ML17221A560

Forwards Revised Responses to Questions 2,6 & 7 Re Replacement of Spent Fuel Racks,Per NRC 870716 Request for Addl Info & Subsequent Discussions W/Util

ML17221A560
Person / Time
Site:	Saint Lucie
Issue date:	12/23/1987
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-0554, RTR-NUREG-0612, RTR-NUREG-554, RTR-NUREG-612 L-87-537, NUDOCS 8712280232
Download: ML17221A560 (31)

Text

REGULATO INFORMATION DISTRIHUTIQN STEM (RIDB)

ACCESSION NHR: 8712280232 DOC. DATE: '7/12/23 NOTARIZED: NO DOCKET 0 FACIL: 50-335 St. Lucie Planti Unit ii Florida Power 5 Light Co. 05000335 AUTH. NAME AUTHOR AFFILIATION WOODYi C. O. Florida Power 5 Light Co.

REC IP. NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards revised responses to Questions 2> 6 5 7 re replacement of spent fuel racks'er NRC 870716 request for addi info 0 subsequent discussions w/util.

DISTRIHUTION CODE: A001D COPIES RECEIVED: LTR l ENCL L SIZE:

TITLE: OR Submittal: General Distribution NOTES:

REC IP IENT COPIES REC IP IENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 0 PD2-2 PD 5 5 I

TOUR GNY i E 1 INTERNAL: ARM/DAF/LFMH 1 0 NRR/DEBT/ADS 1 1 NRR/DEST/CEH 1 NRR/DEBT/MTH NRR/DEBT/RSH 1 1 NRR/DOE*/TBH 1 1 NRR/ S/ ILRH 1 1 OGC/HDB2 1 0 REG IL 01 1 1 RES/DE/EIH 1 1 EXTERNAL: LPDR 1 NRC PDR 1 1 NSIC 1 1 TOTAL NUMHER OF COPIES REQUIRED: LTTR 20 ENCL 17

0 P II t

1 II

P. 0 X 14000, JUNO BEACH, FL 33408.0420

~ps>lb'~

>i

~

DECENHER 8S 198~

L-87-537 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gent lemen:

Re: St. Lucie Unit I Docket No. 50-335 S ent Fuel Rerack By letter L-87-245, dated June l2, l987, Florida Power & Light Company (FPL) submitted a proposed license amendment to permit replacement of the spent fuel pool racks at St. Lucie Unit I to ensure that sufficient future capacity exists for

'storage of spent fuel.

By letter dated July l6, l987 (E. G. Tourigny to C. O. Woody) the NRC Staff requested additional information in the area of plant systems it needed to continue its review of this proposed license amendment. By letters L-87-374, dated September 8, l987, and L-87-4I9, dated October l9, l987, FPL submitted responses to this request. As a result of subsequent discussions with the NRC, it was determined that FPL's responses to questions 2, 6, and 7 would need to be revised. Attached are FPL's revised responses.

tf additional information is required, please contact us.

Very truly yours, C. O. Woo Executive ice President COW/E JW/gp Attachments cc: Dr. J. Nelson Grace, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant 8712280232 871223 PDR P

ADDCZ OSOOO335 PDR E JW4/050/ I an FPL Group company

4 I 4

41 4 k

P II IV pW 4 k

~ 4' ~ ~

4 W 4

I I lk 4 P It J H

~ 4 1

W 4

h 4

4 h

QUESTION 82 QUESTION: Ke submittal specifies that the spent fuel pool cooling system consists, in part, of two spent fuel pool cooling system pumps and one heat exchanger. Since all heat exchangers require maintenance at some point in time, it is not clear how the spent fuel pool will be cooled during the time that the single spent fuel pool cooling system heat exchanger is out of service for maintenance. Provide a discussion of how the spent fuel pool is cooled and the temperature is maintained below the Technical Specification temperature limit for the following conditions:

l. 'Jhe heat exchanger is out of service for plugging of failed tubes.

2. 'Ihe heat exchanger is out of service for cleaning of the tubes.

3. 'Ihe heat exchanger is out of service for replacement.

RESPONSE: Due to the low pressure, low temperature service environment, and the carefully controlled water chemistry, (of both the CCW and SFPCS) the SFP heat exchanger will only infrequently require maintenance. Maintenance operation such as cleaning and tube plugging can normally be scheduled in advance for periods of low pool decay heat loads (just before a normal refueling). 'Ihe spent fuel pool heat exchanger is designed for the life of the plant, therefore its replacement is not postulated. However, in the extremely unlikely event that replacement is required, alternate means of providing cooling to the SFP could be implemented during replacement.

If there were no cooling provided to the SFP such that boiling were to occur, redundant sources of makeup water to the SFP are available, as discussed in St Lucie Unit 1 FSAR Amendment No 6, Section 9.1.3.4 and in Section 3.2.3 of the St Lucie Plant Unit No 1 S nt Fuel Stora e Facilit Modification, Safety Analysis Report, transmitted via letter L-87-245 dated June 12, 1987.

Boiling of the water in the SFP is adddressed in Section 9.1.3.4 of the FSAR and in Section 3.2.2.4 of the Safety Analysis Report. ('there is no Technical Specification temperature limit for the SFP.)

Attachment 6 contains the Spent Fuel Pool Boiling Analysis. 1he results of this analysis demonstrate that the maximum calculated thyroid dose is much less than 1% of the 10 CFR 100 limit of 300 rem and the Mole body and skin does arq completely insignificant. 1herefore, no off-site dose limits will be exceeded if boiling occurs in the spent fuel pool.

t JPEWR&7 %37 Rev 4 Page 10 QUESTION 86 QUESTION: For each step in the replacement of the racks','"pr'ovide drawings which show the location of the spent fuel and the movement of the fuel which demonstrates that no heavy load will be carried over spent fuel or over any rack which contains spent fuel. (Note:

This information can be placed on the set of drawings provided in response to question 5).

RESPONSE: 2he attached drawings show the various locations of the spent fuel during the rack installation. Described below is the sequence of rack removal and installation.

Relocate all the fuel assemblies presently stored in existing Racks 1 and 2. Following the removal plan (see Question 85),

remove Racks 1 and 2 from the pool as shown in Figure l.

2 ~ Following the installation plan (see Question 45), install new Racks Fl, G2 and Gl as shown in Figure 2.

3~ Relocate all the fuel assemblies stored in existing Racks 3, 4, 5 and 6 to the southernmost cells of new racks Fl, G2 and Gl as shown in Figure 3.

4~ Following the removal plan (see Question 85), remove existing Racks 3, 4, 5 and 6 as shown in Figure 3.

5. Following the installation plan (see Question 05), install new Racks D2, C3 and Cl as shown in Figure 4.

6. Relocate all the fuel assemblies stored in existing Racks 7, 8, 9, 10, 13 and 14 into new Racks D2, C3 and Cl as shown in Pigure 5.

7~ Following the removal plan (see Question 85), remove existing Racks 7, 8, 9, 10, 13 and 14 as shown in Figure 5.

8. Pollowing the installation plan (see Question S5), install new Racks D3, B2 and C4 as shown in Figure 6.

9. Relocate all the fuel assemblies stored in existing Racks 11 and 12 into new Rack D3 as shown in Figure 7.

10. Following the removal plan (see Question P5), remove existing Racks 11 and 12 as shown in Figure 7.

Following the installation plan (see Question f5), install new racks Bl, C2, A2, Al, Dl, Hl, E2 and El as shown in Figure 8.

0093L/0017L

1 l

1

~ )

e I pep rp T's

~ /'

'I i~

~Qp', ~ ~, ~ ~'

a ro .e

\

\

P

~

~

1

. JPE-LR-87-037 Rev 4 Page 12 CASK AREA 62 Gl 5

I 7X7 X7 7X7 0 C~

7X7 ~7 X 7 ~i7X7 r IP.(

r14 /z l2 7X7 7X7 7X7 7X7

= NEW RACKS i=EXISTlNG RACKS gg= FUEL FIGURE 2 (REvl)

TITLE: SEQUENCE OF RACK REMOVAL AND INSTALLATION nuctaar uuUtu conauucuon ENCLOSURE: 2 OF 8

JPE-LR-87W37 RELOCATED FUE.L ASSEM BLIES rage 13 FROM RACKS 5 THRU 6 G2

'ASK AREA R EViOVE e

ROV'S 7X7 I I

7X7 I I

7ZV i 7X7 r lo 9 r8 y7 7X7 r

i 14 r7X7 z7X7.

r12 iX II s 7X7 7X7 7X7 7)l7

=NEW RACKS

" ":=EXISTING RACKS Q=FVEL FIGLJRF a (,,)

TITLE: SEQUENCE OF RACK REMOVAL AND INSTALLATION flUCJRBl UQlltL4 COflSCfLICQOllI ~

ENCLOSURE: 3 OF 8

I I~

P ~ ~ I I ~

JPE-LR-87-037 Rev 4 Page 14

. CASK FI 62 Gl AREA D2 CI 6

g7 7X7 ~'+ 7X (X7: 7X7 7X7 7X7 li 'pl 7X7 I

=NEV RACKS

": =

FX IST ING RACKS FIGURF 4 (,,)

N TITS: SEQUENCE OF RACK RE)iOVAL AND XNSTALLATZPN nuclear uautu cunecrucaan I ENCLOSURE: 4 OF 8

\

~

I

~ ~ k a ~ a o

~ 0 a ~

P" p7 p e + ~

Q

<t~

E'~ ~ ~

'4 ~ a 0 0 0 0 %e e

JPE-LR-87-037 Rev 4 Page 16 CASK G2 GI AREA D2 CI

.D3 82 7X7

= NEW RACKS

=

EX ISTING RACKS FIGURE 6,,

TITLE: SEQUENCE OF RACK REMOVAL AND INSTALLATION nuaaar uut>cu ccnsaucaon' ENCLOSURE: 6 OF 8

~ ~ 0~ ~

~ ~

'r~i~ rp 0 j

~ ~

0 i 5

lgr'g

/" ~i ~ ~

~ ~ t g 00 o

0 ~0

'l'4 t

JPE-LR-87-037 Rev 4 Page 18 CASK AREA Gl gD2 C3 CI E2 El D3 C2 DI B2 Bl A2 Al

=NORW RACKS

=FUEL FIGURE 8 I

- TITLE: SEQUENCE OF RACK REMOVAL AND INSTALLATION nuuaar uaueu conmrucuan I l

ENCLOSURE: 8 OF 8

QUESTION 47 Provide additional discussion on the temporary crane that will be used to handle the spent fuel storage racks within the fuel building. As part of the discussion, specify 1) whether the crane is single failure proof; 2) whether the temporary crane will be carried over the spent fuel pool; and 3) the load path that the fuel building crane will take to carry the temporary crane. Provide drawings of the temporary crane and pertinent details which demonstrates the crane's compliance with the guidelines of NUREG-0612 and NUREG&554. If the temporary crane is not single failure proof, provide the results of the analysis which demonstrates that the appropriate safety factors have been applied to the crane and that load handling by the crane will be in accordance with the guidelines of NUREG-0612.

RESPONSE: 'Ihe temporary crane is of double girder, symmetrically mounted double hoist construction designed per the guidelines of the Crane Manufacturer's Association of America (Standard CMAA70) ~

In addition, structural members in the crane structure meet the stress requirements of NUREG 0612. The crane traverses the pool in the north-south direction over rails which run adjacent to the pool edges resulting in an approximately 34'-6" crane span. %he rack installation procedure calls for heavy loads to be restricted to the regions of the pool where there is no fuel stored in the racks. 1he crane is not single failure proof, and is not qualified to handle fuel. It may, however, be used to perform dummy gage testing on the installed fuel racks to ensure that they meet the required opening size.

'Ihe temporary crane will be used as a staging platform on which to set the old racks while the lifting hardware is changed out to enable the racks to be removed from the fuel handling building.

'jhe spacing between the centerlines of the girders is accordingly set at 74.83 inches to coincide with center-to-center spacing of the old rack support channels. 3he temporary crane will also be used to set the new rack modules at their designated locations in the pool after the new racks are brought into the cask area by the cask crane.

%he cask crane will carry the temporary crane through the hatch door opening in five pieces; two end trucks, two girders and one hoist. 'Jhe 4 foot area in the north end of the pool will be utilized for installing the crane. While the temporary crane will be carried over the spent fuel pool, will over spent fuel.

it not be carried

ik ATTACHMENT 6 Page 1 of 7 RADIOLO ICAL CONSEQU NCES OF A SPENT FUEL POOL LOSS OF CO'COLIN" ACCIDENf 1 ~ IN MODU "TIOV In the course of reviewing the proposeB moiification of the spent fuel storage facility at th St. Lucie Plant Unit 1, the NRC has requesteB FP6L to provide tne results of an analysis to show that no off-site Bose limits will be exceedeB were a loss of the spent fuel pool cooling system to result in th spent fuel pool boiling.

2. METHODOLOGY

2. 1 Assumptions Tne analysis of the radiological consequences of a loss of th spent fu 1 pool cooling system is boundeB by the following assumptions:

1. the reactor has been shut down for s ven days prior to the start of tho accident (this is consistent with [1:3-28]*)

2. one full core has been BischargeB into the spent fuel pool

3. the dyna~ics of the pool boiling are characterized by the data in ref. [1:3-37]

4 ~ the leak rate co fficient for iodine leaking from defe"tive fu 1 rods is the same as for rods in the rea"tor seven days after shutdown [2:391]

(although this is data for a BWR, it is the only actually measured leak rate data available and should be representative of light-water reactors)

5. the 'radionuclide core inventory i.s taken fro~ ref. [3:Table 3.3] +

5. the pool water clean-up system will become inop rative when the pool cooling system fails 7 ~ no credit is taken for the emergency filtration system in the fu 1 haaBling building HVAC system

8. the tritiwa concentration in the pool water is taken from ref. [1:5-14]

• Numbers in square brackets are reference numbers, follow B by page or table number Cnange of fuel vendor and use of higher burnup dictates the use of an updated raBionuclide core inventory.

ATTACHMENT 6 Page 2 of 7

9. the leak rate coefficient for noble gases is five tie s that for lodines f2:399]

10. one percent of the fuel rods in the assemblies in the pool are defective

11. the atmospheric dilution factors (Xi/Q) at th n wrest site boundary distance and at the outer boundary of the low population zone (LPZ) are taken from ref. [8:2.3-24J

12. the leakage of halogens and noble gases from spent fu 1 from previous refuelings is negligible

13. the concentrations of halogens and noble gases in the sp nt fu 1 pool at the start of the accident are negligible

'4. a11 halogens* and noble gases leaking from the fuel rods vt.ll evolve directly into the atmosphere following pool boiling

15. the eater level inside the fuel pool ~l.ll b 23 feet above the top of the fuel at the start of the a=cident

16. sufficient make-up eater vill be provide9 to keep the fu 1 covered vlth eater at all times

17. the a"cident is presumed to last 30 days +

• Bee "Discussion", below.

Tnis assumption is consistent with analyses of other design basis accidents involving prolonged releases, su h as a reactor loss-of-coolant accident (LO"-A) ~

2.1.1 Di,s ussion The assumption that all halogens leaking from the fuel rods are released directly to the atmosphere is highly conservative. It is supported by th general principle that dissolved gases become less soluble with in"reasing temperatures, ~1th the solubility becoming ess ntially sero st the boiling point of tho solvent. Furthermore, since the cleanup system is assumed to have failed, there is no mechanism for removing the build-up of radioidines except evaporations The concentration in the ester vill thus increase @bile the accumulation in the air ut.ll be removed by the ventilation system.

Eventually, a quasi-static equtlibrium vill be achieved the concentration in the eater ~111 reach a maximum value and the rate of volatilization from the pool vill be equal to the rate of release from the fuel.

ATTACHMENT 6 Page 3 of 7 Tnis simplifie3 model leads us to conclude that, re~ardless of the a"tual partition co fficient, all the escaping io5ines util eventually enter the atmosphere. A more detailed r

calculation of this non-equilibrium situation

'~ould result in someinat lo~.r release rates, and therefore lour radiolo~ical impacts ~ Su"h a calcula ion would take credit for non-volatile fra"tions of the raiioiiine and also vould sprea0 the releas s over a longer time period, resulting in lover con"entrations due to raliosctive decay. Performing such s calculation 9oes not appear to be )ustified, however, since reliable data for the behavior of radioiodine in boiling paol eater is not readily available.

2.1.2 Input data Atmospheric Dilution Factors (Xi/Q)

Loca t ion Time After Accident XX/Q (seclm )

Exclusion area bomb ar y 0 - 2 hrs 8.558-5

• Lou population zone 0 - 8 hrs 7.97E-6 8 - 24 hrs 4.26".-6 1 - 4 days 9.53E-7 4 - 30 days 3.688-7

• 8-558-5 ~ 3.55 x 10 Fission Product Core Inventory:

Nuclidea " Half-life Core Inventory (Ci)

During op ration 7 days after shutdovnc 131I 8.02 days 7. 84E7 4.28E7 132I 78.2 hoursd 1.12Z8 2. 53" 7 133xe 5.25 days 1.4788 5. 8387 10.7 years 1.0185 1 01E5 Nuclides significant to the present analysis bSee ref. [7]

Values calculated, based on equilibrim inventory and radioactive decay Half-life of parent tellurian-132 (132Te)

The half-life of I is 2.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />s- Hovever, it is in secular cquilibriun vl.th its parent nuclide, Te. Tnerefore, the decrease of its activity kn the core seven days after shutdown vt.ll be almost exactly the same as the decrease of the activity of its parent.

ATTACHMENT 6 Page 4 of 7 2.2 WEI'833 3F CALCULATIOV Tne data in referen"e {2] inlicate that th>> fra" tional leak rate of a given nuclide from the defective fuel rods seven days after shutdown would depend primarily on its ch mical prop rties an9 secon9arily on its half-life. Tnat report also observes that the ratio of iodine to noble gas releases during op ration was approximately 1:5 for nuclides with similar half-lives ~ Tne present analysis assumes that the leak rate coefficient of I is the sate as that given for I (the data indicates that it would be somewhat smaller) ~ Tne rates of Xe and Kr are calculated bas d on the data in that report.

The absorbe9 dose commitm nt to the adult thyroid for radioiodines at a given location is calculated as follows:

D Ii rX B Ri(e ki )o - e i )1) ki D<~ adult absorbed dose due to nu"lide i during tive period 5

~ core inventory of nu"lide i at commencement of pool boiling f ~ fractional leak rate of damaged fuel rod

( )

r ~ fraction of failed fuel

.01 (1X)

X~ Atmospheric dilution fa"tor (Xi./Q) during time p riod ) from start of a"cident (sec/m3)

B~ ~ adult breathing rate during time period j f4:1.4-2) 1.25 m3/hr (0 - 8 hrs from start of accident)

.625 m3/hr (8 - 24 hrs )

.833 m3/hr (1 30 days )

R< Adult dos conversion factor (rads/curie inhaled) [5:25.5j 48 6 for 131 1 5.35E4 for 1321 ki decay constant of nu"lide i (hr )

t~o ~ time from start of accident to beginning of period 5 t~l ~ time fro~ start of accident to end of period 5 Tne dos s to the whole body and to the skin from noble gases are calculated as follows:

h ATTACHMENT 6 Page 5 of 7 IifrX R1(e ~i )o - e ki jl)

D i) k All terms have the same meaning as in the previous equation, except:

R< ~ semi-infinite clou9 dos conversion factor [6:1 109-21J f ~ 4 ~ 35E 9 se" ( Xe)

The atmospheric tritium concentration is calculated as follows:

C ~ PcX~

C ~ tritium concentration in air (uCi/m3)

F boil-off rate of pool water

~ 4,377 ml/sec c ~ tritium concentration in pool water

.077 uCi/ml

3. RESULTS 3.1 Dose Commitment at Exclusion brea Boundary (EAB)

Nu"lide Adult absorbed dose commitment at EAB (rem)

Tnyroid Skin Whole Body 131I 0.117 132I 2.50K-3 133'5Kr 1. 51E-5 1.458-5 1.15K-6 1 38"-8

~

Total 0. 119 1.62E-5 1.458-5 3.1.1 Tritium concentration st EAB During the first two hours following the accident, the maximus tritium con"entration st the Exclusion Area Boundary (EAB) is calculated to be 2.98-8 uCi/ml (2.9E-2 uCi/m3) in air. This value is well below the maximum permissible concentration (MPC) for tritium, which is 2E-7 uCi/ml. The MPC is the limiting concentration for a chronic exposure. Since an individual at the EAB is postulated to be exposed for a maximum of two hours, the contribution of tritium to his total dose is insignificant.

3.2 Dose Commitmen at LPZ The following calculations were performed for the maximum exposed individual at the LPZ outer boundary during the 30-day period of the accident.

ATTACHMENT 6 Page 6 of 7 Nu"lihe A]ult ahsorbe4 dose coz~itm nt at LPZ (re~)

Tnyroid Skin lAole Body 1311 4. 121 132I 1. 92E-3 133~e 1 ~ 89'5 1.82E-5

~

85Kz 2 '9 -6 3.488-8 Total 0.123 2 '8E"5 1.82E-5 3.2.1 Tritium concentration at LPZ During the first eight hours following the a=cident, the maximus time-averagei tritium con"entration at the LPZ boundary is calculated to be 2.7E-9 uCi/ml in air. This value is sell below the MPC for tritium. Since the atmospheric dilution factor (Xi/Q) used in this calculation is applicable to the first tvo hours of the accident only, and since in any cas the accident is postulate9 to last only 30 days, the contribution of tritium to h total dose is insignificant.

4~ CON"LUSZONS The maximus calculated thyroi0 dos is much'less than 1X of the 10 CFR 100 limit of 303 rem. The Aaole-boiy and skin doses are completely insignificant.

REFERENCES

1. Florid'a Pover 8 Light Company: "St- Lucie Plant Unit No. 1, Spent Fuel Storage Facility Modification Safety Analysis Report, Docket No. 5'3-335.

2. Eickelpas"h, N., and R. Hock: "Fission Product Release after Reactor Stutdeun", LhEA-SS-17S/i9, in Erperience free ~Operetin end Fuelling Nu=lear Pover Plants, Vienna, 1974

3. Exxon Nuclear Company, Inc.: "St Lu"ie Unit 1 Radiological Assessment

~

of Postulated Accidents". XN-NF-84-85(P), R v. l.

4. V.S. Atomic Energy Commission, Regulatory Guide 1.4: Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors". Rev. 2, 1974.

5'.S. Nu"lear R adulatory Commission, Regulatory Guide 1.25: "Assunptions us d for Evaluating the Potential Radiological Consequences of a Fuel

ATTACHMENT 6 Page 7 of 7 Handling A"cident in th Fuel Handling and Storage Facility for Boiling and Pressurize) M~ter R a" tors" ~

6 J.S. Nuclear Regulatory Co~mission, Regulatory Guide 1.109: "Calculations of Annual Doses to Non fro~ Routine R leases of R actor Effluents for the Purpose of Evaluating Compliance witn 10 CFR Part 50, Appendix I" ~

Rev. 1, 1977 ~

7. Lederer, C. M., and V. S. Shirley, eds: Table of Zsotop s, 7th Ed ~, 1978

8. Florida Pow r 5 Light Company: "St Lu" ie Plant

~

- Unit No ~ 1, Updated Final Safety Analysis Report", Amendment No. 6.

7

(

'V r