ML20248D653
| ML20248D653 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/31/1989 |
| From: | Bordenick B, Patricia Jehle NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| References | |
| CON-#389-9006 OLA, NUDOCS 8908110101 | |
| Download: ML20248D653 (35) | |
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00CKETED-U9iRC UNITED STATES OF AMERICA o NUCLEAR REGULATORY COMMISSION. .gf AUG -7: f01 :26 "W lBEFORELTHE ATOMIC SAFETY'AND LICENSING APPEAL BOARD
~ . In. the Matter of Docket No.c50-335-OLA:
FLORIDA POWER AND LIGHT-COMPANY: .(SFP Expansion)
(St..LuciePlant,' Unit.No.;1)' )
- n
~,.
i i!RC STAFF'S BRIEF OPPOSING INTERVENOR'S APPEAL OF INITIAL DECISION AUTHORIZING SPENT FUEL POOL RERACKING Patricia A. Jehle Counsel for NRC Staff
~.4 Bernard M. Bordenick a~
s Counsel for NRC Staff July 31,1989 8908110101 890731 PDR ADOCK 05000335 g . >
O eda 7}
50
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of Docket No. 50-335-OLA FLORIDA POWER AND LIGHT
' COMPANY (SFP. Expansion)
.(St.LuciePlant;UnitNo.1)
NRC STAFF'S BRIEF OPPOSING INTERVENOR'S APPEAL OF INITIAL DECISION AUTHORIZING SPENT FUEL P0OL RERACKING Patricia A. Jehle Counsel for NRC Staff Bernard M. Bordenick Counsel for NRC Staff July 31,1989
. TABLE OF CONTENTS Page TABLE OF AUTHORITIES ........................................... 11 I. INTRODUCTION .............................................. 1 II. STATEMENT OF. FACT 3 ........................................ 2 III. ARGUMENT .................................................. 4 A. The Licensing Board Did Not Err in Finding that the Combined Effects of Heat and Radioactivity Have'Been Adequately Studied
, and Did Not Err in Concluding that No Changes in the Neutron Attenuation of Boraflex Are Anticipated ...................................... 5 B. The Licensing Board Did Not Err in Relying on the Testimony of the Staff and Licensee Witnesses which Dealt with the Suitability of the Use of boraflex in Region I of the Spent Fuel Pool .......... 8 C. The Licensing Board Did Not Err in Authorizing the Spent Fuel Reracking and in Finding that the Region I Racks Do Not Utilize a New and Unproven Technology .................................. 13
- 1. The Licensing Board Correctly Relied on the Staff's and Licensee's Analyses of Dimensional Changes of Boraflex ............................. 13
- 2. The Licensing Board Did Not Err in Finding i that the In-Service Surveillance Program Will Detect Any Radiation Effects Beyond those Expected and Accommodated in the Rack Design .... 16 3.- The Licensing Board Did Not Err in Finding that the Presence of Gaps in Boraflex Panels Would
.. Not Significantly Reduce Neutron Keff Values I Above .95 ....................................... 19 4 The Licensing Board Did Not Err it Finding that Normal Use of the Spent Fuel PM 91 Produce !
Low Gamma Radiation Levels ...................... 23 i
- 5. The Licensing Board Did Not Err in Finding that the Licensee's Criticality Calculations and NRC Staff Review of the Calculations Are Adequate ... 24 l
l IV. CONCLUSION ................................................ 28
l TABLE OF AUTHORITIES Page ADMINISTRATIVE DECISIONS:
Florida Power Unit No.1), and Light, Company LBP 29 NRC (St.(slip Lucie op.Plant,)
(May 9, 1989) ......................................... passim Florida Power and Light Company (St. Lucie Plant, Unit No. 1), LBP-88-27, 28 NRC 455 (1988) ............. 4
, Florida Power and Light Company (St. Lutte Plant, Unit No. 1), ALAB-893, 27 NRC 627 (1988) .............. 3 Florida Power and Light Company (St. Lucie Plant, Unit No. 1), LBP-88-10A, 27 NRC 452 (1988) ............ 3 Georgia Power Company, et al. (Vogtle Electric Generating Plant, Units 1 and 2), ALAB-872, 26 NRC 127 (1987) ..................................... 4, 5 Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power 18 NRC Plant,(Units 1 and 2), CLI-83-32, 1309 1983) .................................... 24 Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), ALAB-728, 17 NRC 807 (1983) ..................................... 24 Philadelphia Electric Co. (Limerick Generating Station, Units 1 and 2), ALAo-819, 22 NRC 681 (1985) ........... 9, 12 Wisconsin Electric Power Co. (Point Beach Nuclear Plant, Unit 2), ALAS-78, 5 AEC 319 (1972) ............. 12, 13
, REGULATIONS:
j 10 C.F.R. 6 2.714 (1988) ................................ 3 10 C.F.R. 5 2.732 (1988) ................................ 24 10 C.F.R. 5 2.762(b)(1) (1988) .......................... 4 l 10 C.F.R. $ 2.762(c) (1988) ............................. 2 1
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Page.
REGULATIONS (Continued):
1
' 10'C.F.R..Part 2,' Appendix A, i V(d)(1)-(1988)l .......... 24 10C.F.R.f50.91(a)(4)(1988) .......................... 3 MISCELLANEOUS:-
"An Assessment of Boraflex Performance in Spent-Nuclear-
=,'
- Fuel Storage Racks," Electric Power Research Institute,
.EPRI NP-6159,- Final Report (December 1989) ........... 7, 8, 11, l 12, 13, 14, :
15, 19 i
" Consideration of Issuance of Amendment to Facility !
Operating License. and- Proposed No Significant Hazards; _ .i Consideration Determination and Opportunity for Hearing; Florida Power and Light Co. " ,
52 Fed. Reg. 32,852 (August 31,~1987} ................. 2 l
t Memorandum and Order (Dismissing Contention 5),
-(unpublished)(May'31,1988) ..........................- 3 ;
s NUREG-0800, Standard Review Plan, sec. 9.1.2, j
" Spent Fuel Storage", Rev. 6 (December 1988) ............ 20
.l
" Safety Evaluation by the Office of Nuclear Reactor ,
l Regulation Relating to the Reracking of the Spent i Fuel Pool at the St. Lucie Plant, Unit No. I as i Related to Amendment No. 91 to Unit 1 Facility l OperatingLicenseNo.DRP-67,"(March 11,1988) ....... 17 ,
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Y __ _- ___ _- _ _ _. _ _ _ _ _ . _ .8
.r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of )
) Docket No. 50-335-OLA FLORIDA POWER AND LIGHT )
COMPANY ) (SFPExpansion)
(St. Lucie Plant, Unit No.1) )
NRC STAFF'S BRIEF OPPOSING
., INTERVENOR'S APPEAL OF INITIAL DECISION
' AUTHORIZING SPENT FUEL P0OL RERACKING I. INTRODUCTION On May 9,1989, the Atomic Safety and Licensing Board (Licensing Board) presiding in the captioned proceeding issued an Initial Decision (Authorizing Spent Fuel Pool Reracking) II in which it found that the Florida Power and Light Company (Licensee) had met its burden of proof and was entitled to judgment on all contentions subject to a condition imposed 2/ and that License Amendment No. 91 to License No. DPR-67 should
-1/ Florida Power and Light Company (St. Lucie Plant, Unit No. 1),
LBP-89-12, 29 NRC (slip op.) (May 9, 1989). [ Hereinafter
. InitialDecision].
2/
The Licensing Board imposed the following condition: "in the event thatanyoftheRegion1Boraflextestcoupgnsaresubjectedtogamma irradiation equal to or greater than 1 x 10 rads, Licensee is directed to prepare within 30 days a study program to be approved by l the NRC staff and performed by the Licensee to assess the effect of the irradiation on the integrity of the Boraflex panels. The study l program should include blackness testing or a state-of-the-art .
g equivalent approved by the NRC staff . . ." Initial Decision of f l May 9, 1989 at 38; see generally 36-40. The condition was not l proposed by the NRC statt; however, neither the Staff nor the l
l (FOOTNOTE CONTINUED ON NEXT PAGE) l ,
l I
t __
remain in full force and effect. The Intervenor's brief in support of his appeal of the Initial Decision was filed before the Atomic Safety and Licensing Appeal Board on June 16, 1989. Intervenor's Appeal of Initial Decision (Authorizing Spent Fuel Pool Reracking). [ Hereinafter Intervenor'sAppealBrief]. The Nuclear Regulatory Comission staf f files this brief in opposition to the Intervenor's appeal pursuant to 10 C.F.R.
$ 2.762(c) of the Commission's regulations.
For the reasons discussed below, the NRC staff believes that the Licensing Board was correct in finding for the Licensee in this proceeding. The Staff, therefore, urges the Appeal Board to affirm the Licensing Board's Initial Decision of May 9,1989.
II. STATEMENT OF FACTS The factual background underlying the instant appeal is as follows:
Licensee on June 12, 1987 requcsted an operating license amendment to allow expansion of the spent fuel storage capacity from 728 to 1706 fuel assemblies at St. Lucie Plant, Unit No.1 (St. Lucie Unit 1). A notice of consideration of the issuance of the proposed amendment and an opportunity for a hearing was published in the Federal Register on June 7, 1987 by the U.S. Nuclear Regulatory Commission [ hereinafter NRC or Commission].
" Consideration of Issuance of Amendirent to Facility Operating License and
, Proposed No Significant Hazards Consideration Determitiation and Opportunity for Hearing; Florida Power and Light Co.," 52 Fed. Reg. 32,852 (FOOTNOTE CONTINUED FROM PREVIOUS PAGE Licensee challenges the imposition of this condition. See Testimony of Edmond G. Tourigny, trar. script at 552 [ hereinafter Tourigny, Tr.
at ].
l
- (August 31,1987). Mr. Campbell Rich [ hereinafter Intervenor], by a letter to the Secretary of the NRC on September 30, 1987, requested that a I public hearing on the amendment be held. The NRC staff (Staff) and the Licensee filed responsive pleadings on November 4, 1987, and November 9, 1987, respectively, which argued that the Intervenor's letter did not meet the requirements of 10 C.F.R. 5 2.714 and that therefore the request should be' denied. By a Licensing Board Memorandum and Order dated November 13, 1987, the Intervenor was permitted to file a Request for Hearing and Petition for Leave to Intervene [ hereinafter Petition]. The Petition was filed Janun y 15, 1988, and Intervenor proffered sixteen contentions for admission in this proceeding.
On March 11, 1988, pursuantto10C.F.R.550.91(a)(4),theStaff issued a Final No Significant Hazards Determination and issued Amendment 91 to Facility Operating License No. DPR-67 allowing expansion of the spent fuel pool capacity. A prehearing conference was held on March 29, 1988,'to hear oral argument from the parties concerning the admissibility i
of the proposed contentions. By Order and Memorandum dated April 20, 1988, the Licensing Board granted the petition to intervene and ruled on the contentions. Florida Power and Light Co. (St. Lucie Plant, Unit 1);
LBP-88-10A, 27 NRC 452 (1988). Seven contentions were admitted:
Contentions 3, 4, 6, 8, 9, 11, and 15, which were renumbered as Admitted Contentions 1 through 7. Id. at 456-69. This Order was affirmed by the Atomic Safety and Licensing Appeal Board. Florida Power and Light Co.
(St. Lucie Plant, Unit No. 1), ALAB-893, 27 NRC 627 (1988).
On May 31, 1988 Contention 5 was dismissed in a Licenting Board Memorandum and Order because the Intervenor had not indicated his
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intention to pursue the contention. Memorandum and Order (Dismissing Contention 5)at2-(unpublished)(May 31,1988). The Intervenor requested
- that Contention 2 be withdrawn on the ground of mootness. Contention 2
- was dismissed "with prejudice as moot" on July 17, 1988.
The Licensee filed on August 5,1988 motions for summary disposition of all the contentions. Each motion was supported by the Staff. By Memorandum and Order dated October 14, 1988, the Licensing Board c-determined that there was no genuine issue of material fact as to Contentions 1, 4, and 5. FloridaPowerandLightCo.(St.LucieNuclear Power Plant, Unit.1), LBP-88-27, 28 NRC 455. (1988). The Licensing Board granted sumary disposition as to portions of Contentions 3 and 7, and denied sumary disposition as to all of Contention 6. Id.
Evidentary hearings were held in Stuart, Florida on January 24-26, 1989. The Intervenor never challenged the professional qualifications of any witness. Although the Intervenor did not present direct testimony, he did conduct extensive cross-examination. The Licensee's direct case consisted of testimony of a panel composed of Dr. Stanley E. Turner, Chief Scientist for Holtec International; Edward J. Weinkam, III, Principal Engineer for the Nuclear Licensing Section of Florida Power and Light Company; and Dr. Krishna P. Singh, President of Holtec International. The
. Staff's direct case consisted of testimony presented by a panel composed of Dr. James Wing, a chemical engineer at the NRC; Mr. Edmond G. Tourigny, the NRC project manager for the St. Lucie plant; and Dr. Laurence I. Kopp, a nuclear engineer at the NRC.
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III. ARGUMENT The Intervenor does not clearly identify the errors of fact or law which are the subject of his appeal as is required by 10 C.'F.R.
62.762(b)(1). It is not sufficient that an appeal brief merely state that the appellant disagrees with the Licensing Board's findings; the appellant must state clearly the reasons for the disagreement.
Georgia Power Company (Vogtle Electric Generating Plant, Units 1 and 2)
ALAB-872,26NRC127(1987). In any event, the Staff believes it has identified the central concerns in the Intervenor's brief and will address these issues below. The Staff believes that.Intervenor's appeal is without merit and that the Licensing Board's decision should be affirmed.
A. The Licensing Board Did Not Err in Finding that the Combined Effects of Heat and Radioactivity Have Been AdeqJately Studied and Did Not Err in Concluding that No Changes in the Neutron Attenuation of Boraflex Are Anticipated The Intervenor alleges that the synergistic effects of heat and radioactivity on 83oraflex in spent fuel pools have not been well-studied and that the Licensing Board decision was not supported by facts.
Intervenor's Appeal Brief at 1-3. This is simply not true. The Staff and the Licensee reviewed a substantial body of data from tests conducted in
. 1979-1981 at the University of Michigan's Ford Nuclear Reactor. The tests were designed to evaluate the performance characteristics of Boraflex.
Testimony of Krishna P. Singh Contention 3 and 6 following transcript 139,at13-17[hereinafterSinghContentions3and6,ff.Tr.139];
Testimony of James Wing on Contention 3, following transcript 110, at 2-6
[hereinafterWingContention3,ff.Tr.110]. Dr. Turner performed analyses on the Ford M 1 ear Reactor data. Singh Contentions 3 and 6, ff.
Tr. 139, at 16. Dr. Turner also measured dimensional changes of actual
surveillance coupons from spent fuel pools. The surveillance coupors had been exposed to radiation and spent fuel pool environments. See $tanley E. Turner on Contentions 3 and 6, following Transcript 139, at 2, 10
[hereinafterTurnerContentions3and6,ff.139];SinghContentions3and 6, ff. Tr.139, at 16; Testimony of James Wing, transcript at 436-37, 442-43 [ hereinafter Wing, Tr.). The Licensing Board was justified in relying on the evidence presented regarding these tests, which demonstrate that Boraflex is suitable for use as a neutron absorber in spent fuel pool environments. Singh Contentions 3 and 6, ff. Tr. 139, at 13. The environment of the spent fuel pool will not have a significant negative effect upon Boraflex, based on results of tests to examine its long-term integrity. Testimony of Stanley E. Turner, transcript at 368 [ hereinafter Turner,Tr.].
Prior to accepting Boraflex as a neutron absorber material, the NRC required that the material be tested under physical conditions that greatly exceeded the severity of the environment to which the material would be exposed in actual use. Singh Contentions 3 and 6, ff. Tr. 139, at 14. The testing included heat aging and long-term exposure to borated water and irradiation. M. Some of the test results are abstracted below:
- 1. Boraflex exhibited excellent heat aging characteristics.
At temperatures up to 350 F, the hardness increased to barely 64 Shore A over a 6,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> exposure period. The i polymer passed a Mil-1-16923 Thermal Shock Test, thus demonstrating a significant margin of safety against temperature changes in the pool. '
- 2. The effects of long-term exposure of Boraflex material to high temperccure borated water were also evaluated to determine its stability under aggravated environmental conditions. Boraflex was placed in a boric acid solution (3,000 ppm boron), in a prer,sure bomb-type test vessel,
_ ~_
with a constant temperature of 240*F, maintained for an exposure period of over 6,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. The data demonstrated Boraflex's stability under aggravated environmental conditions.
M. The water in the St. Lucie Unit I spent fuel pool, with boron concentrations of about 1720 ppm, hovers around 100"F for most of the time, much below the 240*F in the pressure bomb test. Id. Moreover, Boraflex is never exposed to temperatures in excess of 200'F anywhere in the spent fuel pool. Id. Therefore, heat aging tests at 350'F, as noted
. above, were designed to simulate worse-than-possible scenarios. Singh Contentions 3 and 6, ff. Tr.139, at 14. In addition, water temperatures in the Ford Nuclear Reactor were substantially higher than the tempera-tures existing in the St. Lucie Unit I spent fuel pool. I_d . The tests at the Ford Nuclear Reactor were designed to identify the physical and chemical characterist5 of Boraflex under a variety of radiation levels, radiation rates and severe environments, with Boraflex receiving cumulative exposures of up to 10 12 rads (including 10 11 gamma). Id. at
- 15. No measurable effects were identified on the neutron attenuation capability of Boraflex due to the effects of environment or irradiation during the evaluation of the data. Id. at 15-16; Turner Contentions 3 and 6, ff. Tr. 139, at 2, 10-11.
The Licensing Board was justified in placing little reliance on the reference in the Electric Power Research Institute (EPRI) report cited by the Intervenor, concerning the use of unirradiated material in a test
I-evaluating the effects of heat on Boraflex 3/. Intervenor's Appeal Brief at 3; Footnote 3, supra. The EPRI Report evaluates the early qualification tests conducted _by the manufacturer of Boraflex. EPRI Report on Section 4. The report properly states that~ heat performance data standing alone does not provide particularly meaningful conclusions about the suitability of Boraflex in spent fuel pools. ld,at4-9. The report merely points to the need for synergistic studies of Boraflex, that is, studies which have evaluated the combined effects of heat and radiation on Boraflex. Ld. Such synergistic studies would provide data on the effects of Boraflex in a spent fuel pool environment.
On the basis of the record, the Licensing Board recognized that the EPRI Report does not change the fact that the Staff and the Licensee did
. consider the synergistic effects of heat and radiation on Boraflex.
Initial Decision at 16. The Staff testified that the results of in-reactor Boraflex irradiation studies would include the effects of reactor temperature along with radiation effects. Wing, Tr. at 548-549.
The Licensing Board ' recognized that the combined effects of heat and radiation were understood since temperatures are much higher in a reactor than in'a spent fuel pool; thus, synergistic effects of heat and radiation
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The Intervenor cites one sentence from "An Assessment of Boraflex
" Electric Power PerformanceinSpent-Nuclear-FuelStorageRacks(December Research Institute EPRI NP-6159, Final Report 1988)
[ hereinafter EPRI Report] as support for his claim that the Licensing Board should have found that Boraflex is not suitable for use in Region I of the spent fuel pool. Intervenor's Appeal Brief at 3.
The EPRI report points out that the effects of heat on irradiated Boraflex may be different than on unirradiated material. Id. at 4-9.
The EPRI report does not characterize as meaningless the technical documents which form the basis for the testimony of the Staff and L'icensee witnesses.
would be included in the reported in-reactor irradiation studies. Initial Decision at 16-17. The Licensing Board properly relied on the Boraflex studies and the 240*F test data, as well as the expert testimony that was based on these studies. The Licensing Board was correct in accepting the NRC staff conclusion that no significant teat-induced deterioration of Boraflex or its neutron-attenuation ability was expected. Initial Decision at 17.
B. The Licensing Board Did Not Err in Relying on the Testimony of the
, Staff and Licensee Witnesses which Dealt with the Suitability of the Use of Boraflex in Region I of the Spent Fuel Pool.
The Intervenor challenges the testimony of the Staff and the Licensee's witnesses with regard to the suitability of the use of Boraflex in Region I of the St. Lucie Unit I spent fuel pool. Intervenor's Appea!
Brief at 1-3, 7. The Intervenor also challenges the Licensing Board's reliance on the testimony of these witnesses.
The Licensing Board's reliance on the Staff and Licensee testimony, as regards the suitability of the use of Boraflex in Region I of the spent fuel pool, is fully supported by the record. Furthermore, the Inter-venor's challenge to the Staff and Licensee witnesses is without merit. A witness is qualified as an expert by knowledge, skill, experience, training, or education. Philadelphia Electric Co. (Limerick Generating
. Station, Units 1 and 2), ALAB-819, 22 NRC 681, 732 n. 67 (1985), citing, Fed. R. Evid. 702.
The Staff and Licensee expert witnesses were highly qualified to be i experts and clearly eligible to testify as such in this proceeding.
l Dr. Turner holds a B.S. degree in Chemistry and a Ph.D degree in Nuclear Chemistry and he has over 37 years of experience in nuclear chemistry and L__ _ _ _ _ _ _ _ _ _ _
L nuclear engineering. His work includes 12 years of experience in criticality analysis of high density spent fuel storage racks, as well as 1
the evaluation of Boraflex surveillance coupons and irradiation programs.
Testimony of Stanley E. Turner on Contention 7, following transcript 21, at Exhibit A. Resume [hereinaf ter Turner Contention 7, ff. Tr. 21).
Dr. Singh holds B.S., M.S. and Ph.D. degrees in Mechanical Engineering. He has 22 years of experience as a npchanical engineer and he has nine years of experience specifically. in spent fuel rack tech-
^
nology. Dr. Singh has published articles on spent fuel pool technology and has written numerous articles-on related engineering and mechanics topics. Singh Contentions 3 and 6 ff. Tr. 139 at Exhibit A, Resume.
Although Dr. Singh is a mechanical engineer, through his education and professional. work he has developed a broad knowledge of polymers and metals. Testimony of Krishna P. Singh, Transcript at 144-45 [ hereinafter Singh Tr.].
Mr. Weinkam received B.S. and M.S. degrees in Mechanical Engineering and has 14 years experience in mechanical and nuclear engineering, including six years with the NRC staff. Testimony of Edward J, Weinkam, III on Contention 7, following transcript 21, at Exhibit A. Resume
[ hereinafter Weinkam Contention 7, ff. Tr. 21].
. Dr. Wing holds B.S., M.S. and Ph.D. degrees in Chemistry. He has over 33 years of experience in nuclear chemistry and nuclear engineering,
. including 13 years with the NRC, where his duties have included the assessment of spent fuel pool water and material compatibility and ;
corrosion potential. In addition, Dr. Wing has published technical papers and laboratory reports on nuclear chemistry and radiochemistry.
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Professional Qualifications of James Wing, following transcript 110, at 15
[hereinafterWing,ff.110].
Mr. Tourigny received B.S. and M.S. degrees in Nuclear' Engineering and an M.S. degree in Industrial Engineering. He has over 20 years of nuclear experience including 13 years with the NRC. He has served as a project manager for four nuclear plants while at the NRC and has werked extensively on radioactive waste management. Qualifications and Experience of Edmond G. Tourigny, following transcript 110, at 16 (hereinafterTourigny,ff.110].
Dr. Kopp holds B.S. and M.S. degrees in Physics and a Ph.D. degree in Nuclear Engineering. He has 32 years of experience in physics and ruclear engineering and he has conducted reviews of criticality analyses of fresh and spent fuel storage racks safety evaluations and reviews of reactor core designs during his 23 years at the NRC. Professional Qualifications of Laurence I. Kopp, following transcript 110, at 20 (hereinafter Kopp, ff. 110].
The Intervenor presents'no challenge to the qualifications of the witnesses. The record shows ample support for the Licensing Board's finding that all witnesses had the appropriate expertise to support their l testimony. Initial Decision at 12; see also Wing, ff. Tr.110, at 15; 1
. Tourigny, ff. Tr. 110, at 16; Kopp, ff. Tr. 110, at 20; Turner Contention i
7, ff. Tr. 21, at Exhibit A; Singh Contentions 3 and 6, ff. Tr.139 at Exhibit A; Weinkam Contention 7, ff. Tr. 21, at Exhibit A.
Intervenor has mischaracterized the Licensing Board's view of Dr.
Wing's testimony on the safety of Boraflex. Intervenor's Appeal Brief at 1-2. The Licensing Board properly relied on the uncontroverted testimony
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of Dr. Wing on the Boraflex issues. Initial Decision at 16, 17, 19.
Furthermore, Dr. Wing's testimony was corroborated by Drs. Turner and Singh, as well as the EPRI report. See Singh Contentions 3 and 6. ff. Tr.
139, at 14-16; Turner Contentions 3 and 6, ff. Tr.139 at 2,10; "An Assessment of Boraflex Performance in Spent-Nuclear Fuel Storage Racks",
EPRI NP-6159, Final Report (December 1988) at 6-2, 6-3 (EPRI Report).
During cross-examination, the witness, Dr. Wing, was asked a number of questions concerning details of the reports on Boraflex which are referenced in his written testimony. See Wing Contention 3, ff. Tr. 110, at 2-6. These reports are summaries of several studies designed to evaluate the performance characteristics of Boraflex under a wide range of conditions; these studies were conducted under the direction of the manufacturer of Boraflex. Id.; Singh Contentions 3 and 6, ff. Tr. 139 at 13-17. The details of the Boraflex studies are contained in these primary documents. Wing, Tr. at 413-74. Dr. Wing could not remember the details of the very exhaustive experiments described in tha documents and, although he had based his conclusions on these reports, Dr. Wing did not bring the primary documents with him to the evidentiary hearing. See Wing, Tr. at 445. When the Licensing Board Chairman's statement regarding Dr. Wing's " obvious lack of presentation [ sic]" is taken in this context, it is clear the Licensing Board was referring to the weight to be given the testimony in light of the witness's inability to remember details.
Nonetheless, the record is supported by the corroborative testimony of the Licensee's witnesses and the record amply supports the Board's conclusion that no changes in the neutron attenuation capability of Boraflex are anticipated.
Furthermore, the fact that Staff and Licensee expert witnesses relied I on the opinions of other researchers and on the data collected and l
analyzed by other experts in forming their opinions is not a basis for l* challenging the competency of expert testimony, h Philadelphia Electric Company (Limerick Generating Station, Units 1 and l
2), ALAB-819, 22 NRC 681 (1985), TheAppealBoardhasstated,"[a]n
. expert is, of course, not expected to derive all his background data from experiments which he personally conducts; if that were required, scientific experts would rarely, if ever, be qualified to give any opinion on any subject whatsoever." Wisconsin Electric Power Co. (Point Beach NuclearPlant, Unit 2),ALAB-78,5AEC319,332(1972). Accordingly, the Licensing Board was correct in relying on the testimony of the Staff and i.icensee witnesses.
C. The Licensing Board Did Not Err in Authorizing the Spent Fuel Reracking and in Finding that the Region I Racks Do Not Utilize a New and Unproven Technology The Intervenor argues that the Licensing Board erred in authorizing the reracking of the spent fuel pool because the racks use ? new and unproven technology. Intervenor's Appeal Brief at 4-5. Underlying this challenge is the Intervenor's belief that the Board should have found:
(1) that Boraflex is expected to shrink 3 to 4 percent during normal, in-service use; (2) that the Licensing Soard should have found the Licensee violated the acceptance criteria for continued use of Boraflex; (3) that the presence of gaps in Boraflex panels will significantly reduce neutron attenuation and increase k eff values above 0.95; (4) that normal use of the spent fuel pool will result in high levels of gamma radiation; and (5) that the Licensee's criticality calculations and the NRC staff's
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review of them are inadequate. Intervenor's Appeal Brief at 3-8. The g Licensing Board,'s finding that the utilization of the high density racks designed and fabricated by the Joseph Oat Corporation is not utilization of a.new and unproven technology is supported by the record. Initial Decision at 27.
- 1. The Licensing Board Correctly Relied on the Staff's and Licensee's Analyses of Dimensional Changes of Boraflex.
Boraflex has been tested under carefully controlled conditions to determine its dimensional changes under irradiation. Singh Contentions 3 and 6, ff. Tr.139, at 16. Laboratory data were acquired at the University of Michigan's Ford Nuclear Reactor, and collated and analyzed.
Id.; see also, Wing Contention 3, ff. Tr.110, at 3. Gamma radiation induces cross-linking of the polyaer in Boraflex which leads to shrinkage.
Wing Contention 3, ff. Tr.110, at 3. The saturation of cross-linking in Boraflex occurs at the cumulative dose of approximately 10 10 rads, which is the point at which Boraflex reaches maximum shrinkage. Id. A December 1988 report published by the Electric Power Research Institute (EPRI) has indicated shrinking stops as cross-linking saturates at about 1 x 10 10 rads. EPRI Report at 5-2 to 5-3, 5-10 to 5-12; Turner Contentions 3 and
. 6, ff. Tr. 139, at 14 The Michigan results showed no significant increase in shrinkage of Boraflex at cumulative radiation doses from 5 x 109 to 5 x 10 10 rads. Wing Contentions 3, ff. Tr. 110, at 3.
When Boraflex is irradiated, the maximum shrinkage expected to occur is 2 to 21 percent, based on the results of the Michigan study and on the results of tests performed by Holtec International. Wing Contention 3, ff. Tr. 110, at 3; see also, Turner Contention 3 and 6, ff. Tr. 139, at 14 Turner has measured shrinkage on a substantial number of coupons and
l found between 2 to 2i percent of shrinkage. Turner, Tr. at 216. Results L
published in the EPRI Report project a maximum of 3 or 4 percent dimensional change.in Boraflex. Turner, Tr. 217-18, 356-57 The Intervenor asserts that Dr. Turner produced data indicating Boraflex shrinkage of 4 percent or' greater during in-service use. Intervenor.'s Appeal Brief at 3. Actually, Dr. Turner referred to dimensional changes, not shrinkage, measured on small Boraflex coupons when subjected to cumulative radiation doses of 5 x 10 10 rads, which exceeds the cumultive radiation doses expected in the St. Lucie spent fuel pool. Turner, Tr. at 387,.398; Singh, Tr. at 385-86. In addition, the Licensee and the Staff stated that dimensional changes in small Boraflex test coupons include two processes: shrinkage and edge deterioration effect. Turner Tr. at 368; Wing Tr..at 451. Analyses which have measured 4 percent dimensional changes in Boraflex identified both shrinkage and edge deterioration as responsible for the changes. Id.; see generally Turner, Tr. at 384-87, 402. The Electric Power Research Institute (EPRI) report proposes as a conservative design consideration that a value of 4 percent shrinkage be used in design work, pending the acquisition of additional data. Turner, Tr. at 217; "An Assessment of Boraflex Performance in Spent Nuclear-Fuel Storage Racks," EPRI NP-6159, Final Report, December 1988 at 6-2, 6-4. In
. calculations performed for the St. Lucie Unit I spent fuel pool, the conservative value of 4 percent shrinkage for Boraflex was used. Turner Contentions 3 and 6, ff. Tr. 139, at 7.
The rack materials, including the Boraflex, except in those areas intentionally set aside for in-service, surveillance testii , is not expected to exceed cumulative radiation exposures of 3 x 10 10 rads. Singh
p n
Contentions 3 and 6. ff. Tr. 139, at 16-17. This is less than 3 percent of the equivalent radiation dose given to this material in the laboratory tests described above. H.;SinghContentions3and6.ff.Tr.139at 16-17. The record amply demonstrated that the material has been subjected to testing environments more severe than environmental conditions which will be encountered in Region I of the St. Lucie Unit I spent fuel pool and still retained its neutron absorption capability and physical integrity.
Observed measurements of shrinkage indicated a maximum of 21 percent of shrinkage is expected for normal, in-service use of Boraflex. Turner, Tr. at 216; Turner Contentions 3 and 6, ff. Tr.139, at 4, 7 Wing Contention 3, ff. Tr. 110, at 3. The Intervenor incorrectly assumes that because the Region I racks have been designed to accommodate 3 to 4 percent shrinkage of the Boraflex panels, that shrinkage on that magnitude will definitely occur. Intervenor's Appeal Brief at 3. The Licensee has used oversized Boraflex panels, which will accommodate 3 to 4 percent shrinkage, as a design conservatism. The Licensee and the Staff expect that the Boraflex panels in the Region I racks will exhibit shrinkage no greater than 21 percent. The Intervenor has not established that the Licensee and Staff estimi.tes are invalid, nor has he pointed to any
. evidence which supports his claims that shrinkage of 3 to 4 percent shrinkage is anticipated during normal, in-service use. The Licensing Board properly relied on the Staff's and the Licensee's conclusions about the dimensional changes of Boraflex.
_ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ .I
- 2. The Licensing Board Did Not Err in Finding that the In-Service Surveillance Program Will Detect Any Radiation Effects Beyond those Expected and Accommodated In the Rack Design, i- The Intervenor argues that the Licensing Board erroJ in not recognizing that one of.the " Staff's acceptance criteria" for continued use of Boraflex will be violated and that the Board should have denied the license amendment. Intervenor's Appeal Brief at 3. The Staff points out that the Intervenor erroneously refers to the cited acceptance criterion as the Staff's acceptance criterion. See M. at 3. In fact, the Intervenor is referring to an acceptance criterion which is part of the Licensee's in-service surveillance program for Boraflex; it has received NRC staff approval, but is not an NRC criterion. The in-service surveillance program is a voluntary proposal instituted by the Licensee.to ensure the integrity of Boraflex.
The N% staff discussed the in-service surveillance program of Boraflex in ite Safety Evaluation Report (SER). " Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to the Reracking of the Spent Fuel Pool at the St. Lucie Plant, Unit No. I as Related to Amendment No. 91 to Unit 1 Facility Operating License No. DRP-67" (March 11,1988),
at p. 5; Staff Exh. 1. The relevant section of the SER provides:
To provide added assurance for detection of degradation of the Boraflex, the licensee has committed to conduct a long-term and
. accelerated surveillance test program. Each surveillance coupon (5 inches by 15 inches) containing Boraflex of a R thickness similar to that used in the racks, is encased in a stainless steel jacket, the alloy of which is identical to that ,
used in the racks. The coupon jacket permits wetting and !
venting of the specimen to the spent fuel pool water similar to that of the rack. The long-term coupon examination frequency occurs after irradiation times of 90 days, 180 days, I year, 5 years,10 years,15 years, 25 years and 35 years. The accelerated test coupon examination frequency is after each discharge from the second to ninth discharge rack utilization.
Acceptance criteria for continued use are dimensional changes
L of no more than'2.5% from the original, hardness not less than 90% of the original, and minimal areal density of boron not less than the original.
Id.;. see also Weinkam Contentions 3 and 6, ff. Tr.139, at 4-5; Turner Contentions 3 and 6, ff. Tr. 139, at 16-17 The in-service surveillance program is designed to detect any dimensional changes in Boraflex and the NRC will be informed if changes in Boraflex beyond the Licensee's acceptance criterion occurs. M. The Licensee has also indicated that corrective actions exist for assuring safe fuel storage if unanticipated Boraflex degradation were to occur4 /.
M. The surveillance program requires the examination of test coupons at scheduled intervals. M.
4/
The degraded Boraflex could be evaluated to determine whether the degradation would adversely affect FPL's ability to limit for the St. Lucie spent fuel satisfyIfthe pool. the0.95 pool k'bbuld still satisfy this limit, no further action would be necessary.
Administrative controls could be imposed on the enrichment and/or burnup of fuel to be placed in or adjacent to storage cell locations containing degraded Boraflex to assure that k,ff would remain less than or equal to the 0.95 limit.
Poison material, such as a control element assembly, could be added to any fuel. assembly to be placed in a storage cell with degraded Boraflex. This would reduce the k,ff to less than or equal to the 0.95 limit.
The storage cells with degraded Boraflex could be blocked off to prevent their loading with fuel assemblies.
These options --'and, perhaps, other as well -- would be considered by FPL if Boraflex degradation problems were to occur.
Weinkam, Contentions 3 and 6 ff.139, at 5-6.
_1 1
i One of the acceptance criteria for continued use of Boraflex is that dimensional changes of the coupons deviate no more than 21 percent from the original. M . The Intervenor incorrectly assumes that because f laboratory measurements have indicated 3 to 4. percent dimensional changes in Borc the results of the in-service surveillance program will necessarily indicate dimensional changes in excess of 21 percent. Of course, that is not the case. The Intervenor has' presented no evidence that the Licensee hrs not met its own acceptance criter1a; ne merely a.
alleges that under normal, in-service conditions Boraflex will shrink 3 to 4 percent. If any unexpected degradation occurs the corrective actions set forth in Footnote 4 supra will be triggered.
The in-service surveillance program itself has not been challenged by the Intervenor on appeal. The Licensee and the Staff presented direct tastimony at the January,1989 hearing on the in-service surveillance program which w s unopposed by the Intervenor. Weinkam Contentions 3 and 6, ff. Tr. 139, at 3-6; Turner Contentions 3 and 6, ff. Tr. 139, at 15-17; Tour igny Contention 3, ff. Tr.110, at 6-8. The Intervenor did not present testimony on.this issue, nor did he conduct cross-examination of Licensee and Staff witnesses or otherwise challenge this program. The Licensing Bocrd found the Licensee's in-service surveillance program to be consistent with the surveillance program reconinended by EPRI to evaluate the performance of Boraflex. Initial Decision at 25. The record amply supports the Licensing Board's finding that the in-service surveillance program will detect any radiation effects beyond those expected and accommodated in the rack design.
- 3. The Licensing Board Did Not Err in Finding that the Presence of Gaps in Buraflex Panels Would Not Significantly Reduce Neutron Attenuation and Increase kg ,. Values Aby.;te 0.95.
The Intervenor argues that the pres.ae of emps in Boraflex panels in spent fuel pools would significantly decrease the neutron attenuation capability of the panels and would increase k eff values beyond NRC guidelines and industry standards. Intervenor's Appeal Brief at 6-8.
This argument is without merit. The NRC guidelines and industry standards I which limit the maximum k,ff to 0.95, including all uncertainties, provide a substantial suberiticality margin as a factor of safety to assure conformance with General Design Criterion 62 and to preclude the possibility of a criticality incident in the storage facilities. Turner Contention 7, ff. Tr. 21, at 13; see also NUREG-0800, Standard Review Plan, Section 9.1.2, " Spent Fuel Storage," Rev. 6 (December 1988).
The physical integrity and neutron absorption pruperties of the Boraflex poison material are important in assuring the suberiticality safety of the Region I spent fuel storage rccks under all credible conditions. Turner Contentions 3 and 6, ff. Tr. ?39, at 6. Gaps which might develop in the Boraflex could perturb the local reactivity and might possibly have a significant effect on the system reactivity, depending l upon the size and spatial distribution of the gaps. M. However, l
. numerous small gaps distributed randomly in size and location would have only a minor effect on the system reactivity, well within tolerances already established (since the local reactivity effects would be averaged over the entire rack array). Turner Contentions 3 and 6, ff. Tr. 139, at 6; see also Singh, Tr. at 313.
- a l
The Boraflex panels used in Region I of the St. Lucie Unit I racks
{
are continuous sheets, over 7 inches wide and 143 inches long, more than 6 inches longer than the active fuel length. Singh Contentions 3 and 6, ff.
Tr.139, at 11; Turner Contentions 3 and 6, ff. Tr.139, at 6. The i stainless steel plates enveloping the Boraflex Knels are spot-welded to the stainless steel cdn of the storage cells through cutouts in the
, Boraflex panel every twelve inches along the edges of both sides. M.
The Licensee indicates that upon shrinking, the Boraflex panels may encounter these spot welds, and local stresses might appear along the axial length of the panels. Id. The spot-welds and cutouts do not function as a gap promoting mechanism, but rather as a gap controlling mechanism. Singh, Tr. at 311-14 The purpose of the design is to control the size and location of any gaps wMch may develop in Boraflex panels.
The development of stresses in the controlled gap system is limited; the spot-welds and cutouts will produce self-limited stresses, and controlled gaps. M.at317. Any gaps which form by this controlling mechanism will be smaller than one-half an inch every 12 inches or one-quarter of an inch every 6 inches along the sides of the Boraflex panels. Id. at 323, 333-34.
The maximum shrinkage of Boraflex under long-term irradiation is conservatively bounded at 4 percent. Turner Contentions 3 and 6, ff. Tr.
139, at 6. Conservative calculations for the limiting condition of gaps in all Boraflex panels, coincidentally at the same axial elevations, result in a maximum k,ff of 0.771 under normal operating conditions, or 0.948 assuming the concurrent loss of all soluble boron in the spent fuel pool water. M.at7. The calculations are based on the additional
conservative assumption of an infinite number of fuel assemblies in the storage cells, all of infinite length. M. Even for the limiting condition of gaps appearing in the Boraflex of Region I racks attributable to 4 percent shrinkage, the maximum k,ff remains within acceptable bounds.
E-With consideration of the double contingency principle5 ,/ and w' ith credit given for the soluble boron present, calculations for a hypothetical loss of all the Boraflex resulted in a maximum k of 0.875 eff for Region I which is still well below the limit of k,ff of 0.95. Turner Contentior.s 3 Tnd 6, ff. Tr. 139, at 9. In addition, Dr. Turner calculated the reactivity for an extremely improbable condition which assumed: (1) 4 percent shrinkage for all panels in Region I (144 inch panels with 5.72 inch shrinkage); (2) the formation of 5.72 inch gaps at the mid-axial plane (the most reactive position) in all Region I panels; and (3) no boron in the spent fuel pool water. Turner, Tr. at 366. A keff of 0.992, a value below criticality, was calculated by Dr. Turner for this condition. Turner, Tr. 366, 412. When credit is given to '
5/
~
Section 111.1.2 (Postulated Accidents) of an April 14. 1978 NRC staff letter, providing guidance on criticality analysis, invokes the double contingency principle of ANSI N16.1-1975 for fuel pool analyses and states that:
The double contino-ncy principle of ANSI N16.1-1975 shall be applied. It shall require two unlikely, independent, concurrent events to produce a criticality accident. Realistic initial conditions (e.g., the presence of soluble baron) may be assumed for the fuel pool and fuel assemblies.
(Turner Contention 7, ff. Tr. 21, attli-13.)
l
borating the pool water to 1720 ppm, the reactivity decreases considerably. Id. at.413. The Intervenor incorrectly interprets the applicability of.the double contingency principle to the ab~ove hypothetical. Because the hypothetical condition is so improbable, if not impossible, it is not necessary to consider it an expected condition. The presence of soluable boron in the spent fuel pool water may properly be taken into account. The Licensing Board did not err in finding that the Region I spent fuel pool racks are acceptable.
- 4. The Licensing Board Did Not Err in Finding That Normal Use of the Spent Fuel Pool Will Produce Low Gamma Radiation Levels.
~The Intervenor alleges that the Licensing Board misunderstands the types of fuel to be stored in Region I. Intervenor's Appeal Brief at 5.
The Board stated that the normal use of Region I will make the shrinking of Boraflex and subsequent gap formation nonexistent or minimal. Initial Decision at 37. The normal use of Region I is for the storage of fresh fuel during preparation for a refueling operation. Testimony of Edward J.
Weinkam, transcript at 140 [ hereinafter Weinkam, Tr.]; Turner, Tr. at 350, Fresh fuel, although highly reactive, does not emit gamma radiation; spent fuel does emit gamma radiation, but is much less reactive than fresh
. fuel. Under normal conditions spent fuel will be stored in Region II.
Weinkam, Tr. at 140; Turner, Tr. at 350. Region I will be used to store spent fuel due to an emergency core off-load or when incompletely burned fuel is removed from x he reactor; these circumstances are not normal. :
Turner, Tr. at 350; in fact, the 13 spent fuel assemblies currently l stored in Region I are stored there to provide data on gamma irradiation under the in-service surveillance program. Terner on Contentions 3 and 6, ff. Tr. 139, at 15-16; Weinkam on Contention 3 and 6, ff. Tr. 139, at 5.
t
l lz ,
'Therefore, the Board's conclusion that normal use of Region I will produce.
low gamma radiation levels is accurate and supported by the record.
-5. The Licensing Board Did Not Err in Finding that the K Licensee's Criticality Calculations and NRC Staff Review of the Calculations Are Adequate.
1 The Intervenor states that the Board erred in relying on criticality calculations performed by the Licensee because the Staff did not independently perform its own calculation. Intervenor's Appeal Brief at 7-8. In addition, the Intervenor challenges the accuracy of the Licensee's calculations because they were conducted by the Licensee's own witness. Intervenor's Appeal Brief at 7. The Intervenor believes tM Licensing Board should have found that the Licensee's criticality calculations and the Staff's review of them are inadequate. Intervenor's Appeal Brief at 7-8.
Interrenor's allegations in this regard are without merit. -First, as regards tie adequacy of the Staff's review, the Licensee's application is at issue, not the adequacy of the Staff's review of the application. Id.;
PacificGasandElectricCo.(DiabloCanyonNuclearPowerPlant,UnitsI and 2), ALAB-728, 17 NRC 777, 807 (1983), review denied, CLI-83-32, 18 NRC 1309(1983). While the Intervenor is free to challenge directly the safety of the application amendment by advancing contentions and this appeal, he may not proceed on the basis of allegations that the Staff has somehow failed in its performance. The central issue on appeal is whether or not the Licensing Brard had a basis to conclude that Licensee has met its burden of proof. See 10 C.F.R. I 2.732 and 10 C.F.R. Part 2, Appendix A,6V(d)(1).
Second, the NRC staff did not perform its own criticality analyses for St. Lucie because that is not general practice in an NRC review of a licensee's submittal for a spent fuel pool expansion. Testimony of Laurence I. Kopp, transcript at 495 [ hereinafter Kopp, Tr.].
Additionally, Intervenor's argument overlooks the fact that the Staff reviews criticality calculations by examining several factors: (1)the codes and methodologies used to calculate reactivity; (2) the uncertainty assumptions and configuration patterns; (3) tne amount of Boraflex; (4) fuel enrichment; and (5) the researcher and tne laboratory conducting the analyses. Kopp, Tr. at 535. The Staff reviewed the Licensee's prefiled written testimony and heard the oral testimony on cross-examination of Dr.
Stanley E. Turner on criticality analyses which assumed one-half inch gaps every 12 inches in the Boraflex in the St. Lucie racks. H . The Staff concluded that there would be no adverse effect on k,ff which would violate the acceptance criteria of 0.95. M.
'!he methods and assumptions used by Dr. Turner are widely used in nuclear reactor physics and the Staff finds them acceptable. g.at534.
Dr. Turner has performed many criticality calculations and the Staff is familiar with his expertise. M.at533-35. The Licensee's calculations are not unreviewed. Furthermore, the Intervenor has advanced no basis for questioning the reliability of the criticality calculations or Dr.
l Turner's qualifications to perform these calculations. The Staff's witness, Dr. Kopp, had full understanding of the Licensee's criticality I calculations. Dr. Kopp's experience with criticality calculations is extensive and includes the review of generic criticality calculations and criticality calculations performed by the Staff for the Turkey Point Plant I
f-Units 3 and 4 Kopp, Tr. at 531-537. This experience provides the basis for Dr. Kopp's , professional opinion on the reliability of the critic 0lity calculations for Region I of the spent fuel pool at St. Lucie Unit 1. Id.
Prior to the issuance of the St. Lucie Unit I spent fuel amendment, the de' sign basis k,ff limit was 0.95. Turner Contention 7, ff. Tr. 21, at
- 13. The spent fuel pool expansion did not modify this design basis limit.
Turner Contentions 3 and 6 ff. Tr.139, at 8-9. Thus, the Licensee and r
the Staff established on the record that the amenoment has not decreased the margin of safety for preventing a criticality accident at St. Lucie Unit 1.
The central technical issue is whether unsafe and unpredictable gap formation will occur in Region.I of the St. Lucie Unit I racks. The Staff did not identify a mechanism for gap development based on the Safety Analysis Report submitted by the Licensee. Wing, Tr. at 543-47. The Licensee, however, postulated a possible mechanism for gap development in Region I of the St. Lucie Unit I storag:: racks that is, the use of a cut-out design in conjunction with spot-welding every 12 inches along both sides of the Boraflex panels. Singh, Tr. at 310-14. The Licensee has postulated that gaps may occur in a systematic pattern. M. The '
Licensee's conclusion is based on additional analyses which were not available to the NRC when the Staff reviewed the Safety Analysis Report and prepared its SER and the prefiled written testimony. Tourigny, ir.
507-14. Using a very conservative assumption that one gap of one-half an inch will occur every 12 inches, the Licensee established that the degree of gapping would not effect reactivity in the spent fuel pool. Turner Contentions 3 and 6, ff. Tr. 139, at 6. Furthermore, the NRC staff
evaluated the sworn written testimony and heard the oral testimony of the Licensee's witnesses which described: 1) the rack design in detail; 2) the conservative assumptions concerning gap formation; and 3) the effects on the reactivity of the pool. Kopp, Tr. at 534-36; Tourigny, Tr. at 499-504, 507_-14, 540-48; Wing, Tr. at 554-45. Therefwe, there is no basis for the Intervenor's allegations that the Staff was unaware of confused about the Region I rack design. Intervenor's Appeal Brief at 4-5. The Staff concluded that should the maximum projected gap formation
~
of one gap of one-half inch occur every 12 inches, there will be no criticality concern. Kopp, Tr. at 534-36; Tourigny, Tr. at 540-48; Wing, Tr. at 544-45.
The Licensing Board correctly and fully described the controlled gap system which is part of the Region I rack design. Initial Decision at 37-38. The Intervenor provided no evidence of mechanisms which could l result in the formation of uncontrolled gaps. While the controlled gap l system at St. Lucie Unit 1 is unique and has not been tested under L operating conditions, the design elements and mechanical and physical principles on which the design is based are well-established. Initial Decision at 26. Therefore, the Licensing Board was correct in finding that the utilization of high density racks in Region I is not utilization l of new and unproven technology. Initial Decision at 27. Furthermore, as an additional measure of conservatism, the Licensing Board required
" blackness testing" or an NRC-approved, state-of-the-art equivalent by the Licensee. Initial Decision at 36-38, 39-40. Through blackness testing the size and distribution of gaps larger than one-half inch can be identified. See Tourigny, Tr. at 552; Turner Tr. at 322. This additional
I surveillance program should provide accurate data based on observation..,
which will supplement the Licensee's prediction of gap formation and insure that Boraflex integrity is not compromised and that sub-criticality is maintained in.the spent fuel pool.
The Licensee has established that the use of the Region I racks is [
4 not use of a new and unproven te'chnology. The Staff recognizes, as does the Licensing Board, that the cut-out feature has not been tested under operating conditions. However, the Licensee has provided evidence which i
~
supports its position that the cut-out feature may produce controlled gaps but that such gaps will not compromise safety in the spent fuel pool. The Staff believes that the Licensing Board's authorization of the reracking amendment is amply supported by the record and, thus, should be affirmed.
IV. CONCLUSION The substantial weight of the evidence supports the Licensing Board's (1) reliance on the testimony of the Staff and Licensee witnesses regarding the suitability of the use of Boraflex in Region I of the spent fuel pool, and (2) decision tc authorize reracking of the spent fuel poui.
The Atomic Safety and Licensing Appeal Board should affirm the Licensing Board's Initial Decision of May 9, 1989.
Respectfully submitted, 4 <%
- Patricia A. Jehle Counsel for NRC Staff b
[nBhnrdM.Bordenik Counsel for NRC Staff Dated at Rockville, Maryland this 31 day of July,1989.
L_; ____- __
[-O'aT ii?
y, UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'89 AUG -7 N1 :26 BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD ,
crrt .
OCCi.i 4 .
In the Matter of ) TF M3
) Docket No. 50-335-OLA FLORIDA POWER AND LIGHT )
COMPANY (SFPExpansion)
(St.LuciePlant,UnitNo.1) )
CERTIFICATE OF SERVICE I hereby certify.that copies of "NRC STAFF'S BRIEF OPPOSING INTERVENOR'S APPEAL OF INITIAL DECISION AUTHORIZING SPENT FUEL P0OL RERACKING" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or as indicated by an asterisk through deposit in the Nuclear Regulatory Commission's internal mail system, this 31st day of July,1989:
Thomas S. Moore
- Alan S. Rosenthal*
Atomic Safety and Licensing Appeal Atomic Safety and Licensing Appeal Board Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission '
Washington, D.C. 20555 Washington, D.C. 20555 Howard A. Wilbur* Glenn 0. Bright
- Atomic Safety and Licensing Appeal Administrative Judge Board Atomic Safety and Licensbig Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Coridssion Washington, D.C. 20555 Washington, D.C. 2055L B. Paul Cotter, Jr., Chairman
- Michael A. Bauser, Esq.
Administrative Judge Harold F. Reis, Esq.
. Atomic Safety and Licensing Board Newman i Holtzinger, P.C.
U.S. Nuclear Regulatory Commission 1615 L. Street, N.W.
Washington, D.C. 20555 Washington, D.C. 20036 Richard F. Cole
- Docketing and Service Section*
Administrative Judge Office of the Secretary Atomic Safety and Licen34rw Board U.S. Nuclear Regulatory Commission ;
U.S. Nuclear Regulatory JWaission Washington, D.C. 20555 Washington, D.C. 20555 Campbell Rich Atomic Safety and Licensing 4626 S.E. Pilot Avenue Board Panel (1)* Stuart, Florida 34997 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ;
E
' , f. > l
)
I Atomic Safety and Licensing Richard J. Goddard
- AppealPanel(5)* U.S. Nuclear. Regulatory Comission a b.S, Nuclear Regulatory Coonission Regional Counsel, Region II Washingt.on, D.C. 70;;; 101 Marietta Street, Suite 2900 Atlanta, Georgia .30323 Adjudicatory File
- Atomic Safett and Licensing Board Office of the Secretary
- l U.S. Nur' ear Regulatory Comission U.S. Nuclear Regulatory Comission Washington,Lt.C. 20555 Washington, D.C. 20555 j i
(Lo w -4g-] \8,
/
Patric16 A. Jenle
. Counsel for NRC' Staff' 4
i
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