ML19254D747
| ML19254D747 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 09/28/1979 |
| From: | Copeland J, Newman J BAKER & BOTTS, HOUSTON LIGHTING & POWER CO., LOWENSTEIN, NEWMAN, REIS, AXELRAD & TOLL |
| To: | |
| References | |
| NUDOCS 7910300051 | |
| Download: ML19254D747 (20) | |
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of 9 5
HOUSTON LIGHTING AND POWER CCMPANY 5 Docket No. 50-466 5
(Allens Creek Muclear Generating 5 Station, Unit 1) 5 APPLICANT'S RESPCNSE TO " JOHN F. DOHERTY'S AMENDED CONTENTIONS NUMBERED 4, 11, 15, 16, 20, 21, 36, 38 and 39 .
Houston Lighting & 'ompany (Applicant) hereby submits the following indiviv al responses to the amended contentions filed by John F. Doherty (Intervenor).
Amended Contention 4. ATdS.
In this second amendment to his contention T on ATds, Intervenor has again failed to present a lhtigable issue. Applicant's original response to this contention stated clearly that "...the Applicant must, and indeed is willing to stipulate, that it will comply with whatever NRC requirements are ultimately established with respect to the ArdS generic issue." Intervenor has presented nothing which might further suggest that Applicant's Ards commitment is inadequate.
Moreover, there is notning in Intervenor'r list of modifications requiring " flexibility of design"
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which suggests that whatever changes are necessary to fully implement the generic resolution cannot be accommodated long before construction on the pertinent components and structure begins.1/ Without this assertion and some reasonable basis for it, no contention exists.
Amended Contention 11. Spent Fuel Pool Loss of Water.
Intervenor contends that the spent fuel pools in the fuel handling building and containment building2 / are subject to a complete loss of all water accident, which could result in melting of fuel rods and subsequent radiation release. The only support given for this po,stulation is a government report entitled, " Spent Fuel Heatup Following Loss of Water During Storage", NUREG/CR-0649, 'Allan S.
Benj amin, March, 1979. In response to this exa,ct same
, a contention, offered by petitioners Madeline and hobert Framson, 1/ Only two items on Intervenor's list -- (b) and (g) --
are actual unincorporated changes required by the alternative systems discussed in NUREG/CR-0460. Three have nothing to do with ATWS whatsoever -- (e), (f), and (i); and four are already capable of accommodation in the ACNGS design -- (a),
(c), (d), and (h).
2/ " Spent fuel can be stored both in the Reactor Building and in the Fuel Handling Building. However, fuel will not be stored in the Reactor Building except during periods of refueling, on a temporary basis." ACNGS PSAR, Section 9.1.2.2.
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both Applicant and Staff noted that there is absolutely nothing in the contention demonstrating any relationship between NUREG/CR-0649 and the spent fuel design for ACNGS.
In fact, NUREG/CR-0649 deals only with the consequences to stored spent fuel after a postulated complete drainage of the storage poci -- ar. e"ent whose likelihood the report concludes "is judged to be extremely low." (p. 12) Further-more, the report notes specifically that "(a)ccident initiation mechanisms the probability of occurrence, the magnitude of radioactive release, or the public consequences are not addressed." (p. 11). Hence, NUREG/CR-0649 does not support any allegation that a spent fuel pool loss of. water accident is a probable event, a dangerous event, or an event which
- threatens public health and safety. Consequenth,the contention should be rejected.
Amendment to Contention 15. WIGLE power excursion theory.
Intervenor argues that the reactivity model WIGLE does not properly account for a " rapid increase in reactivity."
Applicant has twice responded that WIGLE is not used to analyze any event for ACNGS. In an attempt to side-step this clear rebuttal, Intervenor now states that the "one dimensional calculation of the scram reactivity" function that is used by Applicant does not compare with " data
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resulting from the neutron burst experiments reported in IN-1370." The assertion is totally irrelevant because the "one dimensional calculation of scram reactivity" referenced on p. 4-11 of ACNGS SER Supplement 2 is used for calculating the operating transient scram reactivity, not for calculating _
the negative scram reactivity in events which produce a
" rapid increase in reactivity."3/ Intervenor's contention that the one-dimensional reactivity function is used for evaluating events resulting in rapid reactivity insertion is clearly contradicted in NEDO-10527, which states: "The primary design method at General Electric for analysis of super-prompt critcal large-core nuclear excursions uses the adiabatic approximation with a two-dimension multi-group flux calculation." Hence, since Intervneor haq confused
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3/ The worst case example of an event which produces rapid reactivity increases is the rod drop accident. The difference in the operating scram reactivity and the rod drop reactivity calculations are described by the ACNGS PSAR, p. 4.3-26.
"The total scram reactivity and scram fanctions which are used for analyzing the rod drop excursions obviously are quite different in nature than those used for the operating plant transients since the most severe rod drop accidents occur in startup and low power ranges where the void distri-bution is either nonexistent or very dissimilar and the control rod patterns vary grectly from those observed at operating conditions. As was the case for the scram charac-teristics for the plant operational transients, the scram characteristics used for analyzing the rod drop accident are strongly dependent on the fuel design and fuel loading pattern."
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reactivity calculations used during normal operation with those used for particular accident conditions, the conten-tion has no real basis and should be denied.
Amendment to contention 16. Steam blanketing of blocked fuel assemblies.
In his May 25, 1979 contentions, Intervenor contended that " fuel rods will cause steam blanketing of the Emergency Core Cooling System (ECCS) coolant" (Contention 16, May 25 Contentions, at 6). Applicant opposed this contention as either an impermissible challenge to the Commission's ECCS regulations or a nonspecific complaint about Applicant's compliance therewith. Intervenor separately contended that
"[t]he design based [ sic] accident for a flow blockage incident is inadequate because it assumes hlockage of but one fueld assembly." (Contention 25, May 25, Cogtentions at 10. )
In its response, Applicant stated that Contention 25 presented a litigable issue.
Now Intervenor in his amendment to Contention 16 seeks to bootstrap Applicant's acquiescence with Contention 25 into a litigable amended Contention 16 by asserting that steam blanketing will result from flow blockage caused by "an object in the reactor working loose." This attempt fails.
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Intervenor does not state why steam blanketing is t probable, or even possible, result of this postulated loose object in the cores /. Intervenor has provided no information, either experimental o.r theoretical, suggesting that the steam blanketing phenomenon exists, much less whether it should be accounted for in the flow blockage analysis. Without some arguable support for the connection between steam blanketing and flow blockage, stated with reasonable specificity, there is no basis to the new conten-tion and it should be dismissed.
Amended Contention 20. Gap conductance. -
Intervenor has, by amending this contention further clouded an already indecipherable dissertation. Intervenor raises a new allegation that fission gas release due to fuel
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rupture during a LOCA will result in lower pellet clad gap conductance. Applicant is at a total loss to understand how a lower gap conductance can result if a rod has ruptured, thereby releasing its fission gases. The rest of the amend-ment makes no sense in light of this basic misconception.
i/ Intervenor's reference to the incident at Fermi 1 is certainly not helpful, since that reactor was a sodium-cooled breeder reactor and could not possibly have had steam blanketing.
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The only intelligible clue ta whatever issues Intervenor may have in mind is a reference to an article in Nuclear Safety; Volume 20, Number 4; July - Aug., 1979;
- p. 418. The " article" is a letter to the editor from Howard Ocken and J. T. A. Roberts of the Electric Power Research Institute suggesting that fission gas release rates for LWR fuel rods are more dependent on temperature than fuel burnup.
The letter is followed by an answer from tne Commission Staff acknowledging the dominant influence and stating that its analysis accounts for temperature sensitivity. Nothing in the letter or reply correlates with the confusing allega-tions and assertions contained in Intervetor's " contention."5/
In sum, this amendment lacks coherency and basis. It should
.be dismissed. A g.
Amendment to Contention 21. Void collapse reactivity.
This contention relies solely on unsupported speculation. Intervenor claims future information may show that the :eactivity inserted by the collapse of voids during overpressure transients is underestimated. This result may cause the NRC staff to impose technical specification restric-tions on Allens Creek which may result in derating of the 5/ The letter and reply are attached to this response.
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plant. This might affect Intervenor's environmental interests if more environmentally unacceptable means are used to generate electricity. Not once in this dependent series of hypothesis does Intervenor supply any reason to believe that any of this conjecture will materialize, let alone the postulated "demino" result. Obviously such a string of speculations does not form the basis for a litigable contention.
Amended Contention 36. Drywell vacuum breaker.
Intervenor's first complaint is that the drywell vacuum breaker sizing problem "is not reported resolved 10 months after the first special pre-hearing conference."
However, Applicant directs Intervenor's attention to page 6-3 of the ACNGS SER Supplement 2 (March, 1979) whe,re the NRC
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' Staff states that "we conclude that the drywell Wacuum relief system design is acceptable."
Intervenor then discusses bypass leakage of vacuum breakers in the drywell wall during an inadvertent starting of the containment spray system (CSS). The Staff postulated inadvertent CSS starting as a design parameter for the containment vacuum relief system, not the drvwell vacuum breakers. In this regard, inadvertent CSS start analysis has been satisfactorily accomplished for ACNGS See SER Supplement 2, pp. 6-3 and 4.
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Finally, Intervenor alleges that mass transfer effects were not accounted for in the analysis submitted by Applicant of the inadvertent operation of containment spray.
Again, this issue is relevant only to containment vacuum breakers. In any event, at page 6-4 of SER, Supplement 2 (c. 6-4) the Staff reports that:
The effects of mass transfer to the sizing of contain-ment vacuum breakers has been considered in our review of the GESSAR-238 Nuclear Island. As reported in
" Safety Evaluation Report as Related to the Preliminary Design of the GESCAR-238 Nuclear Island Standard Design General Electric Company, Supplement No. 2" NUREG-0124 (Supp. 2 to NUREG 75/110) January 1977, we found the method of analysis acceptable provided that, during normal plant operation, the containment, temperature and relative humidity are maintained within containment temperature limits used in the General Electric Company's vacuum breaker spray analysis. The applicant has stated that Allens Creek design will conform to this resolution.
.All,three of Intervenc~'s complaints are plainly without basis; accordingly, the contention and mnendment should be rejected.
Contention 38. RER single failure proof.
Both Applicant and Staff opposed the July 31, 1979 filing of amended contention #38 on the grounds that Intervenor failed to present a litigable issue regarding compliance with NRC regulations and further failed to specify what
" systems interaction" issues are pertinent to the ACNGS RER system. Intervenor has failed to cure these defects in his
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latest filing, excepting that he has deleted all discussion of systems interaction.
Intervenor again asserts that Applicant does not comply with GDC 19 and 34 with regard to "... bringing the reactor to cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" and references NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations" in support of this assertion. This re?iance on NUREG-0578, however, is misplaced. Item 2.2.3 of NUREG-0578 recommends revisions to Technical Specifications which would require that "...the reactor be placed in a hot shutdown condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in a cold shutdown condition by the licensee within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of any tine that it is found to be or to have been in operation.with a complete loss of safety function. . . " . This recommendatign in NUREG-0578
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is a proposal for an administrative penalty upon loss of a safety system (NUREG-0578 at A-63 ) . As such, it is not at all relevant to Intervenor's allegation that the RER system fails to conform to GDC 19 and 34, nor is it relevant to Intervenor's incorrect assertion that GDC 19 and 34 require the RER system must be capable of bringing the reactor to cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The NRC Staff, in response to Intervenor's July 31, 1979 amendment to Contention 38, stated that: "GDC 19 and 34 do not specify any period of time required for cold shutdown capability" and that
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" . . .there is no basis for this part of the contention. . . " .
(NRC Staff Response to Contention #40 and Amendments to Previously Submitted Contentions of John F. Doherty, August 14, 1979). Applicant submits that this objection still holds.
Moreover, Applicant would note that Section 5.5.7.2 of the ACNGS PSAR clearly states the RER system will have enough heat removal to cool down the reactor to 125* F within approximately ,;urs after shutdown. Thus, even if Inter-venor's addir - GDC 19 and 34 was correct, Applicant would comply. hence, there certainly is no litigable issue.
Finally, Intervenor again proposes,a remedy to his concern in suggesting that Applicant be required to meet the requirements of NUREG-0152 (page 5-21) and NUREG-0190 Appendix A (which are the same document). Applicant rep, eats that its
, h design does meet the requirements of NUREG-0152 bage 5-21.
See Section 5.4.5 of Supplement 2 to the Allen's Creek SER.
Thus, CoLtention No. 38 should be rejected as toally lacking basis.
Contention #39. Fuel rod ballooning.
In this amendmenn to a previous untimely contention, Intervenor substantially repeats himself. Both Staff and Applicant opposed this contention previously on the grounds that Intervenor had failed to establish a clear nexus between the fuel ballooning in TMI-2 accident and the possibility of ang fy I
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similar occurrences at ACNGS. Although Intervenor alleges some arbitrarily chosen dimensional and mJterial similarities between Allen's Creek and TMI-2 fuel, he falls to specify the relevance of the supposed similarities to the fuel rod ballooning issue.
As part of amended contention 39, Intervenor now asserts as another untimely contention that Applicant's fuel ruds are not in compliance with 10 CFR Part 50, Appendix K.
Intervenor fails to specify in any reasonably specific manner why this is true, or to specify in what way the ECCS snalyses in the ACNGS PSAR might be in error. It may be that Intervenor challenges the ECCS criteria on the ground that they do not adequately consider fuel ballooning. That
,is an error but, in any event, Intervenor does' pot begin to
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establish the bases for an attack on Commission regulations.
Intervenor has again relied heavily upon NUREG-0537 and ORNL-4752 as a basis for this contention. The Staff concluded in response to Intervenor's previous filing that "moreover, since N" REG-0557 represents both Mr. Doherty's excuse for the late filing and the substantive basis for the contention, the contention must be rejected regardless of lateness because it lacks the nexus between TMI and Allen's Creek necessary to supply the basis required by 10 CFR 2.714." The staff also states that ". . .with regard to
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... ... ~ . . _ . - . . . .
ORNL-4756, Mr. Doherty has made no attempt whatever to relate the test conditions at Oak Ridge to the Allens Creek design of 8 x 8 fuel." Applicant is in full agreement with these two dispositive objections.
Intervenor has failed to cure the several inade-quacies in this untimely contention and, thus, it should be dismissed.
Respectfully submitted, t
A*
OF COUNSEL: J rego g Cpfeland
. Thomad Bindle, Jr. .
BAKER & BOTTS arles G. Thrash 3000 One Shell Plaza 3000 One Shelf Plaza Houston, Texas 77002 Houston, Texas 77002 LOWENSTEIN, NEWMAN, REIS, J. R. Newman AXELRAD & TOLL Harold F. Reis 1025 Connecticut Ave. , N.W. Robert H. Cu]p
,Washington, D.C. 20036 1025 Connecticut %ve., N.W.
Washington, D.C. Q0036 Attorneys for Applicant HOUSTON LIGHTING & POWER COMPANY TB-02:R
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UNITED STATES OF AMERICA NUCLEAR REGU'.a\ TORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of S S
HOUSTON LIGHTING & POWER COMPANY S Docket No. 50-466 S
(Allens Creek Nuclear Generating S Station, Unit 1) S CERTIFICATE OF SERVICE .
I hereby certify that copies of the foregoing Applicant's Response to John F. Doherty's Amended Conten-tions Numbered 4, 11, 15, 20, 21, 36, 38 and 39 in the above-captioned proceeding were served on the following by deposit in the United States mail, po t e prepaid, or by hand-delivery this 28/A day of , 1979.
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Sheldon J. Wolfe, Esq., Chairman Richard Lowerre, Esq.
Atomic Safety and Iicensing Assistant Attorney General Board Panel for the State of Te::as U.S. Nuclear Regulatory Commission P. O. Sox 12548
- Washington, D. C. 20555 Capito1\ Station Austin, Texas 78711 Dr. E. Leonard Cheatum Route 3, Box 350A Hon. Charles J. Dusek Watkinsville, Georgia 30677 Mayor, City of Wallis P. O. Box 312 Mr. Gustave A. Linenberger Wallis, Texas 77485 Atomic Safety and Licensing Board Panel Hon. Leroy H. Grebe U.S. Nuclear Regulatory Commission County Judge, Austin County Washington, D. C. 20555 P. O. Box 99 Bellville, Texas 77418 Chase R. Stephens Docketing and Service Section Atomic Safety and Licensing Office of the Secretary of the Appeal Board Commission U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission Washington, D. C. 20555 Washington, D. C. 20555 R. Gordon Gooch, Esq. Atomic Safety and Licensing Baker & Botts Board Panel 1701 Pennsylvania 7. venue, N. W. U.S. Nuclear Regulatory Washington, D. C. 20006 Commission Washington, D. C. 20555 3 323
Steve Schinki, Esq.
Staff Counsel U. S. Nuclear Regulatory Commi 31on Washington, D. C. 20555 John F. Doherty 4438 1/2 Leeland Houston, Texas 77023 Madeline Bass Framson 4822 Waynesboro Drive Houston, Texas 77035 Robert S. Framson 4822 Waynesboro Drive Houston, Texas 77035 Carro Hinderstein 8739 Link Terrace Houston, Texas 77025 .
D. Marrack 420 Mulberry Lane Bellaire, Texas 77401 Brenda McCorkle 6140 Darnell '.
' Houston, Texas 77074 '\
F. H. Potthoff, III 7200 Shady Villa, 4110 Houston, Texas 77055 Wayne E. Rentfro P.O. Box 1335 Rosenberg, Texas 77471 James M. Scott, Jr.
8302 Albacore Houston, Texas 77074
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I ACCIDENT ANALYS13 417 I-l Table 3 Probabilities of Eventual Death from Different Competing Risks !
All risks except g AI risks nuclear and except nucisar motor vehich A = 2 a 10 " A=10-'
Other 0.976 1.0 0.976 0.976 Motor vehicle 0.024 0.0 0.024 0.024 Nuclear 0.0 0.0 10-* 7 x 10-LHe expectancy, years 72.3 73.6 72.S(-10 s) 72.S(-1 d) quantifying this kind of risk. The above computations are deScient, however, since many fatalities due to nuclear accidents wtll not be immediate. Further work is needed to quantify the ruk of these delayed fatalities.
U G r ,l, REFERENCES I iO, b bil
- 1. W. D. Rowe An Anaromy ofRisk. John Wiley i Sons. Inc., New York,1977.
2.Stanstrcal Abstract of the Umted Starer 1978 ed., U.S. Department of Commerce, Bureau of the Census, CPO.
- 3. Nuclear Resulatory Commusion. Reactor Safety Study: An Acessment of Acesdent RLrkr ir U.S.
Commereist Nuclear Power Mans. NRC Report WASH.1400 (NUREG.75/014), NTIS,1975.
LETTER TO THE EDITOR: COMMENTS ON " FISSION.GAhELEASE FROM FUEL AT HIGH BURNUP" IN VOL.19, NO. 3 Meyer, Beyer, and Voglewede* have proposed that an enhancement factor be applied to existmg vender models when fission gas release (FGR) at bumups greater than 20,000 mwd! metric ton is calculated for licensing purposes. His enhancement factor is derived from FGR ~2rt obtained from liquid metal. cooled fast breeder reactor (U!FBR) fuel. De analyns surres that te intnnsic source of the high FGR measured for some light water reactor (LWR) fuel rods is, per se, the bumup. In denving the enhancement
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factor, the authors c! aim that effects on FGR due to differences in operstmg temperature between LWR and LMFBR fuels have been taken irito account.
j De following comments are submitted to suggest thr.t an alternat2ve factor, teely, fuel operating temperature, is de dominant variable that determines FGR. Recent FGR data
from LWR fuel rods that support this view are presented. It is argued tat temperature effects can be used to rationalize the high FGR values from LWR fuel rods that were attnbuted in Ref. I to high bumups. This note also questions whether the proposed enhancement factor taaly accounts for known differences in operating temperatures between LWR and DiFBR fuels.
Gaseous nsnon products form directly from the nssioning of 23 5U and 2 2'Pu or by the
, radioactive decay of other fission products. Higher bumup, per se, only increases de inventory of Assion gas that potentially can be released. Fission-gas release requires that the gas atoms, which have been generated in the interior of te f tel, reach a free surface of the i
fuel Vanc is m achanisms have been proposed to account for 1e release of Sssion gases from fuel.Inth6 ge ofinterest (i.e., where FGR is on de order of a few percent), te majority of invesdgators support the view 6at migration of fission gases in the form of atoms or NUCt.f AR SAFETY Vol. 2o. No. 4. My-August 1979
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bubbles to a surface that can communicate with te free space of the fuel rod is required.
Such migration occurs by a thermally activated diffusion process that obeys an Arrhenius equation, and this suggests tat temperature plays the central role in the FGR process.
In this light it is instructive to review the methodology used to derive the enhancement factor of Ref.1. The enhancement factor is an exponential function containing three i arbitrary constants. The sole independent variable is burnup. High.burnup LMFBR FGR 2
data obtained by Dutt and Baker served as the basis for evaluating these constants. In calculating FGR at high burnups, an existing vendor modelis first used to :alculate FGR at 20,000 mwd / metric ton. This calculated value is then weighted by the enhancement factor,
- which is evaluated at the burnup of interest. Since bumup is the only independent variable denned by the enhancement factor, effects et operstmg temperature on FGR have not been explicitly taken into account in denvmg this factor. Since, for the same rating, LMFBR operating temperatures are higher dan LWR operating temperatures, the enhancement factor is conservative with respect to the predicted FGR from LWR fuel rods. The enhancement factor is not imposed on low-bumup (<20,000 mwd / metric ton) FGR calculations, it being ugued tat current FGR models are adequate in this range. Some degree of conservatism is assured, however, by requiring that a minimum value of 1% be used for the FGR at 20,000 mwd! metric ton. [.
Fission-gas-release data that bear on tis issue have been obtained from projects sponsored by the Electric Power Research Institute (EPRI). Measurements have been obtained from fuel rods irradiated in seven commercial LWRs, and additional measurements are to be obtained from two oder reactors. The data, together with analyses usmg the COMETHE IllK computer program, support the view that fuel operstmg temperature is the key vanable that determmes FGR. The role of b'urnup is secondary compared with that of i
fuel operating temperature.
Figure I presents the results of the EPRI. sponsored FGR measurements as,s function of '
- burnup.3 Sigmficant FGR (> 1%)is not restricted to high-bumup values. FGR v'alues ranging from 5.7 to 15.3% were observed in 12 rods with bumups less tan 20,000 mwd,\netne ton.
At higher burnups, from 20,000 to 30,000 mwd / metric ton, FGR values ranging from 6.9 to a 24.2% were observed in 26 rods. The factor common to these rods was that they were unpressunzed. This appears to be a necessary, but not sufScient, condition for high FGR smce small FGR values have been reported for other unpressurized rods at bumups to 24,000 mwd / metric ton. Small FGR (< 1%) was observed in all pressunzed rods.
The resultsindicate cat for these unpressurized rods a trans tion from low FGR values to high FGR values occurs over a very narrow bumuprange. His transition region, or threshold for significant FGR, occurs at about 12,000 MWdimetric ton for ie Maine Yankee fuel rods and at about 24,000 mwd!=etric ton for the Oyster Creek fuel rods. Such a FGR dreshold value also argues against the burnup dependence of FGR as propcsed in Ref.1. Companson of FGR data frorn sibling rods irradiated in de Big Rock Pomt reactor
- also supports de u idea of a threshold value for significant FGR. Data for tree pairs of rods (Table 1) show that, for modest differences in burnup of about 15%, FGR values differms by more than a factor of 10 are observed. Phenomenological models of FGR that predict the general trends observed in te Oyster Creek, Mune Yankee, and Big Rock rods (i.e., nearly zero FGR up to a threshold bumup value, followed by a transition to high FGR va!ues over a narrow bumup 8 '
range) have been reported by Dollins and Nichols and by Hargreaves and Collins.' '
The COMETHE-IIIK fuel performance computer program is being used, together with i
appropriate duty cycle and fuel fabrication data, to interpret these FGR data.' These
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-analyses scpport the view that fuel. rod operating temperatureis the key determinant of -- ---
FGR. The factors that appear to be most important in establishing fuel temperatures are the linear heat rat =g, the fuel-clad gap size, and the thermal cenductivity of the gap. The gap -
mze, in tuni,is strongly idcuenced by lad creepdown and fBel densificaticrirates. The data d 1
in Fig. I suggest ths' gap conductmtiy in unpressurized rods can be degraded to such an I-- ~- ~, q_
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O 5 10 15 25 30 35 s Surman GW tone Fig.1 Fismon pa release vs. burnap 44ca obtamed from LWR fuel rods. The maumum fisson-gas release ,
calculated with the .VRC enhancement factor is aim shown. .
r Table 1 Fision.Gm. Release Data from Fuel Rods Irradiated in Big Rock Point Aversy Potet tp burnup, densty, Fhst eycie mwd / Fission.sas Fue4< tad 5 theoretical fasi shankage. D '7 "
Rod number metric too release, % gap, mils density % A L'L 1st 2nd 3rd -
JJ 400002 8,700 13.7 9.5 89.02 2.65 9.2 1K 400001 S,000 0.3 9.5 90.12 2.32 S.2 AB 30001 21J00 6.9 9-13 94.0 1.12 9.0 13.7 4.4 AB 20001 18,300 0.52 9 - 13 93J 0.80 7.3 12.3 4.1 AB 10001* 19,000 7J 9 - 13 92.2 1.5 7.2 13.0 4.3 AB 40001 16,800 0.21 9 - 13 93.75 1.73 6.3 10.9 3.6
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NUC'.IA A SAFETY, vol. 20, No. 4. July-. Augur 1979 9
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.g 1 420 . ACC10ENT ANALYSIS extent that significant FGR wtll result. On the other hand,for tese fuel rods, pressurization
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i is erTective in maintaining high values of thermal conductivity across the gap so that FGR is j kept at values below the minimum 1% value required by Ref. I even at burnups approaching <
] 30,000 mwd / metric ton. <
1 If bumup is not the key determinant of FGR at high bumups, how can one rationalize the LWR data of Ref. I that were offered as evidence to support the FGR enhancement j factor? Since FGR data are presented there to bumups approactung 60.000 mwd! metric a ton, these data retlect the response of LWR fuel fabricated from the late 1960s to the early ,
1970s. During this period, effects resulting from irradiation-induced fuel densification were not recognized by de nuclear industry. Also,some of te data of Ref. I were obtained from fuel rods operated at high ratings, far in excess of dat found in commercial LWRs.' We suggest that high FGR values were observed because high ratzgs 2nd irradiation 4nduced
- densification led to operating temperatures in dese fuel rods that are higher dan would prevad in commercia! LWR fuel rods. These high temperatures, in tum,'ed to high FGR
- values. In de intenm pened the factors responsible for irradiation. induced densification
] have been determined,' and the fuel vendors are routinely fabricatmg densification resistant l i fuel. As this newer fuel achieves higher burnups, and as older fuel is discharged, it is likely <
dat lower FGR values wdl be observed.
Howard Ocken
- Electric Power Research Institute j Palo Alto,Califomia
. and npy a s f@M N lI J.T. A. Roberts Electric Power Research Institute Palo Alto, California -
- \
i REFERENCES
- 1. R. O. Meyer C. E. Beyer, and J. C. Vosiewede, Fisson.Cas Release from Fuet at High Burnup,NucL 1 Safety. 19(6)
- 699 708 (197S).
j 2. D. S. Dutt and R. B. Baker SIEX. .4 Correlated Code for the Prediction of Liquid Metal Fast Breeder
] / eactor /l VF3R/ Fuel Thermal Performance. ERD A Report HEDL.TME.74 55. Hanford Engineenns L velopment Laboratory NTIS, June 19*5.
. 3. J. T. A. Roberts et at, LWR Fuel Performance Pmg sm: Progress us 19 3. Report 57-1024-SR, pp.
3 24 to 3-27. E'ectnc Power Research Insutute, Feerurf 1979. .
- 4. N. Dudey and C. Crouthamel. Multiple Cycle Plutantum Crill:staan P*otect, RP 306. Annual Report T for 1978, Consumers Power Company, a preparation. I I
5.C. C. Dollins and F. A. Nichols, Sweiling and Cas Release in UO, at Low and Intermediate Temperatuns,/.Nuct Mater.. 66: 143 157 (1977).
- 6. R. Hargreaves and D. A. Collms. A Quantitative Model for Fisnon Cas Release and Swelling in Irradiated Uramum Dieude,/. Sr. Nuct Enegy Soc., 15: 311 313 (1976).
- 7. N. Hoppe and G. R. Thomns. EPRI Pon'taan Paper on Fimon Get Release f>om LWR Fuel Rods, in preparation.
S.C. J. Baroch and M. A. Rigdon. Irradiation 3ehavior of CO, at Bumups from 10 to 30 GWd/ tonne U, l in Intematsanal C.cfernce on Nuclear Fuel performance. London. Oct. 13-19.1973 pp. 581 to l 58 4. Bntish Nuclear Energy Socwty. Londca,1973. I
- 9. M. D. Freshley, D. W. Bnte, J. L. Daruel, and P. E. Hart, Irradia: ion Induced Densdication of UO.
P liet Fuel./. Nuci Marer., 62: 138 166 (1976).
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Acc: DENT ANALYSIS 421 AUTHORS' RESPONSE TO THE PRECEDING LETTER
, We have received a copy of Ocken and Roberts' comments on Fission. Gas Release from Fuel at High Burnup, which was published in the November-December 1973 issue of Nuclear Safety. Our response to these comments is as follows.
I j 1. Application of NRC correlation requires knowledge of both fuelburnup and tempersture at and beyond 20.000 mwd / metric ton. We have not assumed that burnup is the dominant variable in the correlation. To the contrary, we believe that temperature is the strongest variable and therefore have retained the users temperature dependence F(TJ at all burnups.
2.If high. temperature t'tssion gas re! ease is a diffusion. controlled process, as Ocken and Roberts suggest, one would expect it to be both dermally acuvated and concentration i degndent. The question is not, therefore, te existence of a burnup dependent effect,
- but its magnitude.
i 3. The authors further suggest that the range ofinterest for fission. gas release is on the order of a few p.rcent. While this may be the case for low burnup prepressurized fuel rods operated at nommal power levels, it is obviously not the case for all fuel designs, particularly those operated at high linear heat ratings of interest in licensing (i.e., the licensed LWR rower limits of around 13 to.15 kW/ft).
Our original publication assumes the role of burnup with respect to ' temperature to be obvious. Perhaps we did not give this point enough emphasts. As evidenced by Ocken and
, Roberts' comments, this potat does not appear to be clearly understood, and the potenttal for misapplymg the NRC correlation exists'. k l
Ralph O. Meyer. Leader 1 Reactor Fuels Section Core Perictmance Branch Division of Systems Safety Nuclear Regulatory Commission
~~
INTERNATIONAL SYMPOS!UM ON MANAGEMENT OF GASEOUS WASTES FROM NUCLEAR FACILITIES l Vienne. Austna. Fets. 18-22,198o T?ns symposium jointly sponsored by the International Atomic Energy Agency and the OECD Nuciear Energy Agency, will provide a forum for :h'e exchange of information on sewnufic, tec.*uucal, and technolog2 cal aspects associated with the gaseous wastes and effluent treatment at riuclear facilines. The papers presssted should represent an aut'ontative account of the ustus of this subject throughout the world in 1930.
Inquanes regarding U. S. partexpanon should be di:teted to J. H. Kane. Confere::ce Specialist. U. S.
Department of Energy. Washmston, D.C. 20545. All other mquir es should go to the Conference
, Secretanat. Conference Sernce Secton, Dmsaon oi External Reiauens. Internanonal Atorruc Energy Agency. P. O. Box 590 Dmmer Ring 11 A.1011. Vienna, Austna. 'i n o 7 i
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NUCt.f AA SAFETY, voa. :o No. 4, Juty-August 1979
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