ML20137K955

Proposed Findings of Fact & Conclusions of Law Re Amends 99 & 93 to Licenses DPR-31 & DPR-41,respectively,supporting Program for Reduction of Pressure Vessel Neutron Bombardment.Amends Should Remain in Effect

ML20137K955
Person / Time
Site:	Turkey Point
Issue date:	01/21/1986
From:	Bauser M, Bouser M, Coll N FLORIDA POWER & LIGHT CO., NEWMAN & HOLTZINGER, STEEL, HECTOR & DAVIS
To:
Shared Package
ML20137K959	List: ML20137K955 ML20137K971
References
CON-#186-814 84-496-03-LA, 84-496-3-LA, OLA-1, NUDOCS 8601240294
Download: ML20137K955 (40)

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BEFORETHEATOMICSAFETYANDLICENSING86ARbM

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In the Matter of ) Docket Nos. 50-250 OLA-1

) 50-251 OLA-1 FLORIDA POWER AND LIGHT COMPANY )

) ASLBP No. 84-496-03 LA (Turkey Point Plant, Units 3 ) (Vessel Flux Reduction) and 4) )

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LICENSEE'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW Norman A. Coll Michael A. Bauser Steel, Hector & Davis Newman & Holtzinger, P.C.

4000 Southeast 1615 L St., N.W.

Financial Center Washington, D.C. 20036 Miami, Florida 33131-2398 (202) 955-6600 (305) 577-2800 Attorneys for Licensee Florida Power & Light Company January 21, 1986 B601240294 860121 PDR ADOCK 05000250

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of ) Docket Nos. 50-250 OLA-1

) 50-251 OLA-1 FLORIDA POWER AND LIGHT COMPANY )

,s )

(Turkey Point Plant, ) ASLBP No. 84-496-03 LA Units 3 and 4) '

) (Vessel Flux Reduction)

)

LICENSEE'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW INTRODUCTION AND BACKGROUND

1. By application dated August 19, 1983, supplemented on September 9, 1983, September 20, 1983, and October 4, 1983, Licensee, Florida Power & Light Company (FPL), requested identi-cal amendments to operating licenses DPR-31 and DPR-41 for its Turkey Point Plant, Nuclear Generating Units 3 and 4 located in Dade County, Florida. The amendments were intended to support Licensee's program for the reduction of pressure vessel neutron bombardment, and consequent embrittlement of the pressure vessel walls, and to remove restrictions imposed when FPL was operating the Turkey Point plant with steam generators having a larger number of plugged tubes than the steam generators now being utilized. For these purposes, modifications to the technical specifications contained in each license were requested to effect: (1) an increase in hot channel factor limit from 1.55 to

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1.62; (2) an increase in total peaking factor limit from 2.30 to 2.32; (3) changes in the overpower delta-T set points and thermal-hydraulic limit curves; and (4) deletion of restrictions and limits placed on operation prior to replacement of the old steam generators.

2. 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. S 50.92 was published in the Federal Register on October 7, 1983. 48 End. Egg. 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.

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. 1/

They also filed comments, contending that the amendments did 1 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 1/ The Center for Nuclear Responsibility described itself as an environmental organization and corporation with its princi-pal place of business in Miami, Florida. It also stated that its members live, use, work and vacation in the immediate vicinity of the Turkey Point units. Request for Hearing and Petition to Intervene, pp. 1-2. Subsequently, Intervenors named four individuals, including Mrs. Lorion, who " reside, work, and vacation" within 25 miles of those units. Amended Petition to Intervene, p. 4 (Jan. 25, 1984).

consideration and the Commission's finding, among other things, that the issuance of the a.iendments will not be inimical to the common defense and security or to the health and safety of the public. Egg 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 Egd. Rea. 3364, January 28, 1984. Under 10 C.F.R. S 50.91(a)(4) the amendments became effective when issued, with any required hearing to be held thereafter.

4. 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 Turkey Point units, with those actually the subject of the instant proceeding. Tr. 50-53.

The earlier issued amendme'nts provided for, among other things, the replacement, during the course of subsequent refuelings of the two units, of Westinghouse 15x15 Low Parasitic (LOPAR) fuel and borosilicate glass burnable absorber rods with Westinghouse 15x15 Optimized Fuel Assembly (OFA) fuel and Wet Annular Burnable Absorber (WABA) rods. These amendments (subsequently referred to 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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O 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 Egd. Egg. 33,080, and were issued on December 9, 1983. 48 Egd. Egg. 56,518 (Dec. 21, 1963). Egg cenerally, Prehearing Conference Order, pp. 2-5 (May 16, 1984);

SER, Staff Exhibit 1, at 3 (Dec. 23, 1983). In our May 16, 1984 Order, 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 (notwithstanding some subsequent confusing statements made in this regard (agg e.o., Tr. 764-67)); (2) no LiceI>ing Board had been convened to address those amendments; and (3) those amendments were not within the jurisdiction of this Board to decide. Egg Prehearing Conference Order, pp. 3-9 (May 16, 1984). For present purposes, however, one result of the core design change amendments is that the Turkey Point units will operate uith both LOPAR and OFA 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 OFA fuel.

5. The Prehearing Conference Order dated May 16, 1984, also granted the Intervenors standing to intervene in this proceding, and ruled on 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

i 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 antici-pated 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 autho-rize 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 reduc-tion in heat transfer coefficient. Thus, the proposed amendment will both significantly reduce the safety margin and significantly increase the probability of serious conse-quences from an accident.

6. In an Order dated August 16, 1985, and following a second prehearing conference held in Coral Gables, Florida on March 26, 1985, we granted Licensee's August 10, 1984, motion for summary disposition with respect to Contention (b), but denied another motion Licensee had filed for summary disposition of Contention (d). Florida Power & Licht Co. (Turkey Point Plant, Units 3 and 4), LBP-85-29, 22 NRC 300 (1985). In our August 16,

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l 1985 Order, we concluded that "three genuine issues as to material facts remain for litigation . "

. . with respect to Contention (d). 22 NRC at 330. These concerned certain para-meters associated with heat transfer in the core under the amendments in question. They were:

1. Whether the DNBR [ departure from nucleate boiling ratio] of 1.17 which the amendments impose on the OFA

[ 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. 3/

2. Whether, if the DNBR of 1.17 does not compensate for thos0 uncertainties, the SRP's [NRC 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.

Id.

7. Accordingly, on September 18, 1985, we issued an Order scheduling an evidentiary hearing to commence on December 10, 1985, and directing the parties to serve written direct testimony by express mail or its equivalent by November 25, 1985, or by hand delivery by November 26, 1985. Thereafter, Licensee filed a second motion for summary disposition of Contention (d) on September 20, 1985, expressly directed at the three remaining 3/ The three uncertainties, which we specifically identified in our August 16, 1985 Order and are described in greater detail below, related to rod bowing; the use of a mixed rather than a homogeneous fuel core; and the use of a certain Westinghouse correlation (WRB-1) to determine DNBR in connection with the OFA fuel used in the core.

7-questions identified above. This motion was denied by order dated November 8, 1985. Our Memorandum dated November 18, 1985, setting forth the reasons for the November 8, 1985 Order, recognized that our August 16, 1985 Order had limited the scope of Contention (d) to the three qttestions. However, the Memorandum stated our view that sufficient doubts had been raised by Intervenors' filings in response to the second motion for summary disposition so that, together with our own concerns, it was still inappropriate to find that there existed no genuine issue of material fact with respect to these questions.

8. Hearings were held in Miami, Florida on December 10-12, 1985 to address the three Board questions and thereby complete consideration of Contention (d). At the hearings the i

Licensee, the Intervenors and the NRC Staff presented one witness each to address all of the questions. In this decision we conclude that the NRC Staff's December 23, 1983 grant of the operating license anendmenta requested by Licensee was proper and l that, accordingly, they should remain in effect as issued.

9. The Licensee's witness was Edward A. Dzenis, who i

l is employed by Westinghouse Electric Corporation, as Manager of l Core Operations in the Nuclear Fuel Division, where he has worked since 1974. Mr. Dzenis demonstrated considerable expertise, i

acquired through education and experience, in the areas of thermodynamics and thermal-hydraulics. Mr. Dzenis has Bachelor i

and Master of Science Degrees in Mechanical Engineering and has taken undergraduate courses in mathematics involving calculus, l

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I differential equations, mathematical statistics, and, as part of a laboratory course, the statistical evaluation of experimental data. At the graduate level, Mr. Dzenis has taken courses in ,

r thermodynamic power conversion cycles, and in the environmental and economic aspects of nuclear power. His duties at Westinghouse have included analyses of the 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. At the completion of voir dire, Mr. Dzenis' testimony, prefiled on November 26, 1985, was received in evidence without objection and incorporated directly into the transcript as if read. Egg Testimony of Edward A. Dzenis Concerning Contention (d), follows Tr. 302, Professional Qualifications and Experience of Edward A. Dzenis; Tr. 293-302.

10. The NRC's' witness was Dr. Yi-Hsiung Hall.  !

Dr. Hsii possesses Bachelor, Master and Ph.D degrees in Mechanical Engineering and demonstrated considerable expertise in the fields of hydromechanics, and thermal-hydraulics. He has taken undergraduate courses in hydrodynamics, thermal-hydraulics, heat transfer, calculus, differential equations, and graduate level courses in hvdrodynamics, heat transfer, thermal-

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-g-l hydraulics, advanced calculus, and complex variables. Addition-ally, Dr. Hsii minored in mathematics during his university programs. Tr. 714-16.

11. Dr. Hsil, who joined the NRC in 1981, is currently a Nuclear Engineer in the Reactor Systems Branch of the Division of PWR Licensing-A in the Office of Nuclear Reactor Regulation.

Prior to his current assignment, he was in the Core Performance i

I Branch of the Division of Systems Integration. He has most recently worked as a technical reviewer on the safety evaluation reports and reload methodology topical reports on core thermal hydraulics submitted by applicants and licensees. From 1968 to 1981, Dr. Hsii worked for Babcock and Wilcox where he performed core thermal-hydraulic design analyses for reactors, and devel-oped computer codes, including a program to calculate core l performance and its improved DNBR. Testimony of Yi-Hsiung Hsii i

Regarding Contention (d), follows Tr. 733, Professional

Qualifications; Tr. 715. Dr. Hs11's professional qualifications as an expert and his testimony, prefiled on November 25, 1985, were admitted into the record without objection. 4/

12. The Intervenors' sole witness was Dr. Gordon D.J.

Edwards. Both his qualifications and the scope of his testimony were the subject of some controversy. As far as his qualifica-i tions are concerned, Dr. Edwards holds a Ph.D degree in mathe-

-4/ Intervenors originally objected to the admission of Dr.

Hsii's prefiled testimony and statement of professional 4

qualifications (Tr. 713) but after voir dire withdrew the objection (Tr. 733).

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matics and a Master's degree in English Literature. He has taught mathematics at the university level for a number of years, and has also t' aught a limited number of courses in biology and chemistry. E.c., Tr. 254-57, 505. He has been involved with a number of governmental studies in Canada concerning various aspects of nuclear power (Tr. 259-62, 269-72), and has taken certain undergraduate courses in thermodynamics and fluid mechanics (Tr. 263-64). He has also questioned technical experts in certain technical fields associated with nuclear power. Tr.

272-73. However, Dr. Edwards generally has no knowledge, skill, experience, training or education in the field of engineering (e,q. Tr. 538), and is admittedly not an expert in areas such as heat transfer, departure from nucleate boiling testing, critical heat flux correlations, or the determination of reactor core operational limits. E.c., Tr. 505-06. He is generally unfamiliar with subchannel analysis, has never performed any experiments or conducted studies concerning departure from nucleate boiling in a PWR, and has never designed or utilized models performing PWR thermal-hydraulic analyses. Tr. 278, 506, 529-30. In addition, Dr. Edwards has reviewed only a limited amount of the documentation associated with the amendments that are the subject of this proceeding, is not generally familiar with the details of the Turkey Point plant, and performed no independent studies, tests or calculations in preparing his testimony. Eigt, Tr. 506-09, 516-18, 524-25. Accordingly,

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Licensee moved to strike Dr. Edwards' testimony in its entirety by a motion which was supported in part by the NRC Staff. Tr.

544, 546-51.

13. What constituted Dr. Edwards' " testimony" involved another area of dispute among the parties. On November 25, 1985, Intervenors served upon the Board and the parties a brief (approximately one and one-third pages, single spaced) document labeled " Outline of Testimony by Gordon Edwards." However, initially at the hearing, no effort was made to treat the document as Dr. Edwards' prefiled testimony. Instead, he was simply asked whether he had testimony to offer on Contention (d) and the three Board questions. Following Dr. Edwards' affirma-tive response, Intervenors' counsel asked him to proceed by "taking them in sequence beginning with Question 1 as posed by the Board." Tr. 445-46. Following objections by the other counsel, Intervenors' counsel explained that he intended to present Dr. Edwards' " Outline of Testimony" to be bound into the record, but that the outline was not "in the depth I would have wished" (Tr. 446-49; agg also Tr. 459-60) and "that a witness shouldn't be limited to what he prefiles and prewrites." Tr.

450. The other parties' objections were, in essence, based upon the Commission's rules of practice, 10 C.F.R. S 2.743 --

requiring'the submission of direct testimony in written form unless the Licensing Board directs otherwise -- and claims of surprise, unfairness and the likelihood of delaying or expanding the proceeding. Egg, RAS., Tr. 446-47. Particularly because of

the likelihood of such delay (Tr. 476), and the failure of Intervenors' counsel even to notify the Board between November 26 and the date of the hearing that he wished to submit testimony in addition to the outline (Tr. 473), the Board refused to permit Dr. Edwards to supplement his testimony orally. Tr. 475-76.

14. Nevertheless, Intervenors offered as part of Dr.

Edwards written testimony two affidavits previously executed by him and filed in this proceeding. One, dated August 30, 1984, addressed both Contentions (b) and (d) and had been attached to Intervenors' opposition to the Licensee's first motion for summary disposition of those Contentions. The second affidavit was dated November 5, 1985, and had been attached to Intervenors' opposition to Licensee's second motion for summary disposition.

The request to treat the first affidavit as testimony was withdrawn. Tr. 488. However, since all parties had had a reasonable opportunity to examine the second affidavit (Tr. 488),

and it addressed the three questions propounded by the Board relating to Contention (d), that affidavit was received in evidence, but subject to voir dire. Tr. 496.

15. Returning to Dr. Edwards' qualifications, we have noted these above in paragraph 12. In addition, earlier in the proceedings the Board had found Dr. Edwards qualified, as an expert interrogator pursuant to 10 C.F.R. S 2.733, to conduct cross-examination in this proceeding (Tr. 288), and Dr. Edwards had demonstrated some knowledge during his questioning of Mr.

Dzenis. 5/ Based upon that demonstration, and in view of the limited scope and qualified language of Dr. Edwards' Outline of Testimony and his November 5, 1985 affidavit, the Board found him

" qualified as an expert," and admitted both those documents into evidence. Tr. 556. The Board also directed that they be incorporated into the transcript as if read. Tr. 903. We note, nevertheless, that Dr. Edwards himself conceded that his exper-tise was limited to the areas of mathematical analysis, calcula-tions of probability, and the use of mathematical models (Tr.

282-83, 542-44), that he had no knowledge, skill, experience, training or education in the field of engineering (Tr. 262, 538),

and that he did not possess any expertise in the areas of heat transfer, departure from nucleate boiling testing, critical heat flux correlations, the determination of operational limits, the evaluation of departure from nucleate boiling ratio, or thermal-hydraulics. Tr. 279-83, 506.

16. We have carefully considered all of the testimony, opinions and evidence adduced at the hearing and 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.

5/ Thereafter, Dr. Edwards also served as expert interrogator of Dr. Hsil during the Intervencrs' cross-examination of that witness.

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17. 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 dis-cussed in considerable detail in our August 16, 1985 Order 4 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, OPA 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, 111 DNBR acceptance limit for another type of fuel utilizing a different CHF correlation (again, in this case, LOPAR fuel with the W-3 correlation). Egg 22 NRC at 323-28. What was not clear to us, however, was how three particular uncertainties a mentioned in the NRC Staff's December 23, 1983 SER (i.e., those related to rod bowing; the use of new OFA fuel assemblies mixed together with LOPAR fuel assemblies during a transition period on the way to a full OFA core; and the application of the WRB-1 correlation to 15x15 array OFA 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 4

reduction in the margin of safety is not significant.

1 Id. at 330.

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First Board Ouestion Whether the DNBR of 1.17 which the amendments impose on the OFA 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.

18. To address this question, it will be helpful to review certain aspects of PWR operation. As we described in some detail in our August 16, 1985 Order, 22 NRC at 323-28, heat is removed from the core of a 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 I 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 proportion to increasing fuel rod

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temperature. The measure of heat transferred in a given time from a unit of rod surface area is called heat flux.

19. If the fuel rods reach a sufficiently high temperature, however, some of the steam bubbles will remain on the rod surfaces and begin to combine, thus resulting in the formation of a steam film. The point at which a film appears is l called departure from nucleate boiling (DNB). Such a film, in i

effect, insulates the fuel rod causing heat that would otherwise

be given up to the coolant to be retained in the rod. Thus, 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, and the actual heat flux (AHF) under those same conditions. 6/

20. It is impossible to predict with perfect certainty what the CHF for a particular fuel in a reactor will be under a given set of conditions. Different experimentally-determined correlations afford varying degrees of assurance with respect to predictions of CHF. Under the NRC Staff Standard Review Plan (SRP), NUREG-0800, however, a minimum ratio between CHF and AHF

-- called the minimum departure from nucleate boiling ratio, or DNBR -- is established such that there is at least a 951 confi-i dence 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 statis-tical measure of conservatism in the selection of a minimum DNBR is often referred to as the 95/95 condition or standard. 7/

6/ As we noted in our August 16, 1985 Order, 22 NRC at 324, it is a long way from DNB to a release of fission products to the environment. 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 itself a closed system. Prudence, nevertheless, suggests i

that DNB -- and transition to a less desirable heat transfer regime -- be avoided.

l 7/ At no time in this proceeding have Intervenors questioned the adequacy of the 95/95 standard. Egg 22 NRC at 328.

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21. Returning to our first question, as to whether-the 1.17 DNBR acceptance limit applicable to the OFA fuel at Turkey Point compensates for the three uncertainties outlined by the NRC Staff in its December 23, 1983 SER, the answer is clearly: No.

In fact, there is no dispute among the parties as to this point.

22. The 1.17 DNBR has been referred to a number of ways, including "1.17 DNBR design limit," " design DNBR," and "DNBR limit." Egg, e.g., Memorandum, p. 5 (Nov. 13, 1985). It may best be referred to as a DNBR acceptance limit. However, they are all the same and, from time to time herein, we may employ one or other of the terms because of its use by a witness on whose testimony a finding is based. The DNBR acceptance limit of 1.17 is generic to all Westinghouse plants utilizing OFA fuel.

Testimony of Edward A. Dzenis Concerning Contention (d), follow-ing Tr. 302, at 3 (hereinafter Dzenis, ff. Tr. 302, at __). To be distinguished from the DNBR acceptance limit is the " safety analysis minimum DNBR," or " calculated minimum DNBR" which is calculated on a plant-specific basis and will be considered later. Egg Dzenis, ff. Tr. 302, at 3; Testimony of Yi-Hsiung Hsii Regarding Contention (d), following Tr. 733, at 8-10 (hereinafter Hsil, ff. Tr. 733, at ).

23. The DNBR acceptance limit of 1.17 for the WRB-1 correlation, which is used in connection with the analysis of all Westinghouse OFA fuel, constitutes, in accordance with the acceptance criterion presented in Section 4.4 of the SRP, the 95/95 bounding value for experimental data. Dzenis, ff. Tr 302,

at 3; Hsil, ff. Tr. 733, at 11. The 95/95 standard contained in the NRC's 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. Dzenis, ff. Tr. 302, at 3-4; Hsii, ff. Tr. 733, at 11-13.

24. The 1.17 DNBR acceptance limit, however, does not and is not intended to compensate for the three uncertainties referred to in the Board's question, i.e.: rod bow, mixed LOPAR/OFA fueled core, and the application of the WRB-1 correla-1 tion to the 15x15 OFA 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 predic-tion 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 l

from statistical analysis applying the acceptance criterion of 95% probability at 95% confidence, as specified in the SRP.

Hsil, ff. Tr. 733, at 3-4.

25. No CHF correlation, however, is, in fact, able to exactly predict the experimental data upon which it is based. If the true CHF value could be calculated and the actual heat flux

precisely known, the exact DNBR could be determined and a design DNBR limit of 1.0 would ensure that DNB would be avoided. Tr.

743-45. However, because CHF is calculated using an empirical correlation developed from experimental CHF data, and because of random variations in the data, the exact CHF can not be predicted. A DNBR limit greater than 1.0 is therefore utilized to account for this uncertainty. The DNBR limit for a correlation is the value which ensures with a 95% probability at 95% confidence level that DNB will be avoided when the DNBR calculated with the correlation is equal to or greater than this value. In the case of OFA fuel, the 1.17 DNBR acceptance limit provides that the 95/95 standard will be satisfied. Hsii, ff.

Tr. 733, at 3-7. As noted in the preceding paragraph, however, the 1.17 DNBR acceptance limit does not and is not intended to compensate for the three uncertainties referred to in the Board's first question, i.e,: rod bow, mixed LOPAR/OFA fueled core, and the application of the WRB-1 correlation to 15x15 OFA array fuel.

Dzenis, ff. Tr. 302, at 4.

26. As described below in connection with the Board's second question, however, these other uncertainties are taken into account in the evaluations of normal and anticipated operational occurrences performed for specific plants. Those evaluations reflect and account for the fact that conditions in specific reactor cores may vary in some respects from experi-i mental conditions which existed when the data upon which the correlation being used was developed. If appropriate, " penal-

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i ties" are then imposed to reflect the specific variations.

However, as demonstrated below, even with these penalties, assurance -- meeting the 95/95 standard -- that DNB will not occur is provided.

Second Board Ouestion 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.

27. Both the Licensee and NRC Staff have answered our second question in the affirmative (e.g., Dzenis, ff. Tr. 302, at 4; Hsii, ff. Tr. 733, at 22), and presented considerable evidence to support that conclusion. Intervenors, on the other hand, simply offered a number of unsupported allegations concerning the validity of these responses, which we address in connection with our discussion below.

28. After carefully reviewing and considering all of the evidence presented by the parties, we conclude that the answer to our second question is properly: Yes. In order to fully understand the details of exactly how the 95/95 standard is met, however, we must first consider what has been variously 1

referred to as the " safety analysis minimum" or " calculated minimum" or " safety analysis calculated minimum" DNBR 8/ and then 8/ These terms, which are used synonymously herein, are not generic to a group of plants, as is a DNBR acceptance limit.

In essence, they all refer to the lowest value of the DNBRs calculated from predictive computer analyses for the spectrum of normal and anticipated operational occurrences (transients) for Turkey Point. Dzenis, ff. Tr. 302, at 5; Hsii, ff. Tr. 733, at 8.

look at each of the three uncertainties which are not included in the calculated minimum DNBR -- and any physical relationship between them -- to determine if they are properly considered.

29. It is pertinent to note that application of uncertainties to results obtained from predictive analysis (in this case, the 1.34 safety analysis calculated minimum DNBR),

rather than to design basis limits (such as the 1.17 DNBR acceptance limit), is common in the engineering field. Dzenis, ff. Tr. 302, at 6; Hsil, ff. Tr. 733, at 16. This approach is also consistent with that in Section 4.4, Part II.2 (at page 4.4-3) of the SRP. Dzenis, ff. Tr. 302, at 6.

30. For all Westinghouse plants, any uncertainty important to the calculation of the minimum DNBR -- but not included in the input to the Westinghouse THINC computer code, which is used along with a CHF correlation, such as WRB-1 to calculate minimum DNBR -- is converted to a " penalty." Tr. 730; Hsil, ff. Tr. 733, at 8-10. The value of the penalty is the result of a separate analysis in which the physical phenomenon being considered has, in effect, been converted to an equivalent DNBR. Id. For example, fuel rod bowing reduces DNBR, but fuel rod bowing is not directly modeled in the THINC computer code.

Therefore, a rod bow penalty is calculated separately. Id. The proper rod bow penalty has been determined to be 5.51. g/ Since a homogeneous core of the 15x15 OFA fuel was assumed in providing 9/ See Section C, infra, for a discussion of how rod bow penalty is determined.

input to the THINC code, a mixed core penalty of 3% was assessed.

Additionally, since any uncertainty associated with the application of the WRB-1 correlation to the 15x15 OFA fuel array was not considered in the THINC code input, a penalty of 21 was also assessed. Hsil, ff. Tr. 733, at 15-16.

31. Generally speaking, there are two methods which might be used to determine whether the 95/95 criterion is met with respect to the calculated minimum DNBR after taking into

! account the appropriate penalties for uncertainties not included in the original computer code analysis. Hsil, ff. Tr. 733, at

! 12. Ett R112 Tr. 393. First, the total penalties may be combined and added to the DNBR acceptance limit (in this case 1.17), and then compared to the calculated minimum DNBR. If the calculated DNBR is greater than or equal to the new DNBR accep-tance limit then the 95/95 standard is met. Id. The second methodology, (and the one used in this case which, as indicated in paragraph 29, is common engineering practice) is to reduce the calculated minimum DNBR by the total penalty and compare it to the DNBR acceptance limit. Id. If the reduced calculated minimum DNBR is greater than or equal to the DNBR acceptance limit, then the 95/95 standard is met. Id. With respect to the 2

second approach, it is common to calculate the margin between the DNBR acceptance limit and the calculated minimum DNBR. Id. The t

DNBR margin is the percentage difference between the calculated minimum DNBR and the DNBR acceptance limit. Id. The DNBR margin is then compared to the total penalty. If the margin is greater

i than or equal to the penalty, then the SRP's 95/95 standard is met. Id. at 12-13. Egg also Dzenis, ff. Tr. 302, at 6. We accept these methodologies as appropriate to determine whether the SRP's 95/95 standard has been satisfied.

A. Safety Analysis or' Calculated Minimum DNBR

32. As we have already pointed out, the " safety analyses minimum" or " calculated minimum" or " safety analysis calculated minimum" DNBR is essentially the lowest value of the DNBRs calculated from predictive computer analysis, using approved computer codes and correlations, for transients. Hsii, ff. Tr. 733, at 8. If the calculated minimum DNBR, after accounting for uncertainties, exceeds or is equal to the design DNBR of 1.17, then the 95/95 standard is met. The calculated minimum DNBR for Turkey Point was obtained through the use of approved computer codes and correlations. Hs11, ff. Tr. 733, at

15. The minimum DNBR for each of the normal and anticipated operational occurrences was calculated using the THINC sub-channel thermal hydraulic. code and the WRB-1 CHP correlation.

Id. The input to the THINC code included a geometry model representing a reactor core, fuel assemblies and sub-channels.

Reactor conditions during the transients included the values of reactor power, pressure, coolant flow rate, inlet temperature and power distribution. In providing these inputs to the THINC code, uncertainties for all of the important parameters were accounted for by using conservative values. Id. The uncertainty value of

each parameter was obtained by using either a bounding value or a value with a 95/95 confidence level. Id. at 9. We note that this is a conservative approach because it is not likely that all adverse effects requiring conservative inputs would occur simultaneously. Accordingly, the calculated DNBR for each transient is lower than the true value expected, and is a conservative value. Id. at 10. It should be reemphasized that the 1.34 calculated minimum DNBR is plant specific for Turkey Point taking into consideration the specific design of the plant and the Turkey Point technical specifications, including those specifications contained in the amendments which are the subject 1 of this proceeding. Egg Dzenis, ff. Tr. 302, at 5.

i j 33. The 1.34 safety analysis minimum DNBR represents the lower bound to the values calculated for the spectrum of normal and anticipated operational occurrences for Turkey Point.

The procedures and techniques employed by the Licensee in determining the calculated minimum DNBR were in accordance with Sections 4.4, Parts II.4 and II.5 and Section 15 of the Standard Review Plan. Dzenis, ff. Tr. 302, at 5. The 1.34 calculated

, minimum DNBR value, which is computed using Turkey Point plant specific reactor parameters, exceeds the 1.17 DNBR acceptance limit value by a margin of 12.7%, as is demonstrated by the l equation l

1.34-1.17 =

1.34 .127 or 12.7%

j Thus, there is a 12.7% margin between the 1.17 DNBR acceptance limit and the 1.34 calculated minimum DNBR. Id. As we discuss in detail below, the three uncertainties not accounted for in the THINC computer code total only 10.5%. 10/ Since the 12.7% DNBR margin is greater than the 10.5% total of the penalties, there is

, a sufficient margin in the 1.34 safety analysis minimum DNBR to compensate for uncertainties associated with rod bow, the mixed i

LOPAR/OFA fueled core, and the application of the WRB-1 correla-tion to the 15x15 OFA fuel. Dzenis, ff. Tr. 302, at 6.

B. Mixed LOPAR/OFA Fueled Core

34. The 1.34 safety analysis minimum DNBR for Turkey Point was calculated assuming a homogeneous core model. Hsil, ff. Tr. 733, at 13-14. A mixed core penalty was applied to this DNBR to account for the fact that the LOPAR and OFA fuels have i

different hydraulic resistance characteristics which affect the I

cross-flow of coolant between the different fuel bundles such that the OFA fuel, which has the higher grid resistance, will I

receive less flow. Hsil, ff. Tr. 733, at 13. This reduction in 10/ Dr. Edwards in his affidavit of November 5, 1985 at 5, implied that the appropriate method for totaling the uncertainty percentages was multiplication rather than addition. Both Dr. Hsil and Mr. Dzenis testified that in the instant circumstances, on an engineering basis, it is

' appropriate to add the penalties to arrive at a combined l value. Hall, ff. Tr. 733, at 21. Egg Dzenis, ff. Tr. 302, at 6. In any event, the multiplication of the uncertainties yields a total of 10.8% (as demonstrated by (1.055) x (1.03) x (1.02) = 1.10838 or 10.8%). This 10.8% figure is still within the 12.7% margin available.

L

flow vas quantified through experiments on the hydraulic charac-teristics of the two types of fuel assemblies. Tr. 312. The hydraulic characteristics established by these experiments were used to determine the percent difference in the DNBR between a homogeneous 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 core geometry. Hsil, ff. Tr. 733, at 14, 17-18.

35. More specifically, the 3% mixed core penalty was derived from a sensitivity study performed particularly for a 15x15 OFA and 15x15 LOPAR mixed fuel core. Hsil, ff. Tr. 733, at

17. The study was done utilizing the NRC Staff approved method described in a Westinghouse topical report. Id. The need for such a study arises from the fact that exact reactor core loading patterns cannot be defined in advance for all reactors for every fuel loading because of their dependence on specific plant operating schedules and the specific design requirements of particular refueling cycles. 11/ Dzenis, ff. Tr. 302, at 7.

Therefore, the sensitivity study is performed with the THINC code by using a homogeneous core model and various mixed core models.

11/ Once a license amendment of this nature becomes part of the operating license there is no need to change it because of particular choices of specific core reload patterns. At the point in time at which the analysis is performed for this type of submittal, the specific loading patterns for the subsequent two or three cycles of plant operation, when the core is composed of more than one type of fuel, are not known. However, a methodology is chosen which utilizes a bounding concept to cover the effects of conceivable loading patterns which may be established during these intervening cycles. Egg Tr. 314-15.

The mixed core analysis included various combinations of

" checkerboard configurations" including the least favorable mixed core configuration where one OFA assembly was completely sur-rounded by LOPAR assemblies. These configurations were selected to envelope all possible configurations included in reload licensing submittals, arriving at a bounding value of 31.

Dzenis, ff. Tr. 302, at 7; Hs11, ff. Tr. 733, at 18; Tr. 383.

36. In this connection, it is relevant to n.ote that it is not necessary, in performing mixed core analysis, to analyze every possible core configuration for the purposes of determining DNBR. Tr. 383-84. Only three basic configurations are relevant to loading two different types of fuelt one type of fuel (OFA) surrounded by the other type of fuel (LOPAR); a checkerboard configuration where one type of fuel assembly alternates with another type of fuel assembly; and a row of one type of assembly adjacent to a row of another type of assembly. Tr. 383-86.

~

Further, for the purposes of determining DNBR, only the condi-tions immediately surrounding the fuel bundle that contains the hot rod are of interest. 'Tr. 384. The analysis performed to evaluate the possible effect on DNBR of variations in flow due to the use of OFA fuel together with LOPAR fuel was proper to determine a bounding value. E.c., Tr. 383 Hsii, ff. Tr. 733, at 17-18.

37. As indicated in paragraph 35, above, the differences in the DNBh with a homogeneous OPA and mixed core models were calculated for the cases analyzed at various reactor

1 operating conditions. The results showed that none of the calculated DNBRs would have to be reduced by more than 3% to accommodate the effect of the transition core. Tr. 313-14.

l Thus, a 31 mixed core penalty was used as a bounding value.

l Dzenis, ff. Tr. 302, at 7, Hsil, ff. Tr. 733, at 18. This 3%

penalty is applied only to the OFA fuel because it has a higher hydraulic resistance than does the LOPAR fuel. No uncertainty is ,

i j applied to the LOPAR fuel because it always receives at least the  !

reactor coolant flow it would have otherwise experienced. Dzenis, '

j ff. Tr. 302, at 7. Because the 31 penalty for the mixed core is I '

a bounding value it, in fact, exceeds the 95/95 standard and, thus, we find that the 95/95 level of confidence is met with respect to the mixed core penalty. 12/

l C. Rod Bow Penalty '

j

{ 38. At Turkey Point, fuel rods are placed in the j reactor core in assemblies consisting of a 15x15 array of fuel i

1 rods. These fuel rods are supported in the assembly by spacer l grids located approximately every two feet of axial elevation. t i

j As the fuel is irradiated, some random horizontal displacement of i

[ '

1

)

j 12/ During the hearing a question was raised concerning the effect of WABA rods on DNBR. Eig., Tr. 843. The i

introduction of WABA rods into the reactor core results, in effect, in an increase in the bypass flow -- around the heat-producing portion of the fuel -- which may influence the DNBR. However, the number of WABA rods allowed at j

Turkey Point -- and the resultant bypass flow -- is limited,

)

and' safety analyses conservatively assume a large, limiting j bypass flow which envelopes the WABA bypass flow. Thus, the i effect of WABA rods is properly accounted for in DNBR j calculations. Tr. 849-50. L i

l 1

l

s t

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. Egg also Dzenis, ff. Tr. 302, at 7-8; Hsil, ff. Tr. 733, at 16. The effect of rod bow on DNBR is applied as a penalty. '

E2g., Dzenis, ff. Tr. 302, at 4-8; Tr. 322, 436.

39. The rod bow penalty is based on direct measure-ments 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 OFA fuel, was derived, based on an analytical method described in a Westinghouse topical report.

Hsii, ff. Tr. 733, at 16-17. This method of deriving the rod bow penalty has been used at most Westinghouse plants. The NRC Staff has reviewed and approved the calculational methods, verifying that the 5.5% penalty figure meets the 95/95 criteria. Tr. 821.

40. Moreover, the method of calculation described in the Westinghouse report represents a conservative upper bound to the effect of fuel rod bowing on the DNBR. The underlying assumption in the rod bow analysis is that the largest rod bowing occurs at the hot channel fuel rod and at the location of the minimum DNBR. Hsil ff. Tr. 733, at 17. In actual operation, minimum DNBR generally occurs in the upper portion of the core, whereas the worst rod bowing usually occurs in the lower portion of the core. Additionally, severe rod bowing generally occurs at the fuel rods having high burnup, whereas the hot channel with l , .

the highest power peaking factor generally occurs with low burnup fuel. 13/ Therefore, the assumptions of the rod bow analysis are conservative. Id.

41. In sum, the evidence in this proceeding supports the conclusion that the 5.5% rod bow penalty meets the 95/95 criterion of the SRP, and we so find. 14/

13/ The value of 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 /MTU. 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 hot 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.

' 14/

~~ In this connection, we note that Intervenors' witness, Dr.

Edwards, conceded -- in response to a hypothetical question

-- that if the rod bow penalty was a result of a conservative judgment as to the value for that penalty, he would not have any reason to believe that the value was insufficient. Tr. 640. In addition, some questions were raised concerning the use of a rod bow penalty of 14.9% in the separate Safety Evaluation for the core design amendments, as opposed to the 5.5% subsequently used in connection with the instant vessel flux reduction amendments. However, the Staff witness explained that the 14.9% rod bow penalty was based on an interim study using a different calculational method and different experimental test data than that used to derive the 5.5% figure at issue in this proceeding. Egg Tr. 811-16. Egg also SER, Staff Exhibit 1, at 3, which explains that:

In the previous Technical Specification change (Amendments 98 and 92) the fuel rod bow effect on DNBR was calculated using an older approved interim method . . . which resulted in a maximum rod bow penalty of 14.9%. This interim method for rod bow penalty calculation l was developed by Westinghouse and approved by the NRC Staff as a conservative calculational method. The Licensee has recalculated the rod bow penalty using a more recent approved

D. Independence of Mixed LOPAR/OFA Fuel Core Hydraulic and Rod Bow Effects

42. The Intervenors have alleged that "It is entirely likely that the rod bow phenomenon might interact in a fairly complicated way with the already complicated non-uniform hydrau-lic resistance phenomenon". Affidavit of Gordon Edwards i Regarding Contention (d), follows Tr. 606, at 5. Nevertheless, Intervenors did not present any evidence supporting this claim.

Egg Tr. 593-94. BoththebtaffandLicenseewitnesses,onthe other hand, indicated that the rod bow phenomenon and the differ-

, ential resistance of the OFA and LOPAR fuels to flow in the mixed 4

core are independent phenomena, and that they are thus subject to separate modeling and the application of independent penalties.

E2g., Dzenis, ff. Tr. 302, at 8; Hsil, ff. Tr. 733, 19-21. ,

43. Both the LOPAR fuel and the OFA fuel are subject '

to the rod bow phenomenon. Tr. 818. Therefore, whether a homo-i geneous core of only one type of fuel or a mixed core is being i

modeled, the rod bow phenomenon must be taken into account.

However, a mixed core configuration does not increase fuel rod bowing or the resultant rod bow penalty on DNBR. Hsii, ff. Tr.

! 733, at 19; Tr. 388-89. For the mixed, transition core of OFA and LOPAR fuel designs, the flow reduction through the OFA fuel is only about 2% to 3%. Hsil, ff. Tr. 733, at 20. A reduction in flow rate of this magnitude does not affect the localized phenomenon of critical heat flux reduction due to rod bow and can l

l method . . . .

l

l l

l be neglected. Id. at 20-21. Additionally, there is no evidence ;

l showing any deflection (additional bowing) of fuel rods due to differences in hydraulic resistance between OFA and LOPAR fuel and resultant cross-flow. Dzenis, ff. Tr. 302, at 8; Tr. 330-31.

Thus, it is acceptable to assume that there is no significant interaction between the effects of fuel rod bowing on critical heat flux and the flow changes caused by a mixed core configura-tion. The rod bow penalty is independent of mixef. core effects.

Hsil, ff. Tr. 733, at 21.

44. It is also apparent that rod bow has no signifi-cant effect on the hydraulic characteristics of the mixed core.

The length of 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. 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 occuppied by the fuel rods. Tr. 328-29. The deflections that occur with rod bowing are, in most cases, only a few hundredths of an inch over an axial distance of approximately 2 feet. The total localized change in flow area is very smooth and very small. The total flow area of the fuel assembly is essentially unchanged. Tr. 329. There are numerous engineering studies concerning the effects of changes in area on flow regime. This change in local flow area is far too smooth and insignificant to cause any hydraulic characteristic change or resulting effect on mixed-core DNBR penalty. Tr. 328-29.

45. In sum, we find that there is no physical basis for mixed core hydraulic characteristics to effect rod bow, and no physical basis for rod bow to affect the hydraulic character-istics of the mixed core.

E. Application of WRB-1 to 15x15 OFA Fuel

46. At the time the amendments which are the subject I of this proceeding were being evaluated by the NRC Staff, the WRB-1 CHF correlation had been approved for application to 15x15 R grid LOPAR fuel, 17x17 R grid LOPAR fuel, and

, 17x17 OFA fuel, with a DNBR acceptance limit of 1.17. Information demonstrating applicability of the WRB-1 correlation to both 14x14 and 15x15 OFA fuel, including actual test data specifically representative of 14x14 OFA 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 OFA fuel, however, the NRC Staff imposed a 2% penalty for the i

evaluation of the Turkey Point amendments as a conservative measure. Esil, ff. Tr. 733, at 6-7, 18-19; SER, Staff Exhibit 1, at 4.

47. Staff review of the additional information has now been completed. As O result, the Staff has concluded that the WRB-1 correlation is also applicable to both 14x14 and 15x15 OFA fuel with a DNBR acceptance limit of 1.17. Esil, ff. 733, at 1

18-19. Accordingly, there is properly no penalty for application t- W * - * *-w '

w

of the WRB-1 correlation to 15x15 OFA fuel, and the 2% uncer-tainty 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 limit -- is correctly 0.0%. Sag, g2g., 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 OFA fuel. E.o., Tr. 325-26. To the contrary, however, based on a consideration of test results and the geometries involved, such a result is not at all unexpected. Actual test results have demonstrated that the WRB-1 correlation is applicable to 15x15 R grid LOPAR fuel, 14x14 OFA fuel, and 17x17 OFA fuel.

Hsii, ff. Tr. 733, at 5-7. 15x15 OFA 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.

Esii, ff. Tr. 733, at 18. On the other hand, 14x14 and 17x17 OFA fuel have mixing grid designs similar to 15x15 OFA fuel, but differ in rod diameter. Hsil, ff. Tr. 733, at 6, 18. According-ly, 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 OFA

r fuel. Thus, it is not surprising -- but, rather, to be expected

-- that the geometry of 15x15 OFA fuel is within the applicability range of the WRB-1 correlation.

49. In summary, the SRP's 95/95 standard is met by assuring that minimum DNBR values calculated for all normal and anticipated operational occurrences, after accounting for all uncertainties, are greater than or equal to the 1.17 DNBR

acceptance limit. The Licensee has properly utilized the THINC computer code to derive a calculated minimum DNBR of 1.34 for normal and anticipated operational occurrences at Turkey Point.

Uncertainties associated with rod bow, the mixed core configura-1 l.

tion, and the application of the WBR-1 correlation to the 15x15 OFA fuel are accounted for by assigning penalties to each uncertainty which total 10.5%. However, the calculated minimum DNBR of 1.34 has a 12.7% margin above the DNBR acceptance limit of 1.17 for the WRB-1 correlation. This 12.7% margin is larger than the total penalty of 10.5% for the three uncertainties in

question. Since the calculated minimum DNBR of 1.34 is a conservative figure that meets the 95/95 standard, and each of the assessed penalties also meet the 95/95 standard or are actually bounding values, the SRP's 95/95 standard is satisfied.

Thus the answer to our second question is: Yes. 15/

i 15/ There were, during the hearing, some concerns expressed over increased peak linear heat generation rates and hotter core temperatures. E.g., Tr. 589-90. The increase in peak linear heat generation rate under the amendments in question, however, is only from 12.8 kW/ft, to 12.9 kW/ft.

This is very small and would result in only a small increase in clad temperature. Accordingly, we find this concern of no actual importance. Egg SER, Staff Exhibit 1, at 2-6, 9.

. . _ . . . , _ - _ . - _ _ , . , . . . --_,.m -.,_,,_-._.~,......,__.m ,__,.-----_w-e-,-, .m _

.- . - -= .. . . _ ... . _ - _ _ _ _ - _

l 1

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

50. As noted above in response to our second question, we have concluded that the SRP's 95/95 criterion has, in fact, j been met with the 1.17 acceptance limit DNBR. Thus, there is no reduction in safety margin for Turkey Point. In addition, it is

, noted that the operation of a nuclear power reactor is a care-

)

fully controlled technical process. There is constant monitoring of the reactor with respect to numerous parameters, including coolant flow rate, pressure, temperature, and core power distri-bution. Accordingly, safety does not depend on any particular individual criterion, or any individual operating procedure. Tr.

423. Based on a thorough review of all of the evidence we conclude 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. 16/

l

--16/ During the hearing, some inquiry was made into the impact on the Licensee of a decision to modify or revoke the license amendments. Tr. 358-63; 826-38. Given our decision that there has been no reduction in safety margin at Turkey Point Units 3 and-4, and that the license amendments at issue in this proceeding shall remain in full force and effect without modification, we need not address this question further. We emphasize, however, that our concern during this proceeding has been to assure that the Turkey Point plants operate in a safe manner. No evidence was presented i

during the hearing to indicate the need for any restrictions

or modifications in the license amendments to provide a l greater margin of safety than is inherent in the license j amendments as granted.

4

_ - _ _. _ _ . _ . _ , _.. _ _~._ _._ . - _ _ . _ _

- .. . ..- ._. . . _ . . . _ _ = - . -. _.

! CONCLUSIONS OF LAW Based upon the foregoing findings of fact and upon l consideration of the entire evidentiary record in this proceed-

)

ing, the Board makes the following conclusions of law:

1. Appropriate NRC Staff-approved methodology has been used in all analyses associated with the amendment application. Computer programs and DNB correlations used in analysis were proper and have been accepted by the NRC Staff.

2. Florida Power & Light Company has fully met its j burden of proof on issues raised in this proceed-ing. There has been no reduction in safety margin. The NRC Staff SRP requires a 95% probabi-4 lity at a 95% confidence level that the hottest rod will not undergo DNB. This design basis j requirement is met at Turkey Point.

j 3. Results and evaluation of the DNB analysis show that all applicable regulatory requirements have been satisfied.

i ORDER i

WHEREFORE, in accordance with the Atomic Energy Act of l 1954, as amended, and she Rules of Practice of the Commission, i and based on the foregoing findings of fact and conclusions of law, IT IS ORDERED that Amendments Nos. 99 and 93 to operating licenses DPR-31 and DRP-41, respectively, issued by the NRC Staff 4

i l

l

s a  %

modifying technical specifications for Turkey Point Nuclear Generating Units 3 and 4 shall remain in full force and effect without modification.

IT IS FURTHER ORDERED, that this Decision shall constitute the final decision of the Commission within thirty (30) days from the date of issuance, unless an appeal is taken in accordance with 10 C.F.R. $ 2.722 or the Commission otherwise directs. See also 10 C.F.R. SS 2.764, 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 in the case of the NRC 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 NRC Staff),

any other party 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 appellant briefs filed. Egg 10 C.F.R. S 2.762.

THE ATOMIC SAFETY AND LICENSING BOARD Robert M. Lazo, Chairman i

Administrative Judge I

l

s O Emmeth A. Luebke I Administrative Judge i

i Richard F. Cole Administrative Judge j

! Dated _ , 1986, j Bethesda, Maryland 1

i i

i 4

I i

4 I

4 i

, . . . , _ . . - . ~ _ _ . _ . _ _ , -