ML20205K624

Proposed Findings of Fact & Conclusions of Law in Form of Initial Decisions Ordering Licensee 830819 & 0909 Applications to Amend Licenses DPR-31 & DPR-41 on Vessel Flux Reduction to Remain in Effect.Certificate of Svc Encl

ML20205K624
Person / Time
Site:	Turkey Point
Issue date:	02/24/1986
From:	Matt Young NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
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References
CON-#186-223 OLA, NUDOCS 8602270599
Download: ML20205K624 (38)

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%'/ DOCKETED USNRC UNITED STATES OF AMERICA 16 FEB 26 #0:16 NUCLEAR REGULATORY COMMISSION

-BEFORE THE ATOMIC SAFETY AND LICENSING BOARD (C 7 NU *. ;ki I;k .

BRANCH In the Matter of )

) Docket No. 50-250 OLA-1 FLCRIDA POWER & LIGIIT COMPANY ) 50-251 OLA-1

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(Turkey Point Plant, Units 3 and 4) ) (Vessel Flux Reduction)

NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN TIIE FORM OF AN INITIAL DECISION Mitzi A. Young Counsel for NRC Staff February 24, 1986 I

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'Jf NRC UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

-BEFORE THE ATOMIC SAFETY AND LICENSING BOARD EM

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00CHETING s ICMu ONE In the Matter of )

) Docket No. 50-250 OLA-1 FLORIDA POWER a LIGIIT COMPANY ) 50-251 OLA-1

)

(Turkey Point Plant, Units 3 and 4) ) (Vessel Flux Reduction)

NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN TIIE FORM OF AN INITIAL DECISION The NRC Staff, in accordance with 10 C.F.R. 6 2.754 and the Board's ruling during the December 12, 1985 hearing session (Tr.

909-10), propescs the following findings of fact and conclusions of law in the form of an initial decision.

I. INTRODUCTION AND DACKGROUND

1. By letters dated August 19, 1983 and September 9,1983, Florida i

Power & Light Company (Licensee) requested amendments to the technical specifications of Licenses DPR-31 and DPR-41 for its 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 I

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operating with the old steam generators having a greater number of plugged tubes than the new steam generators now in use. I-

2. A notice of consideration of the issuance of the proposed amendments and an opportunity for hearing was published in the Federal Register on October 7, 1983. 48 Fed. Reg. 45,862. In response to the notice , the Center for Nuclear Responsibility, Inc. (CNR) and Joette Lorion (Intervenors) filed a timely joint petition to intervene. On December 23, 1983, pursuant to 10 C.F.R. 50.91(a)(4), the NRC Staff made a final no significant hazards determination and issued the proposed amendments.

3. The Intervenors filed an amended petition on January 25, 1984.

17e ruled on the contentions filed then, and on other matters, in our May 16, 1984 Order and admitted only Contentions (b) and (d).

Contention (b) alleges shortcomings in one of the computer models used in the prediction of the temperature of the hottest rod in a reactor core during reflood of the core after a loss of coolant accident (LOCA).

Contention (d) alleges that under the amendments there is an increase in the probability that films of steam will form around the fuel rods l

(departure from nucleate boiling or DNB) during normal operation and l

1_/ 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 j limit from 2.30 to 2.32; 3) to change the overpower delta-T trip set j points and thermal hydraube 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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1 anticipated operational occurrences, thus significantly reducing the margin ;

of safety. 2/

4. 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 which the Licensee made a " didactic presentation" as ordered by this Board concerning issues raised in the parties' summary disposition papers. See LBP-85-29, 22 NRC 300, 306-310 (1985). The Intervenors i

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. _I d .

5. By Order dated August 16, 1085, we granted Licensee's motion for summary disposition of Contention (b), but denied the motion for 2/ Contention (d) states 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 DNDR 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 t.bove the boundary of the nucleate boiling regine 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 amendtrent will both significantly reduce the safety margin and significantly increase the probability of serious consequences from an accident.

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 [ Optimized 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 DFBR 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. h'hether, if that standard is not being satisfied, the reduction in the margin of safety has been cignificant.

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 filed a second motion for summary disposition of Contention (d) on i

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 j Order, dated November 18, 1985. Evidentiary hearings were held in fliami, Florida from December 10 through December 12, 1985.

G. In our Order ruling on summary disposition, we explained the significance of Intervenors' contention by describing conditions surrounding the fuel rods during normal operation and when departure from nucleate boiling (DNB) occurs. Briefly, bubble.s of steam form on the surfaces of the fuel rods under certain temperatures and are swept away by the flow of the coolant during normal operation. This stage of l

boiling at which bubbles form and leave the rod surface is called nucleate l

l boiling and is an efficient means of transfering heat from the rods to the t

5-coolant . When the temperature of the fuel is high enough, departure from nucleate boiling occurs, that is, some bubbles of steam remain on the surface of the rods, coalesce with adjacent bubbles and begin to form a film over the surface of the rod, thereby insulating the rods and impeding the transfer of heat away from the rods. The point at which heat removal away from rods through bubbles of steam begins to decline is called the critical- heat flux (Cl 2). 22 NRC at 323-?5.

7. It is not possible to say with a high level of certainty what the CI!F for a given kind of fuel and operating conditions would be because CHF correlations used to predict CIIF afford different degrees of assurance. Thus the NRC imposes a statistical measure of prudence such thet:

For a given plant, with a given kind of fuel, and a given set of operating conditions, the minimum ratio between CIIF and AHF [ actual heat flux] called the minimum departure from nucleate boiling ratio (DNBR

-- must afford at least a 95% confidence level that there is a 95% probability that DNB will not be reached on the hottest rod in the core during either normal operation or certain abnormal occurrences other than LOCAs. (footnote omitted).

22 NRC at 324-25. t'e call this statistical measure of prudence the 95/95 standard , which is found in the NR C's Standard Review Plan (SRP),

NUREG-0800 (July 1981) at 4.4-2 to 4.4-3, a guidance document for the Staff in the exerciac of its licensing duties. The license amendments at issue here impose different minimum DNBRs on the low parasitic (LOPAR) and optimized fuel assemblies (OFA) fuel in Units 3 and 4 as a result of the differences between the CI!F correlations and their respective DNBR limits (i.e. ,1.3 for LOPAR and 1.17 for OFA). M . at 325.

8. The record on summary disposition led this Board to question whether a DNBR of 1.17 accounts for the three uncertainties, as outlined in the . Staff's SER, associated with rod bowing, the transitional core containing 'OFA and LOPAR fuel and the application of the WRB-1 correlation to 15 x 15 array OFA 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.

II. FINDINGS

9. The Licensee's direct case consisted of testimony by Edward A.

Dzenis (ff. Tr. 302), Mansfrer 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 Engineer-ing. 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 DND, and analyses of other safety criteria. Mr. Dzenis has also been involved in modifica-tions of the THINC code to incorporate new correlations such as the WRB-1 critical heat flux correlation. Professional Qualifications and

Experience of Edward A. Dzenis , ff. Tr. 302; Tr. 293-302. At the completion of voir dire, Mr. Dzenis' testimony, profiled on November 26, 1985, was received in evidence without objection and bound into the transcript .

10. The Staff's direct case consisted of testimony by Dr. Yi-Hsiung lisii, 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. Hsil has a B achelor , Master and Doctorate degree in l'.lechanical Engineering. He has taken undergraduate courses in hydrodynamics, thermodynamics, heat transfer, calculus , differential equations, and graduate courses in hydrodynamics , heat trans fer, thermodynamics, advanced calculus and complex variables. Tr. 714-716. Since he joined the NDC in 1981, he has reviewed safety evaluation reports and reload methodology topical reports on core thermal hydraulics, including CHF correlations, submitted by applicants and licensees. Dr. Ilsii worked for Babcock and Wilcox from 1967 to 1981 where he performed core t

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. IIsil also developed a computer program to calculate core performance and DNBRs.

(Hsil Professional Oualifications, ff. Tr. 733; Tr. 715) . - In addition, pursuant to 10 C.F.R. 5 2.743(g), the Board received Staff Exhibit 1,

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Intervenors withdrew their objection to the admission of Dr. Ilsli's prefiled testimony and statement of professional qualifications after conducting a voir dire examination (Tr. 733).

the NRC Safety Evaluation s 'I orting the amendments, dated !

December 23, 1983, for the pti %se af documenting the NRC's review of the thermal hydraulics associated with the amendments. (Tr. 735-36).

11. The Intervenors' direct case consisted of testimony by Dr. Gordon D. .T . Edwards ( ff. Tr. 606), President of the Canadian Coalition for Nuclear Responsibility and Professor of Mathematics and Science at Van!cr College, f.lontreal, Canada. Dr. Edwards holds a Ph.D.

in Mathematics, has taught university-level mathematics for several years and has limited experience teaching biologv and chemistry. Tr. 254-57, 505. lie 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.

Ilcrever, as Dr. Edwards himself acknowledged, he generally has no knowledge , skill, experience, training or education in the field of engineering (Tr. 538) nor does he consider himself to be an expert in the areas of heat transfer, departure from nucleate boiling testing, critical heat flux correlation , determination of operational limits or evaluation of DNER. Tr. "83.

12. 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 i

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acknowledged that he was not familiar with the mathematical equations or computer models used to evaluate and analyze the DNB and DNDR at Turkey. Point. Tr. 506-07.

13. 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. I 2.733.

14. Licensee objected to Dr. Edwards' interrogation as an expert in that by his own admission he was not qualified by training or experience in thermodynonics, heat transfer, fluid mechanics or thermal hydraulic analysis , all of which topics were central to the narrow issues at the hearing. Tr. 284-85. The Staff did not object to Dr. Edwards' conduct-ing cross-examination as an expert provided that he examined only in those areas of his admitted expertise , that is, mathematics, including mathematical analysis , calculations of probabilities and the use of mathematical models. Tr. 285-86. 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. I 2.733; Tr. 285-86. The Board found 7

Dr. Edwards to be qualified as an expert interrogator pursuant to 10 C.F.R. I 2.733; we declined 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. See Tr. 288-89.

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15. 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 cliciting oral testimony concerning Dr. Edwards'

" response and explanation" to the three Board questions. Tr. 446.

16. Staff and Licensee objected to this procedure as falling outside the Commission's Rules of Practice,10 C.F.R. E 2.743(b), which requires all parties to file written direct testimony in advance of any hearing.

Tr. 446-64. 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 (Tr. 462) 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. I 2.743 and be unfair to

! opposing parties. Tr. 476.

17. The Staff and Licensee objected to Intervenors' subsequent proffer of written direct testimony , which consisted of two affidcVits previously prepared by Dr. Fdwards in response to notions 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. Tr. 478-91. We ruled that the August 30, 1984 affidavit was stale and its introduction contained an element of surprise. Ilowever , the Staff and Licensee had had reasonable i

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opportunity to examine the later affidavit , dated November 5, 1985

(" November Affidavit" or " Edwards Affidavit") . Tr. 487-88. Thus, pursuant to 10 C . T. R . I 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 would be received in evidence provided it withstood voir dire and any motions to strike. Tr. 496.

18. Dased on the evidence adduced through the voir dire examination of Dr. Edwards as a proposed expert witness, Licensee moved to strike Dr. Edwards' testimony in its entirety (Tr. 544) 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 (Tr. 546-551); (2) portions of the November 5 Affidevit were virtunlly identical to, or unduly repetitious of, statements in the Outline (Tr. 548-550); and (3) portions of the testimony purportedly were irrelevant and lacking in probative value (Tr. 550).

4 Despite these objections, we found Dr. Educeds was qualified 19.

as an expert witness in view of the " limited scope and the qualified language" of his testimony and admitted the Outline and the November 5 Affidavit into evidence. Tr. 556.

20. 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 v'ith no knowledge, education, skill, training or experience

in engineering. While Dr. Edwards' familiarity with reactor 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 based on a complete knowledge, or even reading, of all the documents underlying the review that has been performed. Tr. 644; see Tr. 636-37. 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. We now turn to a discussion of the three questions posed.

A. Question 1 Whether the DNBR of 1.17 which the amendments impose on OFA fuel in Units 3 and 4 compensates for the three uncertainties

[ associated with rod bow, the mixed core and the application of the WRB-1 correlation].

21. All the parties agree and we conclude that the answer to our r first question is that the DNBR of 1.17 does not compensate for the noted 4

uncertainties. Dzenis , ff. Tr. 302, at 3; IIsii, ff. Tr. 730, at 22; Edwards Outline, ff. Tr. 606, at 1. The parties disagree, however, as i

to the significance of this answer, l l

22. A DNBR limit for a particular fuel type is the quantity imposed on a CIIF correlation as the specified acceptable fuel design limit 4/ to 4_/ For. example, 10 C.F.R. Part 50, Appendi:: A, Criterion 10, Reactor Design states:

(FOOTNOTE CONTINUED OF PEXT PAGE) 1

ensure at a 95/05 level that the hot fuel rod in the core will not e:;perience DNB during normal operation and anticipated operational occurrences. IIsil, ff. Tr. 733, at 3. The 1.17 "DNBR design limit" (Hsil, ff. Tr. 733, at 3) or "DNBR acceptance limit" (Dzenis , ff.

Tr. 302, at 3) for the WRB-1 correlation is lower than the DNBR limit of 1

1.3 for the W-3 correlation because the WRB-1 correlation can predict CIIF with less uncertainty (lisii, ff. Tr. 733 at 4-5). This 1.17 limit for the WRB-1 correlation is generic to all Westinghouse plants using OFA fuel, Dzenis, ff. Tr. 302, at 3,

23. If the truc CHF value could be calculated and the actual heat flux were precisely known, the exact DNBR could be determined and a design DNBR limit of 1.0 would ensure DNB would be avoided. However,

' because CIIF is calculated using an empirical correlation developed based on experimental CHF data and because of random variations in the data upon which the correlation is based, the exact CIIF cannot be predicted.

i A DNBR limit greater than 1.0 is therefore imposed to account for this

, uncertainty. The DNBR limit for a correlation is the value which ensures I

with a 95/95 level that DNB will be avoided when the plant-specific DNBR calculated with the correlation is greater than this value. IIsli, ff.

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Tr. 733, at 3.

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l (FOOTNOTE CONTINUED FROM PREVIOUS PAGE) l The reactor core and associated coolant , control, and I protection systems shall be designed with appropriate l

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.

24. A DNDR of 1.17 ensures, with a 95 percent probability at a 95 percent confidence level, that the hot fuel rod in the core will not experience DNB during normal operation and anticipated operational occurrences. Hsil, ff. Tr. 733, at 7; Dzenis, ff. Tr. 302, at 3-4. The only uncertainty affecting the 1.17 DNBR limit is that resulting from the inability of the CHF correlation to precisely calculate the experimental ,

data upon which the correlation is based. Hsii, ff. Tr. 733, at 4.

25. The " calculated minimum DNBR" (Hsil, ff. Tr. 733, at 8) or

" safety analysis minimum DNBR" (Dzenis, ff. Tr. 302, at 3) is the lowest value of the DNDP.s calculated from the Licensee's predictive computer analysis using the THINC subchannel thermal hydraulic code for all anticipated transients and a CHF correlation such as h'RB-1. Input to 4

the TIIINC code are reactor conditions during the transients, including the values of reactor power, pressure, coolant flow rate, inlet temperature and power distribution, and a Feometry model representing the reactor core, fuel assemblies and subchannels. The conservative value of each process and design parameter is used by obtaining a value i

that is bounding or has a 95/05 probability / confidence level . The assumption underlying the input of all process and design parameters at their conservative value into the THINC code is that all the adverse effects occur simultaneously. Since simultaneous occurrence is unlikely, this assumption results in a lower calculated DNBR than the true value j expected during each transient. (Hsii, ff. Tr. 733, at 9-10).

l 26. Compliance with the Standard Review Plan Criterion that the het i

! rod in the core will not experience DNB with a 95/95 l

probability / confidence level is assured by (1) calculating a minimum DNBR

using the bounding or 95/95 value of the uncertainties in the process parameter and core design parameters and (2) accounting for uncertainties in the calculational methods used in the DNBR calculation through the use of penalties and thus arriving at a value for calculated minimum DNBR which is greater than the 95/95 DNBR limit of the CHF correlation used (for example, 1.17 for WRB-1) . (Hsil, ff. Tr. 733, at 11; Dzenis at 3-4.)

27. Uncertainties , which were not included in the calculational method , i.e., input in the THINC computer calculations, are treated as penalties. The three penalties pertinent to these amendments were identified in the Staff's SER (Staff Exhibit 1 at 4) as (1) a rod bow penalty of 5.5 percent, (2) a transitional mixed core penalty of 3 percent and (3) a less than 2 percent penalty for the application of the WRB-1 correlation to 15 x 15 array OFA fuel. IIsil, ff. Tr. 10-12, 15-16.

28. In short, the Board finds that a DNBR limit of 1.17 for OFA fuel (a quantity greater than 1.0 to account for the uncertainty associated with the CHF correlation based on experimental data and the random variation in the data) assures at a 95/95 level that the hot fuel l rod in the core will not experience DNB. The fact that the 1.17 D!lBR

! limit does not compensate for the uncertainties associated with rod bow, the mixed core and the application of the WRB-1 correlation to 15x15 OFA fuel is not significant as long as the plant-specific DNBR (or calculated minimum DNBR) calculated from Licensee's predictive analysis , after accounting for appropriate values for uncertainties, is greater than the 1.17 DNBR limit.

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B. Question 2:

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.

29. Licensee and Staff have answered our second question in the affirmative (eg, Dzenis, ff. Tr. 302, at 4; Ilsil, ff. Tr. 733, at 22) and presented considerable evidence to support that answer. Inter-venors, however, generally ouestion the adequacy of Licensee's DNBR analysis , particularly whether its reliance on predictive versus experi-mental analysis was adequate to establish that the SRP criteria were met.

Intervenors call the numerical penalties " educated guesses" and " crude linear approximations based on guesswork." Edwards Outline , ff.

Tr. 606, at 2; see Edwards Affidavit , ff. Tr. 606, at 3. Specifically, Dr. Edwards questioned whether the penalties meet the 95/95 standard (Tr. 634-35) and urged that the calculated penalties should be compared

to empirical data (e.g. , Tr. 643).

30. Our review of the evidence leads us to conclude that the answer to our second question is properly: Yes, the 95/95 standard is met .

31. The Staff testified that, generally, there are two approaches to account for these penaltics. In one approach, a new 95/95 DNBR limit is derived by summing up the CIIF correlation's 95/95 limit (for example 1.17 for WRB-1) and the total penalty. The TIIINC calculated DNBR is then compared to the new DNBR limit. If the calculated DNBR is greater than the new DNBR limit, the 95/95 standard is met. In the second approach, the 95/95 DNBR limit is equal to the CliF correlation's 95/95 limit (e.g.,

1.17) and the THINC calculated DNBR is reduced by the total penalty.

If the reduced calculated DNBR is greater than the 95/95 DNBR limit for the correlation used in the calculation, the SRP's 95/95 standard is more than met. Hsil, ff. Tr. 733, at 12.

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32. With respect to the second approach, the Staff indicates that it j

is a common practice that no actual subtraction of the penalty is done to reduce the calculated DNBR. Rather, the safety analysis is performed without the penalty. The minimum DNBR for each anticipated transient is calculated with the THINC code and the calculated minimum DNBR is used to calculate a DNUR nargin. This DNBR margin is the percentage i difference between the calculated minimum DNBR and the CliF correlation's 95/95 DNBR limit. This DNBR margin is then compared to the total penalty percentage. If the margin is greater than the penalty, the SPP's 95/95 standard is met. (lisii, ff. Tr. 733, at 12-13).

33. The Licensee also testified that the second approach of applying penalties to results obtained from predictive analysis rather than design basis limits such as the 1.17 DNBR acceptance limit is common in the engineering field. This approach is used for all Westinghouse safety i analyses , including those for Turkey Point, and is consistent with the NRC Standard Review Plan. Dzenis, ff. Tr. 302, at 6.

34. The minimum DNBR in the Turkey Point safety analysis was calculated using the THINC subchannel thermal hydraulic code and the WRB-1 correlation for the OFA fuel and the W-3 correlation for the LOPAR fuel. The minimun DNDR was calculated for each of the anticipated transients based on a homogeneous core model. The uncertainties of the process parameters and core design parameters were treated by using the conservative values of these parameters in the input

to the TIIINC computer code to obtain the " calculated minimum" DNBR.

Since a homogeneous core of the 15 x 15 OFA fuel was assumed in providing the input to the TilINC code, a mixed core penalty was assessed. Since the fuel rod bowing effect and the applicability of the WPB-1 correlation to the 15 x 15 OFA was not considered in the TIIINC code input , penalties for these factors were also assessed. The assessment of penalties against the calculated minimum DNBR is not unique to Turkey Point, but is used at all Westinghouse plants. Iisii, ff.

Tr. 733, at 15-16.

35. No party to this proceeding has argued that there is any defect in the 95/95 standard. Intervenors, however, argue that a DNBR limit of 1.17 should not be applied to a mixed core. Edwards, Tr. 622; Edwards Affidavit , ff. Tr. 606, at 3. Intervenors do not suggest any alternative value for DNBR (Edwards , Tr. 648) and their expert performed no independent tests or studies to determine a different number (id . ,

Tr. 623). They contend that it is improper to start with the limit for a homogencous core and then assess penalties for uncertainties, including a l mixed core penalty, when the procedure is "to avoid detailed study of the mixed core." Edwards, Tr. 635.

l 36. We find no evidence that the use of the DNER limit of 1.17, l

l which is generic to OFA fuel, for the transitional mixed core was to avoid detail study of the mixed core, it is acceptable to defir . appropriate DNBR acceptance limit based on experimental data with a particular correlation for a certain fuel type, to perform a plant-specific safety analysis using the THINC code (an NRC accepted subchannel analysis l code) to conservatively input assumptions consistent with the SRP l

e methodology or plant Technical Specifications, and ultimately to reduce that calculated DNDR (to be compared to the DNBR acceptance limit) by the appropriate uncertainties or penalties. Dzenis , Tr. 365-66.

37. As the Licensee testified, it is reasonable to start with the acceptance limit for the OFA fuct since the analyses are performed in advance of loading when the specific core wide loading pattern is not known. Exact loading patterns depend on an individual plant's operational schedule and the design requirements of particular refueling cycles. Dzenis, ff. Tr. 302, at 7; Tr. 374. In addition, the intent of the transitional period is to get to a full core of OFA and to bound the effects of the transitional period so as to obviate the need to submit cycle-by-cycle reanalyses. Dzenis, Tr. 374-75.

38. Accordingly, the Board finds that it is appropriate to use the generic DNBR limit of 1.17, which meets the 95/05 standard for OFA fuel, and assess penalties against the calculated DNBR of 1.34 to determine if the 95/95 acceptance limit is met . We turn now to a discussion of whether the three penalties meet the 95/95 criterion.

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l l 1. Mixed Core Penalty

39. The Licensee used a homogeneous core model to calculate DNBR for a transitional mixed core containing LOPAR and OFA fuel and eccounted for the effects of the mixed core by applying a penalty to the homogeneous core model results. The mixed core penalty accounts for the fact that coexistence of two different fuel designs having different

[

I 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 OFA fuel has higher grid resistance, more flow would be diverted to the LOPAR fuel. Since the plant-specific safety analysis was performed with the assumption of either a whole core of OFA or a whole core of LOPAR fuel, a penalty was cpplied to the OFA analysis resultc to account for this decreased flow, i.e. , the DNBR calculated for a whole core of OFA fuel is reduced by the mixed care 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. (lisii, ff. Tr. 733, at 13-14) .

40. The 3 percent mixed core penalty is based on a sensitivity study using NRC approved methods performed specifically for the 15 x 15 OFA and 15 x 15 LOPAR fuel mixed core. The sensitivity study was performed with the TillNC 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 OFA 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. lisii, ff. Tr. 733, at 17-18; Dzenis, ff. Tr. 302, at 7; Tr. 318.

41. Intervenors maintained at the hearing, as they do in their Proposed Findings (Si 25, 32), that the mixed core penalty does not meet the 95/"5 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 l

l I

1 l

I

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.

42. The mixed core penalty of 3 percent was chosen as the absolute upper bound of mixed core effects based on three core geometries which were chosen to envelope the range of possible geometries during the transitional core: an OFA assembly surrounded by LOPAR fuel, a checkerboard configuration and a row of OFA assemblies adjacent to a row of LOPAR. All other configurations are subsets of these three. Dzenis ,

ff. Tr. 733, at 7; Tr. 383-84.

43. The THINC code, which has been approved for use for about ten years, has been verified to data which shows the code can perform thermal hydraulic analysis. In accordance with the SRP, empirical data was used to verify the code's capability to predict core flow distribution.

The code has not been empirically tested against the mixed core, but as a matter of engineering judgment, it was concluded that the 4.5 percent difference in the flow resistance between mixed and homogencous core is too small to affect the THINC code's capability. Hsil, Tr. 855-59.

44. The hydraulics of the mixed core are simple to model using the code if the resistance of every channel and 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. IJsii, Tr. 754. The Staff also performed an independent calculation using codes similar to the THINC code and verified that Westinghouse's three percent mixed core penalty was the right magnitude. Id. Tr. 729-30.

l

45. 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 mixed core penalty, fisii, ff. Tr. 733, at 13.

46. Staff and Licensee testified that applying the mixed core penalty to the DNDR calculated with a homogeneous core configuration results in a more conservative DNDR than that calculated with a mixed core model. (Ilsil, ff. Tr. 733, at 14; Dzenis , Tr. 384-85). The Intervenors offered no evidence to the contrary.

47. 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 Turkt.y Point , but has also been used at various plants having transitional mixed cores, fisii, ff. Tr. 733, at 14.

48. Dr. Edwards' insistence that the mixed core penalty be verified against measured data may be misplaced . Even measured data has uncertaintics associated with it. lisii, 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 selection 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.

49. Accordingly, the Board finds that the 3 percent mixed core j penalty meets and in 4ct exceeds, the 95/95 standard. The evidence shows that the penalty is derived from a core configuration which

(

envelopes the effects of the transitional core and the small difference in hydraulic resistance between OFA and LOPAR assemblie's, in our opinion, would not affect the ability of the THINC code to calculate local flow conditiobs. Since only conditions immediately surrounding the fuel bundle are important for DNBR analysis, tl'e number of degree of permut'ations of core configurationn is not important. Dzenis, .Tr. 384. Moreover, the effect of assemblies away from d particuh r ' hot assembly under examination is de minimus. Hsil, Tr. 878. 5,/ Qhtis , the SRP's 95/95 criterion is met.

4

3. Rod Bow Penalty

50. The 5.5 percent rod bow penalty was used to account for the fact that fuel rod bowing results in reduction of the critical heat flux and therefore reduction in DNBR. Since the Licensee's safety analysis was Intervenors also testified et the hearing, although it was not

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mentioned in any of their written testimony, that the WABA rods further complicated the analysis of the mixed core and the accept-ability of the miFed core penalty. See Edwards, Tr. 575, 607-13.

Dr. Edwards acknowledged that the NRC Safety Evaluation on the core design amendments limited the number of ti AB A rods in the core but insisted that in his experience with calculational models a small change could be significant. -Id. Tr. 627-28. Dr. Edwards' testi-mony also demonstrated that he Tvas not familiar with the WABA rods anc' did not know whether their effect had been considered.

Tr. 627-31.

Contrary to Intervenors' assertion (Findings at 150), the Staff, in connection with the issuance of prior amendments which approved the use of OFA in the core \ adequately analyzed the. hydraulic effects of the WABA rods and concluded that as lon'g as the number of WABA rods is below the NRC prescribed limit, there is no effect on DNBR.

IIsil, Tr. 850-51; Dzenis, fr. 340. Similarly, we reject Intervenors' suggestion that a DNDR limit of 1.17 is not appropriate for a homogeneous OFA core because of the presence of WABA rods. See Intervenors Proposed Findings at 147.

performed without the assumption of fuel rod bowing in the input to the computer program, the resulting calculated DNBR of 1.34 was reduced by 5.5 percent to account for the rod bow uncertainty. Hsil, ff. Tr. 733, at 16,

51. The 5.5 percent rod bow penalty was derived based on an NRC 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 limit as was statistically demonstrated in WCAP-8691.

Ilsil, ff. Tr. 733, at 16-17; Tr. 8??.

52. The Licensee testified that the penalty is calculated based on a correlation of the measured rod bow of irradiated fuel assemblics. Since the amount of measured rod bow increases with fuel irradiation time, or burnup, the DNBR uncertainty for rod bow also increases with fuel irradiation time, or burnup. The 5.5 percent penalty value corresponds to the highest burnup, or end-of-life burnup, at which DNB is a concern because, at higher burnups, heat generation rates in PWR fuel decrease because of a decrease in the concentration of fissionable isotopes and the buildup of fission product inventory. Thus, the 5.5 percent penalty for rod bow bounds the effects of rod bow. Dzenis , ff. Tr. 302, at 7-8.

l

53. Fuel assemblies with burnups greater than 33,000 megawatt days l per metric ton cannot be the limiting assemblies with respect to DNHR.

i Hsil, Tr. 823. The manner in which the effect of fuel rod bowing on DNBR is applied as a rod bow penalty is also conservative because of the underlying assumption that the largest rod bowing occurs at the hot channel fuel rods and at the location of the minimum DNBR. The minimum

- 25 i DNBR generally occurs at the upper portion of the core whereas the worst rod bowing usually occurs at tile lower portion of the core. Also, rod bowing does not occur at every fuel rod. Data show that severe rod <

bowing generally occurs at the fuel rods having high burnup, whereas the hot channel with highest power peaking factor generally occurs with low burnup fuel. Therefore, the assumptions of the largest rod bowing occurring at the hot channel rods and at the minimum DNBR locations is I

conservative. Hsil, ff. Tr. 733, at 17; Dzenis, Tr. 386-87.

5 4 ., 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 OFA fuel would add uncertainty to the The Intervenors also question the use of a value chosen. Id. Tr. 638.

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. See Findings at FI 32-29. 50.

55. The Staff testified that there is no rod bow data for 15 x 15 arr e.; GFA fuel, but there is 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 ; rrider rod bow magnitude than OFA fuel.

Thus, using this data base for the rod bow penalty is conservative.

' IIsil, 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. IIsif, Tr. 813-16.

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d 1

4 2* m1.-

56. The evidence establishes that the rod bow penalty meets the 95/95 criterion of the SRP. The assumptions regarding burnup and rod how location and the use of data for fuel of similar geometry, but which j has greater rod bow magnitude due to its grid design were appropriate conservatisms. Accordingly, based on the evidence adduced by Licensee i 1

and Staff, the Board finds that the rod bow penalty meets the 95/95 criterion. l l

1

3. Penalty for Application of WRB-1

57. The WRB-1 correlation was originally approved for application to 15 x 15 and 17 x 17 R-Grid LOPAR fuel and later approved for application to 17 x 17 array OFA fuel with a DNBR limit of 1.17. The i Staff identified the lack of specific 15 x 15 OFA CHF data as en element of uncertainty and based on engineering judgment, concluded that the uncertainty would not be greater than 2 percent. Therefore, prior to the generic review of the application of WRB-1 to both the 14 x 14 and the 15 x 15 OFA fuel designs, the Staff used a 2 percent penalty for the j evaluation of the Turkey Point amendment as a conservative measure to i

account for the lack of the specific 15 x 15 OFA CHF data. Hsil, ff. l Tr. 733, at 18-19.

58. Additional CHF test data for the 14 x 14 OFA with a DNBR limit 4

of 1.17 has provideu be20s Aar the applications of WRB-1 to the 14 x 14 I l

OFA. The 15 x 15 OFA and the 15 x 15 R-Grid LOPAR fuel have the i same fuel diameter, rod pitch, heated length and grid spacing. The only l

difference is in the grid designs. While the 15 x 15 OFA and the 14 x 14 '

l l

OFA have the same grid designs, the only difference is in the fuel rod I

! l i i l

l

.n . . . - .

diameters which are 0.422 inches and 0.4 inches, respectively. Since WRB-1 is applicable to both the 15 x 15 R-Grid LOPAR fuel and 14 x 14 OFA and there are small differences between 15 x 15 OFA and the 14 x 14 OFA and 15 x 15 LOPAR fuels, the WRB-1 correlation would likely be applicable to the 15 x 15 OFA. Subsequently, the NRC Staff's generic review of the application of WRB-1 to both the 14 x 14 and 15 x 15 OFA fuel designs determined that there is no need for the 2 percent penalty.

Hsil, ff. Tr. 733, at 19.

59. Intervonors argue that the lacl: cf experimental data concerning 15 x 15 OFA fuel is a basis for this Board to reject the applicability of WRB-1. See e.g., Intervenors' Proposed Findings at i 31. We are not persuaded that such a conclusion is proper. The range of geometries to which the URB-1 correlation is applicable includes 15 x 15 OFA.

Dzenis, Tr. 343-44. The data base upon which the correlation was based included the geometric characteristics of 15 x 15 OFA fuel, such as rod diameter, rod pitch, grid spacings , hydraulic diameters and the grid structure of the OFA assembly. Id.

, 60. In light of the additional CHF test data concerning 14 x 14 OFA fuel and the similarities between the 15 y 15 OFA and LOPAR fuel i

assembly designs , the Board concludes that the evidence demonstrates that the WEB-1 correlation is applicable to 15 x 15 OFA fuel. -

1

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We note, however, that even if thsre is 2 percent uncertainty associated with the applicability of the correlation , the resulting combination of the penalties (less than 11 percent) is within the 12.7 percent margin between the DNBR limit of 1.17 and the calculated minimum DNBR of 1.34. IIsil, ff. Tr. 733, at 24.

S

4. Independence of Mixed Fuel Core Hydraulic and Rod Bow

- Effects and the WRB-1 Correlation Penalty 61.- Intervenors argue that "[i]t is entirely likely that the rod bow phenomenon might interact in a fairly complicated way with the already complicated non-uniform hydraulic resistance phenomenon." Edwards Affidavit , ff. Tr. 606, at 5. Intervenors presented no ' evidence to support this claim . See Tr. 593-94. Both the Staff and . Licensee witnesses , however, indicated that the rod bow phenomenon and the differential resistance of the OFA and LOPAR fuels to flow in the mixed core are independent phenomena, which are subject to separate modeling and the application of independent penalties. Eg , Dzenis , ff. Tr. 302, at 8; Ilsii, ff. Tr. 733, at 19-21.

62. The Staff testified that the penalty for the application of WRB-1 to the 15 x 15 OFA was independent of the rod bow penalty and mixed core penalty because the correlation was developed without the considera-tion of, and was not influenced by, rod bowing or the mixed core con-figuration. Hsii, ff. Tr. 733, at 19. The Licensee agreed that there was no interaction between the mixed core and WRB-1 and testified that the WRB-1 was applicable to mixed core since the flow reduction was within the range of applicability of the ' correlation. Dzenis, Tr. 389-91.

63. A mixed core configuration does not increase fuel rod bowing or the rod bow penalty on DNBR. ( Hsil, ff. Tr. at 19; Dzenis, Tr. 388-89) . Fuel rod bowl:tg reduces the subchannel rod-to-rod gap (gap closure). Test data show that there is no noticeable effect on CIIF 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 bowing) . However, greater gap closure results in a reduction in CHF.

m ._

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 cf the point of contact. Even thouFh the fuel bundle coolant flow rate has an effect on the subchannel CilF 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 rate. Hsil, ff.

Tr. at 19-20.

64. In a transitional mixed core containing fuel assemblies of different design, the flow reduction in the higher resistance assembly is a global (bundle) phenomenon. For a mixed core with OFA 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 assumption, can be neglected. It is the Staff's technical judgment that it is acceptabic to assume that there is no interaction between the effects of fuel rod bowing on CIIF and the flow changes 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. Hsil, ff. Tr. 733, at 20-21.

65. Based on the evidence presented by Staff and Licensee that the penalties are independent and the failure of Intervenors to present any

evidence to the contrary, the Board concludes that the penalties do not interact with each other. Thus, the numerical values of the penalties meet the 95/95 criterion.

5. Summary

66. We agree 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 OFA fuel (2%) is obtained from simple summation and is 10.5 percent.

67. The calculated minimum DNBR for Turkey Point OFA fuel is 1.04 Since the design DNDR limit for the WRB-1 CHF correlation is 1.17, the DNBR margin between 1.34 and 1.17 is 12,7 percent, which is greater than the 10.5 percent total penalty calculated for the plant .

Therefore the SRP's 95/95 standard is met. IIsil, ff. Tr 733, at 21-22; Dzenis, ff. Tr. 302, at 4-6.

68. Intervenors offered no evidence that the penalties did not bound the phenomena, nor did their questioning persuade us that the quantities were not conservative. The Staff and Licensee gave ample testimony to establish that the penalties were bounding values and, in fact, in the case of the mixed core, may yield a DNBR lower than that which occurs in an actual mixed core. Hsil, ff. Tr. 733, at 14; Dzenis, Tr. 384-85.

69. The Board is confident that the Staff and Licensees were competent to offer expert opinions on this subject. Dr. Edwards' perceived role as a " trouble shooter" regarding mathematical modeling (Tr. 70[) assisted the Board in sharpening the issues. On the other hand,- Dr. Edwards' lack of expertise in DNBR analysis and failure to review all the documentation supporting the values of the penalties lead us to reject his claim that they are not 95/95 values. U

70. While conservative engineering approximations may not satisfy the rigors of an applied mathematician's academic diccipline, the Board finds no evidence that the three penalties either interact with each other or do not meet the 95/95 standard. The Board concludes that the Licensee's analysis of DNDR 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 , and the calculated minimum DNBR of 1.34, after accounting for uncertainties, is greater than the DNDR acceptance limit for OFA fuel. Thus, the SRP's 95/95 standard is met.

C. Ouestion 3 Whether, if that standard is not being satisfied, the reduction in the margin of safety has been significant.

7/ Dr. Edwards' suggestion as to the proper calculation of multiple penalties is of no moment. A more precise linear combination of the penalties would yield 1.055 x 1.03 x 1.02 - 1.0 = 0.10838. See Edwards Affidavit, ff. Tr. 606, at 4-5. Regardless of the calcula-tional method used, it is clear that since the calculated minimum DNDR is 12.7 percent higher than the 1.17 DNBR limit, there is sufficient margin to accommodate the total penalty whether it is 10.5 or 10.84 percent.

71. Intervenors maintain that a DNBR limit of 1.17 for the amendment "will cause a significant reduction in the margin of safety..."

Edwards Outline , ff. Tr. 606, at 1; Intervenors Proposed Findings at if 34-39. Based on the record in this proceeding, the Board concludes that the SRP 95/95 standcrd is met by a DNBR limit of 1.17 for OFA fuel end a calculated minimum DNBR of 1.34 for Turkey Point OFA fuel and the use of a DNBR acceptance limit of 1.17 for OFA fuel does not result in a significant reduction in safety margin for the plant.

72. Dr. IIsii testified succintly , "The safety margin is not determined by a specific value of DNBR limit, but a DNBR limit which provides a protection aga.ast DNB with a 95/95 standard. Since both 1.17 for WRB-1 and OF/ ic! and 1.3 for W-3 and LOPAR fuel meet the 95/95 standard and provide the same degree of assurance that DNB will not occu. , there is no reduction in safety margin provided by the 95/95 standard." Ilsli, ff. Tr. 733, at ?4. We agree that the pertinent inquiry is whether the 95/95 standard is being satisfied. Our consideration of the record in this proceeding convinces us that the 95/95 standard is i

being met.

l

73. The Licensee's calculated minimum DNBR of 1.34, if penalized

(

i for uncertainties totalling 10.5 percent, is still greater than the 1.17 neceptance limit. Thus, the calculated minimum DNBR of 1.34 more than meets the 95/95 standard and therefore, the margin of safety provided by the standard has not been reduced. IIsil, ff. Tr. 733, at 24; Dzenis, ff.

Tr. 302, at 5-6.

74. We also note that the plant's Technical Specifications, which specify the limits on the combination of thermal power, pressurizer

l pressure and the highest operating loop coolant temperature (T,y ), were 1

established through the use of the 1.34 DNBR for OFA fuel and 1.3 DNBR for LOPAR. Hsil, ff. Tr. 733, at 23; Dzenis , Tr. 433. This single set of limits for the plant was derived by determining which fuel is more constraining at each point of operation. Dzents, Tr. 433. These safety limits provide assurance that the 95/05 standard is met for both OFA and LOPAR fuel and any change to the safety analysis which would alter the operating limit curves in the Technical Specifications would require prior NRC approval. Hsil, ff. Tr. 733, at 23,

75. During their cross-examination of the Staff's witness ,

Intervenors offered into evidence the December 9, 1983 Safety Evaluation (SE) supporting the amendments authorizing the use of OFA fuel and WABA rods at Turkey Point , which were issued prior to the instant amendments. See og 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 OFA 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 l

as an exhG(t, we allowed Intertenors to en limited questions on the Safety Evaluation to probe whether any inoa.mistency existed. Tr. 781.

I

76. Testimony by the Staff and Licensee persuade us that the 1

reduction in the margin above the DNBR acceptance limit of 1.17 is not significant. The Staff and Licensee testified that the calculated DNBR for l

r=w e -- -*m= -rer -m-- - -r* "-*r - = - - - -r -e- - - --

I l

these amendments was lower due to the increase in peaking factors (F delta H , F sub Q) which makes the hot channel in the core hotter and thus 10 ers DNBR. IIsil, Tr. 810-11; Dzenis , Tr. 341. Further, the Staff testified that the lower DNBR margin of 12.7 percent is not a reduction in a safety mergin because the safety margin is provided by the 95/95 DNDR limit of 1.17. Hsil, Tr. 901-02.

77 In sum, the evidence clearly shows that wnile there may be a ,

reduction in the " operating margin" for the plant, there is no reducticn in the margin of safety as a result of the amendments in this proceeding. l The 95/95 DNBR limit of 1.17 provides the margin of safety and the 1.34 calculated DMER for the amendments, after accour. ting for uncertainties, is greater than the 95/95 Ilmit. 8_/

III. CONCLUSION Based upon the entire evidentiary record in this proceeding, and upon the foregoing fincings of fact, the Board concludes the following:

1. The Licensee's apolvsis of DNBR performed using NRC Staff j approved methodology and compensating for appropriate uncertainties j

demonstrates at a 95 percent probability at a 95% confidence level that the l

l hottest rod will not undergo DNB.

l 1

Because we conclude that there has been no reduction in the margin

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of safety provided by the 95/95 standard , we reject Intervenors suggestion that we delete the amenc'ments. See Intervenors' Proposed Findings at 25-28.

I

2. Contrary to Intervenors' assertion in Contention (d), the margin cf safety for the operation of the Turkey Point Plant has not been reduced by the issuance of the contested amendments.

IV. ORDER WIIEREFORE, in accordance with the Atomic Energy Act of 1954, as amended, and the Rules of Practice of the Commissiva, and based on the forngoing findings of fact and conclusions of law. IT IS ORDERED TIIAT License Amendment Nos. 99 and 93 to License 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 FIJRTilER ORDERED, pursuant to 10 C.F.R. I 2.760, thct 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. I 2."62 or the Commission directs otherwise.

See also 10 C.F.R. Il 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 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,

l 1

the appeal of any other party. A responding party shall file a single responsive brief, regardless of the number of appellants' briefs filed.

Tile ATOMIC SAFFTI AND LICENSING BOARD l

Robert M. Lazo, Chairman Adminstrative Judge Richard F. Cole Administrative Judge Emmoth A. Luebke Administrative Judge Dated at Bethesda, Maryland l this day of , 198C.

Respectfully submitted, Mit A. Young Counsel for NRC Staff Dated at Dothesda, P!aryland this ?4th day of February,1986.

I

O 00(.KE TED UNITED STATES OF AMERICA NUCLEAR REGULATORY COPlflISSION 86 RB 26 #0:16 BFFORE TIIE ATOMIC SAFETY AND LICENSING BOARD f0C TbG .! 51 BRANCd In the Matter of )

) Docket Nos. 50-250 OLA-1 FLORIDA POWER AND LIGHT ) 50-251 OLA-1 COMPANY )

)

(Turkey Point Plant, Units 3 and 4) ) (Vessel Flux Reduction)

CERTIFICATE OF SERVICE I hereby certify that copics of "NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF s.AW IN Ti1E FORM OF AN INITIAL DECISION" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or ac indicated by an asterisk, by deposit in the Nuclear Regulatory Commis-sion's internal mail system, this 24th day of February,1986:

2 ...-

• Dr. Robert M. Lazo, Chairman Norman A. Coll, Esq.

Administrative Judge Steel, liector & Davis Atomic Safety and LicensinF Board 4000 Southeast Financial Center U.S. Nuclear Regulatory Commission Miami, FL 33131-2398 Washington, DC 20555

• Atomic Safety and Licensing Board

• Dr. Emmeth A. Luebke U.S. Nuclear Regulatory Commission Administrative Judge Washington, DC 20555 l

Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission

• Atomic Safety and Licensing Washington, DC 20555 Appeal Board (8)

U.S. Nuclear Regulatory Commission l *Dr. Richard F. Cole Washington, DC 20555 Administrative Judge

! Atomic Safety and Licensing Board

• Docketing t. Service Section U.S. Nuclear Regulatory Commission Office of the Secretary l Washington, DC 20555 U.S. Nuclear Regulatory Commission l Washington, DC 20555 i

Ilarold F. Reis, Esq.

i Newman & IIoltzinger, P.C. Joette Lorion l' 1615 L St. , NW 7269 SW 54th Avenue j

Washington, DC 20036 Micmi, FL 33143

Martin II. Ilodder, Esq.

l 1131 N. E. 86th Street ~

l Miami, FL 33138 b

tzt Al Moung '

Couns. for NRC Staff l

1