and other onsite workers...
Response
The potential radiological exposure for each action item is:
Action 1, Review of Systems No radiological exposure will result from this action.
Action 2, Nondestmetive Examination of Welds l
At Farley 2, a three-loop unit, the licensee estimated that the accumulated dose for examination of emergency core cooling system piping connected to the three RCS cold legs was 3 to 4 person-rem. Farley 2 also has ECCS piping con-nected to the three RCS hot legs, two charging pipes connected to the RCS cold legs, and one spray pipe connected to the pressurizer. Unisolable sections of piping in each of these pipe mns could be subjected to thermal fatigue caused by leaking valves. For four-loop PWRs, there would be two additional ECCS pipes ns.ing a total of 11 sections of connected piping that might require examination.
For a IVR that ray have had all of these sections of piping subjected to exces-sive therral stresses, the expected accumulated exposure in doing the nondestmetive examination would be approximately 11 to 15 perscn-rem. For BWRs, a problem similar to the Farley 2 problem has not been identified and radiological dose has not been estimated.
Action 3, Elimination of the Potential for Excessive Themal Stresses At Farley 2, the licensea instmmented two ECCS pipes connected to the BCS cold leg and estimated that the accumlated doce was 2 to 3 person-rem.
If 11 secticus of piping connected to the RCS piping and pressurizer at a EHR
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5 were instmmented, then the expected accumulated exposure would be li to 17 person-rem.
Continuatien of Owation (vii):
(e) Installation and continuing costs associated with the action, including the cost of facility downtime or the cost of constmetion delay...
Resonnne:
Actions 2 and 3 are to be ccepleted within 3 nonths to 17 months of receipt of the proposed bulletin for plants with operating licenses. Assuming that a licensee has a four-loop IHR, does the nondestructive examination, installs thenrocouples with po9er supplies that are not environmentally qualified, and plans efficiently, facility downtime would be minimal. Assuming that the plant is a four-loop IHR, 11 lines have 4 suspect welds per line, and 2 persons spend 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to prepare for and examine each weld on the average, then the time required for nondestmetive examination would be 704 person-hours. Assuming that 5 themoccuples are installed on each line and 2 persons spend 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to prepare for and install each themocouple on the average, then 220 person-hours are required for a total of approximately 1000 direct person-hours. Assuming that planning, purchasing, and other indirect costs are 1504, of direct costa, then total time charges would be 2500 person-hours. At $20 per hour, cost for manpower would be $50,000. Assuming material and equipment costs are equal to manpower costs, the total cost for a four-loop IVR with 11 suspect lines with 4 welds each would be $100,000.
Continuation of Question (vii):
(f) The potential safety impact of changes in plant or operational complexity, including the relationship to preposed and existing regulatory requirements and staff positions...
Response
Assuming that a licensee elects to install temperature sensors, a modest increase in cperational complexity will result from the need to monitor the additional sensors and take corrective action when limits are exceeded. However, this will again establish the margin of safety intended in Appendix A, GDC 14.
Continuation of Question (vii):
(g) The estimated resource turden on the NBC associated with the proposed z.ction and the availability of such resources...
Resuoncet Licensees would be required to sutait letters confiming that the required actions have been ccepleted and describing the actions taken. NRR would identify a lead project manager to coordinate the review of the licensees' reports by their project managers. The estimted time required for mview by the staff is 240 percon-hours. No requirement for regional review will be necessary.
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Continuation of Owation (vii):
(h) The potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed action...
Restrnme:
The problem may be relevant to all LWRs. Type, design, and age are ret expected to be significant factors with regard to the practicality of the proposec action.
Continuation of Question (vii):
(i) Whether the proposed action is interim or final, and if interim, the justification for imposing the proposed action on an interim basis.
Response
The proposed action is final.
Owation (viii):
f For each evaluation conducted sursuant to 10 CFR 50.109, the proposing Office Director's detemination tcgether with the rationale for the detemination i
based on the considerations of paragraph (i) through (vii) above that:
(a) there is a sutetantial increase in the overall protection of public health and safety or the common defense and security to be derived from the proposal; and (b) the direct and indirect costs of implementation, for the facilities affected, are justified in view of this increased protection.
Pectense :
Because valves often leak, it is likely as a result of fatigue that cracking of piping connected to the reactor coolant systems in a few plante in addition to
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Farley will occur before end of reactor life if preventive action is not taken.
Based on leak-before-break, it is likely that such cracks will be found and I
repaired before the piping fails ccepletely. However, in the unlikely event that repair is not timely, a severe small-break ILCA could occur. We believe that extenditure by the industry of $10,000,000 is reasonable for the increased protection that would result from the proposed tulletin.
Oientien (ix):
For each evaluation conducted for prorosed relaxations or decreases in current requirements or staff positions, the preresing Office Director's detemination, together with the rationale for the detemination based on the considerations of paragraphs (i) thrcugh (vii) atove, that:
(a) the public health and safety and the ecmon defense and security would be adequs+ely protected if the proposed reduction in requirements or positions were 1:slemented, and
7 (b) the cost savings attributed to the action *:ould be substantial enough to justify takte 'he action.
Response
t Relaxations or decmases in current mquirements or staff positions am not Proposed.
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