ML20205F407

Memorandum & Order (Ruling on Summary Disposition Motions).* Util Motion for Summary Disposition of Contentions 3,4,7,8 & 10 Granted,Nrc Motion Re Contention 4 Granted & Util Motion Re Contentions 5 & 6 Denied.Served on 870326

ML20205F407
Person / Time
Site:	Turkey Point
Issue date:	03/25/1987
From:	Cole R, Lazo R, Luebke E Atomic Safety and Licensing Board Panel
To:	FLORIDA POWER & LIGHT CO., NRC OFFICE OF THE GENERAL COUNSEL (OGC)
References
CON-#187-2900 84-504-07-LA, 84-504-7-LA, OLA-2, NUDOCS 8703310156
Download: ML20205F407 (63)

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00tKETED UNITED STATES OF AMERICA USNRC NUCLEAR REGULATORY CCMMISSION ATOMIC SAFETY AND LICENSING BOARD k7 E3 20 N1 49 Before Administrative Judges: pice c; stcRdm CKETitMi a SEFVILL Dr. Robert M. Lazo, Chaiman ERAllCH Dr. Richard F. Cole Dr. Erineth A. Luebke SERVED Mao 2 6 ton 7

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In the Matter of ) Docket Nos. 50-250-OLA-2 )

) 50-251-OLA-2 FLORIDA POWER AND LIGHT COMPANY )

) ASLBP No. 84-504-07.LA (Turkey Point Plant, ) (Spent Fuel Pool Expansion)

Units 3 A 4)

March 25,1987. _

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1 MEMORANDUM AND ORDER (Ruling on Sumary Disposition Motions)

Before us are notions by Florida Power & Light Company (Licensee) and the Staff of the Nuclear Regulatory Comission (NRC Staff or Staff) for sumary disposition of each of the contentions (Contentions 3, d, 5, 6, 7, 8 and 10) raised by the Center for Nuclear Responsibility, Inc.,

and Joette Lorton (Intervenors) which have been admitted for litigation in this proceeding. Based upon our study of the motions, supporting documents and the pleadings filed in response thereto, we grant the sumary disposition motions directed to Contentions 3, 4, 7, 8 and 10.

The motions for sumary disposition of Contentions 5 and 6 are denied.

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2 I. BACKGROUND OF PROCEEDING On June 7,1984, the NRC published in the Federal Register a notice of consideration of the issuance of amendments to the facility licenses for the Turkey Point Plant, Units 3 and 4, and offered an opportunity for a hearing on the amendments. (49 Fed. Reg. 23715) The amendments allow the expansion of the spent fuel pool storage capacity. By Order of September 16, 1985, the Licensing Board admitted the Center for Nuclear Responsibility, Inc. and Joette Lorion (Intervenors) and seven of their proffered contentions (Contentions 3, 4, 5, 6, 7, 8 and 10).

LPB-85-36, 22 NRC 590 (1985).

On January 23, 1986, Licensee filed a motion for summary disposition of each contention raised by Intervenors accompanied by a statement of material facts as to which it is asserted there is r3 genuine issue to be heard and affidavits concerning each contention.

The Staff response to the motion supported Licensee's motion for summary disposition of each contention except Contention 4. NRC Staff Response to Licensee Motion for Summary Disposition of Contentions, February 18, 1986. Therein, the Staff stated that it agreed that Licensee's motion had established that there was no genuine issue to be litigated as to Contention 4 with respect to whether the offsite does guidelines in 10 C.F.R. Part 100 were met. (Id. at 9) The Staff noted, however, that plant personnel safety in the event of spent fuel pool boiling, an issue which the Staff believed to be raised by the contention, was not

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3 addressed by the Licensee and that the Staff was not. prepared (at that time) to pursue its own motion .for sumary disposition of that issue.

(M. at 10) Thereafter, on July 14, 1986, the Staff submitted "NRC-Staff's Motion for Sumary Disposition of the Personnel Exposure Portion of Contention 4." For the reasons set forth in the motion and in the attached affidavits, the Staff asserts that summary disposition of Contention 4 should also be granted concerning the issue of worker exposure.

II. DISCUSSION Legal Standards for Summary Disposition The Commission's Rules of Practice provide that summary disposition of any matter involved in a licensing proceeding shall be granted if the moving papers, together with the other papers filed in the proceeding, show that there is no genuine issue as to any material fact and that the moving party is entitled to a decision as a matter of law. (10C.F.R. 5 2.749(d))I The use of sumary disposition has been encouraged by the 1

The standards for summary disposition under 10 C.F.R. 5 2.749 are similar to those standards for summary judgment under Rule 56 of the Federal Rules of Civil Procedure. Tennessee Valley Authority (Hartsville Nuclear Plant, Units 1A, 2A, 18, and 28), ALAB-554, 10 NRC 15, 20 n.17 (1979); Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Units 1 and 2), ALAB-443, 6 NRC 741, 753-54 (1977).

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h 4-Consnission and the Appeal Board to avoid unnecessary hearings on contentions for which an intervenor has failed to establish the existence of a genuine issue of ' material fact. See, eg . Statement of Policy on Conduct of Licensing Proceedings,- CLI-81-8,13 NRC 452, 457 (1981); Houston Lighting and Power Company (Allens Creek Nuclear Generating Station, Unit 1), ALAB-590,11 NRC 542, 550-551 (1980);

Northern States Power Company (Prair.ie Island Nuclear Generating Plant, Units 1 and 2), ALAB-107, 6 AEC 188, 194 (1973), aff'd, CLI-73-12, 6 AEC 241,242(1973), aff'd sub nom., BPI v. AEC, 502 F.2d 424 (D.C. Cir.

1974). A material fact is one that may affect the outcome of the litigation. Mutual Fund Investors Inc. v. Putnam Management Co., 553 F.2d 620, 624 (9th Cir. 1977).

When a motion for summary disposition is made and supported by affidavit, a party opposing the motion may not rest upon the mere allegations or denials of his answer but must set forth specific facts such as would be admissible in evidence that show the existence of a genuine issue of material fact. (10 C.F.R. 5 2.749(b)) All material facts set forth in the statement of material facts required to be served by the moving party will be deemed to be admitted unless controverted by the statement of material facts required to be served by the cpposing party. (10 C.F.R. 5 2.749(a)) Any answers supporting or opposing a motion for summary disposition must be served within twenty (20) days after service of the motion. d .)

(_I_d If no answer properly showing the existence of a genuine issue of material fact is filed, the decision

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5 sought by the moving party, if properly supported, shall be rendered.

'(10 C.F.R. 6 2.749(b))

III. CONTENTIONS A. Contention 3 states that:

That the calculation of radiological consequences resulting from a cask drop accident are not conservative, and the radiation releases in such an accident will not be ALARA, and will not meet with the 10 CFP [ sic] Part 100 criteria.

The bases for the contention were given as:

The Florida Power and Light Company did not comply with the conser-vative assumption for a cask drop accident that are specified in the Standard Review Plan 15.7.5 (5) and Regulatory Guide 125 [ sic]

(5), in that they used a 1.0 radial peaking factor, rather than 1.65 factor. Thus, the potential offsite dose using the more conservative calculations could cause FPL to exceed the 10 CFP

[ sic] Part 100 criteria.

The Board notes that in its Memorandum and Order dated September 16, 1985, stated that " reference to the ALARA principle is inappropriate because ALARA generally applies to routine operation, not accidents."

The Licensee, in its affidavit by Rebecca K. Carr, dated ,

January 22, 1986, details the nature of the calculations made to

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determine the offsite doses for a postulated cask drop accident at Turkey Point using appropriate peaking factors. The resultant radiation doses are well within the guidelines of 10 C.F.R. Part 100.-

Two cases were calculated. Case 1, at page 6 of the Carr Affidavit, used a peaking factor of 1.65. Case 2, at page 8, used a peaking factor of 1.0. Case 1 controverts the basis given for the contention that a peaking factor of 1.65 was not used. Licensee's affidavit details the assumptions made in the calculation. A typical refueling at either Turkey Point Unit 3 or 4 would consist of less than half of a full core of 157 assemblies. A radial peaking factor of 1.65 was conservatively applied to all 80 freshly discharged fuel assemblies.

No radial peaking factor was applied to the other 1324 assemblies stored in the spent fuel pool which were discharged during previous refuelings and have decayed for at least 18 months. Other assumptions made in Case 1 were consistent with the assumptions in Regulatory Guide (RG) 1.25. A minimum of 1475 hours0.0171 days <br />0.41 hours <br />0.00244 weeks <br />5.612375e-4 months <br /> of decay time was assumed prior to moving a spent fuel cask into the spent fuel pool.

The thyroid and whole body doses were calculated in accordance with RG 1.25, Sections C.3.a and C.3.b. Offsite doses at the closest site boundary from a postulated cask drop would be 27 ren to the thyroid and less than 1 rem to the whole body. These doses are well within the 10 C.F.R. Part 100 guidelines of 300 ren thyroid dose and 25 rem whole body

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dose.2 That is, the doses are less than 75 rem (25 percent of 300) for the thyroid and less than 6 rem (about 25 percent of 25) for the whole body doses. The Part 100 guideline doses are commonly used in the nuclear industry for evaluating the acceptability of accident condi-tions, and SRP Section 15.7.5, paragraph II, states that the doses calculated for cask drop accidents are acceptable if they are well within the Part 100 guidelines.

The Staff responds with an affidavit by Millard L. Wohl dated February 18, 1986. It explains the application of the NRC's Standard Review Plan (SRP), NUREG-0800, Regulatory Guide 1.25, and HUREG-0612

" Control of Heavy Loads at Nuclear Power Plants," which specify assump-tions regarding radial peaking factors which are acceptable to the Staff for use in analysis of cask drop accidents. In Section 2.5.1 of the Safety Evaluation (SE), the Staff performed an independent analysis to determine the offsite radiological consequences of a postulated cask drop accident at Turkey Point. The results of this analysis indicate the offsite doses for such an accident are 26 rem thyroid and less than 0.1 rem whole body at the exclusion area boundary. These doses are well within the offsite radiological guidelines values specified in 10 C.F.R. Part 100. That is, the doses are less than 75 rem (25 percent of 300) -

2 SRP Section 15.7.5, paragraph II.1, defines "well within" as less than 25 percent of the doses in the 10 C.F.R. Part 100 guidelines.

V-8 for the thyroid and less than 6.25 rem (25 percent of 25) for the whole body dose.

The Staff used a radial power peaking factor of 1.2 which is less than the value of 1.65 recommended in RG 1.25 for a single assembly fuel handling accident. The Staff explains why this is conservative for an evaluation involving damage to multiple fuel assemblies.

It was deemed conservative in NUREG-0612 for heavy load drops involving damage to multiple assemblies. The assumed iodine decontamination factor of 100 as reccmmended in RG 1.25 does not allow for plateout of iodine within the fuel assemblies. It does not allow for credit for the chemical species in which iodine is predominantly expected to occur, which is now believed to be chiefly cesium iodine, which is water soluble and therefore would not escape from the pool in a volatile form. Therefore, the Staff believes this decontamination factor to be conservative. This degree of conservatism would easily outweigh the difference produced in radiological consequence estimates made with a radial power peaking factor of 1.0 which was the peaking factor used by the Licensees, in one estimate. The Staff considers the Licensee's analyses, one of which uses a radial power peaking factor of 1.65 and the other a factor of 1.0, both to be suitably conservative and leading to offsite radiological consequences well ithin the guidelines of 10 C.F.R. Part 100. (Wohl Affidavit, Contention 3, t 8, 9)

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Intervenor, Joette Lorion, in her affidavit dated March 19, 1986, states that her main concern in this contention is the fact that the

' Licensee did not use the 1.65 value recommended in RG 1.25 for analyzing a spent fuel handling. The affidavits by Licensee and Staff state that a value of 1.65 was used in the calculation, Case 1, by the Licensee and the results demonstrated compliance with the Commission's criteria.

Accordingly, the Motion for Summary Disposition of Contention 3 is granted.

B. Contention 4 states that:

That FPL has not provided a site specific radiological analysis of a spent fuel boiling event that proves that offsite dose limits and personal [ sic] exposure limits will not be exceeded in allowing the pool to boil with makeup water from other seismic category 1 sources.

The bases for the contention were given as:

FPL used calculations performed for the Limerick Plant to prove that they would not exceed radiological limits in a spent fuel pool boiling accident. FPL should not be allowed to extrapolate Limerick's study for their own, because there are many differences between the two plants which could be critical. For example, saturation noble gas and iodine inventories could be greater for the Turkey Point Plant as a result of the fuel failure and increased enrichment; more than 1 percent of the fuel rods may be defective because of the ASME fuel failure; and the gap activity of noble gases, such as Krypton 85, and fission products, such as radioactive iodine may also be greater for Turkey Point.

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Thiscontentionwasthesubje.ct of two motions forfumnary g] ,e b

~ disposition. Licensee filed its motion'on. January 23, 1986, whkhwas I '

'r, S opposed by'Intervenor in a filing dated @ arch 19,198G and partially

/ n at that time to address theisubject,of plant'perponnel safety and, '

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accordingly recommended not granting Licensee's motion. Subsequently,A '

s y 3 the Staff did review the subject, and on July 14, 1986 submitted its

s requcst for summary disposition of what it saw as the only remaining portion of Contention 4 - the personnel exposure portion. . Licensee , y supported the NRC motion in a filing of August 8, 1986.

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f' not respond to the NRC motion of July 14, 1986. i i} s.f, . \s i br i Q ['

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In its motion, the Licensee argued and the NRC Staff agreed j as does (q this Board that: I . ,:

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(1) the Turkey Point analysis was not an extrapolation, but wat performed using a methodology similar to that used for t'he LiJiehick plant s

and appropriate site-specific and generic assumptions. Both Sted' and Licensee address the assumptions used in the aralysis, particularly those assumptions related to saturation noble gas and iodine inventories,

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' felled fuel percent, and gap activities of noble gases and io' dine (Carr ni %s.

AC'/Javit on Contention 4, at 4-6; Wohl Affidavit on Contention 4, at j' 3-6); ,

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,onsitegleaseb(Licensee Enotion of 1/23/86, n.12, at 13, Wohl Affidavit w

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g (3) the'offsite dose guidelines of Part 100 are met (Carr Affidavit I Wa at 7. Wohl Affidavit on Contention 4, at 4-7).

t The Board concurs with Licensee and NRC Staff that the Turkey Point -

h 4 analysis was not an extrapolation as proposed by intervenors but was a s

technical analysis which used Turkey Point specific or generic input data in an appropriate manner using as a guide the analysis previously f* T performed for the Linerick Plant. The Board finds no fault with this

'; commonly used procedure. The Board also agrees that Part 20 does not apply to onsite accidental releases. (As discussed briefly below the

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O \ subject of worker protection during accident conditions is the subject of 1 Commission Rulemaking and as such can not be the subject of evidentiary hearing'in individual licensing cases. Potomac Electric Power Company f (Douglas Point Nuclear Generating Station, Units 1 & 2), ALAB-218, 8 AEC 79,85(1974); Sacramento Municipal Utility District (Rancho Seco Nuclear

< Generating Station), ALAB-655, 14 NRC 799, 816 (1981). As regards 1

meeting the offsite dose guidelines of 10 C.F.R. Part 100, the Board

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finds that both Licensee and NRC Staff conservatively calculated the doses and adequately demonstrated ccmpliance with khe requirements.

s (See t Board discussion of Use' Of Regulatory Guide 1.25, infra) a '

The NRC Staff motion of July 14, 1986 deals only with the onsite

'j personnel exposure portion of Contention 4. The Staff reviewed Licensee's emergency procedures, health physics program. and radioactivity and water level / temperature monitoring systems and found them to be adequate to a.

protect onsite personnel and minimize the radiological-consequences of a postulated spent fuel pool boiling event. (Wohl Affidavit on personnel 1 exposure at 2-4; Minns Affidavit at 2-10) The Board agrees.

In its response to Licensee's motion, Intervenor argued that Licensee failed to meet its burden on at least two issues. They were: the appropriateness of using Regulatory Guide 1.25 in analyzing for offsite doses and complis dith 10 C.F.R. Part 100. Concerning the use of.

Regulatory Guide 1.25, Intervenors contend that it is generally used for evaluating dose releases of damaged fuel assemblies with burn-up less than 38,000 t1Wd/MTU and its use is not appropriate to calculate offsite doses for the Turkey Point fuel which is allowed to have a burn-up of 5U,000 mwd /MTV. With respect to compliance with 10 C.F.R. Part 100 Intervenor argues that the deficiency concerns the lack of data to prove that onsite doses to workers resulting from a spent fuel pool boiling event will meet the 10 C.F.R. Part 100 limits.

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- As 'regards the two points raised by Intervenor in its objection to Licensee's motion, it is the Board's view that each of these points is ,

adequately addressed in Licensee's and Staff's response and nothing i

- remains to be litigated. Those two points, the appropriateness of using

. Regulatory Guide 1.25 and compliance with 10 C.F.R. Part 100, are discussed below along with a discussion of makeup water sources.

- Use of Regulatory Guide 1.25 Safety Guide 25.(RG 1.25 dated liarch 23,1972), entitled

" ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUEL' HANDLING ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY FOR BOILING AND PRESSURIZED WATER REACTORS," specifically covers fuel

- handling accidents, but the assumptions used in that evaluation may be applied t'o other accidents involving the release of radionuclides such as a spent fuel pool boiling event. .(WohlAffidavitat3-4) Intervenor argues .that the Regulatory Guide inventories are applicable to assemblies

- with burnup up to 38,000 mwd /MTU batch average at discharge. Since .

. Licensee used an upper limit of 50,000 MVJ/MTV, Intervenor says Regulatory Guide 1.25 is not appropriate. The NRC Staff says that Regulatory Guide 1.25 can be used in evaluating accidents involving fuel -

with burnup greater than 38,000 but with modified gap and plenum fractional volatile radionuclide inventories. (Id.at4,5) Licensee performed its analysis using the gap activities specified in Regulatory Guide 1.25 (10 percent of the iodines and noble gases except for 30 r

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r 14 percent of Krypton-85) and a 1 percent fuel failure assumption. The result of. the Licensee's calculation resulted in 0-2 hour doses of 0.28 rem to the thyroid and 0.00018 rem to the whole body at the Exclusion Area Boundary. This compares with 10 C.F.R. Part 100 guideline values of 300 rem thyroid and 25 rem whole body (less than 0.1 percent and 0.001 percent respectively). The NRC Staff made an independent evaluation and calculations using the guidance in Regulatory Guide 1.25 using however the conservative assumption that all the stored fuel assemblies had 22 percent gap iodine activities, corresponding to a burnup of 50,000 mwd /MTU, with the concomitant assumption that each assembly had operated at the maximum allowable heat generation rate for its entire lifetime and calculated a 0-2 hour thyroid dose of 0.85 rem at the Exclusion Area Boundary (less than 0.3 percent of the 10 C.F.R. Part 100 guidelines value for the thyroid). The Staff used a 1 percent fuel failure value, a common assumption in accident analyses. (Wohl Affidavit at 6)

The Board notes that none of the filings submitted by the Staff referred to Staff calculations of the 0-2 hour whole body dose at the Exclusion Area Boundary (EAB) attributable to a spent fuel pool boiling event although both thyroid and whole body 0-2 hour doses associated with various other spent fuel pool accidents were calculated and reported by the NRC Staff (cask drop /tip, construction accidents, fuel handling accident - SE at 2.5.1 through 2.5.3). Considering that Licensee's fuel pool boiling accident release calculations predicted a whole body dose at EAB of 0.00018 rem (a small fraction of the allowable 25 rem) it is not

15 surprising that the NRC Staff might not have repeated ~ the calculation for whole body dose using the-larger gap iodine activities associated with 50,000 mwd /14TU burnup. It is the Board's view that while it would have been helpful in disposing of this issue to have a whole body dose value based on the higher burnup value, it is not necessary in order to reach a conclusion concerning compliance with the whole body dose standard of 25 rem. (The difference in using the higher burnup values would have to account for an almost 14,000 fold increase in whole body dose rate, while the actual difference is unlikely to be larger than a few fold.)

Onsite Doses and 10 C.F.R. Part 100 The dose guidelines in 10 C.F.R. Part 100 address the protection of the public from accidental releases that extend offsite and are not applied to onsite worker exposures. Likewise, 10 C.F.R. Part 20 limits are not used by the NRC Staff as accidental exposure occupational limits, but as limits applicaole to radiological exposures acquired during routine plant operational and maintenance duties. The NRC does not impose onsite personnel exposure limits for accidental releases. Wohl Affidavit on Personnel Exposure at page 2, No. 3. The Commission is L currently considering a proposed rule regarding " Standards of a Protection Against Radiation," 50 Fed. Reg. at 51992 (1985), which specifically deals with protection of workers during emergency and accident conditions. In that proposed rulemaking, the Commission is considering whether to require, as part of the licensing process,

16 licensees to develop contingency plans for accident and emergency conditions.(See 50 Fed. Reg. at 52010 and 52028 (proposed 10 C.F.R. 920.9)) Notwithstanding the absence of regulatory requirements, the Licensee does have guidelines on doses received by emergency workers responding to emergency conditions such as spent fuel pool boiling. (Id.

at 4, Minns Affidavit at 5-10)

The NRC Staff evaluated the effect of a postulated spent fuel pool boiling accident on plant personnel and demonstrated that onsite personnel will be adequately protected. (Wohl Affidavit on Personnel Exposureat2-4,MinnsAffidavitat2-10)

Seismic Category 1 Makeup Water Sources The existing cooling systems for the spent fuei pools are not safety grade and there are no connections to the shutdown cooling system or other safety related cooling systems. The Licensee has committed to 1

upgrade the Spent Fuel Pool (SFP) cooling system such that they will remain functional after a safe shutdown earthquake. The necessary upgrade will be completed by the end of the second refueling outage after approval of the re-racking. (SE2.7) Because the existing cooling ,

system is not safety grade, all pool cooling water was assumed lost following a safe shutdown carthquake. In such an event, boiling would occur after 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the normal heat load condition and after 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the maximum heat load condition (full core offload) for the new

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17 racks. This would result in a boiloff rate of 37 and 72 gpm, respectively. (Letter, Williams to Varga, October 5, 1984, L-84-264) 1 The rerack will result in no significant change in the time to boiling under the presently authorized storage. Until the upgrade is complete,the amount of fuel that will be stored will be less than the capacity of the existing racks. Multiple alternate sources of makeup water.are available until the sjstem is seismically upgraded. Temporary connections can be provided from the fire water system or the primary water storage tank. Additionally, there are two firehouses nearby and could be available in less than one hour and provide makeup water to the pools. Staff concludes and the Board agrees that adequate time is available to provide the necessary makeup water. (Seat 2.7)3 The Board grants both Licensee's and NRC Staff's motions on this contention because there are no genuine issues of material fact related to Contention 4 remaining for litigation.

3 The Board takes note of IE Information Notice No. 87-13, dated February 24, 1987, " Potential For High Radiation Fields Following Loss Of Water From Fuel Pool." This notice has potential relevance to the Turkey Point plant and illustrates a way in which a single failure resulted in substantial loss of water from the spent fuel pool and a potential for a high radiation field caused not by spent i- fuel but by irradiated control rods stored on short hanger rods clipped over the side of the spent fuel pool. Copies of this Notice have been sent to all PWR and BWR license holders.

18 C. Contention 5 states that:

That the main safety function of the. spent fuel pool, which is to maintain the spent fuel assemblies in a safe configuration through all environmental and abnormal loadings,.may not be met as a result of a recently brought-to-light unreviewed safety question involved in the current re-rack design that allows racks whose outer rows overhang the support pads in the spent fuel pool. Thus, the amendments should be revoked.

The bases for the contention were given as:

In a February 1,1985 letter from Williams, FPL, to Varga, NRC, which describes the potential for rack lift-off under seismic event conditions [ sic]. This is clearly an unreviewed safety question that demands a safety analysis of all seismic and hurricane conditions and their potential impact on the racks in question before the license amendments are issued, because of the potential to increase a possibility of an accident previously evaluate [ sic],

or to create the possibility of a new or different kind of accident caused by loss of structural integrity. If integrity is lost, the damaged fuel rods could cause a criticality accident.

In admitting Contention 5, the Board excluded hurricane conditions as a basis and limited the Contention to whether the fuel can be stored safely in view of the potential for a lift-off during seismic events.

The issue as stated by the Board in its Order is whether there is a deficiency in the current rack design and a necessity for a restriction on loading to prevent potential lift-off. In its motion for summary disposition of January 23, 1986, Licensee argues that it performed a seismic analysis of the new spent fuel storage racks for Turkey Point in accordance with appropriate NRC Staff criteria. The results of their analysis demonstrated that the loads and stresses in the rack would be

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-19 within ASME code allowable limits; that the racks would not impact each other or the pool walls as a result of sliding of the racks; and that there would be adequate margins of safety against tilting of the racks.

Licensee argues that because there is no genuine issue regarding these material facts it is entitled to summary disposition of Contention 5.

Its motion is supported by the affidavits of Harry E. Flanders and Leonard T. Gesinski. Mr. Flanders addresses the structural integrity of the Turkey Point Units 3 and 4 spent fuel pool storage racks while the Gesinski affidavit addresses the structural integrity of the fuel assemblies.

The NRC Staff in its response of February 18, 1986 agrees that Licensee's statement of material facts demonstrates that there are no factual issues to be litigated. The Staff offered no opinion on Licensee's conclusions as to' the amount of lift-off without administrative controls while stating that they agreed that the administrative controls, which prohibit the loading of the overhanging rows while the remaining portions of the racks are empty, would preclude rack lift-off during a seismic event.

Intervenors in a filing of March 19, 1986 opposed summary disposition of Contention 5, stating that the new spent fuel racks do not meet Gereral Design Criteria 2 and 61 because without administrative controls on loading there is no assurance that they will withstand an an earthquake or other seismic load. It is Intervenors firm belief that

20 administrative controls should not be a substitute for spent fuel pool equipment meeting NRC criteria. Intervenors do not contest Licensee's and Staff's position that the imposition of administrative controls would preclude lift-off of the racks during a seismic event.

The NRC Staff's issuance of the amendmerts authorizing the spent fuel pool expansion was based upon its review of the results of a Licensee evaluation which showed that the spent fuel pool storage racks would not lift off the pool during a seismic event. In Section 2.3.6 of the NRC Staff Safety Evaluation supporting the amendments, the Staff concluded that the design of the racks satisfied the structural seismic requirements of General Design Criteria ?, 4, 61 and 62 of 10 C.F.R. Part 50, Appendix A. Kim Affidavit at 7.

The Staff agrees that the administrative controls which prohibit the loading of the overhanging rows while the remaining portions of the racks are empty would preclude rack lift-off during a seismic event and thus, the conclusion in the Staff Safety Evaluation remains valid. (Kim Affidavitat6,8) Subsequent to the Staff's Safety Evaluation which was dated November 21, 1984, in a letter dated February 1,1985, Florida Power and Light presented the analysis of the potential for lift-off during a seismic occurrence in the event that the outer rows of the racks which overhang the support pads were fully loaded while the rest of the racks remain empty. This analysis showed that rack lift-off could occur but that the results of such lift-off would be acceptable.

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21 Licensee requested that the NRC review the results of this analysis and concur that the analysis is acceptable. NRC has not reviewed the results of this analysis. In a letter dated February 26, 1985 from the NRC to Licensee, haC stated that Licensee's request for review of the analysis represented a change in the NRC basis supporting issuance of the amendments which authorized the Turkey Point Spent Fuel Pool Expansions. The NRC further stated that Licensee could make such changes without prior NRC approval provided that a review performed in accordance with-the provisions of 10 C.F.R. 5 50.59 determined that neither a technical specification change nor an unreviewed safety question is involved. The NRC also stated that it would not take further action on FPL's request until it received clarification with respect to whether FPL had performed an analysis pursuant to 10 C.F.R. 5 50.59. In a letter dated November 13, 1985, Licenseecdn withdrew its February 1,1985 request for review and stated that it would review any change in the basis supporting issuance of the amendments in accordance with the provisions of 10 C.F.R. 5 50.59.

There is no question in the Board's mind that properly executed administrative controls which effectively preclude the loading of overhanging rows, while the remaining portions of the racks are empty, would prevent rack lift-off during a seismic event. Licensee's seismic analysis of the spent fuel storage racks was performed for two cases.

The first case is predicated upon the existence of administrative controls to prevent the loading of overhanging rows while the remainder

l 22 i of the rack is empty. In the second case it was assumed that administrative controls did not exist and that the overhanging rows were loaded while the renainder of the rack is empty. The Licensee states that in both cases the results of the analysis demonstrated that the fuel rack stresses would be within the ASME code allowable limits and -

that the maximum displacement of the fuel racks would be less than the size of the gap between adjacent racks and between the racks and the pool wall, thus precluding impact between the racks- or between the racks and the wall. In Case 1 it was demonstrated that rack-lift-off from the floor would not occur. In Case 2 a lift-off of 0.18 inches was predicted to occur. On this point Licensee stated that lift-off of free standing racks under seismic conditions is not uncommon and the structural members of the racks are designed to accommodate the stresses produced by lift-off. A minimum factor of safety against overturn in Case 2 was calculated to be 8 which is substantially greater than the 1.5 minimum factor of safety referenced in the NRC position paper ("0T Position, for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978 and amended January 18,1979).

The position of the NRC Staff on the ma.tter of lift-off and horizontal motion of the racks during a seismic event absent administrative controls (Case 2) would be helpful in assessing the need for an administrative controls requirement attachment to the already issued license amendments. The Staff has articulated no position on that issue in any filings before this Board. Two references to Case 2

23 appear -in Staff filings, one being in a footnote stating that it offers no opinion as to the amount of lift-off without administrative controls (Licensee's Material Fact No.15) but because administrative controls are in effect the balance of Licensee's material facts (Hos.1 - 14,16,

17) support summary dismissal of the contention. (Staff Response, n.3, 2/18/86) The other reference is -in the affidavit of Sang Bo Kim in which the following appears..." With the exception of the statements regarding the Licensee's analysis for Case 2 (Fact No.15), the material facts stated in relation to Contention 5 are correct and I concur in the conclusions reached in the supporting affidavits. Since the Staff has not reviewed the analysis for Case 2, I offer no opinion in that regard." The Staff position on the Case 2 matter is that since the basis for the conclusion (of no lift-off) in the SE and appended Technical Evaluation Report (TER) remain valid with the imposition of administrative controls the basis for issuing the amendment also remains valid. Without administrative controls, the licensing basis would not be supported. (2/26/85 Letter Mcdonald to Williams) In this same letter, Mcdonald advised Licensee that under 10 C.F.R. 50.59 changes can be made without prior Commission approval if the proposed change, test or experiment does not involve a change in technical specifications or an unreviewed safety question. As a result of that letter, Licensee withdrew its request for Staff review stating that Licensee will review any change in the basis supporting issuance of the amendments in accordance with the provisions of 10 C.F.R. 5 50.59.

[-

24 l A brief review of the Staff's Safety Evaluation and appended Technical Evaluation Report raises a question in the Board's mind as to whether the seismic analysis considered overhanging rows. (See TER, Fig. I and 2, at 6-7). It appears that the structural analysis upon which the license amendment was issued was an analysis of the new racks.

-in a fully loaded condition, a condition which normally would produce the largest stresses. The conditions found by Westinghouse and

-described in Williams' February 1,1985 letter to NRC's Varga were not known by the NRC Staff and apparently not a consideration in the issuance of the amendment. There are no fuel loading restrictions related to overhanging rows in the Technical Specifications or in the

-language.of the Amendment.

It appears to the Board that there are sufficient doubts as to the basis for the issuance of the Amendments, particularly the structural analysis involving the safe shutdown earthquake and various loading conditions other than fully loaded and involving the overhanging rows, conditions which the Staff has apparently not evaluated, that sunmary disposition of this Contention must be denied.

D. Contention 6 states that:

The Licensee and Staff have not adequately considered or analyzed materials deterioration or failure in materials integrity resulting from the increased generation and heat and radioactivity, as a result of increased capacity and long-term storage, in the spent fuel pool.

~ k -.

25 1

The bases for the contention were given as:

The spent fuel facility at Turkey Point was originally designed to store a lesser amount of fuel for a short period of time. Some of the problems that have not been analyzed properly are:

(a) deterioration of fuel cladding as a result of increased exposure and decay heat and radiation levels during extended periods of pool storage.

(b) loss of materials integrity of storage rack and pool liner as a result of exposure to higher levels of radia-tion over longer. periods.

(c) deterioration of concrete pool structure as a result of exposure to increased heat over extended periods of time.

The Licensing Board limited the phrase "long-term storage" to the

'" storage period authorized by the amendments." (MemorandumandOrder, September 16, 1985, at 14)

The Licensee details the calculations of water temperature and radiation dose to the spent fuel pool structures in affidavits by Dr. Gerald R. Kilp, Rebecca K. Carr, Eugene W. Thomas, and Daniel C.

Patton. Under normal conditions, the temperatures are not expected to exceed 143 F and will usually be less. Temperatures of the water in the pools could reach boiling during a postulated loss of cooling accident.

For materials stored for 40 years in the spent fuel pools, the

26 cumulative gamma dose was calculated to be 1.9 x 1010 Rads and the.

cumulative neutron fluency was calculated to be 4.8 x 10 13 n/cm2 . Alpha and beta radiation are not a concern because they do not have an ability to penetrate materials deeply enough to appreciably affect the structural integrity of these materials. (Kilp Affidavit, 11 5-6; Patton Affidavit, 11 9-11, 13-15)

The fuel assemblies and cladding can be stored safely in excess of 40 years in the Turkey Point spent fuel pool. The fue'l assemblies for Turkey Point are composed of Zircaloy, Type 304 stainless steel, and Inconel. The fuel assemblies and cladding are designed to withstand the radiation levels and heat loads present in a reactor. The levels in the spent fuel pools are much less severe. The neutron flux in the spent fuel pools, which is the cause of virtually all of the radiation induced changes in Zircaloy, Inconel, and stainless steel, is eight orders of magnitude less than those in the reactor core during full power opera-tion. This radiation will have an insignificant impact on the integrity of the fuel cladding. The corrosion rate of Zircaloy is approximately 1/100,000 inch per year at 500 F and at the higher heat fluxes in a reactor. It is substantially lower at the much lower temperatures pre-dicted for the spent fuel pools. The corrosion rate for 304 stainless steel has been shown not to exceed 6/10,000 inches per 100 years in an oxygenated borated water environment similar to that in the spent fuel pools. The corrosion rates for Inconel are at least as low as those for stainless steel. Stress-corrosion cracking, hydriding, and galvanic

27 attack are not expected to have any impact on the structural integrity of the-materials. Spent fuel has been stored safely for more than three decades. No measurable changes due to corrosion or hydriding and no loss of integrity has been found after hot cell examination of fuel stored for more than 10 years. Thus, over a 40 year period in the spent -

fuel pool, corrosion would not have any appreciable impact on the struc-tural integrity of the fuel assemblies and cladding. (KilpAffidavit, 11 7-14; Ref. 5, A.B. Johnson)

The spent fuel racks may be expected to maintain their material integrity under the conditions expected in the spent fuel pools for Turkey Point. The spent fuel storage racks are constructed of Type 304 stainless steel and contain a neutron absorbing material called Bora-flex. Type 304 stainless steel is virtually immune to corrosion at spent fuel pool temperatures. The neutron radiation levels in the spent fuel pool are orders of magnitude below those levels sufficient to produce any appreciable impact upon the structural integrity of stain-less steel. Boraflex is a silicone-based polymer containing boron carbide. Tests have shown that Boraflex retains its neutron attentua-tion capabilities after being exposed to an environment of borated water and gamma and neutron radiation levels substantially exceeding those anticipated for 40 years of fuel storage at Turkey Point. (Kilp Affi-davit, 11 15-19)

28 No appreciable deterioration or loss of integrity of the spent fuel pool liner will occur as a result of its long-term exposure to heat and radiation levels in the Turkey Point spent fuel pool. The spent fuel pool liner plate is composed of Type 304 stainless steel. This stain-less steel is known to withstand the more severe conditions in the nuclear power plant core. Gannu radiation has a negligible effect or.

the mechanical properties of non-organic materials such as stainless steel. Neutron irradiation tests have damonstrated that stainless steel can withstand neutron fluences orders of magnitude higher than those predicted for the spent fuel pools. Stainless steel maintains its integrity and long-term stability at temperatures in excess of 1000*F, which is far above the temperatures expected ~in the spent fuel pool.

(Carr Affidavit on Contention 6, 11 4-5, 7; Thomas Affidavit, 11 11-12, 16)

No appreciable materials degradation of the reinforced concrete pool structures at Turkey Point is expected. The spent fuel pool struc-ture consists of reinforced concrete, which is a material commonly used-in the nuclear industry. Concrete structures can withstand neutron fluences orders of magnitude above those expected in the Turkey Point spent fuel pool. Gcmma radiation has a negligible effect on the mechanical properties of concrete. Temperatures below approximately 300 F have an insignificant effect on the properties of the type of concrete materials used in the Turkey Point spent fuel pool structures, and the reinfcrcing steel in the structures maintains its integrity and

29 stability atLtemperatures far above that which will be experienced by the Turkey Point spent fuel pool structures. (Carr Affidavit, 11 6-7; Thomas Affidavit, 11 11, 13-16)

The thermal stresses imposed on the pool structure were analyzed by the Licensee. The thermal effects of the increased capacity of the spent fuel pools results in only minor variations in the original design condition. The most severe thermal loads on the structures are caused by the difference between the ambient temperature outside the pool and the temperature of the pool water. Thermal stresses resulting from this differential were calculated assuming a boiling temperature for the. pool water and a steady state outside ambient temperature of 30*F. This is extremely conservative given the south Florida location of Turkey Point and the time required to develop a steady state temperature gradient in the 3-foot thick pool wall. Using methods addressed in American Con-crete Institute Committee Report 349, stresses in the walls and floor of the spent fuel pool were calculated and shown to be within the licensing condition imposed on the original design. An analysis was conducted to determine the effects of thermal, hydrostatic and hydrodynamic loads on the liner plate system. This analysis also showed that there would be no loss of function of this liner. (Thomas Affidavit, 11 6-10)

Licensee concludes that the fuel assemblies and cladding are ,

designed to withstand the conditions in the reactor, which are far more severe than those in the spent fuel pools, The structural materials

l 30 l

used in the storage racks, pool liner, and pool structure are widely used in the nuclear industry and have a demonstrated ability to with-stand the radiation levels and heat loads expected in the Turkey Point spent fuel pools, Consequently, no appreciable deterioration of the materials in the pool is expected, and the materials will maintain their functional integrity.

The Staff responds with an affidavit by Bernard Turovlin dated July 1, 1986. It addresses any potential for loss or material integrity due to the increased fuel storage capacity and the longer term storage of some fuel elements. The Staff agrees that there are no genuine issues to be litigated in Contention 6 and supports Licensee's motion for sunnary disposition.

Each of the Turkey Point spent fuel pools consists of a reinforced concrete pool with a stainless steel liner and racks constructed of stainless steel. The new spent fuel storage racks approved by the amendments are also constructed of stainless steel, but contain sheets of Boraflex, a neutron absorbing material on. the outer surface of the storage cells and between the cells. The Boraflex sheets are held in place by a thin-walled stainless steel wrapper. The other metallic materials in the pool are the Inconel and Zircaloy parts of the fuel element assemblies. (Turovlin Affidavit,17) i l

i

31-The Staff previously approved the use of stainless steel and con-crete in the spent fuel pools through'the end of the- plant's life.

Experiments have shown that stainless steel, as well as the Inconel and Zircaloy, can be exposed to radiation much greater than can be reason-ably expected at the rack structure without significant degradation.

There is no evidence of storage pool liner or concrete structure degra-

- dation due to radiation in spent fuel pools worldwide. (Turovlin Affidavit, 1 10-14) The extended storage of aged elemerits in the pool due to re-racking does not contribute significantly to the heat load in the spent fuel pool. (Turovlin Affidavit, 11 3, 5) The materials in the pools can be exposed to temperatures up to 180*F for extended periods without detectable degradation. Tests show that the Boraflex, the neutron absorbing material, will not undergo significant degradation during its expected service life. The materials in the spent fuel _ pools will not degrade significantly as a result of the expanded storage

' capacity and extended storage. (TurovlinAffidavit, 11 10-14)

Section 2.2.2 of the Staff's safety evaluation concludes that the pools comply with GDC 61 and 62 of 10 C.F.R. Part 50, Appendix A. The environmental compatibility and stability of the materials in the spent fuel pools is adequate based upon test data and actual operating experi-ence and the Licensee's materials surveillance program to assure that no unexptcted corrosion or degradation will compromise rack integrity is acceptable.

-,m- , m

32 Intervenor, Joette Lorion, state.s that the matter remaining at issue in Contention 6 is concerned with the effect that.long-term stor-age of spent fuel will have on the materials _ integrity of the spent fuel pool and its contents. Intervenor claims that experience with spent fuel storage is immature. While Intervenors do not have the document by A. B. Johnson, referenced by Licensee (Kilp Affidavit, Ref. 5), that states that fuel has been stored safely for more than 3 decades, they do have an article by Mr. Johnson, Spent Fuel Storage Experience, Nuclear Technology, Vol. 43, mid-April 1979. In this article Dohnson states that the world's oldest Zircaloy clad fuel in water storage is approach-ing 19 years, and the oldest stainless steel clad fuel about 12 years.

That experience with spent fuel has been favorable so far. Intervenor states that the spent fuel that has been stored safely to date had a burn-up that did not exceed 39,000 mwd /MTU and that Turkey Point fuel is allowed to have a burn-up of 55,000 mwd /MTU. (Intervenor's Response to Licensee's Motion at 3-4) l Intervenor asserts that safe long-term storage of the fuel in question should not be based only on past experience with spent fuel pools; it should be based on additional studies.

( The Board did acquire a copy of le A. B. Johnson report (Kilp Affidavit, Ref. 5). Based on this report, Kilp states: " Visual obser-vations and radiation monitoring of pool water documented in a compre-hensive review of light water reactor fuel behavior during pool storage

33 by A. B. Johnson (Ref. 5), demonstrate that spent fuel has been stored safely for more than three decades." With regard to the three decades, at page 5 of the Johnson report we read: "The-technology for handling spent fuel has developed over 35 years. It has been satisfactory in almost all respects. The exceptions include an occasional fuel handling accident (Table 5), corrosioi of some fuel pool components (carbon steel and in some cases, aluminum alloys), and occasional leaks in pool liners." With regard to long storage times, Johnson states: "If storage times of the spent fuel inventory are expected to extend into the 20-to-100 year time frame, there is an increasing. incentive to determine whether any slow degradation mechanisms are operative."

Application of surveillance procedures and monitoring of the pool water and air above the pool are important to determining whether degradation is taking place over the lengthened years of service and any changes in the storage mode.

The Board denies the motion for summary disposition of Conten-tion 6. The Board asks the parties to address the matter of the modes and effectiveness of surveillance of materials and the monitoring of the fuel storage pool and contents to provide a measured basis for safety during the extended period of use.

34 E. Contention 7 states that:

That there'is no assurance that the health and safety of the workers will be protected during spent fuel pool expansion, and that the NRC estinates of between 80-130 person rem will meet ALARA requirements, in particular those in 10 C.F.R. Part 20.

The bases for the contention were given as:

FPL's estimates of between 80-130 rem / person are much higher than the NRC's estimate for reracking of 40-50 person / rem [ sic], and much higher than experience at other nuclear plants. Thus, there

[ sic] estimates are not ALARA.

Contention 7 was admitted by the Board limited in scope to the basis stated by the Intervenors. (Memorandum and Order, September 16, 1985, at 15)

Intervenor, Joette Lorion, in her response of March 19, 1986, opposes Licensee's motion for summary disposition of Contention 7.

Intervenor is concerned that the dose received by workers in the rerack-ing of the spent fuel pool Unit 4 will not be as low as reasonably achievable or ALARA. This is based in part on the fact that certain systens that Licensee said would be operable during rerack and would help them meet ALARA were not operational during the reracking of Unit 3. Intervenor states that the memorandum dated March 15, 1986, from P. Bemus to H. Thompson in which NRC Region II expressed concern that the cleanup and leakage detection systems would not be operational

E e

35 during rerack. This indicates that the dose the workers receive will be-higher than originally anticipated.

Intervenor challenges the value of 13.2 person / rem collective dose for reracking of Unit 3. Intervenors do not accept this value based on the brief charts and mrm/hr dose estimates in Licensee's affidavit (DanekAffidavit). Intervenor alleges that-the NRC Staff health physi-cist has not examined this data and that acceptance of this figure in the affidavit of Minns was based on second-hand information from the resident NRC inspector who is not a health physicist.

Intervenor expresses concern that NRC information shows performance in the spent fuel pool area and in radiological controls is declining.

A report issued on February 5,1985, " Systematic Assessment of Licensee Performance," July 1, 1983 through October 31, 1984, points out that Licensee performance in the areas of radiological controls and in spent fuel pool activities has declined. On page 21 of that report it states, o

" Spent fuel rerack safety analysis submittals for a license amendment assumed operability of the systems described in the submittals and the FSAR. However, many of the systems assumed to be operable required refurbishment and maintenance schedules had not been determined."

Intervenor cites other reports that raise areas of concern regarding operation of the Turkey Point spent fuel pools.

36 We have the Licensee's affidavit by Joseph L. Danek dated January 21,~1986, that details the measures taken during the actual reracking of Unit 3 spent fuel pool to reduce the radiation exposure dose to workers.

(Danek Affidavit, 11 5-31) -In general, radiation exposures were reduced by preplanning of activities and training of workers. The exposure reduction measures for the Unit 3 spent fuel pool reracking included (1) reducing levels of radiation in work areas, (2) reducing the amount of time spent by workers in radiation areas, (3) increasing the distance between workers and sources of radiation, (4) use of shielding and pro-tective clothing, and (5) radiation monitoring. (Danek Affidavit, 1 5)

The Board expects that Licensee will use similar measures to reduce radiation exposure for the Unit 4 reracking.

The Licensee's affidavit by Rebecca K. Carr dated January 22, 1986, also explains in detail why the radiation exposure during actual rerack-ing at Unit 3 was substantially less than those previously estimated by Licensee and approved by Staff. (Minns Affidavit, 11 9, 13) Carr states that the major reason for the lower actual dose rate was the performance of the spent fuel pool cleanup system. The actual dose rate was reduced by the new resin installed prior to the reracking operation and dedicated operation of the system throughout the reracking effort.

(Carr Affidavit, 1 23)

The expansion of the storage capacity of Turkey Point Unit 3 was completed in March 1985. The actual collective occupational radiation

37 exposure incurred during the reracking operation was 13.2 person-rem for the spent fuel pool expansion. (Danek Affidavit, 11 2, 32; Table 1, 2; Carr Affidavit, 11 21-24) This figure is far below the revised estimate of 59 person-rem and the original estimate of 88 to 130 person-rem. The Board concludes that this indicates that the measures taken by Applicant as described in affidavits were very effective in reducing the radiation dose. Whatever the history of spent fuel pool operation may have been in 1984, the low worker radiation dose achieved during the actual rerack of the Unit 3 pool shows that Licensee used effective procedures to reduce the radiation dose.

Spent fuel pool reracking consists of replacing the old storage racks with new racks that have neutron absorbing plates to store more fuel assemblies in a higher density array. In general, the reracking operation consists of several phases, including removal of the old racks, installation of new racks, transfer of spent fuel assemblies from the old racks to the new racks, and support services, such as quality assurance / quality control and health physics services. The reracking operation is cyclical in nature, involving installation of several new racks, removal of the old racks, and installation of new racks in the space vacated by the removal of the old racks. The water level in the spent fuel pools is lowered approximately 8 feet during the rack-handling operations and is restored to normal levels during fuel-handling operations. Once the water level has been lowered, a work platform is installed in the spent fuel pool base for rack-handling

38 activities. Underwater work is performed using long-handled tools.

During those periods when the water level is lowered, the system for cleanup of radioactive contaminants in the spent fuel pool water will not be used. At other times prior to and during the reracking opera-tion, the radioactive cleanup system will operate for sufficient times such that any further reduction in' the radioactivity in the pool water would be minimal. (Carr Affidavit at 8,16; Danek Affidavit,112,14)

Two analyses were performed by the Licensee to estimate the total occupational exposure required for the reracking. As a result of the first analysis, the Licensee estimated that the collective exposure would be about 109 person-rem per unit. Both the original and revised estimates were based on the expected dose rates for each phase of the reracking operation and the expected person-hours required to complete each phase. Conservative assumptions regarding dose rates and person-hours were made in both the original and revised estimates to ensure that estimates would not under-predict actual exposures. For example, major contributors to the dose rates to workers were expected to be the radioactivity in the spent fuel pool water and the radioactivity on the exposed walls. The original estimate conservatively took no credit for removal of radioactivity in the water or cleaning off the walls.

Similarly, when the original estinate was made, the number of person-hours required to complete all of the tasks was difficult to predict accurately, so conservative time estimates were used. (Carr Affidavit, 11 4, 5, 9-14, 20)

L 39 A revised estimate was made by Licensee using different assump-tions. Data on the radioactivity in the spent fuel pool water demon- '

strated that operation of the cleanup system for a short period of time could significantly reduce isotopic concentrations in the water, and the original estimate of the dose rates in and around the spent fuel pool was reduced to account for this. It was determined that some of the activities would be performed in a different area with lower dose rates than originally assumed. The estimated person-hours for certain tasks were reduced after procedures for these tasks were finalized and more -

~

details about these tasks were known. As a result of these changes '

in assumptions, the original estimate of 109 person-rem was revised downward to 59 person-rem per unit for reracking. (CarrAffidavit, 11 15-20)

The Licensee states that they used standard health physics tech-niques to maintain personnel exposures ALARA during the reracking of Unit 3, and similar techniques will be used during the reracking of Unit 4. These techniques include the following:

Preplanning of activities and training of workers. This consisted of several measures, including meetings of all groups involved in the rerackir,g to discuss radiological protection, minimizing the number of workers and activities needed for reracking, training of workers in FPL's radiation protection program, control of work through use of radiation

40 work permits, establishing written procedures to control the reracking activities, and providing on-the-job coverage by health physics. technicians.

Reducing levels of radioactivity in work areas. This was accomplished by operation of the spent fuel pool cleanup system, cleaning radioactive crud off the exposed walls of the spent fuel pool, removal of radioactive crud from off the old storage racks prior to transfer from the spent fuel pool.

Reducing the amount of time spent by workers in radiation areas. This was accomplished by use of remote tools, assem-bling work equipment in low radiation areas when practical, and controlling access to radiation areas.

Ute of shielding and protective clothing. .This consisted of maintaining approximately 15 feet of water in. the spent fuel

< pool to shield workers from the spent fuel, use of protective clothing, and use of respirators when the potential for airborne radioactivity existed.

' Radiation monitoring. This consisted of use of personnel monitoring equipment, permanent area radiation monitoring, permanent airborne radioactivity detectors, and periodic airborne and area radiation monitoring.

9 41 Licensee states that there are no additional measures which are reason-

-able and could have reduced the occupational exposures appreciably

- during the reracking. (Danek Affidavit, 11 4-30)

The Staff supports Licensee's motion for summary disposition of Contention 7. The Staff believes that the Licensee's statement of material facts is consistent with the Staff's Safety Evaluation, dated November 21, 1984, which was prepared in connection with the issuance of the spent fuel pool amendments. (Minns Affidavit, 1 3) In 5 2.6 of its Safety Evaluation, the Staff reviewed the Licensee's measures for main-taining occupational radiation exposures ALARA during the reracking of the Turkey Point spent fuel pools. The Staff concluded that, based upon its review of these measures, the spent fuel pool modifications can be performed in a manner that will ensure that exposures to workers will be ALARA.

10 C.F.R. Part 20 does not impose quantitative limits on collective radiation dose exposures. Licensees, in addition to complying with the requirements of Part 20, must make "every reasonable effort to maintain radiation exposures ... as low as is reasonably achievable." (10 C.F.R. 5 20.1(c)) This is defined to be "as low as is reasonably achievable taking into account the state of technology, and the economics of improvements in relation to benefits to the public health and sa#ety, and other societal and socioeconomic considerations, and in relation to the utilization of atomic energy in the public interest." The ultimate

42 i

responsibility rests with licensees to consider conditions and situa-tions expected, or known to be present, in a particular licensed activ-ity, and to take appropriate dose reducing actions. (MinnsAffidavit, 1 4)

In addition to operating in accordance with the radiation protec-tion standards of Part 20, the Licensee has committed (1) to provide personal dose monitoring instrumentation in accordance with 10 C.F.R.

QS 20.202(3)(b)(1) and 20.103; (2) to use radiation pr'otection equipment which meets the criteria of Regulatory Guide 1.97; (3) to follow Regula-tory Guides 8.8 and 8.10; and (4) to design, construct and operate the

" Spent Fuel Expansion Program" consistent with Regulatory Guide 8.8 in order to ensure that occupational doses will be kept as low as is rea-sonably achievable. (Minns Affidavit, 11 6, 7)

The Licensee's original estimate of the total occupational exposure required to complate the spent fuel expansion was 109 person-ren. In response to a Staff inquiry regarding the collective dose equivalent for these modifications, the Licensee racalculated and revised its estimate to 59 person-rem. The Staff approved the 59 person-rem exposure esti-mate as satisfying the "as low as is reasonably achievable" standard.

The actual experience reracking Unit 3 was much less,13.2 persor, ren.

(Minns Affidavit, 11 8, 9)

~

q

+.

l 43 Based on the manner in which the Licensee will perform the spent fuel pool modifications, the Licensee's radiation protection program,

'the radiation protection measures proposed for the modification task, including' air and airborne radioactivity monitoring, and expe'rience gained from operating nuclear power plants, the. Staff concludes that radiation protection measures are adequate to assure worker protection and that.the spent fuel. pool expansion can.be performed in a manner-which will ensure that doses to workers will be maintained within the ,

limits of Part 20 of the Commission's regulations and as low as is

. reasonable achievable.- (Minns Affidavit, 11 5, 6, 7, 12) - The Staff supports Licensee's motion for summary disposition of Contention 7.

Based on a consideration of information presented in cffidavits presented by Licensee and Staff, the Board determined.that there are no issues of material fact to be lititgated.with respect to Contention 7.

Licensee's motion for sumnary disposition of Contention 7 is granted.

4 F. Contention 8 states that:

l I

l That the high density design of the fuel racks will cause higher heat loads and increase in^ water temperature which could cause a loss-of-coolant accident in the spent fuel pool, which cculd cause a major release of radioactivity to the environment. And, that the decrease in the time that it takes the spent fuel to reach its i.

l boiling point in such an accident, both increases the probability l of accidents previcusly evaluated ano increases [ sic] the chances

. accidents not previously evaluated.

I' k -. . . . .- - - - . -

44 The bases for contention were given as:

(a) The NRC ha'. stated in numerous documents that the water in spent fuel pools should normally be kept below 122*F. The present temperature of_ the water at Turkey Point is estimated to be 127 F. After the reracking, the temperature of the water could rise to 141 on a normal basis, and could reach 180 F with a full core load added. In addition, the time for the spent fuel boiling point.to be reached in a loss of cool-ing accident will go from 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Four hours is clearly not enough time to take action to prevent a major accident in the spent fuel pool from occurring. Thus, the increase in heat and radioactivity resulting from increa-ses [ sic] density will result in an increase in the proba-bility of a major spent fuel pool meltdown occurring.

(b) There is also the possibility that a delay in the makeup emer-gency water, could cause the zirconium cladding on the fuel rods to heat up to such higher temperatures that any attempt at later cooling by injecting water back into the pool could hasten the heat up, because water reacts chemically with heated zirconium to produce heat and possible explosions.

Thus, the zirconium cladding could catch on fire, especially in a high density design, and create an accident not previ-ously evaluated.

Intervenors allege that the increased spent fuel storage density causes a reduction in the margin of safety from a thermal hydraulic and spent fuel pool cooling system safety concern. The Licensee's calcula-ted maximum bulk pool temperature of 143 F for the increased storage does not meet the requirements of SRP % 9.1.3, which states that the spent fuel pool temperature should be limited to 140 F. The increased heat reduces the time it takes for the spent fuel to reach boiling in a loss of cooling accident. The time for boiling has been reduced to 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for a normal load and to 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for a full core off-load. The

45 spent fuel pool cooling system does not meet the requirements of SRP ,

f 9.1.3, if I.1 and III.1.b or General Design Criteria 2 of 10 C.F.R. Part 50, because the cooling system is not designed to provide adequate cooling in a seismic event. Intervenors do not agree with the Staff's decision to issue the amendments before Licensee installs the seismic design changes.

Intervenors also contend that neither the Licensee nor Staff have proven that makeup water will be supplied in a timely fashion in a loss of cooling accident. This creates the possibility that a zirconium water reaction and possible fire or explosion could be caused by a delay in cooling the pool. The Licensee has not met its burden of proof on Contention 8 because they have not demonstrated that increased storage density and storage of more highly enriched spent. fuel with a higher burn-up will not increase both the possibility and consequences of a spent fuel pool loss of cooling accident.

The Licensee, in its affidavit by Danic' C. Patton, dated Janu-ary 22,1986, addresses the generation of heat and the cooling of the spent fuel as a result of the increase in storage capacity of the spent fuel pools. The decay of fission products _ in spent fuel assemblies pro-duces heat which is a function of burnup in the reactor and time after shutdown. The decay heat decreases with time following shutdown.

(Patton Affidavit, 1 3)

S 46 The Licensee's' analysis assumed that 1/3-core would be loaded into the spent fuel pool for the first six refuelings, and,1/2-core would be loaded into the pool for the next 14 refuelings. The decay heat was calculated in accordance with the fiRC's Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling." The total heat load to the pool following refueling is a function of the amount of fuel placed in the pool. The maximum heat load for a nonnal refueling off-load is calculated as a first case to occur with 8-1/2 cores stored in the pool and 1/2-core'off-loaded at refueling. This results in a peak total heat generation rate of 17.9 x 100 BTU /hr. -This raises the pool water temperature to 143*F.

The pool water temoerature is calculated to decrease to 140 F in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reaching the peak and to 130 F in 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> after reaching the peak. At the peak temperature, the water loss from the pool through evaporation is calculated to be ~approximately 1.5 gallons per minute.

This is well within the capacity of the makeup system. (Id., 1 10)

The calculation considers only 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay after reactor shutdown. In reality, fuel off-load at Turkey Point typically does not begin entil well after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown and extends over a period of time such that much of the decay heat is given off before the fuel is placed in the spent fuel pool. Thus, the calculation of the heat load to the pool is conservatively maximized. The heat removal from the pool is conservatively minimized by neglecting heat loss through convection and evaporation from the surface of the pool to the

dIJI Rv

,m 47 fuel' building atmosphere and by neglecting heat conduction .from the water through the walls'of the pool. The calculations show that the temperature ~ of the pool is still below boiling-and the mass loss through '

- evaporation 11s well within the makeup capacity of the system. (Patton

-Affidavit, 1 11)

The Licensee calculated a second case, which involved a full core fuel off-load. In this abnormal case, the analysis. performed for the increase in spent fuel pool capacity-assumed eight cor6s stored in the pool and one core off-loaded. This results in a maximum heat load to

- the pool of 35.0 x 10 6BTU /hr and causes the temperature of the pool water to increase to a peak value of 183"F. At'this temperature, the rate of water loss through evaporation from the surface of the pool is

- conservatively calculated to be 5.5 galloas per minute, which is well below the 100 gpm capacity of the makeup system. The result is that the- ,

pool temperature is below the point of boiling, that the evaporation ,

. rate is well within the makeup capabilities of the system, and that the s

spent fuel pool cooling system is adequate to remove the heat generated from the stored spent fuel. Thus, there is assurance that the spent fuel will be cooled and covered with water at all times. The Licensee's 1 analysis shows the cooling system to be in compliance with the SRP guidelines. (Id., 1 12)

The Licensee perforned an analysis of loss of cooling to the spent

, fuel pool. It showed that, with a?1 positions in the new storage racks

-t , * -- --------y *e--

48 full' with assertblies,' the pool would not begin to boil for 'a nornal off-load 'until a minimum of. 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after loss of cooling and 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after loss of cooling for a full core off-load, which is an abnormal case. (Patton Affidavit, 1 14)

Regarding the seismic issue, Licensee states that the existing cooling system piping and makeup lines for the Turkey Point spent fuel pools are not seismic Category I and have not been designed to remain functional after a safe shutdown earthquake. The Licensee has committed to complete the seismic upgrade of the Turkey Point spent fuel cooling system by the end of the second refueling outage after issuance of the Turkey Point spent fuel pool expansion amendments. Until the seismic upgrade is completed, the amount of fuel that is scheduled to be stored in the pools will be less than the capacity of the pre-existing storage racks, and the spent fuel pool expansion amendments will not result in an increase in tht. amount of c'., ling and makeup necessary for these assemblies. (Id., 11 8, 15)

The Staff responds with an affidavit by John H. Ridgely dated Feb-ruary 18,1986. The Standard Review Plan (SRP) (NUREG-0800) 5 9.1.3 states that the spent fuel pool temperature should be limited to 140*F for the normal maximum spent fuel heat load conditions. This is the heat generated when all storage cells in the storage pool are filled with spent fuel assemblies on the normal refueling schedule. The decay time of the respective batches is based on the anticipated intervals

~

c:

o-49 between . refuelings. The decay time of the most recently. discharged batches is based on the least time interval between shutdown and when refueling commences plus the minimum time required to accomplish the

, discharge. -This is normally assumed to be 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />. The pool

+ temperature of 140 F is determined by the maximum design temperature at which the spent fuel pool cleanup system can normally operate without

~

degradation. (Ridgely Affidavit, 11 4-5) 3 In its Safety Evaluation (SE), dated November 21,'1984, the Staff made an independent calculation,-SE 9 2.7.2, assuming a normal maximum heat load based on a half core discharge. The normal maximum pool water temperature is expected to be 140.8*F. Although the normal maximum pool water temperature of 140.8'F calculated by Staff-is slightly higher than

.the guidance. identified in the SRP, the pool water temperature is

acceptable because-it is based on conservative assumptions regarding h

. core discharge and the temperature only exceeds 140*F for a short period j; of time. In addition, if-there were a loss of cooling to the spent fuel pool, the fuel cladding temperature will not increase to the temperature necessary for a significant amount of zirconium water reaction to occur and there is adequate time for providing makeup water to the pool to prevent spent fuel pool boiling. (Ridgely Affidavit, 1 14)

A sensitivity analysis indicates that the pool temperature is expected to remain above 140*F for approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the

' spent fuel is placed into the spent fuel pool. The Licensee's l

L

50 calculations concerning the normal maximum pool water temperature (143*F) and the anticipated time required until the pool water temperature is less than 140*F (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) are both. higher and based on more conservative assumptions than the Staff's analysis which was per-formed consistent with the SRP and SRP Branch Technical Position ASB 9-2. Licensee's analysis and Staff's independent analysis using similar assumptions used half-core reloads in lieu of the normal one-third core reloads. If normal one-third core reloads were used, the results of both analyses would have been less than the 140*F guideline. The Staff concludes that the Licensee complies with the guidelines of the Standard Review Plan water temperature limit of 140 F. (Ridgely Affidavit, 11 6-7)

A total loss of cooling to the spent fuel pool would result in the peal water temperature increasing to boiling or 212 F. The ability of the spent fuel to produce heat decreases with time. The time required for the spent fuel pool to commence boiling is 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Once boiling starts, the significant pool water loss is due to boil off. The boil off rate is approximately 37 gallons per minute. It is estimated that there is approximately 193,800 gallons of water in the spent fuel pool above the top of the spent fuel storage racks. Based on this water volume it would take approximately three days and 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> from the time the water reaches 212 F before the top of the racks are uncovered.

Thus, it takes a total time of three days, 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> for the pool water to boil and for the pool water level to boil off down to the top of the

n -

'o; b

51 spent fuel racks. Makeup water to the spent fuel pool can be provided from the demineralized water system, the fire water system, the primary water system, .or from the refueling water storage tanks. Given this number of different methods of providing makeup water, the Staff con-cludes that 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is adequate time to initiate makeup to the spent fuel pool before a spent fuel pool would commence boiling. In the unlikely event that makeup water could not be provided within the 7.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, there would be no detrimental effects on the spent fuel for an additional three days and 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. (SE92.7) The n5rmal pressure inside of the fuel handling building is one atmosphere. The maximum anticipated water temperature is 212 F and the maximum fuel cladding temperature would be in the 200-to-300 F range. The zirconium water reaction is not significant at. temperatures less than 1800*F. Therefore there is no problem with the zirconium water reaction in the spent fuel pool. (RidgelyAffidavit,119-11)

The spent fuel pool cooling system consists of one heat exchanger with two pumps and associated valves and piping. One pump is normally operating with the second pump as a spare in the event that the first pump is not available. The Staff notes that this cooling systen is not seismic Category I, safety-related at this time. The Licensee has conraitted to upgrading the cooling system such that it will remain functional after a safe shutdown earthquake. When the upgrading is complete, the spent fuel pool cooling system will meet the guidelines of Regulatory Guide 1.29, Position C.1, which addresses the design of

52 safety-related structures, systems and components with respect to their ability to withstand the safe shutdown earthquake and to remain operational. By meeting this Position, the Staff concludes that the Licensee complies with the requirements of General Design Criterion 2 of 10 C.F.R. Part 50, Appendix A, " Design Bases for Protection Against Natural Phenomena," for protection against earthquakes. (Ridgely Affidavit, 1 12)

The Staff has performed an independent accident eitaluation of the offsite radiological consequences of boiling in the pool and has found the consequences to be a small fraction of the 10 C.F.R. Part 100 guide-lines. (Wohl Affidavit, Contention 4) Based on the small radiological consequences as the result of the pocl boiling, the ability to cope with a single active failure of the pool boiling, the ability to cope with a single active failure of the spent fuel pool cooling pump, and the low probability of having an earthquake until the cooling system is upgraded the Staff concludes that the design meets the guidelines of Regulatory Guide 1.29, Position C.2, which addresses the seismic aspects of non-safety related equipment. Therefore the Staff concludes that the Licensee meets the requirements of General Design Criterion 2 of 10 C.F.R. Part 50, Appendix A.

Based on results of studies and analyses regarding the issues in Contention 8 in affidavits by Licensee and Staff, the Board grants Licensee's motion for summary disposition of Contention 8.

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

V 2 .'- ,.

53' yr G. Contention 10 states:

.That the increase of the spent' fuel pool capacity which includes fuel rods that are more highl ments of ANSI'NI6-1975 [ sic] notyto~ enriched, be met will and cause the require-will increase the

probability that a criticality accident will occur in the spent fuel pool and will exceed 10 C.F.R. Part 50, A 62 criterion.

i -

The bases for. contention were given as:

4 The increase in the number of fuel rods stored and the ' fact that' neny of them may be more highly enriched and have more reactivity will increase the chances that the fuel pool will go critical, and cause a major criticality accident, and perhaps explosion, that will release large amounts of radioactivity to the environment in excess of the 10 C.F.R. Part 100 criteria.

4 Intervenors contend that the storage of the more highly enriched spent fuel and the increased density, which increases the k,ff, should not be allowed because it reduces a critical safety margin to within a fraction of the allowable limit. Intervenors noted'that before 1976, the acceptable k,ff for the 95/95 standard was 0.90. Prudence would j dictate that one leave a significant margin of safety between the calcu-i lated k,ff limit and the standard. The Licensee's newly calculated

. values for k,ff are no longer conservative and the safety margins for spent fuel storage have been significantly reduced. The possibility of a criticality accident occurring in the spent fuel pool becomes more

{ likely. The use of more highly enriched uranium fuel rods with a greater burnup creates more fission products poisons. The storage of 3

I k.

I

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

q y

54-this fuel increases the possibility of a criticality accident and would increase the radiological consequences of such an accident.. (Lorion

~

Affidavit, 11 G6-G8)

The Licensee, in its affidavit by William A. Boyd, dated

-December 14, 1986, describes criticality analyses for the Spent Fuel Pool Expansion Amendments to show that they conform with applicable industry-standards, employ NRC approved methods, and provide results that meet NRC criteria. These analyses demonstrate th'at fuel assemblies of authorized initial enrichments and burnups to be stored in authorized storage patterns in the Region 1 and Region 2 racks will have a k-effective of less than 0.95, including all uncertainties, under both normal and accident conditions.

The spent fuel pool expansion involves the replacement of the original racks with new storage racks in a higher density array. They are designed to accommodate a more highly enriched fuel. The spent fuel pools are divided into two regions. Each region consists of new storage racks which have a different high density fuel storage configuration and a different amount of neutron absorber. The Region 1 fuel racks permit storage of new, unirradiated fuel assemblies with an enrichment of 4.5 percent of Uranium-235. The Region 2 fuel racks permit storage of irradiated fuel assemblies with enrichment of 4.5 percent and their burnups are 39,000 mwd /MTU or higher. (Boyd Affidavit, 11 10-13)

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55

'The design basis for preventing criticality in the expansion of Turkey Point spent fuel pools is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that k-effective of the fuel assemblies in the pools will be less than 0.95.

This satisfies the 0.95 criterion provided in the NRC and ANSI guidance.

The 0.95 criterion was in force prior to issuance of the licensing amendments authorizing the spent fuel pool expansions. Therefore, the amendments did not modify or increase the design basis k-effective limit for Turkey Point spent fuel pools, and thus the amendments will not increase the probability of a criticality accident. The design of the new racks to limit k-effective is such that the minimum separation distances between the stored fuel assemblies is fixed, and includes neutron absorbing Boraflex inserted between the assemblies. This permits safe storage of more highly enriched fuel assemblies in a higher density array. (Boyd Affidavit, 11 14, 17, 18)

The Licensee performed criticality calculations for the new Turkey Point spent fuel storage racks with a computer program called KENO-IV.

It has been verified by comparison with criticality experiment data for fuel assemblies similar to those for which the Turkey Point spent fuel storage racks were designed. The isotopic compositions of the fuel assemblies to be stored were calculated with a computer code called PH0ENIX. The accuracy of the PH0EllIX code has been demonstrated by comparison with measurements made for fuel samples taken from the core of the Yankee reactor. These samples encompass the pellet size and

1

_/*

56 enrichment of the fuel ' proposed for storage in the Turkey Point racks.

The Licensee calculated the decay of fission products and their neutron

capture effects during storage of fuel assemblies with a computer code called CINDER. ~ This has been widely used in the nuclear industry for .

over 20 years, has been well benchmarked by many sources,.andLis acceptable by the NRC. (Boyd Affidavit, 11 19-22; SE 11 2.1.1., 2.1.2 and 2.1.4)

The Licensee used conservative assumptions in the criticality calc-

-ulations. These included assuming that the array of fuel assemblies is infinite in lateral and axial extent; that there are lower concentrations of certain neutron absorbing fission products than actually present in the spent fuel; neglecting neutron capture by the fuel assembly spacer grids and sleeves; neglecting neutron capture by the boron in the spent fuel pool water; and assuming _a spent fuel pool water density and temperature which maximizes the amount of moderation provided by the water. The criticality calculations for the Region 1 racks and the Region 2 racks with 4.5 percent enriched fuel assemblies in a checkerboard arrangement contained additional conservative assumptions. These included assuming that all fuel rods contain 4.5 percent enriched Uranium, neglecting burnable poison in the fuel rods, neglecting neutron absorbing Uranium-234 and Uranium-236, and neglecting fission-product poisons. The calculations accounted for biases and I uncertainties either by (1) assuming worst-case conditions, such as assuming fuel assemblies are centered and assuming minimum width and i

o 57 thickness of poison materials; or (2) by increasing the nominally calculated value of k-effective by the amount of the reactivity effects of the biases and uncertainties, including biases and uncertainties in the analytical methods and the material and mechanical construction -

tolerances of the' sheet metal cell walls, cell center-to- center spacing cell bowing, and Boraflex neutron absorbing properties. The criticality calculations for the Region 2 racks with irradiated fuel assemblies having an equivalent zero burnup enrichment of 1.5 percent also accounted for the biases and uncertainties associated with plutonium reactivity and other reactivities that are a function of 1,rradiation.

(Boyd Affidavit, 11 13, 20-29)

The results of the Licensee's criticality calculations for the new fuel storage racks were a k-effective of (1) 0.9403 (including all uncertainties) for the Region 1 racks; (2) 0.9304 (including all uncer-tainties) for the Region 2 racks with fuel assemblies with an equivalent zero burnup enrichment of 1.5 percent; and (3) 0.8842 for Region 2 racks with 4.5 percent enriched fuel assemblies in a checkerboard arrangement.

This latter value did not include biases and uncertainties. Assuming conservative values for these terms would still result in a k-effective below 0.95 at a 95 percent /95 percent probability / confidence level. All of these values are below the design basis k-effective limit. (Boyd Affidavit, 11 26-32)

I

e o

58 The NRC Staff has adopted the double contingency principle of ANSI (American National Standards Institute) N16.1-1975 which, in effect, states that it is not necessary to consider two unlikely, independent, and concurrent changes in conditions in performing criticality analyses.

The Licensee's criticality calculations for the new Turkey Point spent fuel storage racks were performec' in accordance with the double con-tingency principle. The criticality calculations were performed assum-ing the absence of boron in the spent fuel pool water, which is an acci-dent condition. The presence of borated water is equivalent to a nega-tive reactivity change of about 0.30 in k-effective. Under the double contingency principle, it is unnecessary to postulate an accident y involving both the absence of borated water and another independent change in conditions in the spent fuel pool. The negative reactivity of the borated water in-the Turkey Point spent fuel pools would more than offset an increase in reactivity resulting from other accidents, includ-ing the absence of Boraflex poison plates in the storage racks and changes in the mechanical or geometric configuration of the storage racks or fuel assemblies caused by an inadvertent drop of an assembly, a cask drop accident, an earthquake, and other types of credible acci-dents. Consequently, these accidents would not cause the calculated k-effective to exceed the design basis limits. (Id., 11 33-40)

F

! The Staff supports Licensee's motion for summary disposition of Contention 10 based on an affidavit by Lawrence I. Kopp dated February 18, 1986, and the Staff review of criticality consideration for the

Q C.

59 reracked spent fuel pool in 99 2.1 through 2.15 of the November 21, 1984 Safety Evaluation (SE) on the amendments. The Staff's review of the Licensee's criticality calculations consisted of determining that gen'erally accepted calculational methods, verified by comparison with experiments, were used, and that the assumptions and uncertainties have been treated appropriately. The Staff states that the material facts presented in the " Licensee's Statement of Material Facts As To Which There Is No Genuine Issue To Be Heard With Respect to Intervenors' Con-tentions," dated January 23, 1986, in relation to Contention 10 are correct and concur in the conclusions reached in the supporting affidavit by Licensee. (Kopp Affidavit,12)

The Staff discusses the acceptance criteria. General Design Cri-teria (GDC) 62, " Prevention of Criticality in Fuel Storage and Hand-ling," states that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. The acceptance criterion for assuring that GDC 62 is met is found in Staff's Standard Review Plan (SRP), 9 9.1.2, which requires maintaining a storage array neutron mul-tiplication factor (k,ff) less than 0.95 in spent fuel pools during normal and accident conditions. This is an adoption of the criteria contained in American National Standard Institute (ANSI) N18.2-1973.

Even for accident conditions, the Staff requires spent fuel pools to be at least 5 percent subcritical, or k eff n greater than 0.95, to supply adequate margin to assure that the requirement of GDC 62 that a keff c

O.

60 less than 1.0 is met. Additional Staff guidance for conducting analyses of spent fuel pools is found in the April 14, 1978 letter from Brian Grimes transmitting the NRC "0T Position for Review and Acceptance of Spent Fuel. Storage and Handling Applications." This guidance provides for the use of certain conservative assumptions and consideration of a variety of calculational, mechanical and materials uncertainties in arriving at the k eff value for a given spent fuel storage array. The conservative assumptions are that: (a)thek,ff of the racks be calculated for the highest reactivity fuel anticipated'for storage at the temperature (within pool limits) yielding the highest kei 'f; (b) pure water instead of borated water is in the pool; and (c) the fuel array is infinite in lateral and axial dimensions. (Kupp Affidavit, 11 4, 5)

The Staff summarizes some features of the redesigned fuel storage cells. A strong neutron absorber Boraflex was added to the fuel assem-bly storage canister walls. This allows storage of higher enriched fuel at closer center-to-center spacings. The spent fuel pool was divided into two regions to allow fuel with a maximum uranium-235 (U-235) en-I richment of 4.5 weight percent to be stored. Region 1 will have 10.6 inch center-to-center spacir$g and will be limited to storage of fuel assemblies meeting certain required burnup considerations. Region 2 will allow a larger number of fuel assemblies to be stored at a closer spacing than in Region 1. The allowed enrichment will be lower because of the depletion or burnup of fissionable U-235 with operating time in the reactor. This burnup dependency for spent fuel storage has been

~

.0 :

' Ol L 61 applied previously by various licensees'(Arkansas Nuclear One Unit 1, Fort Calhoun' Unit-.1, St. Lucie. Unit 2, Ginna Unit 1) and has been approved by the NRC. (Kopp Affidavit, 11 6-8)

The results of the Licensee's analysis using calculational uncer-tainties, conservative assumptions and worst case design basis accidents predict a k-effective of 0.9403 for Region 1 and 0.9304 for Region 2.

The Staff reviewed the Licensee's criticality calculations and concluded that acceptable calculational methods were used and we're verified by. (

comparisons against empirical data and that the assumpticr.s and uncer-tainties were treated appropriately consistent with NRC guidance. The results of the analysis for both Regions 1 and 2 meet the acceptance criterion of k-effective less than 0.95, including all uncertainties'at a 95/95 probability / confidence level. The Staff concludes that the criticality aspects of the design of the Turkey Point spent fuel racks and the Licensee's criticality analysis is acceptable. Since critical-ity does not occur for any postulated normal or accident condition, there is no release of radioactivity to the environment and the 10 C.F.R. Part 100 guidelines are met. (KoppAffidavit,119-15)

Based on a consideration of information presented in affidavits presented by Licensee and Staff, the Board determined that there are no issues of material fact to be litigated with respect to Contention 10.

The Licensee's motion for sumary disposition of Contention 10 is granted.

r.

a s

Q' 62 IV. ORDER For all the foregoing reasons and upon consideration of the entire record in this matter, it is this 25th day of March,1987 ORDERED

1. That the Licensee's motion for sumary disposition of Intervenors' Contentions 3, 4, 7, 8 and 10 is granted;

2. That the NRC Staff's motion for sumary disposition of Intervenors' Contention 4 is granted; and

3. That the Licensee's motion for sumary disposition of Intervenors' Contentions 5 and 6 is denied.

APPEALABILITY A denial of a motion for summary disposition is interloct. tory and therefore cannot be appealed. Louisiana Power and Light Co. (Waterford Steam Electric Cenerating Station, Unit 3), ALAB-270, 8 AEC 93, 94 (1974). Since this Order dismissed some, but not all, of the Intervenors' contentions, the Intervenors are still parties to this proceeding; therefore, the dismissal of Contentions 3, 4, 7, 8 and 10 is interlocutory, and any appeal the Intervenors may wish to take from that

{

o 63 dismissal must await the issuance of an initial decision. See Cleveland Electric Illuminatina Co. (Perry Nuclear Power Plant, Units 1 arid 2),

ALAB-736, 10 NRC 165, 166 (1983).

THE ATOMIC SAFETY AND LICENSING BOARD R 4 2 9n.fa,s RobertM.Lazo,Chairmy ADMINISTRATIVE JUDGE

/ ---

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Emmeth A. Luebke ADMINISTRATIVE JUDGE

,1/ . ff Richard F. Cole ADMINISTRATIVE JUDGE Dated March 25, 1987 Bethesda, Maryland i