ML20236D573

NRC Staff Proposed Findings of Fact & Conclusions of Law in Form of Initial Decision.* Proposed Transcript Corrections to 890124 & 26 Proceedings & Certificate of Svc Encl

ML20236D573
Person / Time
Site:	Saint Lucie
Issue date:	03/17/1989
From:	Patricia Jehle NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:
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CON-#189-8316 OLA, NUDOCS 8903230165
Download: ML20236D573 (50)

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hb(N March 17, 1989 ImE ;t p'

' v ,y UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION gg MR 20 P4 :07 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD cn, De In the Matter of )

) Docket No. 50-335-OLA FLORIDA POWER AND LIGHT )

COMPANY ) (SFPExpansion)

)

(St. Lucie Plant, Unit No.1) )

NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM 0F AN INITIAL DECISION

1. INTRODUCTION AND BACKGROUND

1. Florida Power and Light Company (hereinafter Licensee or FPL) on June 12, 1987 requested 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 1).

2. 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 7,1984 by the U.S. Nuclear Regulatory Commissior .

(hereinafter NRC or Commission). 52 Fed. Pg . 32,852 (1987). Mr.

Campbell Rich (hereinafter the Intervenor), by a letter to the Secretary of the NRC on September 30, 1987 requested that a public hearing on the amendment be held. The NRC Staff and the Licensee filed responsive l

pleadings on November 4, 1987 and November 9, 1987, respectively, which contended that the Intervenor's letter did not meet the requirements of 10 C.F.P. 5 2.714 and that therefore the request should be denied. By Board Memorandum and Order dated November 13, 1987 the Intervenor was 0 [ d

e permit'ted to file a Request for Hearing and Petition for Leave to

-Intervene (hereinafter Petition). The Petition was filed January 15, 1988 and Intervenor profferred sixteen contentions to be admitted in this proceeding.

3. On March 11, 1988, pursuant to 10 C.F.R. 9 50.91(a)(4), the Staff 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.

4. A prehearing conference was held on March 29, 1988 to hear oral argument from the parties. By Order and Memorandum dated April 20, 1988 we granted the petition to intervene and ruled on the contentions.

Florida Power and Light Co. (St. Lucie Plant, Unit I), 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. The decision was affinned 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).

5. On May 31, 1988, Contention 5 was dismissed by our Memorandum and Order because the Intervenor had not indicated his intention to pursue the contention. Memorandum and Order at 2.

6. The Intervenor requested that Contention 2 be withdrawn on the j ground of mootness. We dismissed Contention 2 "with prejudice as moot" on July 17, 1988.

7. The Licensee filed motions for summary disposition of all the contentions on August 5, 1988 and each motion was supported by the Staff.

l By Memorandum and Order dated October 14, 1988 we determined that there 1

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was no genuine issue of material fact as to Contentions 1, 4 and 5.

Florida Power and Light Co. (St. Lucie Nuclear Power Plant, Unit 1),

LBP-88-27,28NRC455.(1988). We granted summary disposition to portions of Contentions 3 and 7, and denied summary disposition as to all of Contention 6. Id.

8. Hearings were held in Stuart, Florida on January 24-26, 1989.

At the hearing, the Staff and Licensee presented testimony on Contentions 3, 6 and 7 by means of witness panels which we found were in general agreement as to the merits of the contentions. As is further discussed below the witnesses were properly qualified at the proceeding and the Intervenor did not challenge the professional qualifications of any witness. The Intervenor did not present direct testimony, but attempted to conduct his case by cross-examination.

9. The Licensee's direct case consisted of testimony of a panel composed of Dr. Stanley E. Turner, Chief Scientist for Holtec I International; Edward J. Weir.kam, III, Principal Engineer for the Nuclear Licensing Section of Florida Power and Light Company; and Dr. Krishna P.

Singh, President of Holtec International.

10. Dr. Turner has over 37 years of experience in nuclear chemistry and nuclear engineering. His work includes 12 years of experience in 1

- criticality analysis of high density spent fuel storage racks, as well as 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, at

]. Dr. Singh has 22 years of experience as a mechanical engineer and he has nine yeart of experience specifically in spent fuel rack technology.

I I

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4 Singh on Contentions 3 and 6, following transcript 139, at Exhibit A, Resume i

[ hereinafter Singh Contentions 3 and 6 ff. Tr. 139, at ]. Mr. Weinkam has 14 years experience in mechanical and nuclear engineering including  !

l years with the NRC staff. Testimony of Edward J. Weinkam, III on Contention 7, following transcript 21, at Exhibit A, Resume [ hereinafter  ;

Weinkam on Contention 7, ff. Tr. 21, at ].

11. The Staff's direct case consisted of testimony presented by a panel composed of Dr. James Wing, a chemical engineer at the NRC; Edmond i G. Tourigny, the NRC project manager for the St. Lucie plant; and Laurence I. Kopp, a nuclear engineer at the NRC. Dr. Wing has over 33 years of experience in nuclear chemistry and nuclear engineering, including 13 l years with the NRC where his duties include the assessment of spent fuel pool water and material compatibility and corrosion potential.

Professional Qualifications of James Wing, following transcript 110, at  !

15. Mr. Tourigny has over 20 years of nuclear experience including 13 years with the NRC. He has served as a project manager at four nuclear plants while at the NRC and has worked extensively on radioactive waste  !

management. Qualifications and Experience of Edmond G. Tourigny, following transcript 110, at 16. Dr. Kopp has 32 years of experience in physics and nuclear engineering and he has been involved in the review of  ;

criticality analyses of fresh and spent fuel storage racks safety evaluations and of reactor core designs during his 23 years at the NRC.

Professional Qualifications of Laurence I. Kopp, following transcript 110, at 20. The Intervenor did not challenge the qualifications of these NRC staff witnesses.

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12. This decision is based upon the record in this hearing, and incorporates the findings and conclusions of law that follow. Any proposed findings of fact or conclusions of law submitted by the parties which are not incorporated into this decision are rejected as being unsupported in law or fact or as unnecessary to this decision.

II. FINDINGS OF FACT A. Contentions 3 and 6 j

1. Contention 3 asserts that:

l The Licensee and Staff have not adequately considered or analyzed materials deterioration or failure in materials integrity from the increased generation of heat and radioactivity as a result of  ;

increased capacity in the spent fuel pool during the storage period authorized by the license amendment.

13. Contention 3 raised the issue of whether the effects of heat and radiation in Boraflex and non-Boraflex materials used in the construction of the spent fuel racks result in safety problems. We summarily disposed of the safety issue concerning the use of non-Boraflex materials in the i

pool racks. S.D. Order of October 14, 1988 at 22. The only issue in Contention 3 which remained for consideration at the hearing was whether there are outstanding safety problems with the use of Boraflex due to the long-term effects of heat and radiation. Id.

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2. Contention 6 asserts that:

The proposed use of high-density racks designed and fabricated by the Joseph Oat Corporation is utiliza-tion of essentially new and unproven technology.

14. The Intervenor pointed to two areas of concern regarding the rack design and fabrication: 1) that the development of gaps might l compromise safety, and 2) that the modifications were unproven and  ;

untested. Amended Petition at 8. In our Order of October 14, 1988 we summarily disposed of the non-Boraflex issues in Contention 5, but we directed the Licensee to establish that there are no outstanding safety issues concerning the use of Boraflex in the racks. S.D. Order of October i

14,1988 at 34.  :

I

15. We will address Contentions 3 and 6 together because the contentions themselves are related and the technical evidence presented on i these issues overlaps. j t

3. Boraflex as a Neutron Absorber  !

16. Neutron attenuation in the St. Lucie spent fuel pool is accomplished by using borated water and Boraflex, a neutron absorber I l

material. Singh on Contentions 3 and 6, ff. Tr.139, at 6. Boraflex, l

commonly referred to as a neutron " poison," has been widely used in the I

industry as an effective neutron absorber since 1980. Id. at 6-7. It is .l built into the spent fuel racks to ensure that stored fuel remains at suberitical levels within the margin of safety. Testimony of Edmond G.

Tourigny on Contention 6, following transcript 110, at 10 [ hereinafter i

1

1 *, ,

1 Tourigny on Contention 6, ff. Tr. 110, at _ ]. Boraflex is a polysiloxane elastomer, in which boron carbide particles are uniformly dispersed in a homogeneous, stable matrix. Singh on Contentions 3 and 6, ff. Tr. 139, at

7. When the polymer is exposed to radiation it experiences an increase in hardness and a loss in flexibility, although it maintains its neutron attenuation capability and its physical integrity. Testimony of James Wing on Contention 3, following transcript 110, at 2-3 [ hereinafter Wing i

on Contention 3, ff. Tr. 110, at ___]; see_also, Singh on Contentions 3 and

6. ff. Tr. 139, at 7.

4. Design of the St. Lucie 1 Storage Racks l

17. The manufacturer of the St. Lucie 1 storage racks, Joseph Oat

\

Corpora-tion, has over eight years of experience in the manufacture of j high-density spent fuel racks for reracking projects. Singh on Contentions 3 and 6, ff. Tr. 139, at 4. For each of its reracking projects. Joseph Oat has drawn upon its extensive experience and technological base in material procurement, handling, forming, machining, welding, quality control, quality assurance, relevant non-destructive I

examination, and sound knowledge of the governing codes and standards. i 1.d

18. The principal material used in the construction of St. Lucie 1 racks is austenitic stainless steel. Singh on Contentions 3 and 6, ff.

Tr. 139, at 4. Joseph Oat Corporation has manufactured pressure vessels, heat exchanges and other types of equipment out of austenitic stainless steel used in nuclear power plants for over 20 years. Jd,.at4-5. <

Hundreds of pieces of Oat equipment have been employed in nuclear and non nuclear plants alike for years; not a single case of equipment failure

1 leading to plant shutdown has been ascribed to Oat-supplied equipment. l Singh on Contentions 3 and 6 ff. Tr. 139, at 5. ,

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19. Rigorous quality control procedures have been employed at Joseph Oat Corporation for decades. M. Oat's Quality Assurance Program has l been reviewed by the survey team of the American Society of Mechanical Engineers (ASME) at three year intervals since 1969, and the company has passed all of its ASME surveys. Singh on Contentions 3 and 6, ff. Tr. ]

t 139, at 5.

20. The material, austenitic stainless steel, used in the St. Lucie storage racks has seen decades of established use. Singh on Contentions 3 and 6, ff. Tr. 139, at 6. The designation of the stainless material is Type 304L. d .

J_d The welding material used in joining the structural '

i members has also been utilized for decades. M.

21. Neutron attenuation in the St. Lucie 1 spent fuel pool is accomplished through the combined action of borated water and the widely used neutron absorber material, Boraflex. Singh at Contentions 3 and 6, ff. Tr. 139, at 6. Since the early 1980's Boraflex has been the preferred ,

" poison" material for neutron absorption in spent fuel storage racks within the United States, as evidenced by the fact that over 85% of all such racks ordered by U.S. utilities since 1980 have incorporated  ;

Boraflex. M.at6-7.

22. The physical capabilities of the manufactured product are consistent with most engineered design requirements, and assured minimum Baron-10(B10) loadings varying up to a limit of 0.036 grams B10/cm 2/0.100" of thickness have been produced. Singh on Contentions 3 and 6, f f. Tr.139, at 8.

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23. Consistent with industry practice, the St. Lucie 1 rack modules are of two basic types, commonly referred to as Region 1 and Region 2.

M. All significant construction features of the St. Lucie 1 racks for Region 1 and Region 2 are direct adaptations of established technology, i

y . Region 1 modules are of the so-called " flux-trap" construction, which is an industry standard for modules of this type. M. In this I i

construction, square cross-section tubes are produced by seam welding two  !

identical channels. M . The seam welding equipment and process are examples of standard technology utilized in the manufacturing of Boraflex racks by the Joseph Oat Corporation. Singh Contentions 3 and 6, ff. Tr.

139, at 8.

24. Each Region 1 box (or " cell" or "can," as individual spaces for fuel assemblies within the storage racks are sometimes referred to) is equipped with a continuous sheet of Boraflex on each of its four sides.

Singh on Contentions 3 and 6, ff. Tr. 139, at 8. The Boraflex panel is positioned in place by stainless steel sheathing, which also serves to protect the Boraflex material from accidental dents. M. The stainless steel box surrounded by Boraflex on four sides is also a universally employed technology. M . The boxes are held in a vertical position and connected to each other by longitudinal connector channnels to produce a honeycomb construction. Singh on Contentions 3 and 6, ff. Tr. 139, at 9.

25. Region 2 rack modules are designed to store fuel with a specified minimum burnup. Singh on Contentions 3 and 6, ff. Tr. 139, at

9. These modules are constructed from the same basic elements as the Region 1 racks; basically, a solid baseplate, seam welded boxes, and continous sheets of Boraflex. M. A flux trap gap (water gap) in Region

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L 2 modules is not required and, therefore, is not provided. M. The L

Boraflex panels are positioned in place by the contiguous walls of the boxes, and suitably located peripheral strips. M. All of these rack

e. construction features are routinely used and represent complete j conformance with industry norms. Singh on Contentions 3 and 6, ff. Tr.

1 139, at 9.

, 5. Boraflex Testing Programs

26. The Staff and the Licensee have reviewed a substantial body of data from tests designed to evaluate the performance characteristics of Boraflex. Wing on Contention 3, ff. Tr. 110, at 2-5; Singh on Contentions 3 and 6, ff. Tr. 139, at 13-17. Tne results of these. tests demonstrate that Boraflex is suitable for use as a neutron absorber in spent fuel pool environments. Singh on Contentions 3 and 6, ff. Tr. 139, at 13; see also.

Wing on Contention 3, ff. Tr. 110, at 5-6. The environment of the spent fuel pool will not have a significant effect upon Boraflex, based on tests )

to examine long-term integrity. Testimony of Turner, transcript at 368

[hereinafterTurner,Tr.at ]; see also, Testimony of Edmond G.

Tourigny, transcript at 518 [ hereinafter Tourigny, Tr. at ]. 1

27. Prior to accepting Boraflex as a neutron abscrber material, the NRC required that material be tested under physical conditions which greatly exaggerated the severity of the environment to which the material would be exposed in actual use. Singh on 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 i

Shore A over a 6,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> exposure period. The 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 temperature 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 pressure 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 1 spent fuel pool, with boron concentra-tions of about 1720 ppm, hovers around 100*F for most of the time, much below the 240*F pressure bomb test. M. Moreover, Boraflex is never exposed to temperatures in excess of 200 F anywhere in a spent fuel pool.

M. Therefore, heat aging tests at 350 F, as noted above, were designed to simulate worse-than-possible scenarios. Singh on Contentions 3 and 6 ff. Tr. 139, at 14.

28. Radiation is the driving mechanism for changes in the physical properties of Boraflex. Turner on Contentions 3 and 6, ff. Tr.139, at

11. Radiation exposure tests of Boraflex at total equivalent doses of 12 over 10 rads were performed at the University of Michigan, Ford Nuclear Reactor, during 1979-1981 to identify the physical and chemical charac-teristics of Boraflex under a variety of radiation levels, radiation rates and severe environments. Singh on Contentions 3 and 6, ff. Tr.110, at

15. For example, water temperatures in the Ford Nuclear Reactor were substantially higher than the temperatures existing in the St. ucie 1 spent fuel pool. M.

l 29. The tests confirmed that no significant loss of boron occurs under irradiation to total radiation levels which are in excess of those expected through the expiration of the St. Lucie 1 operating license.

Turner on Contentions 3 and 6 ff. Tr.139, at 11. In some of these tests, Boraflex has been irradiated to accumulated doses in excess of 10 12 rads, of which 10 11 rads were attributable to ganrna radiation and the remainder was due to the concurrent neutron component. J_d . By comparison, Dr. Turner's calculations show that the Boraflex in the St.

Lucie 1 spent fuel storage racks, except for areas set aside for accelerated testing of Boraflex in the Surveillance Program, will be II exposed to radiation dose levels of less than 10 rads through the expiration of the St. Lucie 1 operating license. Turner on Contentions 3 and 6 ff. Tr.139, at 11.

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30. Neutron absorption by Boraflex was measured at various B (boron) loadings, confirming the neutron absorptive characteristics of the material. Singh on Contentions 3 and 6, ff. Tr. 139, at 15,

31. Irradiated samples were evaluated to determine the combined (syergistic) effects of irradiation, heat, pH, water, and boron on a number of physical and chemical characteristics of Boraflex. Singh on Contentions 3 and 6, ff. Tr.139, at 15; Wing on Contention 3, ff. Tr.

110, at 2-6. All evidence from the radiation tests suggest that, at the exposure levels expected in the spent fuel pool, Boraflex retains sufficient bend tolerance to withstand normal operating and refueling conditions in the spent fuel pool. Wing on Contention 3, ff. Tr.110, at 2-3. Evaluatien of the data reveals no discernible effect of either l environment or irradiation on neutron absorption. Singh on Contentions 3 and 6, ff. Tr. 139, at 15-16. The results showed no significant leaching of either boron or halogens and no significant amount of gas evolved from l

irradiated Boraflex samples. Wing on Contention 3, ff. Tr. 110, at 3.

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32. The results of the Boraflex tests demonstrated its suitability i for use as a neutron absorption material and led th3 NRC to license its  !

l use in spent fuel pools and in other neutron absorption applications since 1979. Singh on Contentions 3 and 6. ff. Tr. 139, at 16.

33. Most recently, Boraflex has been tested under carefully controlled conditions to detennine precisely its dimensional changes under irradiation. Singh on Contentions 3 and 6 ff. Tr. 139, at 16.

Laboratory data have been acquired at the University of Michigan's Ford Nuclear Reactor, and collated and analyzed. Id.; see also, Wing on Contention 3, ff. Tr. 110, at 3. Gamma radiation induces cross-lin. king of the polymer in Boraflex which leads to shrinkage. Wing on Contention 3, ff. Tr. 110, at 3. The saturation of cross-linking in Boraflex occurs at 10 the cumulative dose of approximately 10 rads, which is the point'at which Boraflex reaches maximum shrinkage. I_d . A December 1988 report published by the Electric Power Research Institute (EPRI) has indicated 10 shrinking stops as cross-linking saturates at about 1 x 10 rads. Turner on Contentions 3 and 6, ff. Tr. 139, at 14. The Michigan rm ults showed no significant increase in shrinkage of Boraflex at cumulative radiation 9 10 doses from 5 x 10 to 5 x 10 rads. Wing on Contention 3, ff. Tr. 110, at 3.

34. 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 on Contention 3, ff. Tr.110, at 3; see also, Turner on Contentions 3 and 6 ff. Tr. 139, at 14. Results published in the EPRI report point to a projected maximum Boraflex shrinkage of 3 to 4 percent. Turner, Tr. at

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j. l :216-17. Dr. Turner has measured. shrinkage on a substantial number of coupons and found between 2 to 21 percent of shrinkage. Turner, Tr. 216.

The EPRI report has very conservatively suggested for design considerations a shrinkage larger than that which has been observed at this point, pending the acquisition of additional data. Turner, Tr. 217.

In calculations performed for the St. Lucie 1 spent fuel pool, the conservative value of 4 percent shrinkage for Boraflex was used. Turner on Contentions 3 and 6, ff. Tr. 139, at 7.

35. Physical and chemical changes in Boraflex, for example, physical dimensions, hardness, specific gravity and tensile strength reach saturation levels at about 1 x 10 10 rads and no further changes in physical properties occur up to 1 x 10 11 rads gamma. Turner on Conten-tions 3 and 6, ff. Tr. 139, at 14-15. Radiation-induced scissioning and cross-linking transform Boraflex into a hard, ceramic-like material that has been found to be stable under further radiation, at least up to the radiation levels so far accumulated in the tests. Turner at Contentions 3 and 6, ff. Tr. 139, at 15.

36. The maximum cumulative radiation dose to the Boraflex material in the St. Lucie 1 pool for all fuel discharges until the end of the 11 licensed life of the unit is not expected to exceed 2 x 10 rads. Singh on Contentions 3 and 6, ff. Tr. 139, at 16. In fact, none of the rack materials, including the Boraflex, except in those areas intentionally set  ;

aside for special testing, is expected to exceed exposures at 3 x 10 10 rads. I_d. at 16-17. This is less than 3 percent of the equivalent radiation dose given to this material in the laboratory tests described ,

above. Id. Singh on Contentions 3 and 6, ff. Tr. 139 at 16-17.

1 3:. We find that Boraflex is a material with proven neutron absorption characteristics and that minor degradations will not significantly affect the neutron attenuation capability of Boraflex. We conclude that Boraflex is a satisfactory poison material and that it is suitable for use in the St. Lucie 1 spent fuel pool. In fact, the record amply demonstrated that the material has been subjected to testing environments more severe than environmental conditions which will be encountered in the St. Lucie 1 spent fuel pool, and still retained its )

neutron absorption capability and physical integrity.

6. In-Plant Performance of Boraflex l

38. The Staff and the Licensee have reviewed and evaluated the performance records of Boraflex used in the storage racks at four facilities. The Staff learned that gaps (separations) of up to 4 inches had formed in some Boraflex panels at the Quad Cities Nuclear Power Station, Units 1 and 2 in April of 1987. Wing on Contention 3, Tr. 110, at 3. The NRC Information Notice No. 87-43 " Gaps in Neutron Absorbing Material in High-Density Spent Fuel Storage Racks" (September 8, 1987) on page 1:

"[A] lerted recipients to a potentially significant problem pertaining to gaps identified in the neutron absorber component of the high density spent fuel storage racks at Quad Cities Unit 1 [a BWR facility). The safety concern ... [was] that certain gaps might excessively reduce the margin of nuclear subcriticality in the fuel pool."

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Singh on Contentions 3 and 6 ff. Tr. 139 at 9-10. The development of gaps up to 1.4 inches in size was identified at the Grand Gulf Nuclear Station, Unit in November of 1988. Wing on Contention 3, ff. Tr.110, at 3-4.

39. The Quad Cities racks, although built by Joseph Oat Corporation, utilized a different design and fabrication process than the St. Lucie 1 storage racks. Singh on Contentions 3 and 6, ff. Tr. 139, at 10. The Quad Cities racks employ the so-called " cruciform" construction, wherein angles are welded together along the edges in a fixture to form a ,

cruci form. Id. The Boraflex is contained between the faces of the angle, with the cruciform attached to each other by welding along their junction.

Id. This welding must be done remotely, and, as a result, its quality depends on the flatness and straightness of the cruciform surfaces. Id.

The Boraflex panels at Quad Cities and Grand Gulf were fabricated with an adhesive, Dow Silicone 999. Wing on Contention 3, ff. Tr. 139, at 4. The adhesive did not serve a mechanical or operational function in the rack design; it was employed to aid in handling and installing Boraflex during fabrication. Singh on Contentions 3 and 6, ff. Tr. 239, at 12.

40. The Staff and the Licensee have determined that excessive restraint of the Boraflex panels in conjunction with shrinkage of the Boraflex due to radiation, is responsible for the gap formation in the l panels examined at the Quad Cities and Grand Gulf plants. Wing on Contention 3, ff. Tr. 110, at 4; Singh on Contentions 3 and 6, ff. Tr.

139, at 10-11. There is no actuating mechanism for Boraflex panels to develop gaps there if there are no mechanical restraints on it or if the restraints are too small for it to overcome its tensile strength in the

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irradiated condition. Testimony of Krishna Singh, transcript' at 296

]; Testimony of James Wing, transcript at

~

[hereinafterSingh,Tr.at 541-45 [ hereinafter Wing Tr. at ].

41. The exact mechanism causing restraint and leading to gap information has not been confirmed; however, the two mechanisms believed to have a role have not been used in the design and construction of the St. Lucie 1 storage racks. Wing on Contention 3, ff. Tr.110, at 4; Singh on Contentions 3 and 6 ff. Tr. 139, at 11.

42. The Staff postulated that the adhesive used in the Quad Cities .

I and Grand Gulf storage racks is responsible for the restraint on Boraflex panels which led to cracking and gap formation. Wing on Contention 3, ff.

Tr. 110, at 4. . Adhesives were not used in the fabrication of the St. '

Lucie 1 storage racks. Id.; see also, Singh on Contentions 3 and 6 ff.

Tr. 139, at 12. The Licensee did not rule out and adhesives as a mechanism for restraint, but places less weight on it. Singh, Tr. at 326.

43. The Licensee postulated that the fabrication process used in the Quad Cities storage racks caused excessive restraint of the Boraflex panels. Singh on Contentions 3 and 6 ff. Tr. 139, at 10-11. Irradiation caused shrinkage of the Boraflex panels which led to cracking and gap formation on the restrained panels. Id. The cruciform rack design is i used only in storage racks fabricated for boiling water reactors (BWRs);

while pressurized water reactors, such as, St. Lucie 1, use a " box" module fabrication process. Id. at 11. The " box" type construction of storage racks has never developed the excessive restraint problem found in the cruciform design. Singh on Contentions 3 and 6, ff. Tr. 139, at 11.

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44. A full-length Boraflex panel was removed from the Point Beach Nuclear Plant, Units 1 and 2 spent fuel pool and examined for degradation.

Wing on Contention 3, ff. Tr. 110, at 4-5. The Point Beach rack design did not constrain the Boraflex. Singh on Contentions 3 and 6, ff. Tr.

139, at 12. An unconstrained Boraflex panel at Point Beach, which was removed for inspection after extensive exposure, showed no gaps or breakage. I_d . Moreover, the discoloration of the Boraflex samples at Point Beach and their absorption of spent fuel pool water were determined to have no effect on the neutron absorption capability of the Boraflex.

Id.; see also, Wing on Contention 3, ff. Tr. 110, at 5. Physical degradations of Boraflex, including thinning, erosion and breakage, were observed in Boraflex surveillance coupons at Point Beach Nuclear Plant, Units 1 and 2. Wing on Contention 3 ff. Tr.110, at 4. The breakage was apparently due to improper shipping and handling of the coupons. M. The physical degradation of the coupons is considered to be minor and would not reduce significantly the performance of Boraflex as a neutron absorber. Wing on Contention 3, ff. Tr.10, at 5.

45. The Boraflex panels used in St. Lucie 1 racks are continuous sheets, over 7 inches wide and 143 inches long, more than 6 inches longer than the active fuel length. Singh on Contentions 3 and 6 ff. Tr. 139, at

11. In Region 1, the stainless steel plates enveloping the Boraflex panels are spet-welded to the stainless steel can of the storage cells through cutouts in the Boraflex panel every twelve inches on both sides.

Id. On shrinking, the Boraflex panels may encounter these spot welds, and local stresses might appear along the axial length of the panels. M. In Region 2, the poison panels are engineered to be free to contract, and

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l therefore should not develop any gaps within the panels. M. In the St. J Lucie I rack fabrication, the poison panels were not compressed between the stainless steel sandwich sheets at all. Id. In contrast, the Boraflex panels were compressed in a vise-like jig in manufacturing the Quad Cities racks. Id. at 11-12. Also in contrast to what was done with the Quad Cities racks, an adhesive was not used in handling Boraflex during the St. Lucie 1 rack manufacturing process. M.at12. The adhesive used in manufacturing the Quad Cities racks was merely a convenient aid for handling and installing the Boraflex. I_d. It was not intended to, and does not, serve any mechanical or operational function.

M. Eliminating the adhesive from the St. Lucie 1 rack manufacturing process did not alter the basic fabrication methodology in any manner, but i further helped assure that the Boraficx is free to contract. Singh on Contentions 3 and 6, ff. Tr.139, at 12.

46. The experience gained with Boraflex in operating plants indicated that the dual effects of irradiation, namely shrinkage and hardening, are accommodated by appropriate design features. Singh on Contentions 3 and 6, ff. Tr.139, at 12. In St. Lucie I this accommodation has been effected by installing Boraflex panels large enough to accommodate shrinkage; and as noted earlier, the Boraflex panels in the l -

St. Lucie 1 storage racks are at least 6 inches longer than the active fuel length. Id. at 12-13. In addition, the application of adhesive during construction was eliminated, and provision was made in the manufacturing process for the installation of Boraflex with minimal surface loading. Singh on Contentions 3 and 6, ff. Tr. 139, at ?T. The Staff concluded that from a public health and safety perspective, tt l

authorization of the reracking of the spent fuel pool was-acceptable.

l Tourigny, Tr. at 519. Moreover, the Staff has not learned any information I since the amendment was granted in March 1988 which would change its  !

position on granting the amendment; the amendment should be granted.

Tourigny, Tr. 519-20.

47. We conclude that_the design and fabrication of the St. Lucie 1 racks incorporate proven technology for Boraflex installation and positioning. Id. at 13. The underlying causes of excessive restraint, which led to the Quad Cities Boraflex gap anomalies, cited in NRC Information Notice No. 87-43, and BN-87-11, do not exist in the St. Lucie 1 racks. I_d . The rack modules at St. Lucie 1 are based on refinements of established technology in the light of operating experience, and all aspects of their construction embodies proven design concepts and well-established fabrication techniques. Singh on Contentions 3 and 6, ff. Tr. 139 at 13.

7. Effects of Gaps in Boraflex on Pool Criticality

48. The Intervenor raised the issue that the presence of gaps in Boraflex panels in spent fuel pools would significantly decrease the l neutron attenuation capability of the panels and would increase keff values beyond NRC guidelines and industry standards. l l

49. The federal regulation governing the safe storage of reactor l l

fuel assemblies, from the perspective of criticality, is General Design i Criterion 62 of Appendix A to 10 C.F.R. Part 50, which states that:

" Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by the use of geometrically safe configurations." Turner on Contention 7, ff. Tr. 21, at 11.

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~

j l

50. The primary NRC guidelines for the acceptable implementation of General Design Criterion 62 are provided in NUREG-0800, Standard Review Plan, Section 9.1.2, " Spent Fuel Storage, which-states that the NRC Staff will accept storage racks if "the center-to-center spacing between fuel assemblies and any strong fixed neutron absorbers in the storage racks is sufficient to maintain the array, when fully loaded and flooded with nonborated water, in a subcritical condition. Ak eff n t greater than 0.95 for this condition is acceptable." Turner on Contention 7, ff.

Tr. 21, at 11.

51. ANSI N16.1-1975, " Nuclear Criticality Safety in Operations with i

Fissionable Materials outside Reactors," established the " double contingency principle". Turner on Contention 7, ff. Tr. 21, at 11-12.  ;

The principle is stated as follows:  ;

Process designs should, in general, incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.

ANSI N16.1-1975.

52. The NRC guidelines and industry standards which limit the maximum k eff to 0.95, including all uncertainties, provide a substantial sub-criticality 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 on Contention 7, ff. Tr. 21, at 13.

L 53. The physical integrity and neutron absorption properties of the l Boraflex poison material are important in assuring the criticality safety of the St. Lucie 1 spent fuel storage racks under all credible conditions.

Turner on Contentions 3 and 6, ff. Tr. 139, at 6. Gaps which might L-___-______-___-__

)

develop in the Boraflex could perturb the local reactivity and might l possibly have a significant effect on the system reactivity, depending upon the size and spatial distribution of the gaps. M. However, l numerous small gaps distributed randomly in size and location (as observed l at Quad Cities) 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 on Contentions 3 and 6, ff. Tr. 139, at 6.

54 In Region 1 the stainless steel plates enveloping the Boraflex panels are spot-welded to the stainless steel can of the storage cells through cutouts in the Boraflex panel every six inches; on shrinking, the Boraflex panels may encounter these spot welds, and local stresses might appear along the axial lengths of the panels. Turner on Contentions 3 and 6, ff. Tr. 139, at 6. The maximum shrinkage of Boraflex under long-term irradiation is conservatively bounded at 3 to 4%. M. Conservative calculations for the limiting condition of gaps in all Boraflex panels, coincidentally at the same axial elevations, result in a maximum keff f 0.771 under normal operating conditions, or 0.948 assuming the concurrent loss of all soluble boron. Turner on Contentions 3 and 6, ff. Tr. 139, at 7. 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 1 racks attributable to 4% shrinkage, the maximum k eff remains within acceptable bounds. M. The calculations were performed using the KENO-IV and CASM0-2E, which are bench marked against critical experimental data for configurations closely j

representative of actual rack geometry, as required by Regulatory Guide 3.41. Turner on Contention 7, ff. Tr. 21, at 13-14

55. In Region 2, gaps would significantly affect reactivity only if they were to coincidentally appear at the top few inches of the active fuel elevation in all Boraflex panels (because of the axial burnup distribution which causes the most reactive segment of the fuel assembly to occur at the top). Turner on Contentions 3 and 6, ff. Tr.139, at 7.

To assure that such gaps cannot form, the Boraflex panels in Region 2 are longer than required to provide the extra length sufficient to accommodate the expected shrinkage. M.at7-8. Therefore, shrinkage of Boraflex under irradiation will have no adverse reactivity consequences in Region 2, even with the concurrent accident conditions in which the loss of all  !

soluble boron is postulated. Turner on Contentions 3 and 6, ff. Tr. 139, at 8.

56. For normal operating conditions, the maximum keff is calculated l to be 0.760 and, for the design basis condition (single accident assumed to be the loss of all soluble boron), the maximum k,ff is 0.944. Turner on Contentions 3 and 6, ff. Tr. 139, at 8. The Region 2 calculations were performed assuming that fuel of the minimum allowable burnup had just been l

l removed from the reactor. M . However, during the period of storage in the rack, as the Boraflex is being irradiated, the storage rack reactivity is also continuously and substantially decreasing monotonically to less than 0.90 in 10 years of storage. J_d . This provides an additional safety '

margin to compensate for the reactivity effects of any gaps that might develop. Turner on Contentions 3 and 6, ff. Tr. 21, at 8.

l l

l l

l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ __ ]

l

. i

57. The NRC staff did not perform its own criticality analyses for l St. Lucie because that is not general practice in an NRC review of a i licensee's submittal for a spent fuel pool expansion. Testimony of Laurence I. Kopp, transcript at 495 [ hereinafter Kopp, Tr. at ]. The ,

Staff did not review the criticality analyses performed by Dr. Stanley E.

Turner, which assumed one-half inch gaps every 6 inches in the Boraflex in the St. Lucie racks, because these were not available at the time the Staff conducted its review. Kopp, Tr. at 530; see generally Tourigny, Tr.

at 507-14. However, a Staff member reviewed the written direct testimony and heard the oral testimony of Dr. Stanley E. Turner on these criticality '

analyses and concluded that there would be no adverse effect on k eff which would violate the acceptance criteria of 0.95. Kopp, Tr. at 535. Dr.

Kopp also reviewed generic criticality calculations and criticality calcu-lations performed by the Staff for the Turkey Point Plant Units 3 and 4 which verified these findings. Kopp, Tr. at 531-537. i

58. Although the various parameters such as U-235 enrichment, fuel assembly center-to-center spacing and B-10 areal density in these calculations are not identical to those in the St. Lucie spent fuel racks, they are similar enough so that one would not expect changes in reactivity due to postulated varying gap size to be significantly different. Kopp,

)

Tr. at 532. The base reactivity assuming no gaps would be dependent on i l

all of these parameters. Kopp, Tr. at 536. Changes in reactivity due to a given gap size, however, would not be dependent on these parameters.

Kopp, Tr. at 536.

59. An analysis of the reactivity consequences of a physically impossible condition, namely, the total loss of Boraflex in the storage

racks, was analyzed to demonstrate the very large reactivity margin in the St. Lucie 1 pool. Turner on Contentions 3 and 6, ff. Tr. 139, at 9. With consideration of the double contingency principle and with credit for the  !

soluble boron present, calculations for the hypothetical loss of all the Boraflex resulted in a maximum k,ff of 0.875 for Region 1 and 0.831 for Region 2, both of which are still well below the limit of K eff 0.95.

Turner on Contentions 3 and 6, ff. Tr. 139, at 9.

60. Prior to the issuance of the St. Lucie 1 spent fuel amendment, I the design basis k,ff limit was 0.95. Turner on Contention 7, ff. Tr. 21, at 13. The spent fuel pool expansion did not modify this design basis j limit. Turner on Contentions 3 and 6, ff. Tr. 139, at 8-9. Thus, the Licensee and the Staff have established on the record that the amendment l has not decreased the margin of safety for preventing a criticality accident at St. Lucie 1.

61. The central issue to be resolved is whether unsafe and unpredictable gap formation will develop in the St. Lucie 1 racks. The Staff did not identify a mechanism for gap development based on the FSAR submitted by the Licensee. Wing, Tr. at 543-47. The Licensee, however, f has postulated a possible mechanism for gap development in the St. Lucie 1 storage racks; that is, the use of a cut-out design in conjunction with 1

spot-welding every 12 inches along the sides of the Boraflex panels.

Singh, Tr. at 310-14. The Licensee has concluded that gaps may occur in a i

l systematic pattern. & The Licensee's conclusion is based on additional analyses which were not available to the NRC when the Staff reviewed the FSAR and prepared its SER and the written direct testimony. Tourigny Tr.

at 507-14 Using a conservative assumption that 1/2 inch gaps every 6 l \

i

c.

I racks, was analyzed to demonstrate the very 'large reactivity margin'in the L

St. Lucie 1 pool. Turner on Contentions 3 and 6, ff. Tr. 139, at 9. With consideration of the double contingency principle and with credit for the soluble boron present, calculations for the hypothetical loss of all the Boraflex resulted in a maximum k,ff of 0.875 for Region 1 and 0.831 for Region 2, both of which are still well below the limit of K eff 0.95. I Turner on Contentions 3 and 6, ff. Tr. 139, at 9.

60. Prior to the issuance of the St. Lucie 1 spent fuel amendment, the design basis k,ff limit was 0.95. Turner on Contention 7, ff. Tr. 21, at 13. The spent fuel pool expansion did not modify this design basis limit. Turner on Contentions 3 and 6, ff. Tr. 139, at 8-9.- Thus, the Licensee and the Staff have established on the record that the amendment has not decreased the margin of safety for preventing a criticality accident at St. Lucie 1.

61. The central issue to be resolved is whether unsafe and l unpredictable gap formation will develop in the St. Lucie 1 racks. The Staff did not identify a mechanism for gap development based on the FSAR l

submitted by the Licensee. Wing, Tr. at 543-47. The Licensee, however, l

has postulated a possible mechanism for gap development in the St. Lucie 1 storage racks; that is, the use of a cut-out design in conjunction with  ;

spot-welding every 12 inches along the sides of the Boraflex ' panels.

Singh, Tr. ot 310-14. The Licensee has concluded that gaps may occur in a systematic pattern. Id. The Licensee's conclusion is based on additior,a1 analyses which were not available to the NRC when the Staff reviewed the FSAR and prepared its SER and the written direct testimony. Tourigny Tr.

at 507-14 Using a conservative assumption that 1/2 inch gaps every 6

inches, the Licensee established that the degree of gapping would not effect reactivity in the spent fuel pool. Turner on 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 rear'ivity of the pool. Kopp, Tr. at 534-36; Tourigny, Tr. at 499-504, 507-14, 540-48; Wing, Tr. at 544-45. The Staff concluded that should the maximum projected gap formation, of 1/2 inch gaps every 6 inches, occur there will be no criticality concern. Kopp, Tr. at 534-36; Tourigny, Tr. at 540-48; Wing, Tr. at 544-45.

62. The Intervenor raised the issues that the Boraflex racks at St.

Lucie I spent fuel pool 1) use a significantly modified design and are l

essentially the result of new technology and fabrication process; 2) that the design is unproven and untested; 3) that gap formation problems reported with in-service Boraflex panels at other plants are unresolved; and 4) that gap formation, the separation of neutron absorbing material, may compromise safety. The Licensee established that both the design and the materials used to fabricate the St. Lucie 1 spent fuel pool racks are based on established technology which has been tested. The Licensee also established that the reported incidents of gapping have been resolved and that gapping will not compromise safety.

8. Licensee's In-Service Surveillance Program.

63. Boraflex is a satisfactory neutron absorber, capable of performing its intended function of criticality control. Turner on Contentions 3 ano 6, ff. Tr. 139, at 15. The Licensee and the Staff

\

expect Boraflex to perform satisfactorily, however, an in-service ,

surveillance program has been instituted by the Licensee. Tourigny on Contentions 3, ff. Tr. 110, at 7. Although accelerated testing has been an accepted and necessary methodology in the nuclear and other industries, the use of long-term and synergistic effects cannot be completely simulated. Turner on Contentions 3 and 6, ff. Tr. 139, at 15. Therefore, an in-service surveillance testing program will be conducted at St. Lucie

1. Testimony of Edward J. Weinkam, III on Contentions 3 and 6, following i

transcript 139, at 3 [ hereinafter Weinkam on Contentions 3 and 6. ff. Tr.

139, at ].

64. The Licensee and the Staff presented direct testimony on the in-service surveillance program which was unopposed by the Intervenor.

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.

65. The Licensee's testing and in-service surveillance program for the Boraflex in the St. Lucie 1 spent fuel pool storage racks is embodied in the St. Lucie plant procedures. Weinkam on Contentions 3 and 6, ff.

Tr. 139, at 3. The surveillance program is designed to verify the physical characteristics and neutron absorbing properties of the Boraflex used in Region 1 and 2 of the racks. Id. at 3.

66. Boraflex sample coupons are placed in both Regions 1 and 2.

Weinkam on Contentions 3 and 6, ff. Tr.139, at 3. The St. Lucie surveillance program uses large (approximately 15-inch long) coupons which are mounted in stainless steel jackets, representative of actual rack materials and configurations. Turner on Contention 3 and 6, ff. Tr. 139,

t

\

l l

at 15. Two separate trains (or trees) of coupons in each region of the 4

spent fuel pool are used; one train is exposed to accelerated radiations l levels and the other is exposed to normal radiation levels. Ld.at15-16.

Coupon samples from each of the trains will be removed on a pre-determined schedule, for timely detection and confirmation of the onset of any l k

degradation. M.at16. These coupons will be examined for the following properties:

1. Visual examination intended to reveal any surface  !

or excessive edge deterioration that might appear and to provide supporting information to assist in interpreting any degradation suggested by other measurements.

2. Dimensional measurements to provide a continuing measure of Boraflex shrinkage. The length measurement is of particular importance as an f indicator of the potential for gap formation in excess of that accommodated in the design.  !

3. Neutron attenuation measurements will be made for i establishing area density to confirm that boron is not being lost from the Boraflex. Although previous irradiation tests indicate that boron is retained, l this is perhaps the single most important measure of the ability of Boraflex to continue to serve its intended function.

4. Neutron radiography provides supporting information on neutron attenuation and is intended to reveal any

_ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ - - b

l non-uniformities in the boron distribution within l the Boraflex that might not be uncovered in the attenuation measurements. l S. Shore A hardness measurements will be performed on a continuing basis. Although the Boraflex is expected to become fully hard in the first few cycles of irradiation, continued measurement is intended to uncover any softening or friability as an indicator or excessive degradation.

6. Weight and specific gravity measurements are supporting measurements intended to reveal any significant loss of Boraflex material or the development of more open porosity than expected.

Turner on Contentions 3 and 6, ff. Tr. 139, at 16-17,

67. A number of corrective options exist for assuring continued safe fuel storage if Boraflex degradation problems were to occur. Weinkam on Contentions 3 and 6, ff. Tr.139, at 5. These include the following:

(1) The degraded Boraflex could be evaluated to determine whether the degradation would adversely affect FPL's ability to satisfy the 0.95 K,ff limit for the St. Lucie spent fuel pool. If the pool could still satisfy this limit, no further action would be necessary.

(2) 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.

"o (3) 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.

(4) The storage cells with degraded Boraflex could be blocked off to prevent their loading with fuel  !

l assemblies.  !

Weinkam on Contentions 3 and 6, ff. Tr.139, at 5-6.

68. The Staff concluded that the Licensee's in-service Boraflex surveillance program,. coupled with the materials design requirements, ensures both the safe use of Boraflex in the spent fuel pool and the protection of the public health and safety. Tourigny on Contention 3, ff.  ;

-Tr. 110,-at 8. Furthermore, the Staff concluded that blackness testing.  !

will be ineffective in detecting gaps in Boraflex which are i inch or less in size. Tourigny, Tr. at 552. No blackness testing is needed at the St.  !

1. Lucie 1 spent fuel pool, because the maximum predicted gaps of f inch  !

(using the Licensee's conservative estimate) would not be detectable by blackness testing. Tourigny, Tr. 552. For these reasons the Board will not require that the Licensee perform blackness testing as a license l condition.

l L _ - ---_- -- -- -

(.

s, B. Contention 7

1. Contention 7 asserts: I "That the increase of the spent fuel pool capacity, which includes fuel rods that are more highly l i

enriched, will cause the requirements of '

ANSI-N16-1975 not to be met and will increase the probability that a criticality accident will occur in the spent fuel pool and will exceed 10 C.F.R. Part 50, Appendix A, General Design Criteria 62."

69. We agreed with the Licensee that the St. Lucie spent fuel racks conform to safe and conventional practice within the industry and conform to Commission regulations and guidance. S.D. Order of October 14, 1988 at

36. And we found no credible criticality accident condition had been postulated by the Intervenor's contention or its bases. M. Our order required that the Licensee provide an explanation for two issues: 1) the measures used to prevent erroneous fuel assembly insertion in the storage racks; 2) why 002 fuel enriched to 4.5 weight percent U-235 cannot achieve criticality in the absence of a moderator. M.

70. The Intervenor presented no witnesses or testimony on Contention l

. 7 at the hearing and attempted to make his case by cross-examination. l

2. Principles Underlying Criticality Analysis

71. The term " fissile material" refers to material the atoms of which are capable of being split or fissioned with the attendant production of large quantities of heat energy (the useful product from the reactor) upon the capture (absorption) of neutrons. Turner on Contention l

)

l 7, ff. Tr. 21, at 5-6. The priniary fissile material in new fuel assemblies of most nuclear power reactors, including St. Lucie 1, is a nuclide of uranium called uranium-235. Id. at 6. In natural uranium, the ,

uranium-235 nuclide is present at a concentration less than 1 percent by weight, with almost all of the remainder being the uranium-238 nuclide.

M. To be useful in a light-water nuclear power reactor, natural uranium is enriched in uranium-235. M. The nuclear fuel utilized at St. Lucie 1 may be enriched up to 4.5 percent by weight of uranium-235, with almost all of the remaining 95.5 percent being the uranium-238 nuclide. Turner on Contention 7, ff. Tr. 21, at 6.

72. In general, when a neutron is absorbed by uranium-235, there is a high probability that uranium-235 will undergo fission, resulting in the release of energy, fission products and more neutrons. Turner on Contentions 7, ff. Tr. 21, at 6. These neutrons, in turn, can (1) be absorbed by uranium-235 or other fissile nuclides, (2) be absorbed by uranium-238 nuclides, resulting in virtually no additional fission, (3) be absorbed non-productively by non-fissile materials called " poisons" (resulting in no additional fission), or (4) escape without being absorbed (i.e., leakage, which also results in no additional fission). . Turner on Contention 7, ff. Tr. 21, at 6.

73. As a practical matter, not all neutrons released as a result of fission will cause additional fissions. Turner on Contention 7, ff. Tr.

21, at 6-7. Uranium-238 nuclides, poison materials and leakage inhibit the fission process by reducing the number of neutrons available to cause firsions. Id. at 7. If fewer neutrons are being produced as a result of fission than are leaking and being absorbed, the fission process will not

i 1

l sustain itself; this condition is called "subcriticality." M. In j contrast, if the rate of neutron production as a result of the fission process is equal to the rate of neutron absorption and leakage, the I

fission process will sustain itself, and the condition is referred to as

" critical." Turner on Contention 7, ff. Tr. 21, at 7.

74. The term " effective multiplication factor" is defined as the ratio of the number of neutrons per unit of time produced in the fission process, to the number of neutrons per unit of time absorbed and escaping.

Turner on Contention 7, ff. Tr. 21, at 7. The effective multiplication factor, commonly called k-effective (or keff), is a measure of the ability of a system to sustain a fission reaction. Id. Criticality occurs whenever the effective multiplication factor reaches or exceeds a value of  !

i 1.0 because at least as many neutrons are being produced as are being lost  !

by absorption and leakage. M. For a keff less than 1.0, the fission rate cannot be sustained. M. The margin below a k,ff of 1.0 is the safety margin to criticality, and this subcritical margin is the difference between a k eff of 1.0 and the k,ff of a given system. M.

NRC guidelines for fuel storage racks require that the maximum keff' including all known uncertainties, be equal to or less than 0.95, a value which provides a substantial suberitical margin. Turner on Contention 7,  ;

i

- ff. Tr. 21, at 7-8.

75. One of the most important characteristics of radioactive elements is the decrease in radiation intensity with time as individual )

i nuclides decay. Turner on Contention 7, ff. Tr. 21, at 8. In spent fuel assemblies, the intensity of radiation - and the inventory of radioactive nuclides - begins to decline from the moment the reactor shuts down (e.g.

for refueling). Id. The total inventory of radioactive elements will 1 4

decrease rapidly in the time period following reactor shutdown due to decay of elements with shorter half-lives, ld. For example, for spent  !

fuel placed in the pool 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown, the radioactive inventory will decrease by more than 50 percent in 30 days. Ld. It is this fact of radioactive decay in aged fuel that explains why reracking of a spent fuel storage pool will not result in a proportional increase in radiation intensity or heat generation rates, but will, in fact, result j only in minor increases as aged fuel accumulates in the pool. Turner on Contentions 7, ff. Tr. 21, at 8-9. j

76. During operation of a nuclear power plant, the fission process j within the fuel produces large quantities of heat (a desired product), and new atomic species called fission products (an undesired product). Turner on Contention 7, ff. Tr. 21 at 9. Those fission products which absorb neutrons non-productively are often referred to as poisons, meaning that they inhibit the fission process by reducing the reactivity (k,ff) of the fuel. Id. Most of the fission products are stable and remain in the fuel as neutron absorbing poisons. M. However, a significant fraction of the fission products are unstable, producing the radioactivity commonly associated with spent fuel. [d . Radioactive decay occurs as the unstable atoms emit radiation in a spontaneous transformation to a stable element of the same or different chemical species, often with a greater neutron absorbing (or poisoning) effect. M . In general, the radioactive decay of the radioactive nuclides formed in the fuel during reactor operation results in a substantial and continuous decrease in k eff during the period I

o-_ -

l' j

of time the fuel elements are in storage. Turner on Contention 7, ff. Tr. l 21, at 9-10.

77. The reactivity (k,ff) of fuel assemblies is affected primarily by two factors: (1) the quantity and enrichment of the uranium-235 in the fuel, and (2)'the quantity of neutron-absorbing materials (poisons) present. Turner on Contention 7, ff. Tr. 21, at 10. Increasing the fuel enrichment increases the reactivity of the fuel, as does increasing the density of fuel assemblies in the spent fuel pool or decreasing the concentration of poisons. M . For this reason, Region 1 of the St. Lucie 1 racks is designed to safely store fuel assemblies of the highest enrichment permitted on site. M. Conversely, the reactivity of fuel elements in a spent fuel pool can be decreased by decreasing the enrich- l ment of uranium-235 in stored fuel assemblies; by decreasing the density of the stored fuel assemblies; or by increasing the concentration of poisons. Turner on Contention 7, ff. Tr. 21, at 10.

78. In this latter regard, neutron absorbing poisons may be intentionally installed in the storage racks to reduce the system's reactivity. Turner on Contention 7, ff. Tr. 21, at 10. In the St. Lucie 1 racks, the Boraflex is used as a poison material, and the soluble boron in the pool water acts as additional poison material. M. During  !

operation in the reactor core, uranium-235 is consumed (depleted) in the fission process and the effective fuel enrichment is decreased, resulting l in reduced reactivity. M. In addition, fission products accumulating in the fuel further reduce the reactivity. M. These effects enable Region 2 of the St. Lucie 1 racks to be designed with less added poison material (Boraflex) than Region 1. Turner on Contention 7, ff. Tr. 21, at 10-11.

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3. Prevention of Erroneous Fuel Assembly Insertion

79. The spent fuel racks at St. Lucie 1 are divided into two regions. Weinkam on Contention 7, ff Tr. 139, at 4. The NRC's Technical Specifi-cation (T.S.) 5.6.1.b specifies in which of the two regions fuel assemblies must be placed. Tourigny on Contention 7, ff. Tr. 110, at 13.

Region 1 can be used to store any new or spent fuel. M. Region 2 can be used to store only new or spent fuel meeting the requirements of the

" Initial Enrichment vs. Burnup Requirements for Storage of Fuel Assemblies in Region 2", curve in T.S. 5.6.1.b, Figure 5.6.-l. Id. The storage of fuel in Region 2 is restricted to specific burnup levels depending on initial enrichment and, in effect requires that almost all new fuel be stored in Region 1. Tourigny on Contention 7, ff. Tr. 110, at 13,

80. The Standard Review Plan requires the licensee to develop and employ a system which prevents improper fuel assembly insertion through the use of administrative controls, or physical restraints, or by a combination of both. SRP 9.1.2 " Spent Fuel Storage," NUREG-0800.

Tourigny on Contention 7, ff. Tr. 110, at 12.

81. The Licensee's fuel-handling procedures ensure that fuel assemblies placed in the St. Lucie 1 spent fuel racks are inserted only at intended rack positions, and are appropriate to the initial enrichment and burnup conditions of the fuel. Weinkam on Contention 7, ff. Tr. 21, at 3.

Each fuel assembly arrives at St. Lucie 1 with a unique serial number which is engraved on it. M. The serial number remains visible regard-less of storage location within the pool to facilitate identification.

M. The Licensee tracks the location of a fuel assembly throughout its life by its serial number. Weinkam on Contention 7, ff. Tr. 21, at 3.

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C l j L

l

82. Fuel is moved to, and inserted into, a spent fuel rack cell location with a. spent fuel pool machine which consists of a rolling. bridge j which spans the pool, and a fuel lifting device. Weinkam on Contention 7, ff. Tr. 21, at 4 The fuel lifting device may be positioned by a spent i

' fuel machine operator over any rack cell location in Regions 1 or 2. M.

at 4-5. Each cell location within the racks is identified by a region-unique index system, which uses a grid for Region 1 and another for Region 2. M.at5. Fuel assemblies are tracked within the pool by maintain.q records of their serial numbers on maps indicating the cell locations and associated alpha-numeric index codes where the assemblies are located. Weinkam on Contention 7, f f. Tr. 2",. at 5. Location of new ,

and burned fuel assemblies, stored in the spent fuel pool racks, are tracked by se v 1 numbers which are reported in fuel status report records and spent fuel pool fuel locations maps. g.at5-6. The transfer of assemblies to predetermined locations is conducted by an NRC-licensed operator under the direction of the licensed Control Room operator. M.

at 6-7. The proper storage spent fuel in Region 1 or Region 2 is based on burnup of the fuel, g.at4.

83. Following refueling, an independent verification (by a remotely controlled camera) of the location of the fuel assemblies in the reactor core and the spent fuel pool is conducted, and fuel status records are j updated to reflect any assembly location changes. g.at7. In addition, I an audit of the spent and new fuel in storage must be completed at least annually in accordance with 10 C.F.R. Part 75. Weinkam on Contention 7, j ff. Tr. 21, at 7-8.

i

84. The Board concludes that the Licensee's fuel-handling procedures i ensure the fuel assemblies will only be inserted into pool storage rack cell locations that have been specifically assigned to them. The procedures meet the guidelines of the Standard Review Plan and will ensure against improper storage of fuel assemblies.

4. Criticality and Moderation

85. A moderator is any material composed of light elements sufficient to scatter and decrease the velocity of neutrons without absorbing many of the neutrons. Turner, Tr. at 60. The only moderator used in the St. Lucie 1 spent fuel pools is water. Id.

86. The Intervenor attempted to establish that air, wood, concrete, and zirconium cladding are also effective moderators which have the ability to moderate neutrons in the pool and, therefore, affect the criticality and fission process. Turner, Tr. at 60-62; Kopp, Tr. at 116-119. The Intervenor was not successful because the Licensee and Staff witnesses established that while theoretically the materials may moderate a reletively small number of neutrons in practice the moderation by these materials would be negligible and insignificant. Turner, Tr. at 62; Kopp, Tr. at 116-119,

87. To explain why criticality is impossible with fuel enriched up

- to 4.5% in the absence of a moderator, it is necessary to outline important facts regarding criticality and the fission process, as follows:

(1) The fission process involves the absorption of neutrons in the uranium-235 nuclide (or other fissible nuclides) with the resultant release of energy and additional high energy or " fast" neutrons.

(2) The fissioning process of the uranium-235 nuclide is caused principally by low-energy or " thermal" neutrons. The probability of higher energy neutrons being absorbed in the uranium-235 nuclide and causing fission is very small.

(3) In fuel of 4.5% enrichment, the remainder of the uranium is primarily (95.5%) the uranium-238 nuclide, a strong neutron absorber especially for the higher energy neutrons.

(4) Criticality occurs whenever the effective multiplication factor (K,ff), defined as the ratio of the number of neutrons per unit-time produced in the fission process to the number of neutrons per unit time being absorbed or escaping, reaches or exceeds a value of 1.0.

Turner on Contention 7, ff. Tr. 21, at 19-20.

88. In the absence of a moderator to slow down the fast neutrons released in the fission process, the fraction of neutrons absorbed in uranium-235 is small compared to the fraction absorbed in uranium-238.

1 Turner on Contention 7, ff. Tr. 21, at 20. The net effect is a very low value for K,ff, of the order of 0.5 or less, even for an infinite array of fuel (few neutrons absorbed in uranium-235 to produce more neutrons and a large. fraction of the neutrons non-productivity absorbed in uranium-238).

Id. Furthermore, the absence of a moderator in an actual system of finite size greatly increases the leakage of the high energy neutrons, resulting 1

in an even greater reduction in K,ff. Turner on Contention 7, ff. Tr. 21, at 20.

l. 89. Figure 17 of TID-7028, USAEC, 1964 illustrates that uranium l

metal enriched to 5 weight percent U-235 or less cannot be made critical l

l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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i in a dry system. Testimony of Laurence I. Kopp on Contention 7, following l

l transcript 110, at 14 [ hereinafter Kopp on Contention 7, ff. Tr. 110, at l

_]. Since U0 2 fuel is more diluted than pure uranium metal, the necessary enrichment of U0 2

.for criticality with no moderator would be j even higher than 5 weight percent U-235. I_d . The USAEC document reports that a minimum enrichment of 5.6% is required to achieve criticality in a dry system. Turner on Contention 7, ff. Tr. 21, at 20.

90. One of the long established and well accepted principles of criticality safety is that of " moderator control," whereby the control (i.e., exclusion) of moderating material provides complete assurance that criticality cannot occur in systems of low-enriched uranium. Turner on Contention 7, ff. Tr. 21, at 20-21. This principle assures that fresh fuel assemblies may be safely handled and stored in air (i.e., dry l condition, without moderator), knowing that criticality is not possible without substantial moderating material present. M.at21.

91. In order for criticality to occur in the absence of moderating I material, it would be necessary to enrich the uranium to a much higher level than that of the St. Lucie 1 fuel, or to use other fissile materials that do not contain significant quantities of neutron absorbers, such as the uranium-238 present in St. Lucie 1 fuel. Turner on Contention 7, ff.

Tr. 21, at 21. The 1947 criticality incident at Los Alamos, in which Dr.

Slotin was fatally injured, involved experiments with highly enriched plutonium in a form capable of attaining " dry" criticality, but which has no relation to the low-enriched St. Lucie 1 uranium fuel. M. Similarly, atomic weapons utilize plutonium or possibly highly enriched uranium with little extraneous absorber material. g.;seealso, Turner,Tr.at66-69.

r With the low-enriched uranium used in the St. Lucie 1 fuel or with an array of St. Lucie 1 fuel assemblies, criticality is impossible in the absence of a moderator, which is water in the case of St. Lucie 1. Turner on Contention 7, ff. Tr. 21, at 22.

92. The Intervenor sought to establish that there were circumstances under which spent fuel could go critical. The Licensee and Staff witnesses refuted this suggestion. Spent fuel in a pool cannot form a critical mass or reach criticality, even if it melted and slumped into the bottom of a dry pool. Turner, Tr. at 78. The assumption that a spent  !

fuel pool would lose all its water and that the fuel could meet and slump is beyond the St. Lucie 1 design basis. Kopp, Tr. at 123. Furthermore, without a moderation, fuel in any configuration, even with zirconium or other impurities in it, will not reach criticality. Id. at 123-124 l

l III. CONCLUSIONS OF LAW Based upon the foregoing findings of fact and upon consideration of the evidentiary record compiled in this proceeding we reach the following  ;

conclusions of law:

1. Intervenor's Contentions 3, 6 and 7 are without merit.

2. The design of the spent fuel storage racks for the St. Lucie 1 facility, together with the surveillance program proposed by the Licensee i

and approved by the Staff, provide reasonable assurance that the public l health and safety will not be endangered.

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9 IV. ORDER In accordance with the Atomic Energy Act of 1954, as amended, the Rules of Practice of the Commission, and based on the foregoing findings of fact and conclusions of law, IT IS ORDERED that License Amendment No.

91 to License No. DPR-67, issued by the NRC Office of Nuclear Reactor <

Regulation on March 11, 1988, authorizing the expansion of the St. Lucie 1 spent fuel pool, shall remain in full force and effect without modification.

IT IS FURTHER ORDERED, that in decordance with 10 C.F.R. $$ 2.760, 2.762, 2.764, 2.785, and 2.786, as amended, this initial Decision shall become effective immediately and will constitute, with respect to the matters resolved herein, the final decision of the Commission thirty (30) days after issuance hereof, subject to review pursuant to the above-cited Rules of Practice. Any party may take an appeal from this Initial Decision by filing a notice of Appeal within ten (10) days after service of this Decision. Each appellant must file a brief supporting its position on appeal within thirty (30) days after filing its Notice of Appeal (forty (40) days if the Staff is the appellant). Within thirty (30) days after the period has expired for the filing and service of the briefs of all appellants (forty (40) days in case of the Staff), a party who is not an appellant may file a brief in support of, or in opposition to, any such appeal (s).

l Respectfully submitted, e

behH '

Patricia A. Jehle Counsel for NRC Staff Dated at Rockville, Maryland this 17th day of March, 1989.

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APPENDIX A j l

' LICENSEE'S WITNESS LIST Dr. Krishna P. Singh, Contentions 3 and 6 Dr. Stanley E.' Turner, Contentions 3, 6 and 7 Edward J. Weinkam, III, Contentions 3, 6 and 7

[The applicant's prefiled written direct testimony appears in the record ff. Tr. 139]

STAFF'S WITNESS' LIST Dr. Laurence I. Kopp, Contentions 3 and 7 Edmond G. Tourigny, Contentions 3, 6 and 7 Dr. James Wing, tantention 3

[The staff's prefiled written direct testimony appears in the record ff.

Tr.110]

I

4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ce N . 50-335-0LA f FLORIDA POWER AND LIGHT OMPANY (SFP Expansion)

(St. Lucie Plant, Unit No. 1) )

NRC STAFF PROPOSED TRANSCRIPT CORRECTIONS The NRC Staff proposes the following transcript corrections in the above-captioned proceeding:

Date Page Line Proposed Correction January 24, 1989 109 6 Substitute the word " Wing" for the word " Wade".

114 2 Strike the words "I believe" where they first appear.

114 3 Strike the first four words on this line.  !

1 120 16 Substitute the word " losing" for the word "using".

124 20 Substitute the word "has" for the word "is" and substitute the words " cross section" for the word

" process".

128 23 Delete the "U" between the word " plutonium" and the numbers "239"..

Date- .Page Line ' Proposed Correction 129 10 Substitute.the word "they" for the word "we".

129 16 Delete the second "this" on this line and the word."a" between the word "be" and the word " pure".

Janua ry 26, 1989 Index Between The word "Edmont"'should lines 5 be "Edmond",

and 6 430 11 Substitute the word "one" for the word "on".

437 23 Delete the "?" at the end of this line.

439 15 Insert the word "the" between the word "to" and the word

" eleventh".

439 22 Substitute the word

" composition" for the Word " Conversion".

442 13 Substitute the word "at" for the word "for".

442 22 Substitute the word "were" for the word "was"..

453 18 Add the word "to" at the end of this line after the word "have".

455 18 Delete the "?" at the end

- of this line after the word

" Cities".

January 26, 1989 Index Between The word "Edmont" should lines 3 be "Edmond".

L and 4 1

' .3 -

Date. Page Line Proposed Correction 467 18 Substitute the word.

" cutout" for'the word

" cutoff" 478 4 Add the letters "un"'

on the word " irradiated" to create the word "unirradiated".

483 11 Substitute the word "scissioning".for the word " seasoning".-

483 12 Substitute the word "scissioning" for the word " seasoning".

489 18 Substitute the word

" plate" for the word

" late" 499 6 Delete the word "two" between the words "Those" and " members".

500 5 Substitute the word "we" for the word "we've".

520 5 Delete the letter "Q" at the beginning of the line and connect the text on this line to the text on line 4 536 9 Substitute the word l "show" for the word "so".  ;

544 25 Substitute the word  ;

" produce" for the words

^

" deal with".

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Date Page Line Proposed Correction 548 9 Substitute the words  !

"on that" for the words I "of what" at the end of  !'

the line.

551 13 Insert the word "it" between the words "that" and "has" at the beginning of this line. .

J Respectfully submitted, I

f% k ernard M. Bordenick Counsel for NRC Staff Dated at Rockville, fiaryland this i 7 +h day of tlarch,1989.

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L_________._.

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. 4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

,89 NAR 20 P4 :07 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD l

uf f !t? r 5:

Docxi e ;, . , nu In the Matter of iH% H Docket No. 50-335-OLA FLORIDA POWER AND LIGHT i l

COMPANY (SFP Expansion)

(St. Lucie Plant, Unit No. 1) l CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM 0F AN INITIAL DECISION" 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, or as indicated by a double asterisk by use of express mail service this 17th day of March,1989, or as indicated by a triple asterisk by hand-delivery, this 20th day of March, 1989:

B. Paul Cotter, Jr. , Chairman *** Glenn 0. Bright ***

Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555  !

Richard F. Cole *** Michael A. Bauser, Esq.***

Administrative Judge Harold F. Reis, Esq.

Atomic Safety and Licensing Board Newman & Holtzinger, P.C.

U.S. Nuclear Regulatory Commission 1615 L Street, N.W.

Washington, D.C. 20555 Washington, D.C. 20036 Atomic Safety and Licensing Docketing and Service Section*

BoardPanel(1)* Office of the Secretary U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 Washington, D.C. 20555 Atomic Safety and Licensing Campbell Rich **

Appeal Panel (5)* 4626 S.E. Pilot Avenue U.S. Nuclear Regulatory Commission Stuart, Florida 34997 Washington, D.C. 20555 l

[i; ,

Adjudicatory File

• Richard J. Goddard

• Atomic' Safety and Licensing Board U.S. Nuclear Regulatory Comission U.S. Nuclear Regulatory Comission Regional Administrator, Region II Washington, D.C. 20555 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 (J4:2th < - j Patricia A. Jehle Counsel for NRC Staff i

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