ML20207J513
| ML20207J513 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 07/24/1986 |
| From: | Cole R, Lazo R, Luebke E, Luebke F Atomic Safety and Licensing Board Panel |
| To: | |
| References | |
| CON-#386-124 84-496-03-LA, 84-496-3-LA, OLA-1, NUDOCS 8607290148 | |
| Download: ML20207J513 (43) | |
Text
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x' RY UNITED STATES OF AMERICA NUCLEAR REGULATORY COP 9tISSIg y y ATOMIC SAFETY AND LICENSING BilARD GFFI Before Administrative Judg N E @0: cl'0' W N,_
7 Dr. Robert M. Lazo, Chairman Dr. Richard F. Cole 3MED JUL61986 Dr. Emmeth A. Luebke In the Matter of Docket Nos. 50-250-OLA-1
)
50-251-0LA-1 FLORIDA POWER & LIGHT COMPANY
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)
ASLBP No. 84-496-03 LA (Turkey Point Plant, (Vessel Flux Reduction)
Units 3 & 4)
July 24, 1986 INITIAL DECISION (Operating License Amendment)
APPEARANCES Martin H. Hodder, Miami, Florida, for the Intervenors, Center for Nuclear Responsibility and Joette Lorion.
Norman A. Coll, Miami, Florida, and Michael A. Bauser, Washington, D.C.,
for the Applicant, Florida Power and Light Company.
Mitzi A. Young and Mary E. Wagner for the United States Nuclear Regulatory Commission Staff.
I I.
INTRODUCTION AND BACKGROUND By letters dated August 19, 1983 and September 9,1983, Florida i
Power and Light Company (Licensee) requested amendments to the technical B607290148 860724
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specifications of Facility Operating Licenses DPR-31 and DPR-41 for its Turkey Point Plant, Nuclear Generating Units 3 and 4, two pressurized water nuclear reactors located in Dade County, Florida. The amendments were to support the Licensee's program for reduction of neutron bombardment (vessel flux), and consequent embrittlement, of the pressure vessel walls, and to remove restrictions imposed when the Licensee was operatine with the old steam generators having a greater number of plugged tubes than the new steam generators now in use.1 Notice that the Commission was considering issuance of the amendments, of their proposed content, and of the fact that the Commission had made a proposed determination of no significant hazards consideration in conformance with the standards contained in 10 C.F.R. 9 50.92 was published in the Federal Register on October 7, 1983. 48 Fed. Reg. 45,862. The notice sought public comments on the proposed determination and advised the public of its right to seek a hearing and intervene in the proceedings.
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Specifically, the Licensee requested (1) to increase the hot channel factor limit from 1.55 to 1.62; (2) to increase the total peaking factor limit from 2.30 to 2.32; (3) to change the overpower delta-T trip set points and thermal hydraulic limit curves; and (4) to delete restrictions and limits which allowed the old steam generators to operate with tubes plugged in excess of five percent. NRC Safety Evaluation, December 23, 1983 (Staff Exhibit 1) at 1.
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3 On November 4, 1983 in response to the notice, the Center for Nuclear Responsibility and Joette Lorion jointly petitioned for leave to intervene and requested a hearing. They also filed comments, contending that the amendments did involve a significant hazards consideration.
Nevertheless, on December 23, 1983, the Commission issued the requested amendments pursuant to a final determination of no significant hazards consideration and the Commission's finding, among other things, that the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
See Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 99 to Facility Operating License No. DPR-31 and Amendment No. 93 to Facility Operating License No. DPR-41, Florida Power and Light Company (Dec. 23,1983) (SER), Staff Exhibit 1; 49 Fed. Reg. 3364, January 28, 1984.
Under10C.F.R.650.91(a)(4)theamendmentsbecame effective when issued, with any required hearing to be held thereafter.
Intervenors filed an amended petition on January 25, 1984. A prehearing conference was held in Homestead, Florida on February 28, 1984. During that conference all parties were provided an opportunity to file briefs concerning Intervenors' request to consolidate the consideration of another set of amendments,2 issued earlier for the 1
2 Amendment No. 98 to Facility Operating License No. DPR-31 and Amendment No. 92 to Facility Operating License No. DPR-41.
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4 Turkey Point units, with those actually the subject of the instant proceeding. The earlier issued amendments provided for, among other things, the replacement, during the course of subsequent refuelings of the two units, of Westinghouse 15 x 15 Low Parasitic (LOPAR) fuel and borosilicate glass burnable absorber rods with Westinghouse 15 x 15 Optimized Fuel Assembly (0FA) fuel and Wet Annular Burnable Absorber (WABA) rods. These amendments (subsequently referred to by this Board as the " core design change" amendments, as opposed to the instant
" vessel flux reduction" amendments) were publicly noticed on July 20, 1983, 48 Fed. Reg. 33,080, and were issued on December 9, 1983. 48 Fed.
Reg. 56,518 (Dec. 21, 1983).
See generally SER, Staff Exhibit 1, at 3 (Dec.23,1983).
In our May 16, 1984 Prehearing Conference Order (unpublished), we denied combined consideration of the two separate sets of amendments noting, among other things, that:
(1) no petitions to intervene had been filed in connection with the core design change amendments; (2) no Licensing Board had been convened to address those amendments;and(3)thoseamendmentswerenotwithinthejurisdictionof this Board to decide. For present purposes, however, one result of the core design change amendments is that the Turkey Point units will operate with both LOPAR and 0FA fuel (i.e., with mixed, rather than homogeneous, fuel in the core) until, as a result of future refuelings, the LOPAR fuel has been entirely replaced with 0FA fuel.
The Prehearing Conference Order dated May 16, 1984, also granted the Intervenors standing to intervene in this proceeding, and ruled on I
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Intervenor contentions and other matters. Only Contention (b) and Contention (d) were admitted.
Contention (b) alleged shortcomings in one of the computer models which is involved in the prediction of the temperature of the hottest fuel rod in the reactor core as part of the analysis of loss of coolant accidents. Contention (d) alleged, in effect, that, under the amendments, it is significantly more probable that a steam film will form around a fuel rod during normal and anticipated operational occurrences, resulting in a significant reduction in safety.
In full, Contention (d) reads as follows:
The proposed decrease in the departure in the nucleate boiling ratio (DNBR) would significantly and adversely affect the margin of safety for the operation of the reactors. The restriction of the DNBR safety limit is intended to prevent overheating of the fuel and possible cladding perforation, which would result in the release of fission products from the fuel.
If the minimum allowable DNBR is reduced from 1.3 to 1.7 [ sic: 1.17] as proposed, this would authorize operation of the fuel much closer to the upper boundary of the nucleate boiling regime. Thus, the safety margin will be significantly reduced. Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the departure from the nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. Thus, the proposed amendment will both significantly reduce the safety margin and significantly increase the probability of serious consequences from an accident.
Licensee filed motions for summary disposition of the two contentions on August 10, 1984 which were supported by the Staff and opposed by Intervenors. Because we found the pleadings and the balance of the written record incomplete for reaching a decision, we held a prehearing conference in Coral Gables, Florida on March 26, 1985 during
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6 which the Licensee made a " didactic presentation" as ordered by this Board concerning issues raised in the parties' summary disposition papers.
SeeLBP-85-29,22NRC300,306-310(1985). The Intervenors and Staff were given the opportunity to cross-examine the Licensee's witnesses during the conference and were afforded the opportunity to respond or rebut.
Id.
By Order dated August 16, 1985, we granted Licensee's motion for summary disposition of Contention (b), but denied the motion for summary disposition of Contention (d) and limited the scope of the litigation on Contention (d) to the following three issues:
1.
Whether the DNBR [ departure from nucleate boiling ratio] of 1.17 which the amendments impose on the OFA [0ptimized Fuel Assembly] fuel in the Units 3 and 4 compensates for the three uncertainties outlined by the Staff in its December 23, 1983 SER on the amendments, at 4.
2.
Whether, if the DNBR of 1.17 does not compensate for those uncertainties, the SRP's [ Standard Review Plan's] 95/95 standard, or a comparable one, is somehow satisfied.
3.
Whether, if that standard is not being satisfied, the reduction in the margin of safety has been significant.
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l 22 NRC at 330. Accordingly, we scheduled an evidentiary hearing on these issues to commence on December 10, 1985 and directed the parties to file written testimony to be in hand by November 26, 1985. Order Scheduling Hearing, September 18, 1985.
Subsequently, the Licensee
7 filed a second motion for summary disposition of Contention (d) on September 20, 1985 which was again supported by the Staff and opposed by Intervenors.
By Order of November 8,1985 we denied Licensees' second motion for summary disposition for the reason set forth in a later Order, dated November 18, 1985.
Evidentiary hearings were held in Miami, Florida from December 10 through December 12, 1985.
As noted above, the record on summary disposition led this Board to question whether a ONBR of 1.17 accounts for the three uncertainties, as outlined in the Staff's SER, associated with rod bowing, the transitional core containing 0FA and LOPAR fuel and the application of the WRB-1 correlation to 15 x 15 array 0FA fuel.
If a DNBR of 1.17 did not account for the three uncertainties, we pondered whether that DNBR failed to meet the 95/95 standard and thus resulted in a significant reduction in the margin of safety. 22 NRC at 329-30. We turn now to a discussion of the evidence on three questions we posed.
This Opinion is based upon, and incorporates, the Findings of Fact and Conclusions of Law that follow. Any proposed findings or conclusions submitted by the parties that are not incorporated directly or inferentially in this Initial Decision are rejected as being l
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8 unsupportable in law or in fact or as being unnecessary to the rendering of this Decision.3 II. FINDINGS 1.
The Licensee's direct case consisted of testimony by Edward A.
Dzenis (ff. Tr. 302), Manager of Core Operations in the Nuclear Fuel Division of Westinghouse Electric Corporation. Mr. Dzenis has a Bachelor of Science degree and a Master of Science degree in Mechanical Engineering. He has taken undergraduate courses involving calculus, differential equations, mathematical statistics and statistical evaluation of experimental data and graduate courses in thermodynamic power conversion cycles, and the environmental and economic aspects of nuclear power. Since joining Westinghouse in 1974, his work has included analyses of heat transfer and the fluid flow aspects of reactor fuel assemblies and related components for pressurized water reactors (PWRs), the determination of core operating limits to insure margin for the prevention of DNB, and analyses of other safety criteria.
Mr. Dzenis has also been involved in modifications of the THINC code to incorporate new correlations such as the WRB-1 critical heat flux correlation.
(Professional Qualifications and Experience of Edward A.
3 Intervenors' February 17, 1986 Motion To File Intervenors' Proposed Findings Of Fact And Conclusions Of Law Out Of Time (one working day), is for good cause shown, granted.
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9 Dzenis, ff. Tr. 302; Tr. 293-302). At the completion of voir dire, Mr. Dzenis' testimony, prefiled on November 26, 1985, was received in evidence without objection and bound into the transcript.
2.
The Staff's direct case consisted of testimony by Dr. Yi-Hsiung Hsii, a nuclear engineer in the Reactor Systems Branch of the Division of PWR Licensing-A in the Office of Nuclear Reactor Regulation and formerly in the Core Performance Branch of the Division of Systems Integration. Dr. Hsii has a Bachelor, Master and Doctorate degree in Mechanical Engineering. He has taken undergraduate courses in hydrodynamics, thermodynamics, heat transfer, calculus, differential equations, and graduate courses in hydrodynamics, heat transfer, thermodynamics, advanced calculus and complex variables. Since he joined the NRC in 1981, he has reviewed safety evaluation reports and fuel reload methodology topical reports on core thermal hydraulics, including critical heat flux (CHF) correlations, submitted by applicants and licensees. Dr. Hsii worked for Babcock & Wilcox from 1967 to 1981 where he performed core thermal-hydraulic design analyses for reactors, and developed computer codes in the areas of containment systems, reactor system transients, fuel pin thermal performance analysis and heat transfer. Dr. Hsii also developed a computer program to calculate core performance and DNBRs.
(Hsii Professional Qualifications, ff. Tr.
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10 733; Tr. 715).4 In addition, pursuant to 10 C.F.R. 9 2.743(g), the Board received Staff Exhibit 1, the NRC Safety Evaluation supporting the amendments, dated December 23, 1983, for the purpose of documenting the NRC's review of the thermal hydraulics associated with the amendments.
(Tr.735-36).
I 3.
The Intervenors' direct case consisted of testimony by Dr. Gordon D. J. Edwards (ff. Tr. 606), President of the Canadian Coalition for Nuclear Responsibility and Professor of Mathematics and Science at Vanier College, Montreal, Canada. Dr. Edwards holds a Ph.D.
in Mathematics, has taught university-level mathematics for several l
years and has limited experience teaching biology and chemistry. Tr.
254-57, 505. He has acted as a consultant to a number of Canadian governmental studies concerning reactor safety, and in that regard has both cross-examined witnesses and testified as an expert in his field of expertise, which he considers to be mathematical analysis, calculations of probabilities and use of mathematical models. Tr. 261-62, 272-73, 282-83. However, as Dr. Edwards himself acknowledged, he generally has no knowledge, skill, experience, training or education in the field of l
engineering (Tr. 538) nor does he consider himself to be an expert in the areas of heat transfer, departure from nucleate boiling testing, I
4 Intervenors withdrew their objection to the admission of Dr. Hsii's prefiled testimony and statement of professional qualifications after conducting a voir dire examination.
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11 critical heat flux correlation, determination of operational limits or evaluation of DNBR. Tr. 283.
4.
Dr. Edwards was unfamiliar with the term subchannel analysis, has never conducted DNB tests or DNBR acceptance limits or developed a DNB correlation, and has never designed or used computer models to do thermal-hydraulic analysis of heat transfer and fluid flow aspects of a pressurized water reactor. Tr. 278, 506. Finally, Dr. Edwards acknowledged that he was not familiar with the mathematical equations or computer models used to evaluate and analyze the DNB and DNBR at Turkey Point. Tr. 506-07.
5.
The expert qualifications of Dr. Edwards and the admissibility of his written testimony were challenged by Staff and Licensee. At the outset of the proceeding, Licensee, and Staff to a limited extent,
. objected to Intervenors' request that Dr. Edwards be allowed to act as an expert interrogator, as is permissible under 10 C.F.R. 6 2.733.
6.
Licensee objected to Dr. Edwards' interrogation as an expert in that by his own admission he was not qualified by training or experience in thermodynamics, heat transfer, fluid mechanics or thermal hydraulic analysis, all of which topics were central to the narrow issues at the hearing. The Staff did not object to Dr. Edwards' conducting cross-examination as an expert provided that he examined only in those areas of his admitted expertise, that is, mathematics, l
12 including mathematical analysis, calculations of probabilities and the use of mathematical models. The Staff objected to any interrogation beyond those areas, because the Commission's rules specify that any cross-examination by an expert interrogator "shall be limited to areas within the expertise of the individual conducting the examination or cross-examination." 10 C.F.R. 9 2.733.
The Board found Dr. Edwards to be qualified as an expert interrogator pursuant to 10 C.F.R. 9 2.733.
Not having the benefit of a cross-examination plan, we declined in advance of hearing his questions, to define the limits of Dr. Edwards' expertise for the purpose of examination and permitted Dr. Edwards to conduct cross-examination of both Licensee's and Staff's, witnesses.
7.
The Board also ruled on the limits of Intervenors' direct case. On November 25, 1985, in accordance with our September 18, 1985 order setting the deadline for prefiled testimony, Intervenors served upon the Board and the parties a document entitled " Outline of Testimony By Gordon Edwards" (Edwards Outline), together with a copy of Dr. Edwards' professional qualifications.
On the second day of the hearing at the commencement of their direct case, Intervenors sought to expand the Outline by eliciting oral testimony concerning Dr. Edwards'
" response and explanation" to the three Board questions. Tr. 446.
8.
Staff and Licensee objected to this procedure as falling outside the Commission's Rules of Practice, 10 C.F.R. 9 2.743(b), which requires all parties to file written direct testimony in advance of any
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13 hearing. We too observed that there had been time to prepare an expanded version of Dr. Edwards' testimony and serve it on the parties before the hearing and sustained the objections of Staff and Licensee to the oral supplementation of Dr. Edwards' written testimony on grounds that it would contravene 10 C.F.R. 5 2.743 and be unfair to opposing parties.
9.
The Staff and Licensee objected to Intervenors' subsequent proffer of written direct testimony, which consisted of two affidavits previously prepared by Dr. Edwards in response to motions for summary disposition, claiming surprise and prejudice to the preparation of their cases and lack of good cause for Intervenors' failure to meet the deadline for filing written testimony. We ruled that the August 30, 1984 affidavit was stale and its introduction contained an element of surprise. However, the Staff and Licensee had had reasonable opportunity to examine the later affidavit, dated November 5,1985
(" November Affidavit" or " Edwards Affidavit"). Thus, pursuant to 10 C.F.R. Q.2.743, which provides for the admission of additional written testimony upon a board ruling and if the parties have had a reasonable opportunity to examine it, we determined that the November Affidavit wculd be received in evidence provided it withstood voir dire and any motions to strike.
l 10.
Based on the evidence adduced through the voir dire examination of Dr. Edwards as a proposed expert witness, Licensee moved i
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14 to strike Dr. Edwards' testimony in its entirety and the Staff objected to a large part of Dr. Edwards' proposed testimony, on the grounds that:
(1) the witness had shown he was not competent to testify generally as to the matters at issue in the proceeding other than Dr. Edwards' statements concerning areas of applied mathematics, including statistics and statistical analysis; (2) portions of the November 5 Affidavit were virtually identical to, or unduly repetitious of, statements in the Cutline; and (3) portions of the testimony purportedly were irrelevant and lacking in probative value.
11.
Despite these objections, we found Dr. Edwards was qualified as an expert witness in view of the " limited scope and the qualified language" of his testimony and admitted the Outline and the Novemoer 5 Affidavit into evidence.
12.
In so doing, the Board recognized that the weight to be accorded to Dr. Edwards' testimony is influenced by the fact that he is a mathematician with little knowledge, education, skill, training or j
experience in engineering. While Dr. Edwards' familiarity with reactor i
concepts is impressive for a layman, the depth of his knowledge of engineering problems and ability to evaluate engineering judgments is understandably quite limited. Moreover, by his own admission, his disagreement with the testimony presented by Licensee and Staff is not
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based on a complete knowledge, or even reading, of all the documents
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6 15 underlying the review that has been performed. However, Dr. Edwards was candid and forthright in presenting his testimony as that of a mathematician and not an engineer, and his participation in this proceeding has aided in sharpening the issues in controversy, 13.
We have carefully considered all of the testimony, opinions and evidence adduced at the hearing 6nd have accorded the appropriate weight to the comparative knowledge, skill and experience of the three witnesses. We will now set forth our resolution of each of the questions at issue in this proceeding, seriatim.
In addition, in the course of our discussion we will consider the matters of concern to the Intervenors as we understand them.
14.
As we have indicated, the three questions arose during our consideration of Licensee's first motion for summary disposition of Contention (d). More specifically, as we discussed in considerable detail in ou, August 16, 1985 Order addressing that motion, it was clear to us how a 1.17 DNBR acceptance limit for a certain type of fuel utilizing one critical heat flux (CHF) correlation (in this case, 0FA fuel with the WRB-1 correlation), could provide the same degree of assurance that departure from nucleate boiling would not occur as with a higher, 1.3 DNBR acceptance limit for another type of fuel utilizing a different CHF correlation (again, in this case, LOPAR fuel with the W-3 correlation). See 22 NRC at 323-28. What was not clear to us, however, was how three particular uncertainties mentioned in the NRC Staff's
16 December 23, 1983 SER (i.e., those related to rod bowing; the use of new 0FA fuel assemblies mixed together with LOPAR fuel assemblies during a transition period on the way to a full 0FA core; and the application of the WRB-1 correlation to 15 x 15 array 0FA fuel) were accounted for.
Id. at 328-31. Accordingly, we stated:
The Licensee has the burden of showing in hearing either that the application of a DNBR of 1.17 to the OFA fuel in Units 3 and 4 satisfies the 95/95 [NRC Staff] standard, or that if such application does not, the reduction in the margin of safety is not significant.
Iji at 330.
First Board Question Whether the DNBR (departure from nucleate boiling ratio) of 1.17 which the amendments impose on the 0FA (Optimized Fuel Assembly) fuel in Units 3 and 4 compensates for the three uncertainties outlined by the Staff in its December 23, 1983 SER on the amendments, at 4.
15.
All the parties agree and the Board concludes that the answer to the first question is that the DNBR of 1.17 does not compensate for the uncertainties associated with rod bow, the mixed core and the application of WRB-1 correlation.
Dzenis, ff. Tr. 302, at 3; Hsii, ff.
Tr. 733, at 22; Edwards Outline, ff. Tr. 606, at 1.
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17 16.
It will be helpful to review certain aspects of PWR operation.
Hcat is removed from the core of a nuclear reactor by water flowing around the outside of the fuel rods.
If the temperature of the fuel rods is sufficiently high, bubbles of steam will form on the fuel rod surfaces. These bubbles are then swept away from the rods by the flow of water around them. Once in the bulk flow, the bubbles of steam either condense and disappear or, at a higher temperature, survive in equilibrium with the liquid coolant. The stage of boiling at which bubbles of steam form and leave the surfaces of the fuel rods is called nucleate boiling. During nucleate boiling, the transfer of heat from the rods is efficient and increases in approximate propo,rtion to increasing fuel rod temperature. The measure of heat transferred in a given time from a unit of rod surface area is called heat flux.
17.
If the fuel rods reach a sufficiently high temperature, some of the steam bubbles will remain on the rod surfaces and begin to combine. This results in the formation of a steam film. The point at which a film appears is called departure from nucleate boiling (DNB).
Such a film insulates the fuel rod causing heat that would otherwise be given up to the coolant to be retained in the rod. The heat flux begins to decline. The heat flux at the beginning of this decline is called the critical heat flux, or CHF. To avoid DNB, during normal operation or anticipated operational occurrences, a proper relationship is maintained between what the CHF would be for a given set of conditions,
18 and the actual heat flux (AHF) under those same conditions. DNB does not necessarily result in a failure of cladding, and even if a breach were to occur, any release would only be to the primary coolant system, which is a closed system. This, in turn, is enclosed by the reactor containment building which is designed to avoid release of radioactivity to the environment. The public is kept at a distance by an exclusion area.
In spite of all these protective measuras, it is prudent that DNB, and transition to a less desirable heat transfer regime, be avoided.
18.
It is impossible to predict with complete certainty what the CHF for a particular fuel in a reactor will be under a given set of conditions. Different experimentally-determined correlations give varying degrees of assurance with respect to predictions of CHF. The ratio of CHF to AHF is called the DNB ratio or DNBR.
In the NRC Staff Standard Review Plan (SRP), NUREG-0800, 6 4.4 Thermal and Hydraulic Design, a minimum ratio between the CHF and the AHF, is established such that there is at least a 95% confidence level that there is a 95%
probability that DNB will not be reached by the hottest rod in the core during either normal operation or anticipated operational occurrences.
This statistical measure of conservatism in the selection of a minimum DNBR is referred to as the 95/95 condition or standard.
19.
If the true CHF value could be calculated and the actual heat flux were precisely known, the exact DNBR could be determined and a
i 19 design DNBR limit of 1.0 would ensure DNB would be avoided. However, because CHF is calculated using an empirical correlation based on experimental CHF data and because of random variations in the data upon whicn the correlation is based, the exact CHF cannot be predicted. A DNBR limit greater than 1.0 is therefore imposed to account for this uncertainty and represents a degree of conservatism or margin of safety.
The DNBR is referred to in a number of ways, including "DNBR design limit," " design DNBR," and "DNBR limit."
It is also referred to as a DNBR acceptance limit. The DNBR acceptance limit of 1.17 is generic to all Westinghouse plants utilizing 0FA fuel. The " safety analysis minimum DNBR," or " calculated minimum DNBR" is to be distinguished from the DNBR acceptance limit.
It is calculated on a plant-specific basis.
20.
A DNBR limit for a particular fuel type is the quantity imposed on a CHF correlation as the specified acceptable fuel design limit to ensure at a 95/95 level that the hottest fuel rod in the core will not experience DNB during normal operation and anticipated operational occurrences.
10 C.F.R. Part 50, Appendix A, Criterion 10, Reactor Design states:
l The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
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20 21.
The DNBR acceptance limit of 1.17 for the WRB-1 correlation used in connection with the analysis of all Westinghouse OFA fuel is the 95/95 bounding value for experimental data. The 95/95 standard in Section 4.4 of the SRP will be satisfied by assuring that calculated minimum DNBR values for all normal and anticipated operational occurrences, after accounting for uncertainties, are greater than or equal to the 1.17 DNBR acceptance limit.
22.
The 1.17 DNBR acceptance limit does not and is not intended to compensate for the three uncertainties referred to in the Board's question, namely, the rod bow, the mixed LOPAR/0FA fueled core, and the application of the WRB-1 correlation to the 15x15 0FA array fuel.
Dzenis, ff. Tr. 302, at 4.
The DNBR acceptance limit for a correlation, including the WRB-1 correlation, depends upon the ability of that correlation to predict CHF data.
For every CHF test data point, a CHF prediction is made using the correlation, and a comparison performed between the measured and predicted CHF values. A probability distribution of the measured-to-predicted CHF ratios is obtained for all of the CHF data points. A statistical analysis is then performed to obtain the estimated mean and standard deviation of the measured-to-predicted CHF ratios. The DNBR limit is derived from statistical analysis applying the acceptance criterion of 95%
probability at 95% confidence, as specified in the SRP. Hsii, ff. Tr.
733, at 3-4.
The uncertainties are taken into account in the l
f evaluations of normal and anticipated operational occurrences performed t
21 for specific plants. This is done in connection with the Board's second question that follows.
Second Board Question Whether, if the DNBR of 1.17 does not compensate for those uncertainties, the SRP's 95/95 standard, or a comparable one, is somehow satisfied.
23.
Licensee and NRC Staff have answered the second question in the affirmative (Dzenis, ff. Tr. 302 at 4; Hsii, ff. Tr. 733 at 22).
Intervenors respond in the negative, arguing that the proposed DNBR
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limit of 1.17 does not compensate for the uncertainties identified in the NRC Staff's SER (Edwards, ff. Tr. 606 Outline of Testimony unnumbered p. 1).
24.
The three uncertainties of concern as outlined by the NRC Staff in its December 23, 1983 SER are:
1.
Effects of rod bow phenomena.
l 2.
Thermal hydraulic behavior of a mixed core of 0FA and LOPAR fuel.
3.
Applicability of the WRB-1 critical heat flux correlation l
to the 15x15 0FA fuel configuration.
Gordon Edwards, ff. Tr. 606 at 2; Dzenis ff. Tr. 302 at 4; Hsii, ff. Tr. 733 at 16. Uncertainties, which were not included in the l
22 1
calculational method, i.e., input in the THINC computer calculations, are treated as penalties.
The Assigned Penalties Rod Bow Penalty 25.
At Turkey Point, fuel rods are placed in the reactor core in assemblies consisting of a 15x15 array of fuel rods. These fuel rods are supported in the assembly by spacer grids located approximately every two feet of axial elevation. As the fuel is irradiated, some random horizontal displacement of the fuel rods from their normal position occurs. This displacement is called " rod bow."
Rod bowing can result in a reduction in CHF and, therefore, a reduction in the DNBR.
Tr. 320-22.
See also Dzenis, ff. Tr. 7-8; Hsii, ff. Tr. 733, at 16.
The effect of rod bow on DNBR is applied as a penalty.
E63t.,Dzenis, ff. Tr. 302, at 4-8; Tr. 322, 436.
26.
The rod bow penalty is based on direct measurements of fuel assemblies from operating reactors representing a wide range of burnups and other conditions. Tr. 323. A value of 5.5% for the rod bow penalty for 0FA fuel, was derived, based on an approved method described in a Westinghouse topical report. WCAP-8691, Rev. 1, Fuel Rod Bow Evaluation. This method has been used for most plants of Westinghouse design. The penalty derived using this method is a 95/95 tolerance t
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23 limit as was statistically demonstrated in WCAP-8691. Hsii, ff.
Tr. 733, at 16-17; Tr. 822, 823.
27.
The value of a 5.5% DNBR corresponds to the highest burnup at which DNB is a concern. This is because, at higher burnups, heat generation rates in PWR fuel decrease due to a decrease in the concentration of fissionable isotopes and the buildup of fission product inventory. Dzenis, ff. Tr. 302, at 8.
For the purpose of calculating the rod bow penalty, the maximum burnup used is 33,000 MWD /MTV.
By the time a fuel rod exceeds a burnup of 33,000 MWD /MTU it is not capable of achieving limiting peaking factors (becoming the hottest rod). SER, Staff Exhibit 1, at 3.
Therefore, the value of 5.5% DNBR represents a conservative upper bound to a range of rod bow effects.
28.
Intervenors' witness questioned whether the rod bow penalty meets the 95/95 criterion. Edwards, Tr. 634-37. Dr. Edwards testified that the lack of data for 15 x 15 0FA fuel would add uncertainty to the value chosen.
Iji.Tr.638. The Intervenors also question the use of a 5.5 percent rod bow penalty for the instant amendments instead of the 14.9 percent penalty applied in a previous Safety Evaluation (December 9,1983) issued in connection with the earlier core design amendments.
29.
During their cross-examination of the Staff's witness, Intervenors offered into evidence the December 9,1983 Safety Evaluation 1.
24 (SE) supporting the amendments authorizing the use of 0FA fuel and WABA rods at Turkey Point, which were issued prior to the instant amendments.
See, e.g., Tr. 764-782. Through the introduction of the December 9, 1983 SE, Intervenors sought, in part, to establish that the safety margin for the two Turkey Point reactors had been significantly reduced since the 1.56 calculated DNBR under the previous amendments provided a 25 percent margin over the 1.17 acceptance limit for 0FA fuel, whereas the 1.34 calculated minimum DNBR for the amendments contested here allows only a 12.7 percent margin above the acceptance limit. See Tr.
775, 812. Although we declined to receive the December 9,1983 SE as an exhibit, we allowed Intervenors to ask limited questions on the Safety Evaluation to probe whether any inconsistency existed. Tr. 781.
30.
The Staff testified that there are no rod bow data for 15 x 15 array 0FA fuel, but there are extensive data on 15 x 15 LOPAR fuel which has a geometry similar to 15 x 15 0FA fuel, but has a stronger Inconel spacer grid and therefore a greater rod bow magnitude than 0FA fuel.
Thus, using this data base for the rod bow penalty is conservative.
Hsii, Tr. 818.
In addition, the use of a 5.5 percent rod bow penalty instead of the 14.9 percent penalty was appropriate since the 5.5 percent penalty was based on an improved calculational method which was approved by the NRC Staff. Hsii, Tr. 813-16.
31.
Testimony by the Staff and Licensee persuade us that the reduction in the operating margin above the safety margin DNBR e
25 acceptance limit of 1.17 is not significant. The Staff and Licensee testified that the calculated DNBR for these amendments was lower due to the increase in peaking factors (F delta H, F sub Q) which makes the hottest channel in the core hotter and thus lowers DNBR. Hsii, Tr.
810-11; Dzenis, Tr. 341. Further, the Staff testified that the lower operating DNBR margin of 12.7 percent is not a reduction in a safety margin because the safety margin is provided by the 95/95 DNBR limit of 1.17.
Hsii, Tr. 901-02.
32.
The evidence establishes that the rod bow penalty meets the 95/95 criterion of the SRP. The assumptions regarding burnup and rod bow location and the use of data for fuel of similar geometry, but which has greater rod bow magnitude due to its grid design were appropriate conservatisms. Accordingly, based on the evidence adduced by Licensee and Staff, the Board finds that the rod bow penalty meets the 95/95 criterion.
Mixed Core Penalty 33.
The Licensee used a homogeneous core model to calculate DNBR for a transitional mixed core containing LOPAR and 0FA fuel and accounted for the effects of the mixed core by applying a penalty to the homogeneous core model results. The homogeneous core model safety analysis calculations produced a minimum DNBR value of 1.34 to which any penalties must be applied. The mixed core penalty accounts for the fact
,..,.n,
f 26 that coexistence of two different fuel designs having different hydraulic resistance characteristics affects the cross flow between the different fuel bundles in such a way that the fuel design having the higher grid resistance will have less flow. Since the 0FA fuel has higher grid resistance, more flow would be diverted to the LOPAR fuel.
Since the plant-specific safety analysis was perforwed with the assumption of either a whole core of 0FA or a whole core of LOPAR fuel, a penalty was applied to the 0FA analysis results to account for this decreased flow, i.e., the DNBR calculated for a whole core of 0FA fuel is reduced by the mixed core penalty. No penalty was applied to the LOPAR fuel since a mixed core configuration is advantageous to LOPAR fuel in that more flow is diverted to the LOPAR fuel.
(Hsii,'
ff. Tr. 733, at 13-14).
34.
The reduction in flow through the OFA fuel was quantified through experiments on the hydraulic characteristics of both the OFA and LOPAR fuel assemblies. Dzenis at Tr. 312. The hydraulic characteristics established by these experiments were used to determine the percent difference in the DNBR between a homogenous core and a mixed core for various reactor conditions. These calculations indicated that a 3% DNBR reduction, applied to the OFA fuel, was sufficient to bound all effects for the transition (mixed) core geometry. Hsii, ff. Tr. 733 at 13, 14, 17-18; Dzenis ff. Tr. 302 at 7.
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4 27 35.
The 3 percent mixed core penalty is based on a sensitivity study using NRC approved methods performed specifically for the 15 x 15 0FA and 15 x 15 LOPAR fuel mixed core. The sensitivity study was performed with the THINC code by using a homogeneous core model and various mixed core models, including the worst mixed core configuration where one OFA assembly is completely surrounded by LOPAR assemblies.
The difference in the DNBR calculated with a homogeneous 0FA model and mixed core models are calculated for the cases analyzed at various reactor operating conditions. The results showed the maximum difference is less than 3 percent. Thus, a 3 percent mixed core penalty is used as a bounding value. Hsii, ff. Tr. 733, at 17-18; Dzenis,.ff. Tr. 302, at 7; Tr. 318.
36.
Intervenors maintained at the hearing, as they do in their Proposed Findings (1125, 32), that the mixed core penalty does not meet the 95/95 criterion (Edwards, Tr. 634-35) and that studies on the mixed core were " hypothetical" because they were mathematical, unconfirmed by physical measurements and derived from testing that was not reflective of large scale or full core measurements.
Id. Tr. 573-74. Based on the evidence presented and the proper weight to be accorded the testimony of the witnesses, the Board concludes that the 3 percent penalty appropriately bounds the effects of the transitional core.
37.
The mixed core penalty of 3 percent was chosen as the absolute upper bound of mixed core effects based on three core geometries which
,,-.,----.-.,-...w
28 were chosen to envelope the range of possible geometries during the i
transitional core: an 0FA assembly surrounded by LOPAR fuel, a checkerboard configuration and a row of 0FA assemblies adjacent to a row of LOPAR. All other configurations are subsets of these three. Dzenis, ff. Tr. 302, at 7; Tr. 382-85.
38.
The least favorable configuration from a thermal-hydraulic and mixed core penalty viewpoint was the case where a single OFA assembly was surrounded by eight LOPAR assemblies. The conservatism of the 3%
bounding estimate for mixed core effects derives in part, at least, from the fact that fuel loadings are planned to proceed in one-third core increments. Such large increments would virtually preclude the likelihood that the above-described least favorable configuration (a 1 of 9 ratio) would result. The 3% penalty is reported to bound all of the fuel assembly configurations studies, including the unlikely geometry of an 0FA assembly completely surrounded by LOPAR fuel assemblies. Dzenis, at Tr. 382-385; Hsii, at Tr. 877-878.
39.
The THINC code, which has been approved for use for about ten years, has been verified by data which show the code can perform thermal hydraulic analysis.
In accordance with the SRP, empirical data were used to verify the code's capability to predict core flow distribution.
The code has not been empirically tested against the mixed core geometry, but as a matter of engineering judgment, it was concluded that the 4.5 percent difference in the flow resistance between a mixed and a i
-n
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e
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29 honogeneous core is too small to affect the THINC code's capability.
Hsii, Tr. 855-59.
40.
The hydraulics of the mixed core are simple to model using the code if the resistance of every channel at every location and the total flow rate is known. A resistance network can be developed and the flow distribution through the core can be calculated.
Hsii, Tr. 754. The Staff also performed an independent calcuidtion using codes similar to the THINC code and verified that Westinghouse's three percent mixed core penalty was the right magnitude.
Ijf. Tr. 729-30.
41.
The Staff testified that a more precise approach to calculate the minimum DNBR for a mixed core would be to perform the calculations with a model representing the mixed core. However, using a homogeneous core model to calculate the mixed core minimum DNBR is also acceptable as long as the effect of a mixed core on DNBR is accurately accounted for by a suitable quantity for the mixed core penalty. Hsii, ff. Tr. 733, at 13.
42.
Staff and Licensee testified that applying the mixed core penalty to the DNBR calculated with a homogeneous core configuration results in a more conservative DNBR than that calculated with a mixed core model.
(Hsii, ff. Tr. 733, at 14; Dzenis, Tr. 384-85). The Intervenors offered no evidence to the contrary.
,,,-,-----.n-
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30 43.
The NRC has approved the homogeneous core approach and mixed core penalty for Westinghouse plants on a generic basis. This approach is not unique to Turkey Point, but has also been used at various plants having transitional mixed cores. Hsii, ff. Tr. 733, at 14.
44.
Dr. Edwards' insistence that the mixed core penalty be verified against measured data may be misplaced.
Even measured data have uncertainties associated with them. Hsii, Tr. 748. The prevailing concern is whether the penalty is conservative.
In light of the relatively simple process of analyzing the hydraulics of the mixed core, the analysis and results of the worst case configuration, and the small difference in the hydraulic resistance, we are confident that the calculated penalty bounds the effects of the transitional core.
45.
The Board finds that the 3 percent mixed core penalty is sufficiently conservative, that it is not unreasonable to presume that it meets the 95/95 standard.
Applicability of WRB-1 Correlation l
t 46.
At the time the amendments which are the subject of this proceeding were being evaluated by the NRC Staff, the WRB-1 CHF l
correlation had been approved for application to 15x15 R-grid LOPAR fuel, l
17x17 R-grid LOPAR fuel, and 17x17 0FA fuel,
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31 with a DNBR safety margin acceptance limit of 1.17.
Information demonstrating applicability of the WRB-1 correlation to both 14x14 and 15x15 0FA fuel, including actual test data specifically representative of 14x14 0FA fuel, had been submitted to the NRC Staff for review.
In the absence of either a completed generic review or particular test data specifically representative of 15x15 0FA fuel, however, the NRC Staff imposed a 2% penalty for the evaluation of the Turkey Point amendments as a conservative measure. Hsii, ff. Tr. 733, at 6-7,18-19; SER, Staff Exhibit 1, at 4.
47.
Staff review of the additional infonnation has now been completed. As a result, the Staff has concluded that the WRB-1 correlation is also applicable to both 14x14 and 15x15 0FA fuel with a DNBR safety margin acceptance limit of 1.17.
Hsii, ff. 733, at 18-19.
Accordingly, there is properly no penalty for application of the WRB-1 correlation to 15x15 0FA fuel, and the 2% uncertainty previously assigned -- even though it can be accommodated within the 12.7% margin between the 1.34 safety analysis minimum DNBR and 1.17 DNBR acceptance l
limit -- is correctly 0.0%.
See, e.g., Dzenis, ff. Tr. 302, at 8.
48.
During the hearing, the Intervenors, while not identifying any deficiencies in the analysis employed, expressed some surprise that the WRB-1 correlation should be applicable to 15x15 0FA fuel.
E_.g.,
l l
l l
32 Tr. 325-25. To the contrary, however, based on a consideration of test results and the geometries involved, such a result is not all unexpected. Actual test results have demonstrated that the WRB-1 correlation is applicable to 15x15 R-grid LOPAR fuel, 14x14 0FA fuel, and 17x17 0FA fuel.
Hsii, ff. Tr. 733, at 5-7.
15x15 0FA fuel has the same fuel diameter, rod pitch, heated length and grid spacing as 15x15 R-grid LOPAR fuel; the only difference is in the grid designs. Hsii, ff. Tr. 733, at 18.
On the other hand, 14x14 and 17x17 0FA fuel have mixing' grid designs similar to 15x15 0FA fuel, but differ in rod diameter. Hsii, ff. Tr. 733, at 6, 18. Accordingly, test results demonstrating applicability of the WRB-1 correlation to the three types of fuel listed immediately above essentially encompass all of the physical aspects of 15x15 0FA fuel. Thus, it is not surprising -- but, rather, to be expected -- that the geometry of 15x15 0FA fuel is within the applicability range of the WRB-1 correlation.
Independence of Mixed Fuel Core Hydraulic and Rod Bow Effects and the WRB-1 Correlation P'1alty 49.
Intervenors argue that "[i]t is entirely likely that the rod bow phenomenon might interact in a fairly complicated way with the 1
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33 already complicated non-uniform hydraulic resistance phenomenon."
Edwards Affidavit, ff. Tr. 606, at 5.
Intervenors presented no evidence to support their claim. See Edwards at Tr. 593-94. Both the Staff and Licensee witnesses, however, indicated that the rod bow phenomenon and the differential resistance of the 0FA and LOPAR fuels to flow in the mixed core are independent phenomena, which are subject to separate I[.l.,Dzenis, modeling and the application of independent penalties.
j ff. Tr. 302, at 8; Hsii, ff. Tr. 733, at 19-21.
50.
The Staff testified that the penalty for the application of WRB-1 to the 15 x 15 0FA was independent of the rod bow penalty and mixed core penalty because the correlation was developed without the consideration of, and was not influenced by, rod bowing or the mixed core configuration. Hsti, ff. Tr. 733, at 19. The Licensee agreed that there was no interaction between the mixed core and WRB-1 and testified i
that the WRB-1 was applicable to a mixed core since the flow reduction was within the range of applicability of the correlation. Dzenis, Tr. 389-91.
51.
A mixed core configuration does not increase fuel rod bowing or the rod bow penalty on DNBR.
(Hsii, ff. Tr. 733 at 19; Dzenis, Tr.388-89).
Fuel rod bowing reduces the subchannel rod-to-rod gap (gap I
closure). Test data show that there is no noticeable effect on CHF when the gap closure is less than 54 percent (gap closure is defined as the percent of reduction from the straight rod-to-rod gap due to rod
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34 bowing). However, greater gap closure results in a reduction in CHF.
The exact mechanism of the adverse rod bow effect on CHF is not known but the evidence from the bow-to-contact test data suggests that the reduction in CHF due to rod bow is a highly localized phenomenon caused by the starvation of coolant in the vicinity of the point of contact.
Even though the fuel bundle coolant flow rate has an effect on the subchannel CHF without rod bowing, the test data show that the " bow effect parameter" (a measure of the difference between the unbowed CHF and bowed CHF) is not noticeably affected by the coolant flow rata.
Hsii, ff. Tr. 733 at 19-20.
52.
It is also apparent that rod bow has no significant effect on the hydraulic characteristics of the mixed core. A fuel rod is over 12 feet long.
It is supported about every two feet by a grid structure which serves as the structural element of the fuel assembly. Dzenis at Tr. 328. The distances between adjacent fuel rods are approximately an eighth of an inch, with the vast majority of the area of a fuel assembly occupied by the fuel rods.
Id. at Tr. 328-29. The deflections that l
occur with rod bowing are, in most cases, only a few hundredths of an l
inch over an axial distance of approximately 2 feet. The total l
localized change in flow area is very gradual and very small. The total l
flow area of the fuel assembly is essentially unchanged.
Id, at Tr. 329. There are numerous engineering studies concerning the effects of changes in flow area on flow regime. This change in local flow area is far too gradual and insignificant to cause any hydraulic
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Characteristic change or resulting effect on mixed-core DNBR penalty.
Id. at Tr. 328-29.
For a mixed core with 0FA and LOPAR fuel, the flow reduction through the OFA is approximately 2 to 3 percent. The reduction of flow rate of this magnitude would not affect the localized phenomenon of CHF reduction due to rod bow. Thus, although there may be a physical relationship between the reduction in DNBR due to rod bowing and the flow reduction due to fuel bundle hydraulic resistance, the effect is of a lower order and, as a valid engineering approximation, can be neglected.
It is the Staff's technical judgment that it is acceptable to assume that there is no interaction between the effects of fuel rod bowing on CHF and the flow change caused by a mixed core configuration for calculations to determine DNBR. Therefore, the rod bow penalty and the mixed core penalty are independent of each other.
Hsii, ff. Tr. 733, at 20-21.
1 53.
Based on the evidence presented by Staff and Licensee that the penalties are or can be considered independent and the failure of l
Intervenors to present any evidence to the contrary, the Board concludes that it is reasonable to assume that the penalties do not interact with each other and no additional penalties for interactions are required.
Summary 54.
We agree that the SRP's 95/95 standard is met by assuring that the minimum DNBR calculated for all normal operation and anticipated v
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36 operational occurrences, after accounting for uncertainties, is greater than the 95/95 DNBR safety margin design limit. The total penalty for rod bow (5.5%), the mixed core (3%) and the application of the WRB-1 correlation to the 15 x 15 0FA fuel (2%) is obtained from simple sum: tion and is 10.5 percent. The calculated minimum DNBR for Turkey Point 0FA fuel is 1.34.
The design DNBR safety margin limit for the WRB-1 CHF correlation is 1.17, and the reduction in DNBR margin from 1.34 to 1.17 is 12.7 percent, which is greater than the 10.5 percent total penalty calculated for the plant.
Intervenors offered no evidence that the penalties did not bound the phenomena, nor did their questioning persuade us that the quantities were not conservative.
Therefore the SRP's 95/95 standard is met. Hsfi, ff. Tr. 733, at 21-22; Dzenis, ff. Tr. 302, at 4-6.
55.
The Board is confident that the witnesses for the Staff and Licensee were competent to offer expert opinions on this subject.
Dr. Edwards' perceived role as a " trouble shooter" regarding mathematical modeling (Tr. 707) assisted the Board in shrpening the issues. On the other hand, Dr. Edwards' lack of expertise in DNBR analysis and failure to review all the documentation supporting the l
values of the penalties lead us to reject his claim that they are not 95/95 values.
56.
While conservative engineering approximations may not satisfy the rigors of an applied mathematician's academic discipline, the Board finds no evidence that the three penalties either significantly interact
!l'
37 with each other or do not meet the 95/95 standard. The Board concludes that the Licensee's analysis of DNBR and calculated DNBR for all normal and anticipated operational occurrences was performed using NRC approved methods, the three penalties assessed were either a 95/95 value or a bounding value, which equalled or exceeded an equivalent 95/95 standard and the calculated minimum DNBR of 1.34, after accounting for uncertainties, is greater than the DNBR acceptance limit for 0FA fuel.
Thus, the SRP's 95/95 standard is met.
Third Board Question Whether, if that standard is not being satisfied, the reduction in the margin of safety has been significant.
57.
With regard to question 2, the Board found that the SRP's 95/95 standard is met by assuring that the minimum DNBR calculated for all normal operation and anticipated operational occurrences, after accounting for uncertainties, is greater than the 95/95 DNBR design limit. The total penalty for rod bow (5.5%), the mixed core (3%) and the application of the WRB-1 correlation to the 15 X 15 0FA fuel (2%) is obtained from simple summation and is 10.5%. The calculated minimum DNBR for Turkey Point 0FA fuel is 1.34.
Since the design DNBR safety margin limit for the WRB-1 CHF correlation is 1.17, the DNBR margin between 1.34 and 1.17 is 12.7%, which is greater than the 10.5% total penalty calculated for the plant.
(See 9 52, supra). This resulted i.
r
38 from a thorough revicw cf all of the evidence and the Board concludes that there is no reduction in the margin of safety for the Turkey Point units as a result of the license amendments at issue in this proceeding.
58.
In sum, the evidence clearly shows that while there may be a reduction in the " operating margin" for the plant, there is no reduction in the margin of safety as a result of the amendments in this proceeding. The 95/95 DNBR limit of 1.17 provides the margin of safety and the 1.34 calculated DNBR for the amendments, after accounting for uncertainties, is greater than the 95/95 limit.5 III. CONCLUSION Based upon the entire evidentiary record in this proceeding, and upon the foregoing findings of fact, the Board concludes the following:
1.
The Licensee's analysis of DNBR performed using NRC Staff approved methodology and compensating for appropriate uncertainties demonstrates at a 95 percent probability at a 95 percent confidence level that the hottest rod will not undergo DNB.
5Because we conclude that there has been no reduction in the margin of safety provided by the 95/95 standard, we reject Intervenors' suggestion that we delete the amendments. See Intervenors' Proposed Findings at 25-28.
.n w-r,
i 39 2.
Contrary to Intervenors' assertion in Contention (d), the margin of safety for the operation of the Turkey Point Plant has not been reduced by the issuance of the contested amendments.
IV. BOARD NOTIFICATION REGARDING CONTENTION (b)
The record in this proceeding was closed on December 12, 1985.
Tr. 913.
On August 16, 1985, the Board granted the Licensee's Motion for Summary Disposition of Intervenors' Contention (b), which states:
Whether the entirely new computer model used by the utility, for calculating re-flood portions of accidents meets the Commission's ECCS Acceptance Criteria: specifically, whether a 2.2% reduction in re-flood rate is misleading because for a small decrease in re-flood rate, there results a large increase in fuel temperature.
Re-flood rates are critical if below 1 or 2 inches per minute.
On June 30, 1986, the NRC Staff, through Board Notification BN-86-17, provided the Board with a copy of a June 2, 1986 Westinghouse v
Electric Corporation letter and non-proprietary Topical Report which informed the Staff of the need to make some additions and corrcctions to the Westinghouse 1981 Emergency Core Cooling System (ECCS) evaluation model using the FLECHT correlation and the 1981 ECCS evaluation model using the BART computer code. Although the Licensing Board's grant of summary disposition of Contention (d) was based primarily upon the i
1 1
e a
40 former, we considered both in connection with the matter. LBP-85-29, 22 NRC 300 (August 16,1985).
The notification stated:
the staff believes that the rationale underlying the Board's summary disposition order will not be adversely affected by)the new information.
First, the Board's dismissal of Contention (b was based primarily on the ECCS evaluation model calculation using the FLECHT correlation and there is only, at most, a 12 F estimated increase in the previously calculated PCT (i.e., 2152*F). Second, the staff expects that the PCT calculation using the corrected ECCS evaluation model using BART would be below 2200 F.
Thus, the staff expects that a corrected analysis with both models would satisfy 10 CFR Part 50, Appendix K, and 10 CFR 50.46.
However, the Board Notification also states that, "given the maximum increase resulting from the errors," the Staff is considering the actions necessary for interim and continued operation with respect to both Westinghouse plants which will remain within the 2200'F acceptance criterion specified in 10 C.F.R. 9 50.46(b) and plants which may exceed the criterion. The Staff stated it would keep the Board informed of its actions with respect to the matter.
In view of the information provided in Board Notification BN-86-17, the Licensing Board will retain jurisdiction in this matter pending further actions by the Staff with respect thereto.
.i l
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9 t
41 V.
ORDER WHEREFORE, in accordance with the Atomic Energy Act of 1954, as amended, and the Rules of Practice of the Comission, and based on the foregoing findings of fact and conclusions of law, IT IS ORDERED THAT License Amendment Nos. 99 and 93 to Licanse Nos. DPR-31 and DPR-41, respectively, issued by the Office of Nuclear Reactor Regulation on December 23, 1983 shall remain in full force and effect without modification.
IT IS FURTHER ORDERED that the Licensing Board shall retain jurisdiction in this matter pending receipt of information of any further actions by the Staff in regard to Board Notification BN-86-17 dated June 30, 1986.
IT IS FURTHER ORDERED, pursuant to 10 C.F.R. 5 2.760, that this Initial Decision shall constitute;the final decision of the Commission thirty (30) days from its date of issuance, unless an appeal is taken in accordance with 10 C.F.R. 6 2.762 or the Commission directs otherwise.
See also 10 C.F.R. 95 2.785 and 2.786. Any party may take an appeal from this Decision by filing a Notice of Appeal within ten (10) days after service of this Decision. A brief in support of such appeal shall be filed within thirty (30) days after the filing of the Notice of Appeal (forty (40) days if the appellant is the Staff). Within thirty (30) days after the period has expired for the filing and service of the l
42 1
briefs of all appellants (forty (40) days in the case of the Staff), any party who is not an appellant may file a brief in support of, or in opposition to, the appeal of any other party. A responding party shall file a single responsive brief, regardless of the number of appellants' briefs filed.
THE ATOMIC SAFETY AND LICENSING BOARD YW Robert M. Lazo, ChairmarU ADMINISTRATIVE JUDGE
-)
W Richard F. Cole ADMINISTRATIVE JUDGE b-L
'Emeth A. Luebke ADMINISTRATIVE JUDGE Dated at Bethesda, Maryland, this 24th. day of July, 1986.
i APPENDIX A LIST OF EXHIBITS Licensee's Exhibits None NRC Staff Exhibits No.
Description Date Identified Admitted at Tr.
(Rejected) at Tr.
1 NRC Safety 12/23/83 735 736 Evaluation Intervenors' Exhibits None
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.