ML20235V197

Licensee Proposed Findings of Fact & Conclusions of Law.*

ML20235V197
Person / Time
Site:	Saint Lucie
Issue date:	02/25/1989
From:	Bauser M, Buttler J FLORIDA POWER & LIGHT CO., NEWMAN & HOLTZINGER, STEEL, HECTOR & DAVIS
To:
Shared Package
ML20235V201	List: ML20235V197 ML20235V209
References
CON-#189-8205 88-560-01-LA, 88-560-1-LA, OLA, NUDOCS 8903100076
Download: ML20235V197 (58)

Text

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i.0 i t .M U UNITED STATES OF AMERICA 89 1%R -1 P 4.:43 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD h[cS] -

February 25, 1989

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In the Matter of ) Docket No. '50-335 OLA

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FLORIDA POWER AND LIGHT COMPANY ) (ASLBP No. 88-560-01 LA)

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

)

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LICENSEE'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW Co-Counsel Harold F. Reis Joha T. Butler Michael A. Bauser Patricia A. Comella Steel, Hector & Davis Newman & Holtzinger, P.C.

4100 Southeast 1615 L Street, N.W. ,

Financial Center Washington, D.C. 20036 t Miami, Florida 33131-2398 l f

(202) 955-6600 (305) 577-2939 Attorneys for Licensee Florida Power & Light Company

$$f*000N 000 P

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

February- 25, 1989 l )

In the Matter of ) Docket No. 50-335 OLA

) l FLORIDA POWER AND LIGHT COMPANY ) (ASLBP No. 88-560-01-LA) i

)

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

)

LICENSEE'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW

)

I. Introduction and Background

1. On June 12, 1987, Florida Power and Light Company (hereinafter "FPL" or " Licensee") requested a license amendment to 1

allow for expansion of spent fuel pool storage capacity at its St.

Lucie Plant,' Unit No. 1 ("St. Lucie 1"). 1/ The amendment sought i to increase the capacity of the spent fuel pool at St. Lucie 1 from 728 to 1706 fuel assemblies.

2. On August 31, 1987, the U. S. Nuclear Regulatory Commission (hereinafter "NRC" or " Commission") published a notice referencing FPL's application for an amendment, noting that the Commission had made a proposed finding of no significant hazards consideration, and offering the opportunity for a public hearing. >

l On September 30, 1987, Mr. Campbell 52' Fed. Reg. 32,852 (1987).

1/ See Letter from C.O. Woody (FPL) to U.S. Nuclear Regulatory Commission (St. Lucie 1, Docket No. 50-335) (June 12, 1987).

1 a

Rich (hereinafter "Intervenor") addressed a letter to the Secretary of the Nuclear Regulatory Commission requesting that a public hearing be held concerning the spent fuel pool expansion amendment. 2/ In responsive pleadings filed November 4 and 9, 1

l 1987, both the NRC Staff and FPL pointed out that the letter failed to meet the requirements of 10 C.F.R. $ 2.714 and that, Thereafter, pursuant to I therefore, the request should be denied.

a Board Memorandum and Order dated November 13, 1987, Intervenor submitted on January 15, 1988, a " Request for Hearing and Petition for Leave to Intervene" (hereinafter " Amended Petition").

The Amended Petition contained sixteen contentions which Intervenor proposed be admitted in this proceeding.

3. Following completion of its review, the NRC Staff determined that the requested amendment involved no significant hazards consideration, and issued Amendment No. 91 to Facility Operating License No. DPR-67 on March 11, 1988, accompanied by a Safety Evaluation (hereinafter "SE"). 3/ Thereafter, on April 20, 1988, and following a prehearing conference held on March 29, 1988, we issued a Memorandum and Order granting Mr. Rich's Amended i Petition to intervene and admitting Intervenor Contentions 3, 4, 6, 8, 9, 11 and 15, renumbered, respectively, as Admitted 2/ See Letter from C. Rich to Secretary, U.S. Nuclear Regulatory Commission (Sept. 30, 1987).

3/ The SE was received into evidence as Staff Exhibit 1. A list of the exhibits is provided in Appendix A.

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Contentions 1 through 7 (hereinafter " Contention "). Florida l Power and Light Co. (St. Lucie Plant, Unit No. 1), LBP-88-10A, 27 i NRC 452 (1988). The Atomic Safety and Licensing Appeal Board subsequently affirmed that decision. Florida Power and Light Co.

(St. Lucie Plant, Unit No. 1), ALAB-893, 27 NRC 627 (1988).

4. We deferred ruling on Intervenor's contention 5, pending review by Mr. Rich of certain additional information  :

supplied by the NRC Staff. Subsequently, however, the contention I was dismissed, in a Memorandum and Order dated May 31, 1988, because the Intervenor failed to indicate that he intended to pursue the contention. See also LBP-88-10A, 27 NRC at 461-62.

5. On June 24, 1988, the Intervenor requested that he be permitted to withdraw Contention 2, stating:

It has come to my attention that the temporary crane that was installed in the spent fuel pool storage area to facilitate the reracking procedure has been removed from the storage area. In light of this, I would ask that the Board withdraw from contention, admitted Contention 2, which concerns the damages that might have existed from the presence of the crane. Obviously, these concerns are no longer sensible.

Upon consideration of the Intervenor's request, which was unopposed, we issued an Order on July 27, 1988, dismissing l Contention 2 "with prejudice as moot."

6. In a Memorandum and Order issued on October 18, 1988, we partially granted the August 5, 1988, " Licensee's Motion l

l for Summary Disposition of Intervenor's Contentions."

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Specifically, we found that no genuine-issue of material-fact existed and'that Licensee was entitled to judgment as a matter of law with respect to Contentions 1, 4 and 5. Licensee's Motion was also granted as to portions of Contentions 3 and 7, but denied as to Contention 6.- As a result of our rulings, the contentions remaining for hearing may be stated as follows:

• Contention 3 - The possible materials degradation and failure that mightLoccur in Boraflex panels due to heat and radioactivity generated in the spent fuel pool have not been adequately considered or analyzed.

• Contention 6 - The proposed use of Boraflex in the high density spent fuel storage racks-designed and '

fabricated by the Joseph Oat Corporation is essentially a new and unproven technology.

• Contention 7 a) The mechanisms which prevent the-erroneous insertion of a fuel assembly into a storage cell such that the prescription of Standard Review Plan (SRP) Section 9.1.2, Part III.2.b, that it not be possible for "a fuel assembly . . . (to] be inserted'anywhere other than a design location,"1has not been' demonstrated; and b) It has not been shown why criticality will not occur in the spent fuel' pool in the absence of a moderator.

7. Hearings were held in Stuart, Florida over three days, beginning on January 24, 1989. The record was closed on January 26, 1989, and a schedule set for the filing of proposed findings of fact and conclusions of law by all parties.

8. This decision is based upon the record in this proceeding, including that developed during the hearings held on

January 24-26, 1989. We have accorded appropriate weight to the testimony of the witnesses based on their knowledge, skill and experience. We will now set forth our resolution of each of the 1 matters. remaining at issue in this proceeding. Any proposed finding of' fact or conclusion of law submitted by the parties not ~ f incorporated into this decision is rejected as being unsupported in law or fact or as unnecessary to this decision.

II. Analysis A. Contentions 3 and 6

9. In our Memorandum and Order (Ruling on Motions for Summary Disposition), dated October 14, 1988, (hereinafter "SD Order")', we granted summary disposition of Contention 3 with respect'to all of the non-Boraflex issues, but denied summary disposition as to the Boraflex issues. SD Order at 38. We denied Licensee's motion for summary disposition of Contention 6, pertaining to Boraflex, in toto. Id. at 39. As a result of our rulings, and as indicated earlier, the Boraflex contentions which remained for hearing may be stated as follows:

• Contention 3 - The possible materials degradation i and failure that might occur in Boraflex panels due to heat and radioactivity generated in the spent fuel pool have not been adequately considered or analyzed.

• Contention 6 - The proposed use of Boraflex in the high density spent fuel storage racks designed and t fabricated by the Joseph Oat Corporation is essentially a new and unproven technology. i

3 Because the issues raised by these contentions deal exclusively with the use of Boraflex as a neutron. absorber in high density spent fuel storage racks, and largely overlap,.the contentions will be considered together.

(1). Use of Boraflex in the'St. Lucie 1 spent fuel racks 1 10. Neutron attenuation in the St. Lucie 1 spent fuel racks is accomplished through the combined action of borated water and a widely-used neutron absorber material, Boraflex.

Commonly referred to as a neutron " poison," Boraflex is an i

j effective entrapper of neutrons. (Testimony of Dr. Krishna P. .j Singh.on Contentions 3 & 6, following Transcript 139, at 6-7 (hereinafter "Singh on Contentions 3 & 6, ff. Tr. 139, at

").)

11. Boraflex is produced by uniformly dispersing boron carbide particles in a methylated polysiloxane elastomer, which l serves as the matrix element. (Singh on Contentions 3 & 6, ff. l Tr.'139, at 7.) The polymeric silicone encapsulant entrains and  ;

fixes fine particles of boron carbide in a homogeneous, stable j matrix. (Id.) The function of Boraflex is to provide a boron-curtain between fuel assemblies. (Singh, Tr. 196.) The thickness of the Boraflex used in a particular application is determined by the loading -- the required Boron-10 (or "B-10")

areal density. (Singh, Tr. 182.) If a greater loading per square inch is needed, the thickness of the Boraflex sheet is j

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increased, leaving the ratio of boron carbide to elastomer unchanged. The performance of the elastomer is, thus, unaffected. (See Singh, Tr. 182-85.)

l 12. The St. Lucie 1 spent fuel racks are of a two-region design, utilizing a different B-10 loading in each region, as discussed in Section II.B.(2), infra. Basic principles of the fission process necessary for understanding the criticality and reactivity considerations associated with the St. Lucie 1 racks are discussed in Section II.B.(1), infra. These include the concept of " reactivity" and the " effective multiplication factor"

("k-effective" or "k-eff").

13. The primary concerns of the Intervenor in this proceeding were that: (a) not enough was known about the effects of heat and radioactivity in a spent fuel pool environment to allow the use of Boraflex as a neutron absorber; and (b) the storage racks themselves, including their utilization of Boraflex, were of a new and unproven design. We would note at the outset that the term "new and unproven technology" is somewhat imprecise and subjective. When we admitted Contention 6, we understood its gravamen to be that there was insufficient

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information available -- as a result of both testing and in-service experience -- to support a finding that the utilization of Boraflex as a neutron absorber in the St. Lucie 1 spent fuel storage racks provides reasonable assurance that the health and safety of the public will not be endangered. See, e.g., Northern

State Power Co. (Prairie Island Nuclear Generating Plant, Units 1 and 2), ALAB-455, 7 NRC 41, 44 (1978), remanded on other grounds, Minnesota v. NRC, 602 F.2d 412 (D.C. Cir. 1979); 10 C.F.R. S 50.40(a) (1988). We have proceeded accordingly.

14. In considering whether the license amendment granted by the NRC Staff may remain in effect, we must determine, 1

for each of the factual issues remaining in dispute, whether the preponderance of the evidence supports the Licensee's position.

See Pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), ALAB-763, 19 NRC 571, 577 (1984), review J declined, CLI-84-14, 20 NRC 285 (1984). On the basis of the record before us, we conclude that it does.

(2) Test and in-service experience with Boraflex as a neutron absorber

15. The Boraflex in the St. Lucie 1 racks is subjected to heat and radiation. The water in the St. Lucie 1 spent fuel pool, containing a concentration of about 1720 ppm soluble boron, is generally maintained at approximately 100* F. (Singh on Contentions 3 & 6, ff. Tr. 139, at 14.) The Boraflex is never exposed to temperatures in excess of 200' F anywhere in the St.

Lucie 1 spent fuel pool. (Id.) l

16. Except in certain areas set aside for accelerated testing, (discussed in Section II.A.(7), infra), the Boraflex in the St. Lucie 1 storage racks will be exposed to radiation doses of less than 10 11 rads through the expiration of the St. Lucie 1 i l

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operating license. (Testimony of Dr. Stanley E.' Turner on.

Contention 3 & 6, following Transcript 139, at 11, Figure 1 at 21 (hereinafter " Turner on Contentions 3 & 6, ff. Tr. 139, at

").) 4/. In the areas. set aside for accelerated testing, Boraflex will receive the equivalent of 20 years' normal exposure in about 3 to 4 years, and can accumulate up to 2 x 10 11 rads over the anticipated lifetime of the racks. (Turner on Contentions 3 & 6, ff. Tr. 139, at 12.)

17. Before approving the use of Boraflex as a neutron absorber in spent fuel storage racks, the NRC Staff required testing 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 & 6, ff. Tr. 139, at 14.) The testing included heat aging, long-term exposure to borated water and irradiation. (Id.) The testing program was comprehensive and the results sufficient for the NRC Staff to approve the use of Boraflex in spent fuel pools and other neutron absorption applications, beginning in 1979. (Singh on Contentions 3 & 6, ff. Tr. 139, at 16.)

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18. In the heat aging tests, Boraflex was exposed to temperatures up to 350' 'r', simulating worse-than possible scenarios with respect to temperatures to be encountered in the 4/ Figgre 1 shows the(Turner accumulated dose to be less than 3.5 x 10 rads gamma. on Contentions 3 & 6, ff. Tr.

139, Figure 1 at 21.)

St. Lucie 1 spent fuel pool. (Singh on Contentions 3 & 6, ff.

l Tr. 139, at 14.) The effects of long-term exposure to high l

temperature borated water were also evaluated to determine the stability of Boraflex under aggravated environmental conditions.

(Id.) Boraflex was subjected to a boric acid solution (3,000 ppm boron), in a pressure bomb-type test vessel. A constant l temperature of 240' F was maintained for an exposure period of over 6,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. (Id.) The data showed Boraflex to be stable.

(Id.)

19. Extensive radiation tests on Boraflex have been conducted at the University of Michigan's Ford Nuclear Reactor, first in 1979-1981 as part of the qualification of Boraflex for us,e as a neutron absorber in spent fuel racks; and, subsequently, following the discovery of gaps in Boraflex used in the Quad Cities spent fuel storage racks, to determine the dimensional changes of Boraflex under controlled conditions. (See Singh on Contentions 3 & 6, ff. Tr. 139, at 15-16.; Turner on Contentions 3& 6, ff. Tr. 139, at 10.) Holtec International performed tests and analyses of Boraflex following the discovery of the Quad Cities gaps. (Singh on Contentions 3 & 6, ff. Tr. 139, at 16.)

20. The 1979-1981 tests exposed Boraflex to total 11 rads gamma.)

equivalent doses of over 10 12 rads (including 10 (Singh on Contentions 3 & 6, ff. Tr. 139, at 15.) This test program was designed to identify the physical and chemical

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characteristics of Boraflex under a variety of radiation levels and severe environments. (Id.) For example, the water tempera-tures'in the Ford Nuclear Reactor were'substantially greater than the temperatures existing in the St. Lucie l' spent fuel pool.  ;

1 (Id.) Samples were also irradiated in dry air; in solutions of delonized water, some with soluble boron, some without; and in a .

spent fuel pool. (Wing, Tr. 437, 442-43.) 5/

21. Neutron absorption by Boraflex was measured at various B-10 loadings, and confirmed the' neutron absorptive characteristics of the material at neutron energies represen-tative of neutrons which could cause fission. (Singh on Contentions 3 & 6, ff. Tr. 139, at 15.) The test results show that after exposure to 1.03 x 10 11 rads of gamma radiation in an environment of borated water, Boraflex maintains its neutron attenuation capability. (Testimony of James Wing on Contention 3, following Transcript 110, at 2 (hereinafter " Wing on Contention 3, ff. Tr. 110, at ").)

22. Irradiated samples were evaluated for the effects of radiation on a number of physical and chemical character-istics. (Singh on Contentions 3 & 6, ff. Tr. 139, at 15.)

Evaluation of test data revealed no discernible effect of either environment or irradiation on neutron absorption. (See, e.g.

5/ The results of these tests will thus reflect any synergistic effects from both heat and radiation, even though the results are not specifically reported as such. (See Wing, Tr. 548.)

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Singh on Content. ions 3 & 6, ff. Tr. 139, at 15-16; Singh, Tr.

344.)

2 3. - Subsequent to the discovery of gaps at Quad cities, Boraflex was tested'under carefully controlled conditions to determine its dimensional changes under. irradiation. (See Singh on Contentions 3 & 6, ff. Tr. 139, at 16; Turner.on Contentions 3 & 6, ff. Tr. 139, at.2, 10.) Laboratory data were acquired at the University of Michigan's Ford Nuclear Reactor, collated and analyzed. (Singh on Contentions 3 & 6, ff. Tr. 139, at 16.) Dr. Turner performed, or had performed under his direction, measurements on irradiated samples of Boraflex, from both tests and actual surveillance coupons from reactor spent fuel pools, which had undergone significant, concurrent exposure to both radiation and typical spent fuel pool environments.

(Turner on Contentions 3 & 6, ff. Tr. 139, at 2; Singh on Contentions 3 & 6, ff. Tr. 139, at 16.) y/

24. The experimental irradiation programs have shown that,-upon irradiation, Boraflex undergoes shrinkage and becomes a hard ceramic-like material, with increased compressive strength and reduced ductility. (Turner on contentions 3 & 6, ff.

Tr. 139, at 10; see Wing, Tr. 478.) These observations have been confirmed in the measurements on irradiated surveillance coupons removed from spent fuel pools. (Turner on Contentions 3 & 6, ff.

y/ The results of Dr. Turner's study are reported in Exhibit 9.

Tr. 139, at 10.) However, irradiation, alone, does not cause the Boraflex to crack, or fracture, or gap. (Singh, Tr. 195, 197.) 7/

25. Shrinkage results from cross-linking of the polymer in Boraflex, which is induced by gamma radiation. (See Wing on Contention 3, ff. Tr. 110, at 3.) As the accumulated radiation dose increases, cross-linking becomes saturated and no further shrinkage occurs. (Id.) The NRC Staff estimated that saturation of cross-linking in Boraflex occurs at a cumulative i dose of about 10 10 rads, the point, therefore, at which Boraflex attains maximum shrinkage. (Id.) This estimate is supported by

.the University of Michigan tests, which showed'no significant difference in Boraflex shrinkage at cumulative radiation doses from 5 x 10 9 to 1 x 10 10 rads. (Id.) In addition, physical and chemical changes in Boraflex (as measured by physical dimensions, hardness, specific gravity and tensile strength) also reach i saturation levels at about 1 x 10 10 rads. (Turner on Contentions 3 & 6, ff. Tr. 139, at 14-15.) No further changes in 11 rads gamma. (Id.)

physical properties occur up to 1 x 10 After. irradiation to levels at which saturation occurs, Boraflex has been found to be stable under further irradiation, at least up to the radiation levels so far accumulated in the tests, which 7/ We discuss mechanisms for gap formation in Section II.A.(3),

infra.

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"" are well beyond the levels expected at St. Lucie 1. (Turner on Contentions 3 & 6, ff. Tr. 139, at 15.)

26. Measurements on Boraflex coupons irradiated in spent fuel pools revealed the maximum shrinkage to be 2 to 2-1/2 percent. (Turner on contentions 3 & 6, ff. Tr. 139, at 6-7; see Wing on Contention 3, ff. Tr. 110, at 3.) 8/ EPRI's report projects the maximum shrinkage to be 3 to 4 percent, based on considerations other than observations. (Turner, Tr. 217-18, 357-58.) 9/

8/ There were uncertainties in the various measurements because of coupon size. The coupons used in the study were very small. (Singh, Tr. 385.) Edge' deterioration effects were encountered, making it difficult to obtain accurate measurements. (See Turner, Tr. 386; Singh,.Tr. 386.) In its report (Exhibit 1; see Appendix A, infra), entitled "An Assessment of Boraflex Performance in Spent-Nuclear-Fuel Storage Racks," EPRI NP-6159 (December 1988), the Electric Power Research Institute (hereinafter "EPRI") discarded those samples showing edge deterioration effects as being of no value in determining shrinkage. (See Turner, Tr. 401.)

9/ In criticality calculations examining the reactivity effects of postulated gaps in the Boraflex used in the St.

Lucie 1 spent fuel racks, Dr. Turner assigned a conservative value of 4 percent shrinkage. The results of these calculations are described in Section II.A.(5), infra.

Except for the differences in maximum projected shrinkage (which are based on different considerations), the results reported in Dr. Turner's study (Exhibit 9) are in general agreement with the results reported by EPRI. (Turner on Contentions 3 & 6, ff. Tr. 139, at 14.) Some of the principal results and recommendations of EPRI's report are the following:

1. Shrinking stggsrads.

as cross-linking saturates at j about'1 x 10 (continued...)

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27. Based on the tests that have been performed, the environment in the spent fuel pool should not have any  ;

I significant effect on the Boraflex. (Turner, Tr. 368.)

Radiation produces the dominant effect. (Id.) Except for gap i

formation, changes in Boraflex structure have significance only to the extent that they cause or result in the loss of boron and, hence, reduce the effectiveness of the material in controlling reactivity. (Turner on Contentions 3 & 6,'ff. Tr. 139, at 11.)

Tests have confirmed that no significant loss of boron occurs under irradiation to levels in excess of those expected through the expiration of the St. Lucie 1 operating license. (Id.)

28. Not only has Boraflex been tested extensively under appropriate conditions to determine its effectiveness as a neutron absorber, there is also considerable in-service experience regarding its effectiveness. Since the early 1980's Boraflex has been the preferred " poison" material for neutron 9/(... continued)

2. Maximum shrinkage is about 3 to 4 percent.

3. No loss of boron has been observed.

4. Shrinkage and gap problems can be accommodated in design. Newer designs should anticipate and accommodate irradiation changes.

5. Boraflex appears to be a satisfactory poison material.

6. Meaningful surveillance programs are appropriate to ensure long-term performance.

(Turner on Contention ~s 3 & 6, ff. Tr. 139, at 14.)

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• absorption in spent. fuel storage racks within the United States, as evidenced by the fact that over 85 percent of all such racks ordered by U.S. utilities since 1980 have utilized Boraflex.

(Singh on Contentions 3 & 6, ff. Tr. 139, at 7, Table B at 19.)

Industry's experience with Boraflex has been satisfactory. The record before us reveals no in-service experience which has shown Boraflex to be unacceptable for use as a neutron absorber in spent fuel storage racks.

29. As discussed in the next section, Boraflex gaps discovered in spent fuel racks have no safety significance.

(3) Occurrence of gaps in Boraflex following in-service irradiation

30. As we indicated in our October 14 Memorandum and Order, gaps have been found in the Boraflex absorber materials used in the Quad Cities Plant's high density spent fuel storage racks. SD Order at 21. More recently, gaps have been found in the Boraflex panels used in high-density racks for the Grand Gulf Nuclear Station spent fuel pool. (See Wing on Contention 3, ff.

Tr. 110, at 3.)

31. As we discussed in the immediately preceding section, experience with Boraflex has shown that it shrinks upon irradiation. The shrinkage, itself, does not cause gaps to form in the Boraflex, as experience with a panel of Boraflex from the Point Beach Nuclear Power Plant has demonstrated. (See Singh, I s

Tr. 197, 296-97.) However, both the NRC Staff and Licensee's

)

1 witnesses postulate that gaps may occur if the Boraflex is restrained mechanically in too rigid a fashion, thus preventing free contraction when the Boraflex shrinks upon irradiation.

(See Turner on Contentions 3 & 6, ff. Tr. 139, at 10; Singh, Tr.

197, 296-97; Wing, Tr. 453; Wing on Contention 3, ff. Tr. 110, at 4.) The NRC Staff maintained that it did not.have sufficient information to determine conclusively what caused the gap formation in the Boraflex used in the Quad Cities and Grand Gulf racks, but believed that the adhesive, Dow Silicone 999, used in fabricating the racks could have physically restrained the Boraflex, inducing the gap formation in conjunction with the shrinkage due to irradiation. (See Wing on Contention 3, ff.

Tr. 110, at 4.) Dr. Singh believed that it was more likely that the fabrication process itself -- rather than the adhesive, which was a fabrication aid -- led to restraint of the Boraflex panels and their subsequent gap formation following shrinkage. (See Singh on Contentions 3 & 6, ff. Tr. 139, at 10-12.) In particular, Dr. Singh believed that the welding process used in constructing the Quad Cities cruciforms was a direct cause of the restraint. (See Singh on Contentions 3 & 6, ff. Tr. 139, at 10; Singh, Tr. 218-20.) 10/

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The same method of manufacture was used for the Grand Gulf racks. (Turner, Tr. 346.) The St. Lucie 1 racks are constructed differently. We discuss their construction, in Section II.A.(4), infra.

32. Gaps of up to 4 inches have been'found in some Boraflex panels at the Quad Cities Nuclear Power Station, Units 1 ]

d These and 2. (Wing on Contention 3, ff. Tr. 110, at 3.)

Boraflex panels received an estimated gamma radiation dose of 10 9 rads. (Wing on Contention 3, ff. Tr. 110, at 3.)- _ Gaps of up to 1.4 inches were found in Ruraflex p.inels at the_ Grand Gulf Nuclear Station. (Id.) The gaps did not occur at the same plane at either Quad Cities or Grand Gulf, but were distributed, with some panels showing multiple gaps. (Turner, Tr. 367-68.)

33. The gaps:in the Boraflex panels in the Quad Cities and Grand Gulf racks have been evaluated for their impact

-on reactivity. Criticality calculations have been performed for each of the racks. This was necessary because gaps which might develop in the Boraflex panels could perturb the local reactivity and might possibly have a significant effect on the system reactivity, depending on the size and spatial distribution of the gaps. (See Turner on Contentions 3 & 6, ff. Tr. 139, at 6.)

Under conservative assumptions, which assumed larger gaps than those observed, the NRC safety margin (k-eff not in excess of 0.95) was not exceeded. (See Turner on Contentions 3 & 6, ff.

Tr. 139, at 5-6; Turner, Tr. 350-52, 170-71.)

34. Other anomalies have also been found in Boraflex materials. Physical degradation, including thinning, erosion, and breakage, were observed in Boraflex surveillance coupons at the Point Beach Nuclear Power Plant, Units 1 and 2. (Wing on 1

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Contention 3, ff. Tr. 110, at 4.) Wisconsin Electric Power Company, the Point Beach licensee, was of the opinion that improper shipping and handling of the small coupons caused the observed breakage. (Wing on Contention 3, ff. Tr. 110, at 4-5.)

The Prairie Island Nuclear Generating Plant, Units 1 and 2, experienced physical degradation in a Boraflex coupon, similar to that found in an irradiated Point Beach panel removed from the spent fuel pool, which included discoloration and embrittlement along the edges. (Wing on Contention 3, ff. Tr. 110, at 5.) The NRC Staff determined that these minor degradations would not affect the neutron attenuation capability of Boraflex panels.

(Wing on Contention 3, ff. Tr. 110, at 5.) Blistering has also been reported, probably as a local defect in manufacturing, but the blistering occurs only as a local, not bulk, swelling.

(Singh, Tr. 195.) Thinning, change of color and open porosity also may occur; however, the ability of the irradiated Boraflex to maintain a uniform boron carbide dispersion remains intact.

(See Singh on Contentions 3 & 6, ff. Tr. 139, at 7; Singh, Tr.

195-98.)

(4) Potential for gap formation in the St. Lucie 1 spent fuel racks i

35. The St. Lucie 1 storage racks have been designed to accommodate safely any Boraflex shrinkage which might occur, without the formation of excessive gaps. Conservative analyses have been performed demonstrating that any gaps that might form

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in the Boraflex panels in the St. Lucie 1 spent fuel racks would not increase k-eff above the limiting value of 0.95 (discussed in Section II.A.(5), infra); and that, in the unlikely event of unexpectedly large gaps, or even the complete loss of Boraflex, the presence of soluble boron in the St. Lucie 1 spent fuel pool ensures that k-eff will be maintained well below the limit of 0.95 (discussed in Section II.A.(6), infra).

36. Two approaches have been used in the design of the St. Lucie 1 spent fuel racks which will limit the fornation 1

of gaps in the racks. One design approach -- used for Region 2 -- ensures that gaps will not develop because the Boraflex is 1

allowed to shrink freely along its entire length. (See Singh,  !

Tr. 296-97.) The other design approach -- used for Region 1 --

i permits unrestrained shrinkage up to a point. It then operates j to assure that any gaps that might develop as a result of further shrinkage are small and occur at defined, distributed locations.

(See Singh, Tr. 211-12, 312.) Thus, any reactivity effects would be very small. 11/ Moreover, as a precaution following the discovery of gaps at Quad Cities, the use of adhesive was eliminated from the St. Lucie 1 rack fabrication process. (See Singh, Tr. 212-13; 413-14; Singh on Contentions 3 & 6, ff. Tr.

139, at 13.) Also, the Boraflex panels used in the St. Lucie 1 racks are about 6 percent longer than the active fuel length of 11/ Numerous small gaps have only a minor effect on system reactivity. (See Turner, Tr. 313.)

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- 136 inches, thus providing a margin of safety for expected shrinkage. (See Singh, Tr. 210, 420; Turner on Contentions 3 &

6, ff. Tr. 139, at 7.)

37. The essential characteristic of Region 1 rack construction is that there are two panels of Boraflex between two facing storage fuel assemblies, and between the two panels there is a gap of water, known as a flux trap. (Singh, Tr. 207.)

38. A flux trap is not required for Region 2 modules j and, therefore, is not provided. (Singh on contentions 3 & 6,  !

ff. Tr. 139, at 9.) Instead, Region 2 racks employ the picture frame design and construction, which is a standard industry approach to constructing storage racks. (See Singh, Tr. 210-11.)

Storage cells are connected to each other at the edges, and spacer strips are placed around the corners to form a picture frame in which the Boraflex is placed. (Id.) Each strip is approximately twice the thickness of the Boraflex in order to provide free contraction of the Boraflex should it shrink under .

irradiation. (Singh, Tr. 210-11, 213, 320.) Because the Boraflex panels in Region 2 are engineered to be free to contract, gaps should not develop within the Region 2 panels.

(Singh, Tr. 296-97, 320; Singh on Contentions 3 & 6 ff. Tr. 139, at 11.) In the absence of any physical restraint, the Staff does not anticipate separation or gap formation in the Boraflex panels. (Wing on Contention 3, ff. Tr. 110, at 4.)

39. In the flux-trap construction used for Region 1, square cross-section tubes are produced by seam welding two l identical channels. (Singh on Contentions 3 & 6, ff. Tr. 139, at 8.) 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 i each of its four sides. (Id.) The Boraflex panels are affixed  !

in position by stainless steel sheathing, which also serves to protect the Boraflex material from accidental dents. (Id.) The boxes are held in a vertical position and connected to each other by longitudinal connector channels to produce a honeycomb construction. (Singh on Contentions 3 & 6, ff. Tr. 139, at 9.)

$' 40. In Region 1, the stainless steel plates enveloping the Boraflex panels are spot-welded to the stainless steel can of the storage cells through the cutouts in the Boraflex panels every six inches. (Singh on Contentions 3 & 6, ff. Tr. 139, at 11.) The cutout is a scallop, semicircular in shape. (Singh, Tr. 305-06; Weinkam, Tr. 309). The scallops are found along the length of the Boraflex at the edges on both sides. (See Singh, Tr. 308-10, 418.) At the cutouts the Boraflex can move by an amount defined by the clearance between t'h'e scalloped cutout and the spot-weld. (Singh, Tr. 310-11.) By

.having these spot welds and cutouts at repeated intervals along the length, there are well-defined locations where a panel may be allowed to separate -- form gaps -- due to shrinkage. (Singh, i

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . . _ _ _ _ . _ _ _ . _ (

Tr. 311.) The presence of the scallops guarantees that any gaps will form only at the locations of the cutouts because this is where the cracks which precede the gaps will initiate. (Singh, Tr. 311-12.) When the stress level at the location of the cutout exceeds the tensile strength of the material, then the gap will develop, thus making the stresses go to zero. (See Singh, Tr.

333.) The size of the gap will be proportional to the free length between the locations of the scallops. (Id.) The scallops occur every 12 inches on the same side of the panel, but the scallops on the two sides are staggered so that there is a cutout and spot-weld every 6 inches along the entire length of the panel considering both sides. (See id.; Singh, Tr. 418.)

Assuming a 4 percent shrinkage of the Boraflex panels, any gaps forming 12 inches apart would not exceed 1/2 inch; and any gaps forming 6 inches apart would not exceed 1/4 inch. (See Singh, Tr. 333-34; 363-64.) Under these conservative assumptions the required safety margins would be met, as we discuss in Section II.A.(5), infra.

41. Initially, the Boraflex in Region 1 racks can move freely. It is free to shrink until it encounters a spot-weld.

(See Singh, Tr. 310-11.) The Boraflex should encounter the spot-welds if it shrinks 2 percent. The resulting restraint could lead to cracking, followed by gapping, if any further shrinkage  !

occurs. (See Sintjh, Tr. 564-66.) l

_ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ 1

1

42. Within this context, we would note that the NRC Staff's SE 12/ does'not~contain-specific discussion of gap formation in the St. Lucie 1 spent fuel racks. In fact, the Staff has maintained throughout that gapping will not occur, in either Region 1 or Region 2, because the Boraflex panels are.not sufficiently restrained to result in gaps in the event of shrinkage. (See, e.g., Wing on Contention 3, ff. Tr. 110, at 4; Wing, Tr. 544-45.) This position is somewhat different than that presented by Dr. Singh, indicating that some gapping is possible; indeed, expected. (See, e.g., Singh', Tr. 415.) 13/ However, we need not decide between " gaps" or "no gaps." This is because, as will be discussed in detail in Section~II.A.(5), infra, the effect of gap formation has been fully evaluated. Conservative analysis shows that even with the gaps the Licensee's witnesses  !

' believed might form, at 4 percent shrinkage, the racks will meet the,NRC shutdown standard. There is no information in the record l providing any basis for assuming that greater than 4 percent shrinkage will or can occur. Accordingly, the precise matter of 1_2/ Staff Exhibit 1.

13/ Initially, it appeared that this difference in views might have been the result of some NRC Staff confusion over the details of inner-Region 1 cell design and fabrication.

(See, e.g., Wing, Tr. 456-62; Tourigny, Tr. 522-25.)

However, even after it was apparent that the Staff clearly understood all of the technicalities associated with Region 1 cell configuration, the Staff maintained its position that gapping would not occur because the Boraflex panels were not sufficiently restrained to result in gap formation'in the case of shrinkage. (See, e.g., Tourigny, Wing, Tr. 539-46).

l b__-- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ __ ._ J

i l

j whether or not gaps will form in the St. Lucie 1 racks is not consequential to protection of public health, safety and the environment.

(5) Effect on reactivity of gaps in the Boraflex panels in the St. Lucie 1 spent fuel racks

43. Conservative calculations were performed for Region 1 to obtain the maximum possible k-eff under an assumption of 4 percent maximum Boraflex shrinkage for the limiting condi-tion of 1/2 inch gaps in the Boraflex every 12 inches at the same height everywhere. (Turner on Contentions 3 & 6, ff.

Tr. 139, at 7, Table 1 at 19.) The calculations were based on the additional conservative assumption of an infinite number of fuel assemblies in the storage cells, all of infinite length.

(Turner on Contentions 3 & 6, ff. Tr. 139, at 7.) The calculations were performed using the methods' described in I Section II.B., infra. (See Turner on Contentions 3 & 6, ff. Tr.

139, at 7.) The calculations show a maximum k-eff of 0.771 under normal operating conditions. (Turner on Contentions 3 & 6, ff. i Tr. 139, at 7, Table 1 at 19.) Under the design basis condition of concurrent loss of all soluble boron in the St. Lucie 1 pool water, the calculations show a maximum k-eff of 0.948. (Id.)

Thus, for the limiting condition of gaps, attributable to 4 percent shrinkage, occurring in the Boraflex in the Region 1 l racks, the maximum k-eff remains within acceptable bounds.

(Turner on Contentions 3 & 6, ff, Tr. 139, at 7.)

44. In Region 2, gaps would significantly affect reactivity only if they were to simultaneously appear at the top few inches of the active fuel 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 & 6, ff. Tr. 139, at 7.) To ensure that such gaps cannot form, the Boraflex panels in Region 2 are fully free to contract and are longer than required to provide the extra length sufficient to accommodate the expected shrinkage. (Turner on Contentions 3 & 6, ff. Tr. 139, at 7-8; see-Singh, Tr. 296-97.)

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 & 6, ff. Tr. 139, at 8.) For normal operating conditions, the maximum k-eff is calculated to be 0.760 and for the design basis condition (single accident assumed to be the loss of all soluble boron), the maximum k-eff is 0.944. (Turner on Contentions 3 & 6, ff. Tr.

139, at 8, Table 1 at 19.) The Staff sees no criticality concerns because the Staff's acceptance criterion (k-eff no greater than 0.95) has not been violated. (Kopp, Tr. 535.)

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(6) The effect on reactivity assuming complete loss of Boraflex in the St. Lucie 1 spqnt fuel racks

45. In this section we consider the reactivity consequence of the hypothetical loss of all Boraflex.

Dr. Turner, who provided relevant analysis, believes that this condition is not credible because it is physically impossible.

However, he included the calculations to indicate the very large reactivity margin that is available to ensure the criticality safety of the St. Lucie 1 storage racks. (See Turner on Contentions 3 & 6, ff. Tr. 139, at'9.)

46. As discussed in Section II.B.(3) infra, the double contingency principle of the American National Standard Institute (or " ANSI") standard, ANSI N16.1-1975, endorsed and invoked by the definitive NRC letter of April 14, 1978, provides that operations with fissionable materials outside reactors are acceptable if at least two unlikely, independent and concurrent accidents are necessary before a criticality incident would be possible. (See Turner on contentions 3 & 6, ff. Tr. 139, at 9.)

In invoking the double contingency principle, the NRC letter specifically provides that credit for soluble boron is permis-sible under accident conditions. (See id.) With consideration of the double contingency principle and credit for the soluble boron present, calculations for the hypothetical loss of all the Boraflex resulted in a maximum k-eff of 0.875 for Region 1 and 0.831 for Region 2, both of which are still well below the limit

of 0.95. (Turner on Contentions 3 &'6, ff.'Tr. 139, at 9, Table 2 at 20.)

(7)- The in-service surveillance program for St. Lucie 1

47. While it is not expected that.the aqueous pool-d environment will have an effect on the performance of.the Boraflex, it is not'possible to guarantee 30' years' performance at'this point in time. (Turner, Tr. 370.) Ultimate proof of the long-term suitability of Boraflex for its entire residence in the pool requires actual, long-term operation. (Turner, Tr. 372.')

Al'though accelerated testing programs have~been accepted 1 throughout the nuclear industry, slow, long-term and synergistic effects can never be completely simulated. (Turner on Contentions 3'& 6, ff. Tr. 139, at 15.) Therefore, an in-service j surveillance program will be conducted at St. Lucie 1 to monitor the integrity and performance of Boraflex on a continuing basis.

(Id.) 14/

48. The FPL testing and in-service surveillance program for the Boraflex in the St. Lucie 1 spent fuel pool storage racks is embodied in St. Lucie plant procedures.

(Testimony of Edward J. Weinkam, III on Contentions 3 & 6, ,

14/-~ EPRI recommends the conduct of surveillance programs to ensure long-term performance. (Turner on Contentions 3 & 6, l

ff. Tr. 139, at 14.) The St. Lucie I surveillance program is consistent with the program described by EPRI in its '

study with respect to all parameters relevant to the performance of Boraflex as a neutron absorber. (Turner on Contentions 3 & 6, ff. Tr. 139, at 17.) ,

s

following Transcript 139, at 3 (hereinafter "Weinkam on contentions 3 & 6, ff. Tr. 139, at ").) These procedures I provide instructions for the removal and testing of Boraflex coupons which are maintained within the spent fuel pool. (Id.)

The program is designed to verify the physical characteristics

. and neutron absorbing properties of the Boraflex poison material  !

utilized in both Regions 1 and 2 of the St. Lucie 1 high density spent fuel pool storage racks. (Id.)

49. The Boraflex used in the surveillance program is I

representative of the Boraflex within the storage racks: it is l of the same composition, produced by the same method, and certified to the same criteria as the production lot material.

(Weinkam on Contentions 3 & 6, ff. Tr. 139, at 4.) Each sample coupon is the same thickness as the poison employed within the storage system; is not less than 5 inches in width, and 15 inches in length; and is encased in a stainless steel jacket which is L made from an austenitic stainless steel alloy identical to that utilized in the storage racks and which has been formed so as to .

i l encase the poison material and fix it in a position with tolerances similar to the design utilized in the racks. (Weinkam on contentions 3 & 6, ff. Tr. 139, at 4.) The jacket permits wetting and venting of the specimens in a manner similar to that which occurs in an actual rack environment. (Weinkam on i Contentions 3 & 6, ff. Tr. 139, at 4.)

i k _ _ . _ _ _ _ _ . _ _ _ _ . - _ . _ _ _ _ _ _ . _ _ _

50. Two separate trains (or trees) of encased coupons are being used in each region of the racks -- one exposed to high radiation' levels and the other exposed to radiation levels representative of normal use. (Turner on Contentions 3 & 6, ff.

Tr. 139, at 15-16; Weinkam on Contentions 3 & 6, ff. Tr. 139, at i 5.) In the long-term test, coupons will be surrounded by the j i

same spent fuel assemblies during the entire irradiation period; while in the accelerated test, coupons will be surrounded by l freshly discharged spent fuel assemblies at each refueling.

(Weinkam on Contentions 3 & 6, ff. Tr. 139, at 5.) The long-term test coupon examination frequency is after nominal irradiation times of 90 days, 180 days, 1 year, 5 years, 10 years, 15 years, 25 years and 35 years. (Id.) The accelerated test coupon will >

be examined after each discharge of spent fuel into the spent fuel pool from the second discharge to the ninth discharge after rack installation. (Id.)

51. When the Boraflex coupons are removed, they will be subjected to careful examination, as follows:

a. Visual examination 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.

b. Dimensional measurements to provide a continuing measure of Boraflex shrinkage.

The length measurement is of particular importance as an indicator of the potential for gap formation in excess of that accommodated in the design.

l l

c. Neutron attenuation measurements to establish' areal density to confirm that boron is not being lost from the Boraflex. Although previous irradiation tests indicate that boron is retained, this is perhaps the single most important measure of the ability of Boraflex to continue to serve its intended function.

d. Neutron radiography to provide supporting information on neutron attenuation. It is intended to reveal any non-uniformities in the boron distribution within the Boraflex that might not be uncovered in the attenuation measurements,

e. 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 of excessive degradation.

f. Weight and specific gravity measurements to support measurements intended to reveal any significant loss of Boraflex material or the  ;

development of more open porosity than expected.

(Turner on Contentions 3 & 6, ff. Tr. 139, at 16-17; see Weinkam on Contentions 3 & 6, ff. Tr. 139, at 4.)

52. These observations are designed to evaluate the continuing suitability of Boraflex for use in the St. Lucie 1 spent fuel pool as a neutron absorber. (Weinkam on Contentions 3

& 6, ff. Tr. 139, at 4.) The principal evaluation criteria, within the precision of the measurements, are: dimensional changes in width and length no more than 2.5 percent from the original; hardness not less than 90 percent of the material after initial irradiation; and no change in the boron content of the

material from the original. (Weinkam on contentions 3 & 6, ff.

Tr. 139, at14-5.) If the observations reveal changes in material properties of the Boraflex significantly in excess of that expected (that is, if the Boraflex shows evidence of degrada-tion), the Licensee will initiate additional evaluation and, if necessary, additional tests. (Weinkam on Contentions 3 & 6, ff.

Tr. 139, at 5.)

53. The intensity of radiation emitted from a fuel assembly decreases.very rapidly during the first few months of spent fuel storage and most of the dose to the Boraflex in the spent fuel pool is accumulated during this time interval.

(Turner on Contentions 3 & 6, ff. Tr. 139, at 11-12, Figure 1 at 21.) A Boraflex panel in the St. Lucie 1 spent fuel pool will have accumulated a dose of approximately 1 x 10 10 rads; and at twenty years, the accumulated dose will be less than-3 x 10 10 rads. (Turner on Contentions 3 & 6, ff. Tr. 139, at 12, Figure 1 at 21.) However, cells in two small areas, which have been designated for accelerated testing of Boraflex and which will contain one set of surveillance coupons each, will intentionally receive freshly discharged spent fuel at each refueling, so that the Boraflex in the surveillance coupons will accumulate the equivalent of 20 years normal use in about 3 to 4 years, and can accumulate up to 2 x 10 10 rads over the anticipated lifetime of the racks, which is a small fraction (approximately one-fifth) of the dose to which the material has been tested. (Turner on

l

. Contentions 3 & 6, ff. Tr. 139, at 12.) These accelerated doses are adequate to reveal any degradation of the Boraflex in ample time to take any necessary corrective action. (Id.)

54. The Licensee has available a number of options for  !

taking corrective action to ensure continued safe fuel storage should Boraflex degradation problems occur. They include the i

following: l

a. The degraded Boraflex could be evaluated to determine whether the degradation would adversely affect FPL's ability to satisfy the 0.95 k-eff limit for the St.

Lucie 1 spent fuel pool. If the pool .

could still satisfy this limit, no l further action would be necessary.

b. Administrative controls could be imposed on the enrichment-and/or burn-up of. fuel to be placed in or adjacent to storage cell locations containing degraded Boraflex to ensure that k-eff would remain less than or equal to the 0.95 limit.

c.. 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-eff to less than or equal to the 0.95 limit.

d. The storage cells with degraded Boraflex  ;

could be blocked off to prevent their loading i with fuel assemblies. j i

(Weinkam on Contentions 3 & 6, ff. Tr. 139, at 5-6.) l

(8)- Technology used in the St. Lucie l' spent fuel racks

55. As we stated above, the Intervenor in this proceeding was concerned that the use of Boraflex in the St.

Luc e 1 spent fuel racks, manufactured by the Joseph Oat Corporation, represented an essentially new and unproven tech-1 nology. As our findings of fact in Section II.A.(2), supra, .l

' demonstrate, the test and in-service-experience to date supports the conclusion that Boraflex can serve as an effective neutron i absorber in spent fuel rack e. applications, such as at St. Lucie 1.

Insofar as the other rack materials are concerned, their l l

effectiveness in spent fuel storage rack applications is well-  !

established. Indeed, we granted summary disposition of Costention 3 as to all non-Boraflex materials in the spent fuel

-pool, including those used in the St. Lucie 1 storage racks. See l

SD Order at 38.

56. As to the manufacture of the St. Lucie 1 racks, l

since 1980, the fabricator -- the Joseph Oat Corporation -- has had extensive experience in the fabrication of spent fuel storage racks which use Boraflex; and before that, in the 1970's, in the fabrication of "new fuel racks," which employ the same tech-nological base as spent fuel racks. (See Singh on Contentions 3

& 6, ff. Tr. 139, at 4-5, Table A at 18.) Moreover, the Joseph Oat Corporation has decades of experience in the fit-up, cleaning and handling of stainless steel components, and in the welding l

I L

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L ,

processes which must be used in fabricating from stainless steel in sheet metal' form, such as in storage rack applications. (See L Singh on Contentions 3 & 6, ff. Tr. 139, at 5-6.) Furthermore, l all of the significant features of.the.St. Lucie 1 racks for both Region 1 and Region 2 are direct adaptations of established technology. (Singh on' Contentions 3 & 6, ff. Tr. 139, at 8.)

57. In sum, the use of Boraflex in the St. Lucie 1 spent fuel racks is well-grounded in experience and established technology.

B. Contention 7 l

58. Contention 7 reads as follows: .

i That the increase of the spent fuel pool capacity, which includes fuel rods that are l more enriched, will cause the requirements of ANSI-N16-1975 not to be met and will increase r the probability that a criticality accident i will occur in the spent fuel pool and will exceed 10 CFR Part 50, A 62 criterion.

Florida Power & Light Co. (St. Lucie Plant,' Unit No. 1), LBP 10A, 27 NRC 452, 471, aff'd, ALAB-893, 27 NRC 627 (1988). As indicated in Part I, supra, the issues pertinent to the contention which remained for hearing after our ruling on the Licensee's motion for summary disposition can be stated as follows:

1 a) The mechanisms which prevent the erroneous insertion of a fuel assembly into a storage cell, such that the prescription of Standard Review Plan (SRP) Section 9.1.2, Part ,

III.2.b,'that it not be possible for "a fuel l assembly . . . (to) be inserted anywhere other 1 1

• \

i than a design location," has not been demonstrated; and b) It has not been shown why criticality will not occur in the spent fuel ,

pool in the absence of a moderator.

(1) The process of nuclear fission

59. In order to examine the issues raised in Contention 7, as well as to provide additional background information pertinent to the discussion of Contentions 3 & 6, supra, it will be helpful to consider some of the basic principles of the fission process. 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. The primary fissile material in the new fuel assemblies of most nuclear power reactors, including St. Lucie 1, is a nuclide of uranium called uranium-235. In natural uranium, the uranium-235 nuclide is present at a concentration of less than 1 percent by weight, with almost all of the remainder being the uranium-238 nuclide. To be useful in a light-water nuclear power reactor, natural uranium is enriched in uranium-235. 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.

(Testimony of Dr. Stanley E. Turner on Contention 7, following

. 1 Transcripts 21, at 5-6 (hereinafter " Turner on Contention'7, ff.

Tr. 21, at ").)

60. In general, when a neutron is absorbed by uranium-235, there is a-high probability that the uranium-235 will undergo fission, resulting in the release of energy and addi-tional neutrons. These neutrons, in turn, can: (a) be absorbed  :

by uranium-235 or other fissile nuclides; (b) be absorbed by uranium-238 nuclides, resulting in virtually no additional fi6sion; (c) be absorbed non-productively by non-fissile material called " poisons"'(resulting in no additional fission); or (d) escape without being absorbed (i.e., leakage, which also results in no additional fission). (Turner on Contention 7, ff.

Tr. 21,.at 6.)

61. In a practical system, not all neutrons released as'a result of fission will cause additional fissions. Uranium-238 nuclides, non-fissile poison materials, and leakage all inhibit the fission process by reducing the number of neutrons available to cause additional fissions. If fewer neutrons are being produced as a result of fission than are leaking and being absorbed, the fission process will not sustain itself. This condition is called "subcriticality." In 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 fission process will sustain itself, and the condition is referred to as

" critical." (Turner on Contention 7, ff. Tr. 21, at 6-7.)

l

62. The term " effective multiplication factor" is defined as the ratio of the number of neutrons per unit time produced in the fission process, to the number of neutrons per unit of time absorbed or escaping. The effective multiplication factor, commonly called k-effective (or "k-eff"), is a measure of the ability of a system to sustain a fission reaction. Criti-cality occurs whenever the effective multiplication factor reaches or exceeds a value of 1.0, because at least as many neutrons are being produced as are being lost by absorption and leakage. For a k-eff of less than 1.0, the fission rate cannot be sustained. The margin below a k-eff of 1.0 is the safety margin to criticality. This subcritical margin is the difference between 1.0 and the k-eff of a given system. NRC guidelines for fuel storage racks require that the maximum k-eff, including all known uncertainties, be equal to or less than 0.95. This provides a substantial suberitical margin. (Turner on Contention 7, ff. Tr. 21, at 7-8.)

63. During the operation of a nuclear power plant, the fission process within the fuel produces a large quantity of heat (a desired product), and new atomic species called fission products (an undesired product). Those fission products which absorb neutrons non-productively are often referred to as

" poisons," meaning that they inhib!t the fission process by reducing the reactivity (k-eff) of the fuel. Most of the fission products are stable and remain in the fuel as neutron-absorbing

poisons. However, a significant fraction of the fission products is unstable, producing the radioactivity commonly associated with spent fuel. 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. In general, the radioactive decay of the radioactive nuclides which were formed in the fuel during reactor operation results in a substantial and continuous decrease in k-eff during the period of time the fuel elements are in storage. (Turner on Contention 7, ff. Tr. 21, at 9-10.)

64. The reactivity (k-eff) of fuel assemblies is affected primarily by two factors: (a) the quantity and enrichment of the uranium-235 in the fuel, and (b) the quantity of neutron-absorbing materials (poisons) present. Changes in k-eff can be produced by several different mechanisms. Increasing the fuel enrichment increases the fuel's reactivity, as does increasing the density of fuel assemblies in the spent fuel pool, or lowering the concentration of poisons. Conversely, the reactivity of fuel elements in a spent fuel pool can be decreased by decreasing the enrichment of uranium-235 in stored fuel assemblies; by decreasing the density of the stored fuel assemb31eSe or by increasing the concentration of poisons. With respect to the last factor, neutron-absorbing poisons may be intentionally installed in the storage racks to reduce the

i 8

l system's reactivity. This is accomplished in the St. Lucie 1 racks by the installation of Boraflex as a neutron-absorbing or poison material. Soluble boron is also utilized in the spent fuel pool water as additional poison material. During reactor operation, uranium-235 contained in the fuel within the core is consumed (depleted) in the fission proceas and the effective fuel enrichment is decreased, thus resulting in reduced reactivity.

In addition, fission products accumulating in the fuel further reduce the reactivity. (Turner on Contention 7, ff. Tr. 21, at 10-11.)

(2) Two-region spent fuel storage rack design

65. The expanded fuel storage racks at St. Lucie 1 ar of a two-region design. Region 1 is designed to accommodate fresh unirradiated fuel. Region 2 is designed for storage of spent fuel of a minimum specified burnup that, in turn, depends upon the initial enrichment of the particular fuel batch. Each of the two regions of the fuel storage racks, therefore, has different design criteria, provides for a different B-10 loading in the Boraflex, and utilizes a different fuel assembly spacing.

(Turner on Contention 7, ff. Tr. 21, at 3-4.)

66. Region 1 modules are designed to safely accommodate fuel of the highest reactivity anticipated to be stored in the pool; i.e., fresh unburned uranium fuel, enriched to 4.5 weight percent in the uranium-235 nuclide. Since fresh l

l

i fuel is more highly reactive than burned fuel, fuel of any burnup may be stored in Region 1 with assurance that the k-eff will be less than the maximum design case. Region 1, therefore, is intended to provide safe storage for fresh fuel, to accommodate a full core off-load when required, and to store fuel whose ,

burnup does not satisfy the criteria for storage in Region 2 of the pool. Region 1 utilizes Boraflex absorber sheets containing a nominal B-10 loading of 0.0238 grams per square centimeter '

(two Boraflex sheets between adjacent cells in a conventional flux-trap configuration). The storage cells are arranged on a nominal center-to-center spacing of 10.12 inches, which is adequate to assure a k-eff less than 0.95 in the absence of all soluble boron, including all known uncertainties, with the most reactive fuel present in all the Region 1 cells. (Turner on ,

Contention 7, ff. Tr. 21, at 4-5.)

67. Region 2 modules are provided for the purpose of storing spent fuel removed from the reactor. Because the reactivity of the fuel assemblies decreases substantially as burnup is accumulated (and the fissile material is depleted),

Region 2 is designed to safely store fuel of 4.5 weight percent initial enrichment which has accumulated a burnup of at least 36,500 megawatt days per metric ton of uranium (MWD /MtU). A similar minimum, or limiting fuel burnup, has also been estab-lished analytically for fuel of lower initial enrichments, and i

these data define the bounding condition of burnup for acceptable i i

I i

1

A e i I

storage in Region 2. For any given initial enrichment, fuel assemblies with burnup equal to or greater than the bounding l

condition may be safely stored in Region 2, while assemblies having less than the minimum required burnup will be stored in Region 1. Region 2 of the storage pool utilizes cells on a nominal 8.86 inch center-to-center spacing with a single panel of Boraflex between cells providing a nominal B-10 loading of 0.0097 grams per square centimeter. (Turner on Contention 7, ff.

Tr. 21, at 5.)

(3) NRC criticality criteria and the double contingency principle

68. The federal regulation governing safe storage of reactor fuel assemblies, insofar as criticality is concerned, is General Design Criterion 62 of Appendix A to 10 C.F.R. Part 50.

This provision states that "[c]riticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by the use of geometrically safe configura-i tions." The primary NRC guidelines for the acceptable implemen-tation 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. A k-eff not greater than 0.95

e: _ 43 _

for this condition is acceptable."- (Turner on Contention-7, ff.

Tr. 21, at 11.)

69. ANSI N16.1-1975, " Nuclear Criticality Safety in

. operations with Fissionable Materials outside Reactors,"

establishes the " double contingency principle," subsequently adopted and incorporated in NRC guidelines. This 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.

(Turner on contention 7, ff. Tr. 21, at 11-12.)

70. - Additional related guidance is provided in industry standards and NRC regulatory guides, including:

(a) ANS 8.17-1984, " Criticality Safety Criteria for the l Handling, Storage and Transportation of LWR Fuel Outside Reactors"; and (b) USNRC Regulatory Guide 1.13, " Spent Fuel Storage Facility Design Basis," Rev._2 (proposed), December 1981.

I (Turner on Contention 7, ff. Tr. 21 , at 12.)

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71. The most definitive clarification of NRC guidance is provided in the April 14, 1978 letter from the NRC to all l power reactor licensees transmitting the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,"

which sets forth in detail the NRC acceptance criteria for spent fuel storage pools.Section III.1.5 emphasizes that the " neutron multiplication factor in spent fuel pools shall be less than or i

l-

equal to 0.95, including all uncertainties, under all conditions." (Emphasis in original). Section III.l.2 (Postulated Accidents) of the April 14, 1978, letter also invokes the double contingency principle of ANSI N16.1-1975 for fuel pool analyses, stating that:

The double contingency principle of ANSI N 16.1-1975 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 boron) may be assumed for the fuel pool and fuel assemblies.

(Turner on Contention 7, ff. Tr. 21, at 12-13.) i

72. As previously noted, industry standards and NRC ,

guidelines limit the maximum k-eff to 0.95. As will be evident from the subsequent discussion, the design and criticality safety analysis for the new spent fuel storage racks at'St. Lucie 1 were-performed in accordance with and conform to the definitive guidance contained in the April 14, 1978 letter. In addition, it is relevant to note that prior to issuance of the St. Lucie 1 spent fuel pool expansion amendment, the design basis k-eff limit was 0.95, and remains unchanged. The spent fuel pool expansion amendment did not modify this design basis limit. The amendment has not decreased the margin of safety preventing a criticality accident at St. Lucie 1. (Turner on Contention 7, ff. Tr. 21, at 13.)

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(4) Criticality analysis of the St. Lucie 1 spent fuel pool

73. Two independent calculational methods were used 1

J to determine the k-eff of the St. Lucie 1 fuel storage racks, KENO-IV, and CASMO-2E. Both of these methods were benchmarked.

against critical experiment data for configurations as l representative as possible of actual storage rack geometry, in accordance with Regulatory Guide 3.41 and the guidelines of the April 14, 1978 NRC letter. (Turner on Contention 7, ff. Tr. 21, at 13-14.)

74. The nominal k-eff calculated for an infinite array of Region 1 assemblies was 0.9313, without credit for the (redundant) reactivity control provided by the soluble boron in the pool water. The corresponding value for Region 2 is a nominal k-eff of 0.9114. (Turner on Contention 7, ff. Tr. 21, at 14.)

75. The NRC acceptance criteria for criticality analyses require consideration and inclusion of all known uncertainties in the calculation of k-eff. With all uncertain-ties included, and assuming the absence of any soluble boron, the maximum possible k-eff values are 0.941 for Region 1 and 0.944 for Region 2. (Turner on Contention 7, ff. Tr. 21, at 14-15.)

Since all uncertainties are included in the maximum calculated k-eff values, the safety margins specified in the NRC acceptance criteria are conservative. With soluble boron in the pool water

.. i.

(normal operating conditions), the' maximum reactivities-in Regions 1 and 2 are 0.767 and 0.760, respectively. (Turner on Contention 7, ff. Tr. 21, at 15.) .

(5) Boraflex and soluble boron as independent

- and redundant methods of reactivity control

76. Once the racks are manufactured and installed in the storage pool, maintaining a k-eff of less than 0.95 is accom-plished by restricting the fuel enrichment stored in Region 1 to 2

4.5 percent uranium-235, and by restricting the fuel stored in Region 2 to those assemblies which have accumulated the required minimum burnup as required in the plant Technical Specifications.

The racks in St. Lucie 1 are designed to ensure a k-eff of less than 0.95 in the absence of soluble boron in the pool water. In addition, the required concentration of 1720 ppm (parts per million by weight) of soluble boron in the pool water, would preclude a criticality accident even if the Boraflex absorber panels were to be completely lost. Thus, the soluble boron in the pool water and the Boraflex panels in the racks constitute independent and redundant means of preventing a criticality accident. For normal storage conditions, the presence of both soluble boron and Boraflex absorber panels provide a very large a subcriticality margin (k-eff less than 0.80). (Turner on Contention 7, ff. Tr. 21, at 15-16.)

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78. It is theoretically possible, from a physical standpoint, to improperly insert a fresh fuel assembly (or one of less than the requisite burnup) into a Region 2 storage cell. q (See Turner on Contention 7, ff. Tr. 21, at 17-18; Tourigny on Contention 7, ff. Tr. 110, at 13.) It is also physically possible to lower a fuel assembly into the shipping cask handling area, and a small area between the east wall of the pool and rack j modules E y and H y. There are no racks in either area. (Tourigny on Contention 7, ff. Tr. 110, at 12-13; Weinkam on Contention 7,  !

l ff. Tr. 21, at 3-4.)

79. NRC Staff guidance, however, allows for l administrative controls,' utilizing written procedures, to prevent the misplacement'of fuel in the pool. (See Turner on Contention {

7, ff. Tr. 21, at 17-18; Tourigny on Contention 7, ff. Tr. 110, I

at 13.) FPL has developed rigorous procedures to ensure the l proper placement and storage of all fuel, from the time it first arrives on site, until the time it is fully spent and moved to the spent fuel pool for storage and eventual shipment for disposal. (Weinkam on Contention 7, ff. Tr. 21, at 4-9.) These procedures ensure that fuel assemblies will only be inserted into pool storage rack locations that have been specifically assigned to them. (Id.) By ensuring that fuel assemblies are inserted

! only where they should be, the procedures necessarily ensure that they will not be inadvertently inserted where they should not be.

(Weinkam on Contention 7, ff. Tr. 21, at 8-9.) This would I

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include ~ improper cell locations in the wrong region, or outside of the racks altogether. . (Weinkam on' Contention 7, ff.' Tr. 21, at_9.)

80. It is also pertinent to note that, even if a fresh' fuel assembly were to be mislocated within the storage pool in the worst possible location, the maximum k-eff would remain below 0.8,.taking into account the presence of-soluble boron in the pool water. (Turner on Contention 7, ff. Tr. 21, at 18-19; Turner, Tr. 92-93.) Even in the absence-of soluble boron, the disinsertion of a fresh fuel assembly into a Region'2 location would not result in criticality. (Turner, Tr. 92-93.) 15/

Multiple misinsertions would be necessary. (Id.) W;th the prescribed soluble boron in the pool, criticality would not occur l

.even if fresh fuel were misinserted into each and every Region 2 cell. (Turner, Tr. 55-57.)

(7) Criticality considerations assuming the absence of moderator

81. Contention 7 also raises the issue of a'possible criticality accident in the absence of a moderator; i.e., if the pool, for some unknown reason, were to be drained of water. To 15/' Quantitatively, for the worst possible dislocation of fuel in the pool, k-eff would remain less than 0.8, or more than 0.2 away from criticality at k-eff = 1. The soluble boron in the pool contributes 0.184 to shutdown margin. (See Turner on Contentions 3 & 6, ff. Tr. 139, at 19.) This, of course, is less than the more than 0.2 increase in k-eff necessary to reach criticality.

1. ~ l consider'this matter, it is necessary to recall certain-facts regarding criticality and the fission process, discussed-earlier. These may be summarized'as follows:

(1) The fission process involves the absorption of i neutrons in the uranium-235 nuclide (or other fissile nuclides) with the resultant release of  !

energy and additional high energy, or " fast,"

neutrons.

(2) Fissioning of the uranium-235 nuclide is caused principally by low-energy or " thermal" neutrons.

The probability of higher energy neutrons.being absorbed by a uranium-235 atom and causing fission l is very small.

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

(4.) Criticality occurs whenever the effective multiplication factor (k-eff) -- 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.

82. In the absence of a moderator to slow down the fast neutrons released in the fission process, the fraction of ,

l neutrons absorbed in uranium-235 is small compared to the fraction absorbed in uranium-238. The net effect is a very low value for k-eff, 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-productively absorbed in uranium-238). Furthermore, the absence ,

of a moderator in an actual system of finite size would greatly ,

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increase the leakage of the high energy neutrons, resulting in-an even greater reduction in k-eff. (Turner on Contention'7,'ff.

Tr. 21, at 20; see Turner, Tr. 58-60.)

83. In fact, 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. This principle assures that fresh fuel assemblies may be safely handled and stored in air (i.e., dry condition, without moderator), knowing that criticality is not possible without substantial moderating material present. (Turner on Contention 7, ff. Tr. 21, at 20-21.)

84. In order for criticality to occur in the absence of moderating 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 the St. Lucie 1 fuel. With the low-enriched uranium used in 1

the St. Lucie 1 fuel or with an array of St. Lucie 1 fuel assemblies, criticality is impossible in the absence of a moderator (water in the case of St. Lucie 1). (Turner on Contention 7, ff. Tr. 21, at 21-22; Turner, Tr. 62-67; Tecti.nony of Lawrence I. Kopp on Contention 7, following Transcript 110, at 14 (hereinafter "Kopp on Contention 7, ff. Tr. 110, at ").)

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85. Studies'have shown that, for an enrichment'of less that 5 percent, criticality is impossible absent.a moderator. (Turner on contention 7, ff. Tr. 21, at 20; Kopp on Contention 7, ff. Tr. 110, at 14.) Even assuming the minimal moderating effect of zirconium and air, and all fissionable nuclides which'might be present, criticality would be impossible in the'St. Lucie 1 Upent fuel storage pool without water as a moderator, regar(Ab's of its mass or configuration. (Turner, Tr.

62-67, 81-87, 94-93, 101-04.)

III. Conclusions of Law Based on the foregoing findings of fact and upon co sideration of the entire evidentiary record in this proceeding, we make the following conclusions of law:

1. Appropriate, NRC Staff-approved methodology has been applied.in analyses associated with evaluating the subject amendment. The results of the analyses reveal that the new spent fuel storage racks at St. .Lucie 1 satisfy all' applicable NRC criteria.

2. Contrary to Intervenor's assertion in Contention 3, the Licensee and Staff have adequately considered the integrity of Boraflex to serve as a .

fixed neutron absorber in the St. Lucie 1 spent fuel racks. The Boraflex in the St. Lucie 1 spent fuel racks is an appropriate use of fixed, neutron-absorbing material to control criticality.

Florida Power & Light Company's in-service surveillance program for the Boraflex is an effective method of ensuring that the material i performs its intended function, and provides adequate opportunity for remedial action, if

. necessary.

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3. The mechanisms utilized by Florida Power & Light Company at St. Lucie 1 to prevent the disinsertion of a fuel assembly into a storage cell are such that Section 9.1.2, Part III.2.b of the NRC l

Standard Review Plan is satisfied.

l 4. Florida Power & Light Company has fully met its l burden of proof on issues raised in this proceeding. The subject amendment does not constitute a reduction in safety margin in that the 0.95 criteria for k-eff is unchanged and satisfied for the St. Lucie 1 spent fuel racks.

5. The design of the spent fuel storage racks for St.

Lucie 1, together with Florida Power & Light Company's surveillance program, provide reasonable assurance that the health and safety of the public will not be endangered.

IV. ORDER WHEREFORE, 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 l

modification.

IT IS FURTHER ORDERED, that this Decision shall constitute the final decision of the Commission within thirty (30) days from the date of issuance, unless an appeal is taken in accordance with 10 C.F.R. S 2.762 or the Commission otherwise directs. See also 10 C.F.R. SS 2.764, 2.785 and 2.786. Any party may take an appeal from this Decision by filing a Notice of I

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ )

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Appeal within ten (10) days after service of this Decision. A brief in support of such appeal shall be filed within thirty (30) days after the filing of the Notice of Appeal (forty (40) days if the appellant is the NRC Staff). Within thirty (30) days after j the period has expired for the filing and service of the briefs of all appellants (forty (40) days in the case of the NRC Staff),

any other party may file a brief in support of, or in opposition to, the appeal of any other party. A responding party shall file a single responsive brief, regardless of the number of appellant briefs filed. See 10 C.F.R. 5 2.762.

Respectfully submitted, IAA Co-Counsel Hl=af Harold F. Reis f )!

g. ,(b?a.u.*,

Michael A. Bauner John T. Butler Patricia A. Comella Steel, Hector & Davis 4100 Southeast Financial Newman & Holtzinger, P.C.

Center 1615 L Street, N.W.

Miami, FL 33131-2398 Washington, D.C. 20036 Telephone: (305) 577-2939 Telephone: (202) 955-6600 Counsel for Florida Power & Light Company

+.

'E L Appendix A l

A. Exhibits received into evidence for'their' evidentiary value. i L

(See Tr.'568.) ']

Exhibit 11: Letter, dated June 12, 1987, '

from C. O. Woody, Florida Power & Light Co.,

to U.S. NRC Document Control Desk, regarding L "St. Lucie 1, Docket No. 50-335, Proposed License Amendment, Spent Fuel Rerack," (L 245), and Attachments, including, " Florida Power.& Light Co., St. Lucie Plant - Unit No.

1, Spent Fuel Storage Facility Modification, Safety Analysis Report, Docket No. 50-335."

Exhibit 12: ' Letter, dated October- 20, 1987, from C. O. Woody,. Florida Power & Light Co.,

to U.S. NRC Document Control Desk, regarding "St. Lucie Unit 1, Docket No. 50-335, Spent Fuel Pool'Rerack - Boraflex'and Pool Cleanup,"'and Attachment, " Responses to NRC' Letter dated September 1, 1987 (E., G.

Tourigny to C. O. Woody)."

B. Exhibits received "into the record for the purpose of elucidating the examination that has taken place." (Tr.

427.)

Exhibit 1:- Electric Power Research Institute, "An Assessment of Boraflex Performance in Spent-Nuclear-Fuel Storage Racks," EPRI NP-6159 (December 1988).

Exhibit 2: U.S. NRC, " Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to Fuel i Enrichment Increase, System Energy Resources, Inc.,

Grand Gulf Nuclear Station Unit 1, Docket No. 50-416."

Exhibit 3: Letter, dated November 21, 1988, from John G. Cesare, Jr., System Energy Resources, Inc., to U.S. NRC Document Control Desk, regarding " Grand Gulf Nuclear Station, Unit 1, Docket No. 50-416, License No. NPF-29, Request for Additional Information,  ;

Criticality Analysis for Cycle 4, AECM-

a-o  :

l 88/0233," and Attachment to AECM-88/0233,  !

" Response to NRC Question Regarding GGNS Boraflex Gap Surveillance Program."

Exhibit 4: Bisco Products, Inc., Technical Report No. NS-1-002, "Boraflex Neutron Shielding Material Product Performance Data" (Revision 0, 8/25/81).

Exhibit 5: Letter, dated February 11, 1987, from C. W. Fay, Wisconsin Electric Power Co.,

to U.S. NRC Document Control Desk, regarding

" Docket Nos. 50 .266 and 50-301, Results of Examination of Poison Insert Assemblies Removed from the Spent Fuel Storage Racks, Point Beach Nuclear Plant, Units'l and 2,"

and Attachment, "Results of Boraflex Examination, Point Beach Nuclear Plant."

Exhibit 6: Letter, dated January 29, 1988, from C. O. Woody, Florida Power & Light Co.,

to U.S. NRC Document Control Desk. (L-88-38),

regarding "St. Lucie Unit 1, Docket No. 50-335, Sper.t Fuel Rerack" (L-88-38), and Attachment, " Florida Power & Light Company, St. Lucie Plant-Unit No. 1, Spent Fuel Storage Facility Modification, Safety Analysis Report, Docket No. 50-335" (Revision 1).

Exhibit 7: Letter, dated May 5, 1987, from M. S. Turbak, Commonwealth Edison, to A.

Bert Davis, U.S. NRC,' Region II, regarding

" Quad Cities Station Units 1 and 2, Spent Fuel Storage Racks, NRC Docket Nos. 50-254 and 50-265," and Attachment, Northeast Technology Corp., Report No. NET-042-01,

" Preliminary Assessment of Boraflex Performance in the Quad Cities' Spent Fuel ,

Storage Racks" (Revision 0, 4/10/87).  ;

l Bisco Products, Inc., Techn3 cal

) Exhibit 9:

} Report No. NS-1-050, " Irradiation Study of

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Boraflex (Interim Report)" (Revision 1, 11/25/87).

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o C. Exhibit received'into evidence "as a vehicle for reviewing Dr. Singh's testimony in the aspects that are illustrated by the drawing." (Tr. 427.)

Exhibit 10: Dr. Singh's drawing illustrating various aspects of the St. Lucie 1 Region 1 rack design.

t D. Staff Exhibits " stipulated as exhibits by the parties" and received into evidence for their evidentiary value. (Tr.

433).

Staff Exhibit 1: Letter, dated March 11, 1988, from E. G. Tourigny, U.S. NRC, to W. F.

Conway, Florida Power & Light Co., regarding "St. Lucie Unit 1 - Issuance of Amendment re:

Spent Fuel Pool Expansion (TAC No. 65589), j including Attachments, " Florida Power & Light. j Company, Docket No. 50-335, St. Lucie Plant Unit No. 1, Amendment to Facility Operating License (Amendment No. 91, License No. DPR-67, March 11, 1988); Revised Technical Specifications; and " Safety Evaluation by the Office of Nuclear Reactor Regulation relating i to the Reracking of the Spent Fuel Pool at the St. Lucie Plant, Unit No. 1, as related j to Amendment No. 91 to Unit 1 Facility  ;

Operating License No. DPR-67, Florida Power l and Light Company, Docket No. 50-335." t Staff Exhibit 2: Letter, dated February 29, i 1988, from E. G. Tourigny, U.S. NRC, to C. O.

Woody, Florida Power & Light Co., regarding

" Environmental Assessment and Finding of No Significant Impact - Spent Fuel Pool Expansion, St. Lucie Plant, Unit No. 1 (TAC No. 65587)," and the subject attachments.

E. Rejected exhibit. (Tr. 427.)

Exhibit 8: R. W. Lambert, Electric Power Research Center, Presentation at the 1988 Joint Power Conference, September 28, 1988,  ;

Philadelphia, PA, " Neutron Absorber Materials for Spent Fuel Racks." l l

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