ML20238D432

NRC Staff Proposed Findings of Fact & Conclusions of Law in Form of Initial Decision.* Licensee Seismic Analysis for New Spent Fuel Pool Racks Shows That Rack Design Satisfies Structural Aspects of GDC 2,4,61 & 62.W/Certificate of Svc

ML20238D432
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Site:	Turkey Point
Issue date:	12/24/1987
From:	Matt Young NRC OFFICE OF THE GENERAL COUNSEL (OGC)
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ML20238D394	List: ML20238D392 ML20238D432
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OLA-2, NUDOCS 8801040255
Download: ML20238D432 (52)

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December 24, 1987 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD in the Matter of )

} Docket Nos. 50-250 OLA-2 FLORIDA POWER AND LIGHT ) 50-251 OLA-2 COMPANY )

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(Turkey Point Plant, Units 3 & 4) ). (SFP Expansion)

NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM OF AN INITIAL DECISION The NRC staff, in accordance with 10 C.F.R. 9 2.754 and the Board's ruling at hearing (Tr. 376-77), as amended by Order dated November 4, 1987 and a conference call on November 20, 1987, proposes the following findings of fact and conclusions of law in the form of an initial decision,

l. INTRODUCTION AND BACKGROUND

1. Florida Power & Light Company (Licensee or FPL) is licensed to possess, use and operate Turkey Point Plant, Units 3 and 5, two pres-surized water nuclear reactors located in Dade County, Florida. The amendments would allow the expansion of the spent fuel storage capacity from 621 fuel assemblies to approximately 1404 fuel assemblies.

2. A notice of consideration of the issuance of the proposed amendments and an opportunity for hearing was published in the Federal 0

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Register on June 7, 1984. 48 Fed. Reg. 23,715. In response to the notice, the Center for Nuclear Responsibility, Inc. and Joette Lorlon (In-tervenors) filed a timely joint request for hearing and petition for leave to intervene. On November 21, 1984, pursuant to 10 C.F.R. 50.91(a)(4),

the Staff made a final no significant hazards determination and issued the amendments authorizing expansion of the capacity of the spent fuel pools.

3. Interveners filed a nontimely amended petition on March 7, 1985. We ruled on the contentions and late petition in our September 16, 1

1985 Order and admitted seven of the proffered contentions (Contentions 3, 4, 5, 6, 7, 8 and 10). LBP-85-36, 22 NRC 590 (1985).

4. Licensee filed motions for summary disposition of all the conten-tiens on January 23, 1986 and each motion was supported by the Staff, with the exception of a portion of Contention 4. Intervenor's opposed Licensee's motions and based its opposition on the affidavit of Joette Lorion. The Staff later submitted its own motion for summary disposition of Contention 4 which was supported by Licensee and not responded to by Interveners. \

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5. By Memorandum and Order dated March 25,1987 (unpublished) m.

(hereafter "SD Order"), we determined that there was no genuine issue '

of material fact es to Contentions 3, 4, 7, 8 and 10 and we granted Licensec's motion- for summary disposition. We denied the motion for sum-mary disposition of Contentions 5 (seismic analysis of rack design) and 6 (materials integrity). Hearings were held on these contentions in Miami, Florida on September 15 and 16, 1987. At the hearing, the Staff and Licensee presented testimony on the two contentions via a series of wit- .

l ness panels which we found were in general agreement as to the merits of

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s v .i t, Contentions 5 and 6. We also received into evidence, pursuant to 10 C.F.R. 6 2.743(g), the NRC Safety Evpluation supporting the % mend-

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ments (Staff Exhibit 1). T r. 131. Interveners did not sponsor any di-rect testimony and based their caseon cross-examination of thd Licensee and Staff panAss. I I

. 6. This opinion is based upon, and incorporates, the Findings of

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f pcynd Conclusions of Law that tolicw. Any proposed findings or con-

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clushns submitted by the parties that are not incc porated directly or infere'rtidiy in this inNial Decision are rejected as being insupportable in law or in fact or as be'ing unnecessary to the rendering of this Decision.

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11. FINDINGS A. Contention 5

7. Contention 5 states:

, .s~ \ That the' main safety function of the spent fuel pool which is to maintain the spent fuel assemblies in a l'

safe configuration through all environmental and" ab- ,

t normal loadings, may not be met as a result of a re- '

,5 cently brought to light unreviewed safety question s

S involved in the current rerack destgra t"st allows A racks whose outer rows overhang the support pads in

'e.he spent fuel pool. Thus, the amendments should be revoked.

( , This contention questions whether there is a deficiency in the Turkey i Point rack design and a necessity for a restriction on loading to prevent i

N-potential lift-off during seismic events. SD Order at 18. This concedrl Is based o 'e Licmsee letter which indicated that the structural design of

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c 1/ Interveners did not cross-examine Licensee witnesses on Contention 5 due to Ms. Lorion's lack of preparation. T r. 104.

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the rack, whose outer rows overhang the support pad, could cause the racks to lift-off (o' morts r likely tip-off) from the pool floor during seismic -

events. See Letter from J. W. Williams, Jr., FPL, to Steven A. Varga, s, NRC, dated February ' F, 1985. In our March 25, 1987 Order denying summary disposition, we found that there is no question that properly

+/. executed adrdnistrative controls wbuld prevent rack lift-off during a sels-j i mic event, SD Order at 21, but osserved that "there are sufficient n'

doubts as to the basis for }ssuance of the amendments, particularly the structural analysis involving the safe shutdown earthquake. and various ,

loading conditions other than fully loaded arid involving the overhanging rows, conditions which the Staff apparently has not evaluated." SD Or-der at 24. '

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8. To determine whether administrative centrols on loading sho }d be imposed by means c'Leither a license conditlan or technical specifica-tion requirement for Turkey Point, the Board has applied the guidance of .

the Appeal Board in Podtland General Electric Co. (Trojan Nuclear Plant),

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~ ALAD-531, f NRC 263.(1979). There the Appeal Board- stated:

[T]here is rbither a statutor'y nor rec).datory require-ment that every operational detail set fortb in an appli-cant's safety pnalysis report (or the equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licerrsee when and until chenged with specific Comtrdssion approval. Rather, . . . thti con-

• templation of both .the Act .ind the regulations iL.that technical specificaticos are to be reserved for those. mat-ters as to which the impu<.,ition of rigid conditions er

, limitations upon recctor operation is deemed necessary to obviate the possibi;ity of an abnormal situation' cc event giving rise to an immediate threat to the public. hatth and safety. '

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9 NRC at 273 (footnote omitted). U Consequently, we will determine whether the administrative controls are necessary to prevent an abnormal situation or event which poses an immediate threat to the public health and safety.

9. The Licensee's direct case consisted of testimony of a panel of f 1

Edmund E. DeMario (ff. Tr.103), an advisory engineer in the Commercial i Nuclear Fuel Division of Westinghouse Electric Corporation (Westing-house), liarry E. Flanders, Jr., a Principal Engineer for the Advanced Engineering Analysis Section of Westinghouse's Nuclear Components Divi-slon (ff. Tr.103), and Russell Couldy, a Senior Engineer in FPL's Nu-

) clear Licensing Department (ff. Tr.103) . Mr. DeMario has 26 years of engineering experience which includes over 18 years as a nuclear engi-neer for Westinghouse. Since joining Westinghouse in January 1969, he has been responsible for the mechanical design of advanced fuel assem-blies. DeMario, ff. Tr.103, at 6.

10. Mr. Flanders has been a practicing engineer for about 23 years and since joining Westinghouse in April 1974, he has been responsible for the seismic and structural analysis of the internal components of pressur-ized water reactors and of spent fuel racks. Flanders, ff. Tr.103, at

20. Mr. Couldy has about 10 years of nuclear engineering experience,

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See 10 C.F.R. 5 50.36; Sacramento Municipal Utility District (Rancho 5ec'o Nuclear Generating Station), A LA B-746, 18 NRC 749, 754 n.4 (1983); Commonwealth Edison Co. (Zion Station, Units 1 and 2),

A LA B-616, 12 NRC 419, 422 (1980); Virginia Electric Power Co.

(North Anna Nuclear Power Station, Units 1 and 2), A LA B-578, 11 NRC 189, 217 (1980) .

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• and has ' worked in several capacities at Turkey Point since June 1983.

Gouldy, ff. Tr.103, at 9.

11. The Staff's direct case consisted of testimony by Sang Bo Kim, a structural engineer at NRC, and Daniel G. Mcdonald, Jr., the NRC j project manager for Turkey Point. Mr. Kim has over 24 years of nuclear experience and his duties at the NRC have included the evaluation of the structural and earthquake engineering aspects of safety-related struc-tures, systems and components, and the performance of. independent cal-culations and engineering analyses to confirm applicants' or vendors' assessment of structural integrity and response under pertinent load com-binations, including postulated transient and accident conditions. Kim, ff. Tr.129, at Attachment 1. Mr. Mcdonald has over 25 years of engi-neering experience and is responsible for the coordination and integration of all licensing activities related to Turkey Point, Units 2 and 3.

Mcdonald, ff. Tr.129, at Attachment 2.

, 12. The spent fuel pools for Units 3 and 4 each have a total of twelve free standing racks. Each of the racks have support assemblies on the bottom which include remotely adjustable leveling screws. Four of the twelve racks have the support assemblies located at each of the four corners. The remaining eight racks have support assemblies recessed from the corners. The location of the support assemblies for all the racks was determined by the location of preexisting steel support plates which are permanently embedded in the concrete floors of the spent fuel pools. The racks th'ta have support assemblies recessed from the corners allow storage of fuel assemblies in the rows which overhang the support assemblies with the remainder of the racks empty. The racks are free

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standing, i.e., they are not anchored to the floor nor braced to the pool walls. Kim, ff. Tr,129, at 3-4; Flanders, ff. Tr.103, at 3-4.

13. The structural design of spent fuel pool racks as well as the spent fuel pool must satisfy General Design Criterion 2, " Design bases for protection against natural phenomena." GDC 2 provides that fuel storage be designed to withstand the effects of earthquakes without loss of capa-bility to perform its safety function. In addition, the spent fuel pool and pool storage racks must be designed to assure adequate safety under nor-mal and postulated accident conditions (GDC 61, " Fuel storage and han-dling and radioactivity control") and geometrically safe configurations of the fuel _ storage system should be used in order to prevent fuel criticality (GDC 62, " Prevention of criticality in fuel storage and handling"). Kim, ff. Tr.129, at 4.

14. The NRC review scope and acceptance criteria intended to as-sure conformance with the GDCs is described in the "OT Position For Review and Acceptance of Spent Fuel Storage and Handling Applications,"

dated April 14, 1978, and later amended on January 18, 1979 (OT Posi-tion) . The OT Position specifies acceptable load combinations of weight, temperature and earthquake. For example, only dead and live loads are considered for normal service conditions, and, for accident conditions, thermal and earthquake loads are added. Allowable stress levels increase with the severity of the service level and this is generally the industry practice as evidenced by the latest edition of ASME Code, Section Ill. In addition, the OT Position specifies an allowable safety factor for overturning by referencing Section 3.8.5.11.5 of the Standard Review Plan (SRP), NUREG-0800, in which a range of the safety factors between

-B-I 1.1 to 1.5 are provided depending on load combinations. The OT Position also states that total displacement, including thermal expansion due to temperature as well as movement of the rack due to earthquake (sliding  !

and tilting), should be considered using a detailed non-linear dynamic analysis which demonstrates that displacement is minimal. Kim, ff. Tr.

129, at 5,10-11; Flanders, ff. Tr.103, at 4-6.

15. This Staff criteria allows lift-off or tilting provided that, as stated in the criteria, (a) the factors of safety against tilting (or over-turning) are within the value permitted by Section 3.8.4.11.5 of the SRP and (b) it can be shown that any sliding and tilting motion will be con-tained within suitable geometric constraints such as thermal clearances and that any impact due to clearance is incorporated. Thermal clearance are calculations of the space between the racks after expansion of the racks due to the heat transferred from the spent fuel assemblies. Kim, ff. Tr.129, at 11; Flanders, ff. Tr.103, at 4-6.

16. The Staff's evaluation of Licensee's rack design was performed with the assistance of Franklin Research Center (FRC), the Staff's tech-nical consultant, and published in safety evaluation supporting the amendments. The review consisted of an evaluation of the Licensee's de-scription of the structural configuration of the spent fuel racks as well as the spent fuel storage pool, load combinations, calculations including rack response to an earthquake, resultant stresses in the rack, and compari-son of final stresses with allowable stress limits prescribed in the OT Position. The Staff concluded in Section 2.3.6 of its Safety Evaluation that the design of the racks satisfied the structural aspects of the Appendix A requirements of 10 C.F.R. Part 50 (GDC 2, 4, 61 and 62) i ,

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.because: (a) the Licensee considered all the required loading conditions including earthquakes and accidents; (b) the analysis methods that calcu-late stresses and earthquake response were in accordance with industry practice were acceptable as detailed in FRC's Technical Evaluation Report which is appendec to the Staff's Safety Evaluation; and (c) the resultant stresses and overturning safety factors satisfied the allowable limits spec-

. ified in the Staff OT Position. Kim, ff. Tr.129, at 4-6; Staff Exhibit 1, 6 2.3.6.

17. Subsequent to the Staff's November 21, 1984 Safety Evaluation, Licensee, by letter dated February 1,1985, presented an additional rack earthquake response analysis concerning the loading of the overhanging outer rows. This additional analysis was done as a result of being in-formed by Westinghouse Electric Corporation, the rack vendor, (a) that lif ting of a rack could occur during a seismic event if the outer rows are fully loaded while the rest of the rack is empty and (b) that administra-tive controls on fuel loading would be needed for those spent fuel racks whose outer rows overhang the support pads. Licensee stated that the analysis results demonstrated that the design of racks with fuel overhang continued to satisfy the OT Position in that there are adequate safety margins against overturning and stresses in the racks and pool. in addi-tion, Licensee stated it would provide administrative controls on fuel placement in order to preclude the possibility of rack lift-off. Kim, ff.

Tr.129, at 6.

18. Changes that do not result in a change to the technical specifications or result in an unreviewed safety question may be accom-plished through the use of 10 C.F.R. 50.59 Changes , tests and 1

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experiments." That regulation provides that a licensee may make changes in the facility or procedures as described in the Safety Analysis Report (SAR) and conduct tests or experiments not described in the SAR without prior Commission approval, unless the proposed change, test or experi-ment involves a change in the technical specifications incorporated in the license or an unreviewed safety question. Licensees may also employ pro-cedures and administrative controls. All technical specifications of li-censed nuclear plants include requirements for the development and use of procedures and administrative controls. Mcdonald, ff. Tr. 129, at 6-7.

19. By letter dated February 26, 1985 (Mcdonald, ff. Tr.129 at Attachment 3), the Staff responded to FPL's February 1,1985 request for NRC review of an analysis which showed that the results of lift-off would be acceptable. The Staff stated that although Licensees request for re-view of the analysis represented a change in the NRC basis, supporting issuance of the amendments which authorized the pool expansions, Licensee could make changes without prior NRC approval provided it per-formed a review pursuant to 10 C.F.R. 9 50.59 and determined that nei-ther a technical specification change nor an unreviewed safety question is involved. Accordingly, in a letter dated November 13, 1985, FPL with-drew its February 1,1965 request. Gouldy, ff. Tr.103, at 4-5.

20. In addition to stating that Licensee could institute a change in the use of administrative controls pursuant to a 10 C.F.R. 9 50.59 analysis, the Staff stated that the conclusions in its Safety Evaluation and supporting Technical Evaluation Report (TER) remained valid because administrative controls were initiated prior to any fuel being loaded in the

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SFP racks with overhanging rows and thus, precluded the possibility of any rack lift-off. Mcdonald, ff. Tr.129, at 7-8 and Attachment 3.

21. The Turkey Point Technical Specifications contain provisions concerning procedures and administrative controls. Section 6, "Adminis-trative Controls," generally require the use of procedures and administra-tive controls to assure that all safety-related structures, systems and components remain within their design basis and can perform their safety f unction. In addition, Section 6.8.1, " Procedures," requires that written procedures and administrative policies be established, implemented and maintained that meet or exceed the guidance of the American National Standards Institute ( ANSI) N18.7-1972 as endorsed by Regulatory Guide 1.33, " Quality / Assurance Program Requirements (Operation) . " Under ANSI N18.7-1972, Section 5.3.4.5, " Fuel Handling Procedures," fuel han-dling operations (which would include the movement of fuel in or about the spent fuel pools) must be performed in accordance with written proce-dures. Furthermore, Section 6.8.3 of the Technical Specifications gov-erns the modification of procedures and permits changes if: (1) the intent of the procedure is not altered; (2) the change is approved by two members of the plant management staff, at least one of whom holds a Se-nlor Operators License; and (3) the change is documented, review by the Plant Nuclear Safety Committee and approved by the Plant Manager. M.

at 9-10,

22. The fuel movement procedure for Turkey Point has been revised to include a restriction which prevents loading of racks with overhanging rows while the remainder of a rack is empty. This procedure is current-ly being used at Turkey Point. The procedure involves the use of a fuel i

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i handling data sheet which designates a specific storage location for each fuel assembly (ideritified by number) and the data sheet is derived from information on fuel status boards which contains a diagram of the spent fuel pool that shows the locations of currently-stored fuel assemblies and available storage locations. Gouldy, ff. Tr. 103, at 5-6. The Board l finds the procedure and the TS controls on its revision are adequate to prevent Icading of overhanging rows while the remainder of a rack is empty.

23. Licensee's scismic analysis of the spent fuel storage rack was i

performed for two cases. Case 1 is predicated on the use of administra-tive controls to prevent loading of overhanging rows while the remaining rows of the racks are empty. That analysis considered full fuel loading (fuel assemblies in all storage locations) and various partial loading condi-tions. The Case 2 analysis assumes that fuel assemblies are loaded in the overhanging locations before the remaining locations are loaded (i.e., the absence of fuel loading restrictions). Flanders, ff. Tr.103, et 14-15.

24. The results of Licensee's analysis of Case 2 are consistent with the OT Position. The methodology used to calculate overturning and stresses is the same that reviewed by FRC and the Staff in connection with the issuance of the rerack amendments. The calculational methodol-i ogy included a general purpose computer code which performs rack re-sponse analysis for the duration of an earthquake and the results of the analysis of the loading of overhanging rows (in the absence of administra-tive controls) satisfy the OT Position. Licensee's calculations and tabu-lated rest ts i show that the total displacements can be easily accommodated by the 9 ,os 2 provided between the racks and between the rack and the l l

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pool wall. In addition, the results show the stresses in the rack and pool are within the limits- specified in the OT Position. Kim, ff. Tr.129, at 11-12; Flanders, ff. Tr.103, at 16-18.

25. Specifically, Licensee s calculated factor of safety against over-turning of 8 is greater than the SRP minimum value of 1.1. Thus, the criteria is satisfied and the results indicate that overturning of a rack is unlikely during an earthquake. In addition, Licensee's calculations of a 0.72-inch total combined displacement attributable to seismic motion and thermal growth shows thct the impact or closure of gaps during seismic motion would also be precluded because the cold gap between the fuel racks (space between the racks prior to insertion of spent fuel assembly and thermal expansion) is designed to be not less than 1.10 inches. Kim, ff. Tr.129, at 12-13; Flanders, ff. Tr.103, at 16-18.

26. Consequently, the Staff concluded that administrative controls on fuel loading are no longer necessary for the Turkey Point spent fuel pools. Kim, ff. Tr.129, at 13. The Board agrees.

27. Interveners argue that the Staff failed to adequately discharge its duties because (a) it did not perform a review of the Case 2 analysis until the hearing and (b) it permitted Licensee to perform a 10 C.F.R. 5 50.59 analysis to determine whether removal of the loading controis in-volved an unreviewed safety question despite an alleged pattern of abuse of the provision by Licensee. Interveners also assert that the Staff should have suspended the amendments upon receipt of the Case 2 analy-l l sis in February 1965 and that the amendments should now be suspended until the Staff performs an " independent review" and makes "a final and L- - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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formal determination" that administrative controls are no longer necessary.

Interveners Proposed Findings at 1 14.

28. We reject interveners arguments. The record shows that the Staff adequately reviewed Licensee's Case 2 analysis against the pertinent acceptance criteria and acted in accordance with the regulations by permitting Licensee to perform a 10 C.F.R. 9 50.59 analysis. There is no evidence or record that Licensee has abused this provision nor did Intervenor offer such evidence, in addition, we find that the sworn Staff testimony, which concludes that loading controls are no longer necessary, is adequate for the Board to rule on this contention.

29. In sum, the evidence shows that Licensee's lift-off analysis shows that the fuel rack stresses would be within ASME code limits, the safety factors for overturning are sufficiently larger than the Staff ac-ceptance criteria, and the total displacement due to seismic motion and thermal growth is less than the cold gap between the fuel racks. Thus, the rack design satisfies the structural aspects of GDC 2, 4, 61, and 62; and there is reasonable assurance of safe storage of the fuel in the event of an carthquake. Accordingly, we find that the administrative controls on fuel loading are no longer necessary.

B. Contention 6

30. Contention 6 states:

The Licensee and Staff have not adequately consid-ered or analyzed materials deterioration or failure in materials integrity resulting from the increased gener-ation and heat and radioactivity , as a result of L - - --_____--------- _ _

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f I l 1 increased capaci y and long term storage, in the spent fue,I pool.

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31. In our March 25, 1987 Order, the Board denied summary dispo-sition of Contention 6 and raised an issue as to "the modes and effec-tiveness of surveillance of materials and the monitoring of the fuel storage pool and contents to provide a measured basis for safety during the ex-tended period of use." SD Order at 33. Our question stemmed from Interveners arguments concerning publications by A. B. Johnson entitled

" Behavior of Spent Nuclear Fuel in Water Storage" (BNWL 2256, Septem-ber, 1977) and " Spent Fuel Storage Experience" (Nuclear Technology, Volume 113, mid-April,1979) .

32. We noted that statements in two Johnson publications indicated that the longest water storage of Zircaloy-clad fuel and stainless steel-clad fuel is about 19 years and 12 years, respectively. SD Order at 32. While Johnson stated that the technology for handling spent fuel has developed over 35 years and has largely been satisfactory, Johnson concluded that expected spent fuel storage of 20-to-100 years would be an incentive to determine whether any sicw degradation mechanisms are oper-ative. ,l d , at 33. The Board also acknowledged the Intervenor's obser-vation that spent fuel presently stored at Turkey Point did not exceed 39,000 MWD /MT but that under the amendments the plant could operate until burnup of 55,000 MWD /MT. Id. at 32, 3/ In admitting this contention, we limited the phrase "long term stor-age" to the storage period authorized by the amendments.

LEP-85-36, 22 NRC 590, 598 (1985).

33. A few months after we issued our summary disposition order,

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we were informed by the Staff about new information concerning Boraflex, a neutron absorber material used in the Turkey Point spent fuel pools.

B N 11, " Board Notification Regarding Anomalies in Boraflex Neutron Absorbing Material," dated July 15, 1987. Therein the Staff informed us that its testimony on Contention 6 would address the new information on Boraflex. B N-87-11, at 3. b Because the issues concerning Boraflex are distinct, we will address our findings on spent fuel pool materials other than Boraflex first and analyze the record with respect to Boraflex second.

(1) Materials Other Than Doraflex

34. The Licensee's direct case on this portion of Contention 6 con-sisted of a panel of Wililam C. Hopkins and Eugene W. Thomas from Bechtel Eastern Power Company (Bechtel). Mr. Hopkins, a Nuclear Engl-neer at Bechtel for 16 years who has analyzed the operation of plant sys-tems in radiation environments, testified on the impacts of radiation on the spent fuel pool liner and concrete structure. Hopkins, ff. Tr.163, at 9-10. Ur. Thomas, an Assistant Chief Civil Engineer for Bechtel with over 23 years engineering experience, including over 17 years at Bechtel as a civil engineer in the design and seismic qualification of fossil and nuclear power plants, testified on the impacts of heat on the spent fuel pool liner end concrete. Thomas, ff. Tr.163, at 12-14.

The Staff also stated that it would evaluate whether its response

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favoring summary disposition of Contention 10 was affected by the (FOOTNOTE CONTINUED ON NEXT PAGE)

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35. The Staff'; y

• ness on this portion of Contention 6 was Clifford David Sellers, a Seri Metallurgist at NRC, with over 36 years experi-ence in metallurgy, ind.4 ding the analysis of materials problems caused by radiation exposure and 15 years at NRC conducting reviews and evaluations of the safety of materials used in the construction and mainte-nance of nuclear power plants. Sellers, ff. Tr.188, Professional Qualifications.

36. The new spent fuel storage racks are constructed of Type 304 stainless steel as the load carrying structure and use sheets of Boraflex (held in place by a thin-walled stainless steel wrapper) on the outer sur-face of the storage cells and between the cells as a neutron absorbing

. material. Type 304 stainless steel is also used in the pool liner and rack feet consist of 17-4 PH stainless steel. The pool structure is concrete composed of cement and aggregate and uses reinforcing bars of carbon steel. The fuel assemblies are constructed of Zircaloy fuel cladding, inconel 718 springs, and stainless steel nozzles and bands. Sellers, ff.

Tr.188, at 3; Kilp and Gouldy, ff. Tr. 222, at 4.

37. Redesign of the spent fuel pool racks increases only the storage capacity of the pool and not the frequency or the amount of newly dis-charged fuel to be placed in the pool during each fuel reload cycle. The rerack design does not change the radioactivity of the newly discharged (FOOTNOTE CONTINUED FROM PREVIOUS PAGE) new information. B N 11, at 2. Staff counsel informed us that Staff's position on summary disposition of Contention 10 was not changed by the new information. Young, Tr. 276-77.

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fuel placed in the. storage pool. The proposed pool storage densification will equip each pool with sufficient storage locations and provide adequate -

storage with full core discharge capacity well into the next century based on a conservatively estimated 18-month fuel cycle. Sellers, ff. Tr.188 at 3.

'38. As a result of the expanded storage capacity, there will be a small increase in radiation exposure and radiation heating to spent fuel pool materials. The increased capacity of the spent fuel pool has little affect on the heat load of the spent fuel pool because older assemblies cause a small increase in maximum pool temperature when newly dis-charged fuel is added to the pool. As these old. elements continue to I age, they contribute less and less to the heat load of the pool. The heat load contribution of all the aged elements under normal conditions is not expected to raise the maximum pool temperature after refueling above 143 F and will decrease thereafter. This maximum pool temperature is within NRC guidelines for maximum exposure temperature to concrete.

Sellers, ff. Tr.188, at 10, 11; Kilp and Gouldy, ff. Tr. 222, at 4-5.

39. Licensee performed two sets of calculations to determine the cumulative gamma and neutron exposures of materials stored for over 40 years in the Turkey Point spent fuel pools. One set of calculations assumed each fuel assembly has an average burnup level of 36,000 mwd /MT U , which is conservative based upon burn-up levels at Turkey Point, and the second set assumed an average burn-up of 55,000 mwd /MTU, which is conservative based on the maximum expected burnup of fuel assemblies at Turkey Point. The results for 36,000 and 55,000 mwd /MTU showed that pool materials would receive cumulative gamma

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4 0 10 radiation doses of 1.9 x 10 rads and 2.9 x 10 rads, respectively.

The cumulative .. neutron radiation dose for the two burn-up levels was 4.8 x 10 neutrons /cm ' and 1.7 x 10 neutrons /cm , respectively.- Kilp and Couldy, - f f. Tr. 222 at 5-10.

(a) Effects of Heat and Radiation on the Spent Fuel' Pool Structure and Liner

40. The generation of heat in the spent fuel pool may cause the concrete pool structure and pool liner to be affected by thermal stresses

- resulting from the temperature differential between the pool water and ambient conditions and affect the integrity of the materials due to the increased temperature of the pool liner and . concrete. Thomas, ff.

Tr.163, at 4. Licensee analyzed the effect of thermal stresses on the pool structure and assumed pool water temperatures up to 212oF (boiling), the maximum temperature the pool water could approach during

j. normal and accident conditions. Licensee also assumed ambient temperature as low as 30 F outside the pool- to conservatively account for the heat loads caused by temperature gradients (i.e., large temperature differences on opposite sides of the pool walls). The analysis showed

-that the pool would maintain its structural integrity even under severe thermal stress conditions of postulated bolling combined with the effects of the design basis earthquake. Licensee's analysis of the effects of the thermal, and mechanical effects of increased spent fuel pool capacity on the liner plate system during normal and abnormal conditions demonstrated that the pool liner will maintain its structural integrity. Thomas, ff.

T r.163, at 5-7.

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41. The Staff and its consultant, FRC, reviewed Licensee's analysis of temperature-ind'uced stresses on the- pool structure and liner and-l concluded that the stresses were acceptable. Staff Exhibit 1 (Safety l

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Evaluation at _10,15; Technical Evaluation Report at 25-26).

42. The temperatures associated with radiation due to the increased fuel - storage capacity will not result in significant deterioration of the concrete pool structures or liner. The performance of the concrete in the spent fuel pools and the dismantling of many concrete structures used as shielding has not produced any evidence of degradation due to radiation heating. Temperatures below 300oF have little effect on the concrete and reinforcing steel of the spent fuel pool. Similarly, the pool liner plate, which is made of Type 304 stainless steel, maintains its stability and integrity in temperatures in excess of 1000oF, which is far above expected pool temperatures. Sellers, ff. T r. 188, at 10; Thomas, ff.

Tr.163, at 8-10.

43. Stress corrosion cracking and intergranular corrosion of austenitic stainless steels that are sensitized (e.g., adjacent to welds) are conceivable mechanisms of degradation of components in spent fuel storage pools. However, tests show that stress corrosion cracking of sensitized steels adjacent to welds in the fuel pool liner would be highly localized and would not lead to gross degradation of the liner, in addition, chlo-ride caused strers corrosion cracking is prevented in the stainless steels at Turkey Point by the controls on chloride levels in the fuel pools.

Sellers, ff. Tr.188, at 6-7.

44. The expanded storage capacity will not result in deterioration of the spent fuel concrete structure and liner due to radiation. The pool l

l liner and pool concrete structure, while in place in the pool for the entire storage period, will have the radiation attenuated by distance from the radiation sources (the aged fuel assemblies) and the shielding afforded by the water. Such attenuated exposure would be well below the threshold for radiation damage to the carbon steel in the pool structure and the I 18 stainless steel, which is in the order of 10 and 10 , respectively for neutron exposure. Concrete is used throughout a nuclear plant for its structural support and radiation shielding characteristics. Gamma radiation has a negligible effect on the mechanical properties of concrete.

21 A concrete structure can also withstand neutron fluences up to 10 2

neutrons /cm without loss of material integrity, which is many orders of magnitude higher than the fluence expected in the Turkey Point spent  ;

fuel pool. Reports on the irradiation of concrete have not identified any defects in concrete which can be traced directly to direct radiation damage. Sellers, ff. Tr . 188, at 5, 10, 14-15; Hopkins, ff. Tr.163, at 5-7.

45. Gamma radiation, the predominant source of radiation in the spent fuel pool, has a negligible effect on the mechanical properties of stainless steel in the pool liner through direct radiation damage mecha-nisms. Neutron irradiation tests have shown that stainless steel can withstand neutron radiation levels which are orders of magnitude higher than those predicted in the Turkey Point spent fuel pools without loss of integrity from direct radiation damage. The effect of nuclear heating on stainless steel is negligible at the levels of radiation which will be present in the spent fuel pool. Thus, there will be no loss of integrity of the pool liner. (Hopkins, ff. Tr.163, at 3-5, 7; Sellers, ff. Tr.188, at 5).

46. Based upon the evidence presented by the Staff and Licensee,

\

the Licensing Board finds that the heat-induced stresses in the Turkey Point spent fuel pool concrete structures and stainless steel liners are acceptable, and that the temperature and radiation levels in the spent fuel pool will not result in any loss of integrity or degradation of the pool concrete or liner.

(b) Effects of Radiation and Heat on Fuel Assemblies and Racks

47. Neutrons cause virtually all irradiated induced changes in Zircaloy, inconel and stainless steel used for fuel assemblies. These ma-terials are essentially unaffected by the alpha, beta and gamma radiation which comprise the major fraction of the radiation in the spent fuel pool.

The primary effect of gamma radiation at the levels expected at Turkey loint on these materials is heating and not structural damage. Kilp and Gouldy, ff. Tr. 222, at 5,11-12.

48. The racks containing the first discharged fuel assemblies can be expected to receive the maximum radiation predicted in the pool. The assemblies are exposed to approximately 10 22 neutrons /cm while in the reactor or radiation that is approximately 8 orders of magnitude greater than the 1.7 x 10 neutrons /cm exposures during 40 year storage of fuel with burn-up of 55,000 mwd /MTU. Stated another way, a 40 year storage dose is similar to one second in the operating reactor. Sellers, ff. Tr.188, at 5; Kilp and Couldy, ff. Tr. 222, at 10-12,15-16; Sellers, Tr. 211-12.

49. Although fuel assemblies will be stored for a longer period of time, the Staff does not anticipate a significant increase in the corrosion occurring in the pool because the rates of most corrosion reactions tend

to decrease with time as protective oxide films form on the metals.

Microstructural change can occur with Zircaloy-clad fuel when the hydro-gen proposed by the reaction between zirconium and water diffuses into metal, forming hydride particles or a hydride phase within the Zircaloy cladding, but with stainless steel fuel cladding, such microstructural changes are not likely to occur. In fact , stainless steels are often considered one of the better barriers to hydrogen diffusion. Further, microstructural changes from solid state diffusion processes do not occur below 500oF in stainless steels. Therefore, little or no microstructural changes would occur in the spent fuel pool materials which is attributable to the extended storage. Sellers, ff. Tr.188, at 5-6; Kilp and Gouldy, f f. Tr. 222, at 12-14.

50. Stress corrosion cracking and integranular corrosion can occur in the storage racks steels adjacent to welds but it would be highly local-ized and would not lead to gross degradation of the steel. Test reactors which use Type 17-4 PH stainless steel in control rod drive mechanisms, and inservice surveillance, have shown no degradation at all of this mate-rial after many years of service in water of similar quality to that in the Turkey Point pools, and a temperature of 145 F. In addition, chloride caused stress corrosion cracking and intergranular stress corrosion is prevented in the stainless steel at Turkey Point by controls on chloride levels in the fuel pools. Sellers, ff. Tr.188, at 6-7; Tr.193-94; Kilp and Gouldy, ff. Tr. 222, at 12-14.

51. The primary source of radioactivity release to the spent fuel pool environment is crud that enters the pool with the freshly discharged l fuel, where it is subsequently removed by the pool water purification

system. The increased spent fuel storage should not increase the crud burden on the demineralizers, since crud release occurs primarily within the first few weeks after the fresh fuel is placed in the pool, and is re-moved by the filter-demineralizers in the water purification well before the next refueling. In the absence of evidence that such crud deposits influence the corrosion of stainless steel, there is no mechanism for the t increased storage of spent fuel in spent fuel pools to result in any serious degradation to the spent fuel pool components or the fuel itself.

Sellers, ff. Tr.188, at 7-8.

52. Leakage and disintegration of spent fuel and its cladding while in pool storage is highly unlikely. In the Battelle Pacific Northwest Lab-oratories report B NWL-2256, Dr. Johnson surveyed the information on behavior of spent fuel in pool storage and found no evidence of degrada-tion of spent nuclear fuel during pool storage after times up to 18 years for Zircaloy-clad fuel and 12 years for stainless steel-clad fuel (as of 1977). The results of surveys for the Nuclear Regulatory Commission, performed by Dr. J., R. Weeks of Brookhaven National Labs, since the issuance of Dr. Johnson's report show that stainless steel-clad fuel has been continuously stored in spent fuel pools since the early 1970's with  ;

no evidence of any failures developing in fuel cladding. Sellers, ff.

Tr.188, at 8.

53. While leaking fuel has been stored in a number of fuel pools, )

uranium oxide fuel pellets have displayed the excellent corrosion resis-tance which has prevented any noticeable additional degradation of these fuel pellets in the pool environment both in high purity water BWR type pools, or in the boric acid pools such as exist at most PWR sites. Should

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a defect develop in a fuel cladding in the reactor, the volatile and soluble

. fission products, n'ormally the alkalis and the halogens, would be released I to the reactor coolant and removed by the reactor coolant purification l l

system. Some small amounts of these materials may enter the pool from fuel that developed defects in the reactor, during the first few months ]

after the fuel enters the pool. These (except for the inert gases) would readily be removed by the spent fuel pool water purification system.

Fuel elements are tested for their leak tightness before being placed in the pool so that the plant staff can determine which fuel elements to be placed in the pool contain defects. Sellers, ff. Tr.188, at 9.

54. In the unlikely event that a defect should develop in the fuel cladding during the first few months of pool storage, gaseous and alkall fission products could be released to the pool and the pool environment.

The spent fuel pool monitors, which are used to monitor the spent fuel pool area, and the cleanup system monitors would detect such a release.

Should a leak develop in a fuel cladding several months after it has been placed in the pool (an unlikely occurrence) and after most of the gaseous fission product activity has decayed, the consequences would be less and would differ little from those associated with stored fuel elements contain-ing known defects. Therefore, the proposed long-term storage does not affect the probability that degradation of the fuel will occur in the pool or that significant amounts of fission products would be released to the pool . Further, the possibility of leakage of the spent fuel element clad-ding and/or the disintegration of the spent fuel during storage is remote because of the excellent corrosion resistance of the spent fuel pellets and ,

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l the zircaloy cladding in the spent fuel pool environment. Sellers, ff.

Tr.188, at 9-10,

55. The 40 years of industry experience with wet storage illustrates that it is a fully-developed technology with no associated major techno-logical problems. Fuel elements have been stored continuously for as many as 25 years without evidence that Zircaloy-clad fuel or stainless steel structural elements degrade significantly during wet storage.

Sellers, ff. T r. 188, at 4; Tr.195; Kilp and Couldy, ff. Tr. 222, at 14-17.

56. Stainless steel clad spent fuel has been stored in PWR spent fuel pools more than 18 years. The exposure in the reactor, which is much greater than radiation levels in the storage pools, represents the maximum radiation exposure any stainless steel can accumulate in a spent fuel pool since the steel is directly against the fuel as the cladding mate-rial. Destructive and visual examination of this material produced no evidence of significant degradation of the stainless steel. Relating these observations to the materials of construction for the storage racks, dem-onstrates that they would also not be subject to any significant degrada-tion over long term use, far beyond the present storage time. Sellers, ff. Tr.188, at 11; Kilp and Gouldy, ff. Tr. 222, at 14-17.

57. Zircaloy-clad rods were examined after nearly 21 years of water storage. A comparison of cladding properties with those measured 20 years earlier on rods from the same fuel assembly showed that no detect- ,

i able changes had occurred in corrosion film thickness, cladding mechani-cal properties and fission gas thickness, cladding mechanical properties and fission gas release. Zircaloy-clad fuel elements which were loaded  !

e into Canada's NPD reactor in 1962 are continuing to operate satisfactorily (i.e. with no apparent degradation) after 22 years of exposure to far greater radiation than any element in the Turkey Point spent fuel pools will receive during residence in the pools. Sellers, ff. Tr.188, at 11-12; Kilp and Gouldy, ff. Tr. 222, at 14-17.

58. Survelliance, as used in the context of materials engineering, means the installation of specifically prepared test specimens which are nondestructively removable for testing after exposure to an environment which may degrade certain material properties. As such, no surveillance of spent fuel pool materials is planned. However, in the broader sense, I

spent fuel pool materials are subject to " surveillance." There is monitor-ing of activity in the spent fuel pool building atmosphere and the spent fuel pool cleanup system which is capable of determining the condition of stored spent fuel. The Licensee also periodically performs routine visual observations inside the fuel storage building and subjects the fuel to in-ventory by underwater television. The condition of the ilner is monitored by the installed leak chase system and procedures exist which require a daily check of the system to determine whether leakage has occurred.

In addition, the Licensee maintains spent fuel pool area monitors to continuously monitor the pool areas and the plant's vent monitoring system to monitor total plant airborne radioactivity released (noble gas, loaine and particulate). Further, the Licensee's use of water chemistry specifications aid in prevention of corrosion of materlats in this pool.

Sellers, ff. Tr.188, at 12-13; Kilp and Couldy, ff. Tr. 222, at 17-19, 43; Gouldy, Tr. 301.

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59. The Board finds that the routine surveillance or monitoring currently performed by the Licensee is adequate to assure safety of the fuel storage pool and its contents during the extended storage period authorized by the amendments.

60. The Board is also mindful that the Commission has concluded that spent fuel can be safely stored in reactor spent fuel storage pool for at least 30 years beyond the expiration of a reactor's operating license.

For example, the Commission found that the cladding which encases spent fuel is highly resistant to fuel failure under pool storage conditions and that corrosion would have a negligible effect during several decades of extended storage. Rulemaking On The Storage And Disposal of Nuclear Waste (Waste Confidence Rulemaking), CLI-84-15, 20 NRC 288, 353-57, 366 (1984).

61. Interveners argued at hearing (e.g., Tr. 93) and in their find-ings that Licensee and Staff witnesses base their conclusions regarding the ability of the pool materials to withstand radiation based upon as-sumptions and engineering judgment rather than field experience. They further assert that because predictions regarding extended storage are based on limited operational experience, an extensive materials survell-lance program is needed to adequately protect the public health and safe-ty. Interveners' Proposed Finoings at 11 17-21. The Board disagrees.

62. A similar argument was rejected by the Commission in the Waste Confidence Rulemaking proceeding. The Ccmmission agreed that the basis for confidence that spent fuel will maintain its integrity during extended storage was based on an extrapolation for storage thirty-years beyond a facility's license from current experience, however, it found that "the

e extrapolation is made for conditions in which corrosion mechanisms are well understood" and "[the] extrapolation is reasonable and consistent with standard engineering practice." 20 NRC at 356-57. We also reject i

interveners' argument because the record shows that the mechanisms for  !

spent fuel material degradation are sufficiently understood and the small increases in spent fuel pool radiation exposures and radiation heating will not significantly affect the integrity of spent fuel pool materials. See Findings 11 38-57.

63. In sum, the evidence shows that the materials in the spent fuel pools will not degrade significantly because of the increased pool storage capacity over any term of years foreseeable for storage at individual plants. Stainless steel racks can be used to the end of life of the plant and experiments have shown that stainless steel, as well as the inconel and Zircaloy in the aged fuel assemblies can be exposed to many orders of magnitude of radiation greater than can be reasonably expected in spent fuel pool racks without significant degradation. In addition, there is no evidence that degradation would occur due to the small increases in radiation or heat to storage pool liners or the concrete structure in spent fuel pools as a result of the increased storage. Thus, contrary to the contention, the Licensee and Staff have adequately considered and ana-lyzed degradation in materials integrity as a result of the increased ca-I pacity and the Board concludes that no additional monitoring or surveillance of materials is needed to provide reasonable assurance of safe storage during the extended storage authorized by the amendments.

1 1

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(2) Boraflex Neutron Absorber Panels x

64. The Licensee presented a panel of William A. Boyd and Dr.

Gerald R. Kilp from Westinghouse's Nuclear Fuel Division and Russell Gouldy from FPL. Mr. Boyd, a Senior Engineer in Nuclear Design with over ten years of nuclear experience and responsib;11ty for the reload core designs for Turkey Point and development and coordination of criti- 3 cality analysis for fuel racks, testified on the impacts of postulated gaps ,',

i in Boraflex on K-effective. Boyd, ff. Tr. 222, at Exhibit A. Dr. Kilp, an Advisor Engineer concerning materials performance questions, testified on the integrity of materials in fuel assemblies and storage racks.

Dr. Kilp has over 30 years experience in the evaluation of reactor materi-als which includes the research, development and deployment of materials ( g in Westinghouse reactors, the selection and evaluation of materials for long-term storage of light water reactor fuel, and responsibility for se-lected materials development prog rams. Kilp, ff. Tr. 222, at 45-47.

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65. The Staff panel consisted of Dr. James Wing, Conrad E.

McCrachen and Dr. Laurence I Kopp. Dr. Wing, a Chemical- Engineer with over 32 years experience.as a practicing chemist, including 15 years at the NRC performing independent assessments of compatibility and cor-rosion potential of materials, testified on materials integrity of Boraflex.

Wing, ff. Tr. 339, Professional Qualifications. Mr. McCracken, the Act-Ing Chief of NRC's Chemical Engineering Branch, also testified on this issue. Mr. McCracken has over 29 years experience in chemistry, corro-sion and mechanical systems operation including (since joining the NRC in 1981) the evaluation of materials compatibility and degradation issues at l

NRC licensed plants, and the evaluation of PWR compliance with NRC chemical and corrosion requirements and material compatibility studies for .

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j ' fcssional Coalifications. Dr. Kopp, a Nuclear Engineer who has more N than 28 years of nuclear experierice which includes performing safety evaluations of reactor core designs and criticality analyses of fresh and t

spent fuel rtorage racks, teshfied .9n the criticality aspects of Boraflex.

Kopp, ff. Tr. SM, Professional Qualifkations.

66. Boraflex, a neutron absorbing material or poison used in the spent fuel storage racks, is made by uniformly dispMaing fine particles of boron carbide in a homogenour, stable matrir of s methylated polysiloxane elastomer (a polymer). Kilp and Gouldy, ff. Tr.I 222, at 23.

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t The geometrical configWation of the racks, boron in the spent fuel pool S

water, and *.he use of Boraflex or other poison materials:n the racks are irtdependent methods of preventing spent fuel pools from becoming criti-cal.1 ;%yd , T r. 330-32.

g 67. There are two regions in the Turkey Point spent fuel pools.

The Reg'on 1 racks are designed to hold fuel assernblies with a maximum enrichment of 4.5%. T[e' Reg'on 2 racks are designed to hold fuel assem-blies with a maximw.i heactivity' equivalent to the reactivity of assemblies having an indial enrichment of 1.5%. The Region 1 spent fuel storage rack modules at Turkey Point are each composed of a number of cells with 4

, Boraflexupwals which run along the length of each of the fou. sides of

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the cell. The Region 2 rack trodules have a somewhat similar strtrture, but spacing between indiv? dual cells is smalle and the density of the l Boraflex panels is lower than in the Region 1 racks. Boyd, ff. Tr. 222, at 4-5. ,

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68. , The regulatory requirements of subcriticality is contained in b l General : Design ~ Criterion (GDC) 62, " Prevention of criticality in; fuel stor-I age and handling." GDC 62 states that criticality in the fuel storage end l handling system shall be prevented by physical systems or processes, ,

l preferably by use' of geometrically safe configurations The NRC's accep- ,

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tance. criterion for- dssuring that GDC 62 is met J5 jhund in the Standard Review Plan (SRP), Section 9.1.2, whir;h requires qd!ntaining a st.irage i array. neutron multipilcation factor (keff) less than or' equal to '0.35 in spent fuel pools during normal and acc7 dent cor.ditions. Therefore, sven for act, . conditions, the Staff requires spent fuel pools to' be at least-5 percent subcr!tical (k,77 'no greater than 0.95) to supply adeq'uate . g margin to assure-that .tne requirements of GDC 62 '{N gff less than 1.01. Is ,'

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met. Kopp, ff. Tr. 339, at 3. ,

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69. The Beraflex captuns yeutrons which would;, have otherwise

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.been available for fission and therefore aids in providihn the required subcriticality margin. The subcriticality margin, when considering both f Boraflex panels and the Technical Specification concenrr.ation of 1950 ppm boron in the spent fuel pool water, is approximately 25 percent (h,ff

• O 75). Kopp, ff. Tr. 339, at 3-4.

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70. Boraflex has undergone extensiveEtegung, to determine the ef-i fects of gamma irradiation < ik var!aus e:wironhents and to verify its structural integelty ' and ' suitability as a ndistron absorbing matertsl. Thet

'h. n ji evaluation ' tests have shown tnat Pr.cafier was unaffected -by the pool wa-s..

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ter ' environment and would . nct be ' degraded by corrosion.

j'O Tests tvere 3 performeci at the . University cd Mich!gan, exposing Boraflex up to 1.03' x ll 10 rads of gamma radiaVon with substantial concurrent neutron flux in

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borated water. . These tests indicated that Boraflex maintained its neutron attenuation capabilities after being subjected to an environment of boratad water and gamma irradiation, irradiation caused some loss of flexibility, but would not lead to breakup of the Boraflex. Long-term borated water soak tests at high temperatures also showed that Boraflex withstands a borated water immersion at 240 F for 251 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of abiUty to maintain a uniform distribution of boron carbide. Wing, ff. Tr.

339, at 4-5; Staff Exhibit 1, 6 2.2 at 7; Kilp and Gouldy, ff. Tr. 222, at 23-24.

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71. At the Turkey Point Nuclear Plant, the spent fuel pool water .

temperatures under normal operating conditions are not expected to ex-ceed 143CF, which is well below the 2400F test temperature. In general,

.the rate of a chemical reaction, which could cause material deterioration, decreases exponentially with decreasing temperature. On the basis of these tests, the Stt.ff did not anticipate any significant deterioration of the Boraflex at the pool under normal operating conditions over the de-sign life of the spent fuel racks. Wing, ff. Tr. 339, at 5.

72. Some materials deterioration or failure in integrity of Boraflex has been found in operating nuclear power plants. Subsequent to the l

Staff's review and acceptance of the Turkey Point spent fuel pool racks, anomalies (minor. physical chanqes or gaps) vere identified in some spent fuel pools using Boraflex as reported in Board Notification B N-87-11.

The Wisconsin Electric Power Company provided the Staff with results of the surveillance program for Boraflex used in the spent fuel pools at

• Point Beach Nuclear Plant, Units 1 and 2. The results showed that the c_______________ _ _ _ _ _

2-inch by 2-inch surveillance coupons, which had a maximum exposure of 10 1.6 x 10 rads of gamma radiation, experienced some physical changes in color, size, hardness and brittleness. A full-length Boraflex assembly, 10 which had a maximum exposure of about 10 rads of gamma radiation, showed far less physical changes than the surveillance coupons. Neither the coupons nor the full-length Boraflex assembly showed any. unexpected change in neutron attenuation properties. The Commonwealth Edison Company also submitted the results of recent inspections of the Roraflex used in the spent fuel pools at Quad Cities Station, Units 1 and 2. The inspections discovered numerous gaps in some Boraflex panels which had been exposed to an estimated radiation dose of 10 9 rads. The Boraflex assemblies showed anomalies in the neutron attenuation profiles. Wing, ff. Tr. 339, at 5-6; Kilp and Gouldy, ff. Tr. 222, at 25-26.

73. In addition, one of the Boraflex surveillance coupons (8-inch by 12-inch) at the Prairie Island Nuclear Generating Plant, Units 1 and 2, showed some slight physical changes or degradations similar to the full-length Boraflex panels at the Point Beach Nuclear Plant. Kilp and Gouldy, ff. Tr. 222, at 26.

74. The exact mechanisms that caused the observed physical degra-dations of Boraflex have not been confirmed. The Staff postulated that gamma radiation from the spent fuel initially induced crosslinking of the polymer in Boraflex and produced shrinkage of the Boraflex material.

When crosslinking became saturated, scissioning (a process in which

. bonds between atoms are broken) of the polymer predominated as the ac-cumulated radiation dose increased. Scissioning produced porosity which allowed the spent fuel pool water to permeate the Boraflex material.

Scissioning and water permeation could embrittle the Boraflex material.

Thus, gamma radiation from spent fuel is the most probable cause of the l

physical degradations, such as changes in color, size, hardness, and I brittleness, that were found in the Boraflex material at the Point Beach plant. While the Staff.could not pinpoint the cause of the gap formation in some Boraflex panels at the Quad Cities Station, the Staff thought it conceivable that full-length Boraflex panels which are physically re-strained could experience shrinkage caused by gamma radiation which could lead to gap formation. "'ing , ff. Tr. 339, at 7.

75. Licensee attributed the gap formation in Quad Cities' Boraflex to a rack design and fabrication process which did not allow the Boraflex material to shrink without cracking. Licensee testified that the fabrica-tion process, which required the Boraflex material to be glued along the entire axial length and' firmly clamped in place to the stainless steel fuel rack walls, did not allow for shrinkage of Boraflex, and, as such, gaps developed. In the Turkey Point racks, Boraflex is held to a stainless steel wall by enclosing it to a wrapper. The wrapper is an enclosure which protects the Boraflex from the flow of water and maintains a clear-ance between the Boraflex and the rack cell wall which is large enough to allow shrinkage, but small enough to prevent dislocation of the panel should it become brittle or crack. Although short lengths of adhesive were used to attach the panels to the wrapper for panels produced by an automated process (to provide temporary support during the spot welding process), none of the Region 1 racks, and only some of the Region 2 racks, were fabricated utilizing adhesive to affix the Boraflex panel to

the wrapper or storage cell. Kilp and Couldy, ff. Tr. 222, at 39-40, as corrected; Gouldy, Tr. 242-44.

76. Commonwealth Edison Company (CECO) hypothesized that Boraflex shrinkage caused by irradiation resulted in sufficient tenslie stress to lead to breakage when it was restrained as in the Quad Cities $

spent fuel rack. B N-87-11, enclosure letter dated May 5, 1987. Bisco Products, Inc., the manufacturer of Boraflex material, informed the Staff that the failure of the neutron absorber may be due to the material's h,

properties or, in the case of the Quad Cities racks, some manufacturing deficiencies such as the tearing of the Boraflex panels during handling.

Based on this information, the Staff inferred that gaps may have been formed at Quad Cities before the panels were exposed to any radiation.

Wing, ff. Tr. 339, at 8-9.

77. Gamma radiation-induced crosslinking and scissioning of the polymer in Boraflex can take place in the spent fuel pool racks of the Turkey Point plant in the presence of spent fuels. Because water can permeate into the Bora flex , especially at the edges of the panel, minor degradations, such as changes in color, size, hardness, and brittleness, can be expected. However, the Staff could not predict with certainty whether or not gap formation will occur. Testing at Point Beach and Turkey Point indicates there are no gaps at accumulated levels of irradia-tion higher than at Quad Cities and there is information which suggests that the Quad Cities gaps may be related to fabrication and design of the racks. Thus, it may be inferred that gap formation may result from a combination of shrinkage due to irradiation and fabrication or rack design deficiencies. In addition, the Staff was not certain whether physical

restraints exist in the Boraflex panels at Turkey Point which are sufficient to cause gap formation. Because the Boraflex panels at the Turkey Point plant were constructed from single sheets, the Staff testified that it did not expect that there were gaps in all the Boraflex panels prior to exposure to radiation from spent fuels, unless the panels were damaged by some means. Wing, ff. Tr. 339, at 10,12.

78. Similarly, Licensee testified that since the design and fabrica-tion process used for Turkey Point is more similar to that used for Point Beach (rather than Quad Cities) and those panels were not restrained from shrinking and did not develop any gaps, it would not expect gaps of significant size or extent to develop at Turkey Point. Kilp and Gouldy, ff. Tr. 222, at 40,

79. The Staff is collecting operating experience of Boraflex from plants that use Boraflex, additional test data from the vendor, and fabri-cation information from spent fuel rack contractors. The Staff will evalu-ate the information to arrive at the cause(s) of the observed gap forma tion . McCracken, ff. Tr. 339, at 10.

80. The results of Licensee's testing on 54 Boraflex panels from storage cells in both Region I and Region ll of the spent fuel pool, that are representative of those storage locations which have received an esti-mated radiation dose of 7.8 x 109rads, the highest cumulated exposure to date. The testing had the capability to detect gaps of 1" to li" or greater. Kilp and Gouldy, ff. Tr. 222, at 33, 39. No indication of gaps, voids, or other spatial distribution anomalies was observed. The l results of this testing also verifies that no gaps existed in these 54 Boraflex pant!s prior to exposure to spent fuel, and that probably no l

I

physical restraints exist in these panels. Therefore, on the basis of all available data and information, and if indeed physical restraints do not exist in the Boraflex panels, the Staff believes that gaps will not likely form in the Turkey Point Boraflex panels. Wing, ff. Tr. 339, at 11.

81. Substantial physical degradation can alter the neutron attenua-tion properties of Boraflex and reduce the neutron absorption effective-ness of the Boraflex panels. Consequently, physical degradation can decrease the margin of subcriticality of the fuel pool. Neutron attenua-tion of Boraflex is mainly due to boron-10 (a boron isotope with a mass number of 10) that is present in the boron carbide power in Boraflex. If the spatial distribution of boron-10 is not disturbed, the neutron attenua-tion properties of Boraflex should remain unchanged. Minor physical degradations, such as changes in color, size (shrinkage), hardness and brittleness, that do not disturb the spatial distribution of boron-10, should not alter the neutron attenuation properties of Boraflex. Howev-er, large gap formation in a Boraflex sheet could alter the neutron atten-uation profile. Wing, ff. Tr. 339, at 11-12.

82. The Staff, recognizing that such gaps may not form in the Tur-key Point Boraflex panels since the factors that contributed to gap forma-tion at Quad Cities may not exist at Turkey Point, conservatively applied the limited Quad Cities data -- the only data available indicating the oc-currence of gaps -- to estimate the potential gap size in the Boraflex panels at Turkey Point. Of the 203 Boraflex panels examined at Quad Cities, 31 gaps were found in 28 pa nels , and two three- to four-inch gaps were found among the 31 gaps. Thus, three- to four-inch gaps, the largest gap size identified, were found in one percent of the panels I

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tested and 6 percent of the gaps examined. This largest gap size was found in Boraflex ' panels having a nominal length of 152 inches which 9 Therefore, if the were exposed to 10 rads of gamma radiation.

conditions which resulted in gap formation at Quad Cities are present, the Staff concluded. that Turkey Point will not likely have gaps greater than four inches in approximately one percent of its Boraflex panels. Wing, ff. Tr. 339, at 12-13.

83. At the Staff's request, Licensee performed a sensitivity study to determine the effect of possible gaps in the Boraflex at Turkey Point on the margin of subcriticality. Since Region 1 of the spent fuel pool contains the higher Boraflex loading as well as the smaller subcriticality margin, the sensitivity study conservatively used the Region I spent fuel rack configuration. As an additional conservatism, the calculations did not take credit for the boron in the pool water, i.e., the racks are flood-ed with pure water. The results indicate that for fuel enriched to 4.5 weight percent U-235, the acceptance criterion of k gf7 less than or equal to 0.95 is met for the case of a 2-inch gap at the same elevation in all of the Boraflex panels in the rack. The acceptance criterion is also met for the case of almost a 4-inch gap at the same elevation in one-half of the Bo.raflex panels (2 of 4 panels in each storage cell in Region l) in the rack. Kopp, ff. Tr. 339, at 13-14; Boyd, ff. Tr. 222, at 6-9.

84. The maximum enrichment of the fuel currently used at Turkey Point is only 3.6 weight percent U-235. Licensee estimates that in ap-proximately three years, the maximum fuel enrichment at Turkey Point will be less than 4.1 weight percent U-235. For fuel of 4.1 weight per-cent enrichment, the 0.95 acceptance criterion would be met for a 3.5

f i

inch gap in all the Boraflex panels and a 7-inch gap in one-half of the panels in the rack. Kopp, ff. Tr. 339, at 14; Boyd, ff. Tr. 222, at 6-9.

85. The Staff considers Licensee's assumptions regarding the distri-bution of gaps to be conservative since if gaps were to develop, they would probably not all occur at the same elevation nor throughout the entire storage location within the racks. In Quad Cities, for example, the distribution of gap sizes ranged from 0 to about 4 inches with the maximum size (between 3 to 4 inches) observed in only one percent of the Doraflex panels tested. Therefore, conservatively assuming that the maximum gap size of 4 inches observed at Quad Cities occurs in 50 per-cent of the panels at Turkey Point, k eff f r the storage rack would only be 0.93 for 4.1 weight percent enriched fuel at Turkey Point. In fact, the acceptance criterion of 0.95 would be met with as much as a 7-inch gap in 50 percent of the Boraflex panels for 4.1 weight percent fuel .

Kopp, ff. Tr. 339, at 14-15.

86. In order to confirm that Boraflex is acceptable for continued use, Licensee had originally planned to perform an initial survelliance of Boraflex specimens after about five years of exposure in the spent fuel pool environment, as described in Section 4.8 of the Turkey Point Units 3 and 4 Spent Fuel Storage Facility Modification Safety Analysis Report, dated March 14, 1984. Two types of examinations will be conducted on Boraflex to examine and evaluate its physical and nuclear characteristics.

First, an in-service surveillance program will evaluate the Boraflex speci-mens in both Region i and Region il of the spent fuel pool for physical and nuclear characteristics, including the determination of uniformity of

_ 41 _

l boron distribution and neutron attenuation measurements. Second, a sur-

^

veillance program will detect any spatial distribution anomalies in the full-length Boraflex panels. Wing, ff. Tr. 339, at 15; Kilp and Gouldy, ff. Tr. 222, at 30-33.

87. - The second surveillance program is referred to as " blackness testing . " These tests are performed using a fast neutron source -and thermal neutron detectors. Any gaps in the Boraflex will be detectable by an increase in the number of thermal neutrons reflected back to the detectors. This method has been used satisfactorily in other spent fuel' pool facilities such as the Quad Cites Station Units 1 and 2 to detect spatial anomalies in Boraflex. By retesting at regular Intervals, any changes in the neutron attenuation properties or in the spatial distribu-tion of the boron-10 in Boraqex should be detected and corrective actions taken should it be determined that gaps large enough to violate the k,ff acceptance criterion may occur. Kopp, ff. Tr. 339, at 16; Couldy, ff.

Tr. 222, at 31-32.

88. In early August,1987, Licensee performed baseline blackness testing on the Boraflex panels that have received the highest cumulated radiation exposure to date. Licensee expects to perform future survell-lance testing of the Boraflex panels within approximately three years, or sooner if industry experience indicates a shorter period for surveillance is warranted. In addition, Licensee made a commitment not to store any fuel with an enrichment greater than 4.1 weight percent U-235 prior to completion of the next surveillance. Kopp, ff. Tr. 339, at 16; McCracken, Tr. 375-76; Gouldy, ff. Tr. 222, at 30-33.

89. The Staff believes that the next proposed surveillance should l include a representative sample of panels subjected to a range of radiation exposures to provide reasonable assurance that fuel with enrichment up to 4.5 weight percent U-235 can be stored at Turkey Point and maintain the 0.95 k eff acceptance criterion. McCracken, ff. Tr. 339, at 17.

90. Initial surveillance testing was performed by FPL during the first week of August 1987 in the Turkey Point Unit 3 spent fuel racks.

Storage locations were chosen in which the Boraflex panels would have experienced the highest accumulated gamma doses to date and, therefore, the largest percentage of shrinkage. No indication of gaps or other spa-tial anomalies were observed. The maximum accumulated gamma dose during this testing was estimated by Westinghouse Electric Corporation, 9

the fuel vendor, to be 7.8 x 10 rads. The next surveillance testing of the Boraflex panels at Turkey Point is scheduled in approximately three years (December 1989) when the maximum accumulated gamma dose is esti-10 mated by Westinghouse to be 1.2 x 10 rads. Wing, McCracken & Kopp, ff. Tr. 339, at 17; Kilp and Gouldy, ff. Tr. 222, at 36; Gouldy, Tr. 310-12.

91. Bisco Products, Inc. submitted additional test data of Boraflex on June 25, 1987 and August 26, 1987. The data showed that shrinkage 9 10 in the Boraflex samples at the dose levels of 5 x 10 and 10 rads of gamma radiation was essentially the same, averaging about 2.1 percent.

0 Irradiation at 2.5 x 10 rads showed an average shrinkage of 2.4 percent. The data indicated that no appreciable change in shrinkage of 9 10 Boraflex material occurred between 5 x 10 and 2.5 x 10 rads. The 54 Boraflex canels tested at Turkey Point had an estimated radiation dose of

L.

l 10 7.8 x 10 rads and an estimated maximum dose of 1.2 x 10 rads in 9

three years. These dose levels are within the range of 5 x 10 and 10 2.5 x 10 rads where no appreciable change in shrinkage was found.

The Staff believes that the proposed Turkey Point surveillance interval is adequate. However, the Staff will continually monitor industry experience with Boraflex to determine whether a shorter time interval is warranted.

Wing, ff. Tr. 339, at 17-18.

92. In the event that Licensee's surveillance program detects degra-dation of Boraflex, there are corrective actions avaliable to maintain the 0.95 k,ff acceptance criteria. These actions include: 1) control of the placement of fuel to. increase the effective spacing between assemblies and thus reduce the k eff value; 2) consolidation of two or more fuel assemblies into one cell location by removing the individual fuel rods and replacing them in a more compact configuration to reduce the number of thermal neutrons available to cause fission and thereby reduce the k,ff value; 3) insertion of control rods or burnable poison rods into the fuel assembly to reduce k eff, with appropriate measures to prevent these movable poisons from inadvertently being removed at a later time; and

4) inserting poison panels into the space between the fuel assembly and the cell wall to reduce k eff. M. at 19; Kilp and Gouldy, ff. Tr. 222, at 37-38.

93. In addition to the Boraflex surveillance, Turkey Point Technical Specification 3.17 requires the minimum boron concentration while fuel is stored in the spent fuel pit to be 1950 ppm and Table 4.1-2 requires that the boron concentration be sampled monthly. NRC calculations have shown that under normal storage conditions at Turkey Point with the pool

_y .

water borated to 1950 ppm of boron, all of the Boraflex panels could be removed and the 0.95 k gf7 acceptance criterion would be met, even with 4.5 weight percent enriched fuel . Therefore, the boron concentration and sampling requirements provide additional assurance of safe fuel stor-age between surveillance of the Boraflex. Kopp, ff. Tr. 339, at 18; Boyd, Tr. 267-69, 271, 328-29.

94. Interveners argue that because the blackness tests performed by Licensee do not establish that no gaps exist in the panels since the test could not detect gaps smaller than 1.5 inqhes (Dr. Turner, Tr. 254), the amendments should be suspended until the absence of gaps is proven by an in-depth testing program. Interveners also recommend that the Board direct the Staff to determine if Boraflex is "an unproven material" for spent fuel pool usage and if the use of Boraflex in the ex-panded storage capacity amendment involves a significant hazard. Inter-venors Proposed Findings at 131.

95. As to the first point, the Board agrees that while the testing may not show that no gaps exist in the Turkey Point Boraflex panels, the record is clear that the K-effective limit for either 4.1 or 4.5 percent fuel enrichment would not be exceeded even if gaps smaller than 1.5 inches exist in all the panels in the pool (Kopp, ff. Tr. 339, at 13-15; Boyd, ff. Tr. 222, at 6-9) and that the presence of boron in the pool water alone.is enough to maintain the subcriticality margin. Kopp, ff. Tr. 339, at 18; Boyd, Tr. 267-69. Thus, we find no safety reason for. suspending the amendments. We also reject Interveners' arguments concerning the need for further Staff actions. The record shows that Boraflex remains acceptable for use at Turkey Point in that no safety significant

degradation is expected and there is an adequate surveillance program to monitor its performance. In addition, since the Staff's determination as to whether an amendment involves significant hazards pursuant to 10 C.F.R. G 50.92 determines the timing of any potential hearing (either before or after the action is taken) and this hearing has established that the Boraflex panels does not pose a significant safety concern, there is no need to revisit the issue of significant hazards.

96. Based on the evidence presented by the Licensee and Staff, no safety significant degradation in the Turkey Point Boraflex panels at Tur-key Point will occur. We find that Licensee's surveillance prog rams ,

which include blackness testing, on Boraflex specimens and panels at specified schedules are adequate to detect physical degradations, including gaps, and will provide reasonable assurance that gap formation will be detected in sufficient time to enable Licensee to take corrective actions such that the NRC acceptance criterion of k eff less than or equal to 0.95 is met. Therefore , Licensee and Staff have adequately analyzed the materials integrity of Boraflex and Boraflex material continues to be acceptable for use in safe storage of the spent fuel at the Turkey Point Nuclear Plant.

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

1. The Licensee's seismic analysis for the new Turkey Point spent fuel pool racks shows that the rack design satisfies the structural aspects

of GDC 2, 4, 61 and 62 and thus, there is reasonable assurance of safe storage of fuel in the event of an earthquake.

2. Contrary to Interveners' assertion in Contention 6, the Licensee and Staff have adequately considered materials spent fuel pool integrity during the storage under the expanded capacity.

IV. ORDER WHEREFORE, in accordance with the Atomic Energy Act of 1954, as amended, and 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 Nos.111 and 105 to License Nos. DPR-31 and DPR-41, respectively, issued by the Office of Nuclear Reactor Regulation on November 21, 1984 shall remain in full force and effect without modification.

IT IS FURTHER ORDERED, pursuant to 10 C.F.R. 5 2.760, that this initial Decision shall constitute the final decision of the Commission thirty (30) days from its date of issuance, unless an appeal is taken in accor-dance with 10 C.F.R. 6 2.762 or the Commission directs otherwise. See also 10 C.F.R. 66 2.785 and 2.786. Any party may take an appeal from this Decision liy filing a Notice of Appeal within ten (10) days after ser-vice of this Decision. A brief in support of such appeal shall be filed within thirty (30) days after the filing of the Notice of Appeal, (forty (40) days if the appellant is the Staff). Within thirty (30) days after the period has expired for the filing and service of the briefs of all appel-lants, (forty (40) days in the case of the Staff), any party who is not an

47 -

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

THE ATOMIC SAFETY AND LICENSING BOARD Robert M. Lazo, Chairman Administrative Judge Richard F. Cole Administrative Judge Emmeth A. Luebke Administrative Judge Dated at Bethesda, Maryland this day of ,1987 Respectfully submitted,

/Y Mitz' ..

T/t_.

Young Counsel for NRC Staff Dated at Bethesda, Maryland this 24th day of December,1987 i4

9 December 24, 1987 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD in the Matter of )

) Docket Nos. 50-250 OLA-2 FLORIDA POWER AND LIGHT ) 50-251 OLA-2 COMPANY )

)

(Turkey Point Plant, Units 3 & 4) ) (SFP Expansion)

NRC STAFF PROPOSED TRANSCRIPT CORRECTIONS -

Date Page Line Change 09/15/87 132 6 " A" to "Q" 133 1 "and our" to "for" 134 8 " Indicated" to "as indicated" 136 5 "19" to "10" 141 14 "AIT teach" to "(AIT)"

150 6 "old spots" to " hold spots" 153 1 " based" to " bases" 154 1 "- " to " evaluation" 12 " variation" to " evaluation" 156 11 "two wrap" to "two racks" l

l 189 4 " David" to "Clifford David" 191 18 "zircaloy" to "Zircaloy" i

22 zircaloy" ,to "Zircaloy" 192 21 "zircaloy" to "Zircaloy" 25 "zircaloy" to "Zircaloy" 193 6 "zircaloy" to "Zircaloy" .

193 23 "zircaloy" to "Zircaloy" 194 25 " fools" to " fuels" 195 2 "with" to " began with" 199 2 " equipment" to " equivalent" 19 " ink canal" to "Inconel" 200 12 "an" to "and" 203 6 "but Tithlated" to "not ilthiated" 207 5 "Well," to "Well, there's a" 208 24 "then" to "they" 211 1 "it" to "that" 212 7 "10 to 17" to "10 to the 17th" 8 "18" to "18th" r

217 3 " wrap" to " rack" g 09/16/87 337 17 " inserting an 'e'" to " reversing the 'e' and *p'"

338 15 "McCRACKEN:" to " YOUNG:"

341 7 " wall?" to "all" -

8 "The wall" to "the" 342 1 "with" to "with a" 25 "(Kopp)" to "(Wing)"

347 18 "their formulation" to " gap formation" 25 " application" to " fabrication" 350 25 "what" to "what was" 351 1 " effective" to "effect of" 4 353 18 " barometric" to "parametri:_"

357 4 "the sizt" to "as high as" l 358 21 "a" to "the" .

I 360 25 "in" to "l"  ;

363 17 "fleid" to " fuel" 365 10 ' "value" to " vague" 11 Hin" to "on" 368 19 "a" to "the" 377 19 "just" to "I just" Respectfully submitted, o

Counsel for NRC Staff Dated at Bethesda, Maryland this 24th day of December,1987

% 1 DOLKETED USWC l

. UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSIO[ E 30 P9 M OFFICE 0? S!MtlW Y BEFORE THE ATOMIC. SAFETY AND LICENSitMC 5tgkE9VICI:

In the Matter of )

) Docket Nos. 50-250 OLA-2 FLORIDA POWER AND LIGHT ) 50-251 OLA-2 COMPANY )

)

(Turkey Point Plant, Units 3 S 4) ) (SFP Expansion)

CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN THE FORM OPF AN INITI AL DECISION" and "NRC STAFF PROPOSED TRANSCRIPT CORRECTIONS" 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, by deposit in the Nuclear Regulatory Commission's internal mail system, this 24th day of December,1987:

Dr. Robert M. Lazo, Chairman

• Richard J. Goddard, Esq.

Administrative Judge Regional Counsel Atomic Safety and Licensing Board USNRC, Region !!

U.S. Nuclear Regulatory Commission 101 Marietta St., N.W., Suite 2900 Washington, DC 20555 Atlanta, GA 30303 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board

• Administrative Judge U.S. Nuclear Regulatory Commission 5500 Friendship Boulevard, Apt.1923N Washington, DC 20555 Chevy Chase, MD 20815 Dr. Richard F. Cole

• Atomic Safety and Licensing Administrative Judge Appeal Board

• Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Michael A. Bauser, Esq. Docketing and Service Section*

Newman & Holtzinger, P.C. Office of the Secretary 1(15 L Street, NW U.S Nuclear Regulatory Commission Washington, DC 20036 Washington, DC 20555

,4 t

Norman A. Coll, Esq. Joette Lorlon I. Coll, Davidson, Carter, Smith 7269 SW 54th Avenue Salter & Barkett, P. A. Miami, FL 33143

( Co-Counsel for FPL -

3200 Miami Center.

100 Chopin Plaza Miami, FL 33131 Mitz A.fYodncf f ((

CouiWI for NRC Staff I

l l

l l

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