ML19210C763
| ML19210C763 | |
| Person / Time | |
|---|---|
| Issue date: | 07/12/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1636, NUDOCS 7911200072 | |
| Download: ML19210C763 (49) | |
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,,, / 4 } h ACRS SCsCCMMIm.E CN REACCR DJEL MAY 9, 1979 WASHI!GTCN, D.C.
On May 8, 1979, the ACRS Feactor Fuel Subecmmittee met in Washington, D.C.,
to discuss various items concerning NRC actions on fuel-rated isrues. Notice of dis meeting apoeared on the Federal Register, en M::nday, April 23, 1979.
There were no requests for oral or written statements from memcers of the public and ncne were made at de meeting.
Attachment A is a copy of de meeting agenda. The attendees list is Attachment S.
Attachment C is the tentative schedule for the meeting. Selected slides and handcuts are Attachment D to these minutes. A complete set of slides and handcuts is attached to de office ccpy of these minutes.
CPEN SESSICN (9:30 a.m. - 4:35 p.m.) IN'GCDUC* ION Cr. Shewmen, Subcommittee Chairman, called the meeting to order at 3:30 a.m.
The Chairman explained the purpose of de meeting, and mies and precedures of conducting de meeting, pointing cut that Cr. Themas G. McCreless was de Cesignated Federal Employee in attendance.
Dr. Shewmen said that concerning the discussions of extended fuel burnup, he was interested in focusirg on de NRC regulatory requirements and whether er not de exterded burnup programs will supply the information required by NRC.
Or. Shewmon introduced Dr. Ralph Meyer of de Reactor Ebel Section of de NRC Division of Systems Safety and he cpened de day's presentatiens.
NRC PJEL LICENSI!G CRITERIA IN THE S*ANDARD Rf'CEW PLAN - R. MEYER O!RC-CSS)
Cr. Nyer described the new criteria applied to fuel designs, noting that dese sa:.e criteria wculd be applied in de extended burnup range (>30,C00 %D/M:"J).
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Reactor Fuel May 9, 1979 The objectives of 2e fuel svstem safety review are to assure that:
(1) the fuel system is not damaged as a result of nor.a1 cperatien and antici-pated opestional occurrences; (2) fuel system damage is never so severe as to prevent centrol rod insertion when required; (3) the number of fuel rod failures is not underestimated for postulated accidents; and (4) coolability is always maintained.
Cr. w yer discussed the origin of these four ebjectives e
(Figures Dl-4).
Cr. Meyer defined fuel red failures as a cladding defect that allows the escape of fission gas.
Dr. Shewmen asked if given a Three Mile Island type transient, wuld high burnup cladding be more amenable to balicening?
Or. Meyer replied that radiation effects saturate cut around 5,000 M4/r) and no enange in ballocning behavier is expected at high burnup.
Dr.
Be:hent (ACRS censultant) noted that red balicening also depends en de amount of fissien gas in the red, and it is possible that blewout-type failures may be seen because of the high internal pressures due to enhanced fissien gas release.
Or. Meyer replied that. fissien gas release at high burnup is being studied by NRC.
Dr. Meyer reviewed the varicus damage mechanims (Figures 05-7).
In respense to a questien frca Dr. Sement, Or. Meyer emphasized the need for transient testing of fuel r:ds exposed to high burnup in order to estaclish safety margins.
Dr. Meyer sc11 cited tne Subecemittee's cccments on a procesal to reccmmend the vendors include segmented test rods in lead test assemolies for use by
.NRC in a test reacter (e.g. FSF) for tests on high burnup fuel.
Cr. Shewmen noted that fuel failures have been seen recently in the Lacrosse EWR and Yankee Rcwe PWR. As a result of further discussien, it was decided that 2ese two tcpics w uld be discussed in later presentatiens.
1372 197
Reactor Fuel May 8, 1979 OEPARO' INT CF ENEPG (DCE) EXTENCED PJEL UTILIZATICN PROCPAM - P. I.ANG (DCE)
Dr. Lang said that he would discuss the objectives of the DCE.Nel Utilization Program as well as detail the various projects, either underway or planned.
Normally DCE uses the utilities as prime contractors in these projects, with the vendors acting as subcontractors.
Or. Lang said that the objectives of de CCE Program are to increase uranium utiti:stion, increase power plant productivity, and decrease radiation doses to workers. Se principle focus of the program, however, is high burnup, since this item offers the greatest potential for uranium saving (10-20%) in the near term (by 1963). It was noted however, that there are other improve-ments that can be made to increase uranium utili:stion, such as lattice changes, spectrum shift, and end of cycle stretch-cut, among others (Figure D-8).
Dr. Lang said preliminary estimates indicate that the optimum par burnup is in de range of 45 to 50,000 %D/rJ: optimum SWR ::urnups appear to be in the range of 40 to 45,000 %D/rJ.
Dr. Shewmon asked if de burnup wre increased by approximately 50%, wuld the enrichment have to be increased by a similar amount.
Dr. Lang dicated the enrichment increase would be slightly less than 50% in this case (Figures C9-10).
Technical issues identified by DOE that must be addressed for high burnup fuel include: pellet-clad interaction (PCI) fission gas release and fuel red pressure, corrosion and hydridi.ng, dimensional and structural c~.anges, and in-core fuel management design.
Current DCE projects identified by Cr. Lang include a Consumers Pcwer/Ex. von /
Batt'elle EHL project to demonstrate PCI remedies (annular pellets, annular pellets with graphite coated clad, packed particle fuel), a Commonwealth Edison Project designed to demonstrate PCI barrier and/or clad liner design, a Cuke Pcwer/ Arkansas Power and Light Project to demonstrate 45,000 %D/rJ PAR fuel, and a WA/GE Project, still under negotiation, to improve uranium utilization (saa Ficures D11-14).
Dr. Lang noted that it is DCE's intent to 1372 198
Reactor Fuel.ty 3, 1979 have five programs dat parallel de Duke Power / Arkansas Power and Light /3&W Program that -ould involve the remaining four fuel verders.
EXXCN NUCSR PRESDITATICN - K. '4CCS Before beginning his presentation, Dr. Woods resconded to Cr. Shewmon's questien concerning the fuel failures seen at Yankee Row.
Cr. Woods noted that during a fuel tutage, about two years ago, a plenum spring was found in de reactor.
Subsequent examination did not reveal any damage to the fue2, and the reactor ran for anotner cycle. Daring a fuel cutage last October, the fuel was again examined, following a modest increase in coolant activity. B is examinatien revealed 2ree assemblies suspected of damage. Tao assemclies ere subsequently identified as damaged, the damage being confined to de upper corner rods.
The damage was a result of fretting. te cause of the fretting and failure is not clear at this time. Speculation centers en possible damage during loading of the fuel, or a cross-ficw of coolant resulting from a leak in the core shroud that impinged coolant on de damaged rods. Se fuel will be examined during the next outage to see if this problem is still occurring.
Exxon reviewed deir fuel performance as of March 1,1979 (Figure C-15). Cver 50% of the Exxon fuel had exceeded 10,000 MWC/r burnup.
Cr. Woods reviewed the special design features dat Exxon believes wii2 alicw extention of fuel burnup. Bese design features include strict moisture control, short disned pellets, and thicker claddim. Exxcn has lead test assemclies (:.TA's) in Big Rocx Point, Cyster Creek, and H.S. Robinson.
Exxen is also participatim in PCI-related deveiccment and test programs, wnich include programs _ sponsored by IPRI at Cyster Creek, as well as pcwer ramp tests at de Studsvick reactor in Sweden (Figure C-16).
Dr. Wood said that Exxon is not planning a core reload of high burnup fuel before 1980, at de earliest.
GDJERAL ELEC'"RIC PRESDTlATICN - J. C@RNLEY Ms. Charnley noted that General Electric is taking a cauticus approach *o extending fuel burnup. GE's prime consideration is to assure the fuel re-liability (integritv) is maintained at high burnup.
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Reactor Fuel May 3, 1979 GE's extended exposure program encompasses three projects. Bese projects include:
(1) a high burnup demonstration project with lead test assemblies at Monticello and Peach Bottem reactors; (2) a uraniun utilization st x!y which includes a DCE/TVA PTogram now under negotiatien, and (3) a barrier fuel program which censist of copper and ircenium liner designs to mitigate PCI.
Four LTA's are under irradiation in Cuad Cities Unit 1 to test various barrier designs.
In response to questions from Drs. Shewmen and Sement concerning reconstitution /
inversicn of fuel bundles, Dr. Lang (DCE) indicated dat inversion of SWR fuel is a long-range (appecximately 20 years) uranium utilization option.
vs. Charnley reviewed the details of the Monticello high burnup demonstration program (Figure D-17). Be Monticello program will eventually take two assemblies to a peak pellet burnup of approximately 50,000 *D/T.
Bere will be octh non-destructive and destructive examinaticn of selected fuel rods during, and af ter de program's lifetime.
Major engineering censiderations dat GE has under study for high burnup include:
fission gas release, :ircaloy corrosion, PCI, assembly dimensional changes, and lack of dat3 cOncerning the effect of these Considerations at high burnups.
Or. Shewmen asked if stress corrosion cracking is more likely at high burnups.
Dr. Meyer (NRC) noted that PCI failures, in general, show a burnup dependence, especially early in life.
%T.STINGECUSE PRESENTATICN - O. BL%' MAN Mr. Burnman ncted that Westinghouse is cccpiling an extensive data base on fuel wid high burnup. W has exposed abcut 20,000 fuel rods to burnups in excess of 36,000 %D/rJ (Figure D-19).
Mr. Surman said dat no failures have been seen dat could be attributed to burnup.
In response to a questien frcm Or. Shewmen, Mr. Burman said that most of the failures dat have occurred were the result of hydride moisture failures, and collapsed cladding resulting frcm the use of icw-density fuel.
Mr. Burnman noted that de cojective of de W high burnup fuel performance program is to provide sufficient and necessary fuel per-formance data to assure that no intrinsic raterials performance limitations exist at high burnups (greater than 50,000 %D/rJ-peak pellet). Westinghouse also wants a assure sufficient information exists, to satisfy recuirements of design 1372 200'
Reactor Fuel
'i-May 8, 1979 margins, technical reliability, and licersin; concerns.
Mr. Burman noted that presently Westinghcuse high burnur experience is greater for fuel reds,than fuel usemblies. Key performance areas that require assessment include fissien gas release and clad corresien (as they affect fuel rods) zircaloy cceponent integrity, and grid relaxation (as it effects fuel asserblies).
W has lead test assemblies in de Zicn and Trojan reactors that are designed to address de potential performance proclems noted above (Figure 0-20).
Or. Shewmen asked if Westinghouse plans to subject any high burnup rods to power ramping.
Mr. Burm.an replied that Westinghouse has subjected rods with arcund 23,000 VWC/.rJ burnup to pcwer ramps and has under censidera-tien a program to subject higher burnup rods to pcwer ramps.
Mr. Burnran also said that he does not believe PCI is subject to a burnup deperdence.
BABCOCK AND WILCOX PRESENTATICN - T. COLF. MAN, J.WILLSE B&W described their extended burnup pregram. Bis program censists of tw subprograms; the first subprogram is a DCE/tuke Power Company /3&W prcgram with the cbjective of qualifying the current design Mark-B assembly for a burnup of 40,000 WD/rJ. Results frem de first program will be fed inte a second DCE/ Arkansas Pcwer and Light /3&W subprogram with de objective of developing a extended t:ctnup fuel design fer a burnup of 50,000.C/rJ.
The second sucprogram will also begin irradiation of lead test assem= lies of tne advanced :urnup design. Figure D-21 shows a diagram of de pro-gram schedules. Both of the programs at the Ccenee ard Arkansas Nuclear i reactors wil2 include post irradiatien examinatien involving ncn-destructive and destructive tests in the S&W het cell facility (Figures 022-23).
Se series of fuel utili:atien studies is being conducted as de first task in de ZE/Arkarsas Pcwer and Light impreved fuel design pregram. B&W shewed an example of the expected benefits of an 18 month cycle with a batch average burnup of 45,300 &D/r) (Figure D-24).
3&W is also conducting a parametic study to determine de sensitivity of beginning-of-life design variacles en end-of-life parameters (Figure D-25).
1372 201
Reactor Fuel May 8, 1979 Mr. Wilise discussed de impact on high burnup fuel assembly design of the licensing requirements in Section 4.2 of the Standard Review Plan. B&W be-lieves tna.: their post irradiation examination program will provide the neces-sary data to answer licensing concerns regarding such parameters as:
fission gas release, fuel assembly and fuel rod bow, densification, PCI, and others (Figures D-26). S&W will also rely on the use of their TACO-II fuel code for predictions of such parameters as fuel temperatures and pin pressures.
CCMBCS*"!CN ENGINEERING PRESENTAT CN - M. ANCREWS (CE)
Mr. Andrea stated that additional data is needed for de understanding of fuel behavior. Programs underway at DCE, NRC, industry organitations suc..
as EPRI, and the lead test assembly programs in power reactors all centri-bute to this geal.
Mr. Andrews noted that fuel reliability has been continually improving, i.e.
less fuel failures are being seen.
It was noted that a true-life limit for
- ircaloy-clad CO fuel r ds has not yet been identified. CE believes that 2
more data is required M prove that 45-50,000 WO/ r) burnups are feasiM a.
Mr. Andrews discussed key technical areas that need to be addressed concerning high burnup. tese items include PCI/ SCC, fission gas release, dimensional stacility of rods and assemolies, external clad corrosion, and burnable poisen requirements. Ccementing en de above tcpics, Mr. Andrews noted eat WU data shew a very icw PCI defect rate belew de 12.5 kw/ft. pcwer level. He also acted that fission gas release data shew no enhancement wi d burnup cut to 30,C00 WD/ rJ.
Mr. Andrews did note that dere is a need to judiciously separate de effects of burnup and temperature en fission gas release rates.
In response to a question frem Mr. Mathis, Mr. Andrews ncted dat dere are dree burnable poison systems in use today. Sey are:
(1) ceramic pellets of aluminum oxide containing 3 C, (2) bcr: silicate glass, and (3) gadolinia in CO
- 4 2
1372 202
Reactor Fuel
-d-May 8, 1979 NFC-CSS FIANS KR REVIrd CF HIGH-BURNUP PJEL - R..vEYER Cr..v yer said that CSS has begun to ask for additional information frcra e
applicants in de areas of on-line monitoring for fuel failures, and im-proved post-irradiation surveillance in the reporting of fuel inspections.
Dr..v yer suggested that NRC support routine fuel surveillance at all e
plants with a mandated " floor" on de recuired examination hardware avail-aole at the plant site.
The NRC review of fuel behavior analysis during postulated accidents was detailed (Figure D-27).
Items of particular _importance included fission gas release at high burnup, PCI as a fuel failure mechanism, R'A fuel damage limits, and fuel rod bewing. NRC has taken, or is taking, licensing action concerning fission gas release and rod bow (Figures C28-29). Be RIA and PCI issues are under close scrutiny by the Staff.
Mr. Cean Houston (NRC-CSS) briefly discussed limits in de regulations, primarily envirormental, that will have to be addressed before high-burnup fuel can be shipced, transoorted, stored, etc.
NRC-CCR LEAD TEST ASSE.v8LY RELCAD SCHE *ULE KR HIGH-BURNUP AND RELA *ED CCR REV!rd PRCBLD'S - F. CCFFvAN (CCR)
Mr. Coffman discussed CCR's review of LTA's and high burnup reload appiications.
CCR has taken a tolerant attitude toward LTA insertion in cperating reactors since, among other reasons, there are usually few assenblies involved and they are put in low-pcwer regicns of the core.
Mr. Coffman noted that for transient analysis, censideraticq must be given to de fact that high-burnup asser211es are
- r. ore sensitive to failures of reactivity control ecmponents at beginning-of-cycle.
At end-of-cycle, the concern for high-burnup asser.clies is tnat operating mrgins (e.g., CPR, CNBR) are reduced.
?JEL FAILURES AT V~~v.vCNT YANKEE AND CONNEC"! CUT YANKEE - J. C'4AFNLEY (GE),
R. LCBELLL (NRC-CCR)
Ms. Charnley discussed the recent fuel failures seen at Verment Yankee. "his is the first case of widescale failure of the GE 3 x 3 fuel. A total of 29 bundles have been identified as centaining failed fuel. Figure C-30 details a chronology of the croblem. Be failure mechanism appears to be spalling of 1372 203
Reactor Fuel My 3, 1979 excessive exterior clad oxide. Most of de failures occurred on gadolina bearing rods. Preliminary investigations have not uncovered any obvious failure mechanism (s), nor ha,s this failure method been obser/ed elsewhere. Hot cell examinations of selected fuel assemblies, both non-destructive and destructive, are planned. Vermont Yankee is currently back in operation, with no indication of additional fuel failures.
Mr. Label (NRC-CCR) discussed the fuel failures obser/ed at Connecticut Yankee (W PWR-1925 PH(t)-stainless steel clad). High of fgas activity lead the verrior to suspect failure of high-burnup fuel. During a recent refueling, 36 of 48 Batch 9 assemblies, which were scheduled to be discharged, were found to be leaking. Se vendor also sipped selected assemblies of the remaining batches in the core and all were found to be sound. Figure D-31 shows the location of the failed assemblies in the core.
Investigations by the licensee showed that Batch S fuel was unicue in two respects:
(1) it was de only fuel in the core fabricated by British Nuclear Fuel Limited; (2) it was the only batch to be subjected to a tw-step process during fuel rod fabrication (Figure D-32). S e applicant has characterized the failures as stress rupture type failures with no evidence of corrosion.
Mr. Lobel noted that, as of this date, the utility has no plans for hot cell examination of the failed fuel.
Dr. Bement questioned de utility's conclusion dat failure was by stress rupture. Mr. Crocker felt dat hot cell examination of de failed fuel was necessary for an accurate determination of de failure mechanism.
Or. Mark was of de opinion dat the fuel itself was suspect.
NEW (PRCPCSED) NRC EXCEPTANCE CRI""ERIA ER PJEL ASSEMELY AUCTJFAL RESPCNSE
'"O SEISMIC AND LCC\\ LCACS - R..vEYER (OSS)
Mr. w yer began by reviewing the history of de bases for combining seismic and e
LOCA loads. He noted that the precedent was set with Nord Anna, upon de discovery of the asymnetric blowdown load problem. Criteria developed at North Anna was applied to Diablo Canyon wi-h scme modification (Figure D-33).
~72 204
Reactor Fuel May 8, 1979 Cr. Meyer noted, however, that with de San Cnofre plant revies, currently in progress, NRC, in demanding a collapsed grid assumption,*.culd provide an incentive to use an inferior grid, rather than the strerg grid now scheduled for use in San Cnofre.
Dr. Meyer reviewed the proposed method for analysis of SSE and LCCA loads (Figure D-34).
In resconse to questien frem Cr. Mark, Mr. Kniel (NRC-CSS) stated that be analyses for the seismic and *CCA loads are done separately.
CSS is proposing not to combine SSE and LCCA leads. CSS's reasoting for not ecmcining these loads centers en the belief that structural failure of fuel during a SSE cannet lead to a LOCA, so there is no need to coccine these leads (SSE and LOCA) for de fuel.
Cr. Mark and Mr. Etherington pointed out that an SSE may induce a LOCA however, and these load may have to be dealt with in tandem.
Mr. Kniel noted that there has never been a concise statement frce the NRC that an ear 2 quake will cause a LOCA.
?.e NRC position has been to assure that a LOCA will not occur, given an earth-quake.
Cr. Mark noted dat the Japanese do not combine SSE and LOCA loads.
Dr. Meyer noted that he was informed of this fact upon a recent visit to Japan.
Cr. Meyer summarized the proposed criteria as f0110ws:
(1) use standard conservative methods, (2) add margin only for identified shortccmirgs, (3) use no safety factor, (4) do not combine loads, (5) use average strength values for grids, and (6) use conservative tcunding strength values for other compenents.
Concerning the schedule of the proposed revision to de Standard Review Plan, Cr. w yer said that a draft of de revision, and a value/ impact study, is e
scheduled for completion on May 1.
A 60 day puclic comment period would 3
begin around August 1 of this year. Since R C is not currently scheduling Standard Review Plan actions, issuance of de revision is uncertain.
- n response to a question frem Cr. McCreless (ACRS Staff), Mr. Meyer noted that
, r,I; 205
t Reactor Fuel May 8, 1979 it is a new procedure for a revisien to the Standard Review Plan to be e djected to p611e cc m.ent.
Themeetingwasadjedenedat4:35p.m.
NCr"E : A copy of a transcript of this meeting is v/ailaole in the NRC P lic Docsent Rocm at 1717 H St., N.W., or can be obtained frem Ace-Federal Reprters, 444 North Capital St., N.W., Wash-ington, D.C.
-} f $
poderal Register / vol. 44. No. 7 / Monday. April 23, 1979 / Notices NUCt. EAR REGut.ATORY COMMISSION Advisory Committee on Reactor Safeguards Subcommittee on Reactor Fuet; Meetinc ne ACRS Subcommittee on Reactor Fuel will hold an open meeting on May 8.1979, in Room 104a.1717 H Street.
N.W. Washington. D.C. 20ss5 to discuss various items concerning NRC actions on fuel-related issues. Notice of 'his meeting was published March 23.1979
[44 FR 17837].
In accordance with the procedures outlined in the Federal Register on October 4.1978. (43 FR 45928), oral or written statements may be presented by members of the public. recordmgs wdl be permitted only during those portions of the meeting when a transcript is being kept, and questions may be asked only by memben of the Subcommittee,its consultants, and Staff. Persons desiring to make oral statements should notify the Designated Federal Employee as far in advance as practicable so that appropriate arrangements can be made to allow the neceuary time during the meeting for such statements.
De agenda for subject meeting shall be as iallows: Tuesday. May S 1979, t.30 a.m. undthe conclusion of business.
He Subcommittee may meet in Executive Session, with any ofits consultants who may be present. to explore and exchange their prelimiwy opinions regarding. matters which should be considered durma the meeting and to formulate a report and recommendations to the full Committee.
At the conclusion of the Executive Session, the Subcommittee will hold discussions with representatives of the NRC Staff, and their consultants, pertinent to this review. The Subcommittee may then caucus to determine whether the matters identifled in the initial session have been adequately covered and whether the subject is ready for review by the full Committee.
Further informatien regardmg topcis to be discussed, whether the meeting has been cancelled or rescheduled. the Chairman's ruling on requests for the opportunity to present oral statements and the the allotted therrfor can be
'obtained by a prepaid telephone call to the Designated Federal Employee for ATTACHMENT A-this meeting. Dr. Romas G. McCreless (telephone 202/834-3287) between 8:15 a.m. and 100 p.m EST.
2 207
'17.NTATIVE SCHEDULE OF PRESENTATIOS ACRS REAC'ICR FUEL SUBCCMMITTEE MESTING MAY 8, 1979 WASHINGTON, D.C.
PRESD4TATICN ACTJAL TIME TIME I.
INTRCOUCTICN 10 min 8:30 m P. SHEWMCN, CHAIR W II.
NRC FUEL LICENSING CRITERIA IN THE STANCARD 30 min 8:45 m REVIEW PIAN R. MEYER, NRC-CSS III.
EXTENDED FUEL BURNUP PRESDITATICNS A.
DCE PRESFNTATICN 45 min 9:30 m P. IANG
- BREAK -
10 min 10:50m B.
EVEL VENDORS PRESENTATIONS 1.
EXXCN PRESENTATICN 20 min 11:00 m J. CWSI2Y AND K. WCCDS 2.
GE PRESENTATICN 20 min 11:30 m J. GARNLEY 3.
WESTINGHOUSE PRESENTATION 20 min 12:00 noen D. BUR W 4.
BABCCCK AND WIIfCX PRESENTATICN 20 min 12:30 ;:m T. Col 2AN 5.
CCNSUSTICN ENGINEERING PRESENTATICN 30 min 1:00 pn M. ANEREWS
- LUNCH -
1:45 pn 1377 208 ATTACHMENT C
- CCNTINUATICN CF -
TENTAT!VE SCHEDULE CF PRESENTATICNS ACRS REAC'ICR FUEL SUBCCNMITTEE MEETDG MAY 8, 1979 WASHINGTCN, D.C.
PRESENTATICN AC WAL TIME TIME C.
NRC PRESENTATICtG 1.
ESS PIANS FCR REVIEW CF HIGH-15 min 2:45 pn SURNUP FUEL R. Meyer - DSS 2.
CCR - LEAD TEST ASSDGLY REIIRD 15 min 3:10 pn SCHEDULE FCR HIGH-BLRNUP AND RE-LMED DOR itrVIEW PRGIIMS F. CCFFMAN - DCR IV.
}RC DCR AND MS LICENSING ACTIVITIES A.
RTL FAILURES AT CONN. YANKEE AND VEr.CNT 30 min 3:30 pn YANNKEE R. IABEL - DCR 3.
NEW (PRCPCSED) ACCEPTANCE CRITERIA FCR 30 min 4:15 pn R'EL ASSDGLY STRUC'IURAL RESPONSE 'IO SEISMIC AND I4CA LCADS R. MEYER - DSS V.
AIUCURN 5:00 :xn 1372 209
ACRS SUBCOMMITTEE ON REACTOR FUEL MAY 8, 1979 WASHINGTON, D.C.
ATTENDEES LIST
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ACRS NRC P. Shewmon, Chainnan M. Tokar H. Etherington, Member C, Powers C. Mark, Member J. Voglewede W. Mathis, Member R. Meyer J. Crocker, Consultant M. Houston A. Bement, Consultant S. Kim T. McCreless, Staff
- K. Kniel P. Boehnert, Staff S. Sands D. Bessette, Fellow R. Lobel
- Designated Federal Employee WESTINGHOUSE GENERAL ELECTRIC G. Antaki D. Burman J. Charnley J. McInerney C. Richard INST. OF RADIATION EXXON NUCLEAR PROTECTION G. Owsley L. Wiliberg K. Woods THE TOKYO ELECTRIC NUSCO POWER C0 M. Pitek H. Hamada YANKEE ATOMIC ELECTRIC C0 Ba'4 W. Metevia T. Coleman CE J. Wills.
J. Tulenko M. Andrews E. Coppola M. Marugg C. Brinkman DUKE POWER EG&G R. Snipes J. Crocker G. Swindlehurst NUCLEAR ASSOCIATES INTERNL.
VE"C0 D. Coleman N. P. Wolfhope RAINBOW FAMILY OF LIVING LIG:-
EMBASSY OF JAPAN M. Williamson g ~-) 210 A. Yuki A. Morishima ATTACHMENT B
1.
The Fuel System is Not Damaged as a Result of Normal
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Operation and Anticipated Operationa! Occurrences 10 CFR 50 Appendix A II.
Protection of Multiple Fission Product Barriers Criterion 10 - Reactor Design
... Assure That Specified Acceptable Fuel Design
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URANIUM UTILIZATION EARLIEST POTENTIAL DESIGN IMPROVEMENT IMPLEMENTATION l
PHASE I PRELIMINARY ALTERNATIVES DATE 1983 FOUR BUNDLE ENRICHMENT INITIAL CORE REDUCED GD RESIDUALIN RELOADS (AXIAL SHAPING) 1981 BURNUP OPTIMlZATION (INCREASED BURNUP) 1982 C
REFUELING PATTERN OPTIMlZATION 1981 Z
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.FOUR LEAD BURNUP BUNDLES EXTENDED ONE CYCLE
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INCLUDES ONE 8X8 SURVEILLANCE BUNDLE DISCHARGE ONE BUNDLE (~40 GWD/STU) AT END-OF-CYCLE 7 o
EXTEND OTHER THREE WITH FIVE LOWER BURNUP BUNDLES o
VISUAL EXAM 0F EXTENDED RODS o
NONDESTRUCTIVE AND DESTRUCTIVE BUNDLE EXAMINATIONS o
OTHER ASSEMBLY COMPONENT EXAMINATIONS o
MONITOR AND STORE OPERATING HISTORY DATA DISCHARGE ONE LEAD BUNDLE (~45 BWD/STU) AT END-OF-CYCLE 8 o
DISCHARGE THREE FOLLOW-0N BUNDLES o
EXTEND TWO LEAD AND TWO FOLLOW-ON BUNDLES o
VISUAL EXAMS AND GAMMA SCAN OF EXTENDED FUEL o
MONITOR / STORE OPERATING HISTORY DISCHARGE REMAINING BUNDLES AT END-0F-CYCLE 9 o
PEAK PELLET EXPOSURE ~50 GWD/STU o
NONDESTRUCTIVE AND DESTRUCTIVE EXAMINATIONS o
EXAMINE OTHER ASSEMBLY COMPONENTS JSC:saw/1032 5/8/79
. 7g o-17
MAJOR ENGINEERING CONSIDERATIONS FISSION GAS RELEASE o
EXTEND DATA BASE / VERIFY PREDICTIONS o
POTENTIAL DESIGN IMPACT ZIRCALOY CORROSION o
EXTEND OPERATING DATA o
POTENTIAL MATERIAL /0PERATION FEEDBACK PELLET-CLAD INTERACTION o
KNOWN ACTIVE FAILURE MECHANISM o
EXTEND EXPERIENCE o
REFLECT BARRIER PROGRAM RESULTS ASSEMBLY DIMENSIONAL CHANGES o
INCONEL RELAXATION o
CHANNEL LIFETIME COMPATIBILITY NUCLEAR CHARACTERISTICS o
VERIFY MODEL PREDICTIONS GAD 0LINIA ENRICHMENT / RESIDUAL CONSIDERATIONS o
JSC:saw/1033 5/8/79 7 pg g _ /$
IIIGl BURNUP RR DIS 01ARGED CYCESOF DISCitARGE No.OF AVERAGE max R00 1-131 PLAta REG.
[XPOSURE CYCE-DATE F/A BuRta;P BURNUP (vCI/CM)
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2 2-2/76 33 31,900 35,200 O.02 3
3 3-2B7 37 36,1ti0 42,800-0.00f1 ZORITA 5
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5-10/77 27 33,180 110,000 0.0fi 5
6 6-10/78 8
32,000 37,300 0.025 6
6 6-10/78 13 37/.;00 110,110 0 0.025 GINNA 5
11 6-407 20 31,000 35,000 0.20 TURKEY Poira 3 3
3 3-11 # 6 52 29,600 3II,600 0.03 TURKEY Poira 4 3
3 3-11/77 11 8 28,900 3tl,300 0.02 u
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3 3-9H8 fil 36,000 42,000 0.03 ZION 1 2
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I OPPORTUNITIES FOR EXTENDED BURNUP ON WESTINGHOUSE FUEL Year item 1977 1978 1970 1980 1981 1982 1983 1984 15x15 Assemblies 39,000 Cycle-2 Cycle-3 Zion No.1 0
47,000 55,000 Cycle-4 Cycle-5 Zion Unit 2
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FEED ENRICINENT 2.8 4.1 SWU REQUIREMENTS,
10S,000 100,000 112,000
+5 U0 REQUIRElENTS (STU 0 )
183 270 164
- 11 l
33 38 UTILIZATION (INY/STU 0 )
11.2 12.5 12.5
+ 12 33 l
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ANALYSIS OF FUEL BEHAVIOR AT HIGH BURNUP
Background
Current Restrictions on Mixed-Oxide Utilization and Recycle Technology Have Placed increased Emphasis on Extended Burnups.
Re0ulatory Interest in Fuel Behavior at High Burnup Includes:
1.
Enhanced Fission Gas Release 2.
Increased Potential for Fuel Failure by PCI J
3.
Rod Bowing 4.
Material Property Changes h
5.
Cladding Collapse 6.
Cladding Axial Growth i
N t
7.
Fretting.and Wear 8.
Fatigue 9.
Oxidation and Crud Deposition
[
i-10.
Relaxation of Springs b
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I ANALYSIS OF FUEL BEHAVIOR AT l
HIGH BURNUP B
Status of Fission Gas Release:
1.
All Vendors Asked to Revise Codes by February 1979.
l i-2.
Westinghouse, General Electric, and Babcock &
Wilcox Have Developed improved Codes.
3.
The Combustion Engineering and Exxon Codes With the N RC Fission Gas Model (Nureg-0418) will be Given Additional N RC Review.
4.
The ANS-5.4 Model is Being incorporated in the N RC F R APCON Code.
~
w 5.
An NRC High-Burnup Standard Problem Set will be 9
Available Soon.
l.
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I FUEL ROD BOWING
.,j
1 Background
9' 1.
Bowed Fuel Rods Alter Local Heat Transfer and Neu tronics.
2.
Early Vendor Analytical Models Were inconsistent and Inaccurate.
3.
Conservative Licensing Models Developed by NRC in 1976 Give Excessive Heat-Transfer Penalties.
i Sta tus
., 1.
NRC Letters to Vendors in Summer of 1978 Agreed to Accept More Realistic Models and Presented Guidelines.
~
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' 2.
New Rod Bow Models are Expected from Vendors During Next 2 Years.
4
VERMONT YANKEE OPERATING HISTORY 0FF-GAS ACTIVITY INCREASE MAY 1978 SEVENTEEN LEAKING RODS FOUND SEPT. 1978 PU\\NT RESTART, LOW 0FF-GAS OCT. 1978 0FF-GAS ACTIVITY INCREASED NOV. 1978 REDUCED POWER DUE TO OFF-GAS FEB. 1979 SHUTDOWN FOR SIPPING (24 LEAKING BUNDLES)
MAR. 1979 RESTART WITH REPLACEMENT FUEL APRIL 1979 OFF-GAS ACTIVITY LOW AND STEADY MAY 1979 JSC:BJW/1024 5/8/79 1372 240 0'$$
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o REl.0ADBATCH PAP #EIS 30.
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FtEL PELLET SwPusR Bal Bei BNFL GIF FtEL Pm FABRICATIm Bal Bal Cow / GWF BD1 CLADDINGSwPUER SwERIOR ILEE CUPNN' t
Ava BwNP 10 24.2 33.8 33.5 (TO I%TE)
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THE DIABLO CAllYON ASSUMPTION l.
PERIPHERAL BUNDLES COULD NOT f1EET NORTH ANNA PRECEDENT, 2.
ASSUMED GRIDS Ill PERIPHERAL BuilDLES WERE FULLY COLLAPSED.
3.
FULLY COLLAPSED GRIDS RESULTED IN ONLY 25 F INCREASE 0
IN PCT.
4.
LOW POWER IN PERIPHERAL BuilDLES REDUCED PCT BY MORE 0
THAN 25 F THUS COMPENSATING FOR GRID DEFORMATION.
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ANALYSIS OF LOADS i
A.
INPUT:
USE OUTPUT OF PRIMARY SYSTEM ANALYSIS THAT PRODUCES LARGEST FUEL LOADS.
B.
METHODS:
REVIEW If!CLUDES SAMPLE PROBLEM.
C.
UNCERTAINTY ALLOWAf!CE:
ADD NARGIN FOR STEAM-FLASHING AND PRONOUNCED SENSITIVITY TO INPUT VARIATIONS.
D.
AUDIT:
flRC WILL AUDIT ANALYSIS OF CURRENT PLANT DESIGN WITH INDEPENDENT CODE.
E.
COMBINATION OF LOADS:
NONE.
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