ML18058B392

EA-A-NL-91-169-01 Offsite Dose Calculations of Fuel Handling Accident.

ML18058B392
Person / Time
Site:	Palisades
Issue date:	12/17/1992
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18058B389	List: ML18058B388 ML18058B390 ML18058B391 ML18058B392
References
EA-A-NL-91-169, EA-A-NL-91-169-01R00, EA-A-NL-91-169-1R, NUDOCS 9302040167
Download: ML18058B392 (15)

Text

ATTACHMENT 4 Consumers Power Company Palisades Plant Docket 50-255 EA-A-NL-91-169-01 OFFSITE DOSE CALCULATIONS OF FUEL HANDLING ACCIDENT January 29, 1993 9302040167 930129 PDR ADOCK 05000255 P PDR

PALISADES NUCLEAR PLANT EA-A-NL-91-169-01 ENGINEERING ANALYSIS COVER SHEET Total Number of Sheets 14 Title Offsite Dose Calculations of Fuel Handling Accident INITIATION AND REVIEW Calculation Status Preliminary Pending Final Superseded Q Q x Q Initiated In it Review Method Technically Reviewed Revr Rev Appd Appd CPCo I Description By Detail Qual By Appd By Date Alt Cale Review Test By Date tt'fuca rte.£) #9/ltl'h

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PURPOSE:

The purpose of this analysis is to demonstrate that the offsite radiological doses of a Fuel Handling Accident (FHA), using the Regulatory Guide 1.25 source term, will be within the limits of 10 CFR 100 as defined in the Standard Review Plan 15.7.4. This analysis accounts for increased radial peaking for Cycle 11, as well as potential future increases in radial peaking factor up to a value of 1.8 for the peak assembly.

SUMMARY

OF RESULTS:

The offsite doses from a fuel handling accident were calculated, bounding the maximum radial peaking factors in Cycle 11 and beyond. This was done to accommodate an increase in the peaking factor to 1.76 in Cycle 11. The analysis followed the guidelines and assumptions of Regulatory Guide 1.25 and the Standard Review Plan. The resultant doses were calculated to be 41.19 rem thyroid and 0.12 rem whole body for 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at the site boundary, and 7.23 rem thyroid and 0.02 rem whole body for 30 days at the low population zone distance. These calculated doses are well within the limits of 10 CFR 100, as interpreted by the Standard Review Plan 15.7.4.

q (@consumers.

• • Power . PALISADES NUCLEAR PLANT E-A-NL-91-169-01

.AUUOUln 1'llWUlllM l'WinSS ANALYSIS CONTINUATION SHEET Sheet _2_ Rev # --""o__

Table of Contents 1.0 OBJECTIVE 3

2.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

3.0 BACKGROUND

4 4.0 ANALYSIS INPUT 4 5.0 ASSUMPTIONS ............................................. * . . . . . . 6 6.0 ANALYSIS 6 7.0

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

PALISADES NUCLEAR PLANT E-A-NL-91-169-01 ANALYSIS CONTINUATION SHEET Sheet 3 Rev # _ _,.a_ _

1.0 OBJECTIVE The purpose of this analysis is to demonstrate that the offsite radiological doses of a Fuel Handling Accident (FHA), using the Regulatory Guide 1.25 source term, will be within the limits of 10 CFR 100 as defined in the Standard Review Plan [Ref. 2.1]. This analysis accounts for increased radial peaking for Cycle 11, as well as potential future increases in radial peaking factors up to a value of 1.8.

2.0 REFERENCES

2.1 Regulatory Guide 1.25 Rev 2, "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," March 1972.

2.2 NUREG-0800, USNRC Standard Review Plan, Section 15.7.4 Rev 1, "Radiological Consequences of Fuel Handling Accidents.

2.3 Letter from D.L. Ziemann (NRC) to D. Bixel (CPCo) dated June 21, 1979.

Subject:

Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding the Fuel Handling Accident Inside Containment. Cart/Frame: 2600/0540.

2.4 Letter from D.L. Ziemann (NRC) to D.P. Hoffman (CPCo) dated February 5, 1980.

Subject:

XV Radiological Consequences of Fuel Damaging Accidents (Inside and Outside Containment). Cart/Frame: 2614/0667.

2.5 NED0-24782, "BWR Owners' Group NUREG-0578 Implementation: Analysis and Positions For Plant Unique Submittals," General Electric, August 1980.

2.6 NUREG/CR-1413 ORNL/NUREG-70, "A Radionuclide Decay Data Base - Index and Summary Table," Oak Ridge National Laboratory, May 1980.

2.7 NUREG/CR-5009 PN-6258, "Assessment of the Use of Extended Burnup Fuel in Lightwater Power Reactors," Pacific Northwest Laboratory.

2.8 ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers," July 1978.

2.9 Regulatory Guide 1.4 Rev 2, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," June 1974.

2.10 Palisades Plant Final Safety Analysis Report.

2.11 Palisades Plant Technical Specifications, Section 3.8.3, through Amendment 153.

2.12 EMF-92-178, "Palisades Cycle 11: Disposition and Analysis of Standard Review Plan Chapter 15 Events," Siemens Power Corporation, November 1992.

PALISADES NUCLEAR PLANT E-A-NL-91-169-01 ANALYSIS CONTINUATION SHEET Sheet _4_ Rev # _...;.o__

2.13 E-PAH-92-03, "Core Decay Heat Release Fractions Using NRC Branch Technical Position ASB 9-2" 2.14 Letter from H.G. Shaw (Siemens) to R.J. Gerling dated October 6, 1992.

Subject:

Fuel Internal Rod Pressure at End of Life.

3.0 BACKGROUND

The FHA (Fuel Handling Accident) analysis is the bounding radiological consequence analysis for the dropping of a fuel assembly. For an FHA, the source terms of Regulatory Guide 1.25 [Ref. 2.2]

and Standard Review Plan (SRP) 15.7.4 [Ref. 2.1] are used. This source term amounts to 10% of the assembly iodine, krypton, and xenon found in the gap between the fuel pins and the cladding escaping to the outside atmosphere. The exception to this is Kr-85, where 30% escapes to the outside atmosphere. NUREG/CR-5009 updates the value of escaping I-131 to 12% due to higher fuel bumup.

The limits for offsite dose are 300 rem thyroid and 25 rem whole body, as found in 10 CFR 100.

The Standard Review Plan 15.7.4 [Ref. 2.2] sets the acceptance criteria for a FHA as "appropriately within the guidelines" of 10 CFR 100 and gives the value of 25% of the 10 CFR 100 limits, or 75 rem thyroid and 6.25 rem whole body. In Reference 2.3, the dose of 91 rem thyroid is given and deemed acceptable. In Reference 2.4, the limit is given as 100 rem thyroid for Palisades Nuclear Plant. This value is the licensing basis thyroid dose limit for a FHA at Palisades.

4.0 ANALYSIS INPUT

• 4.1 The breathing rate for offsite doses is 3.47E-4 m3/sec in accordance with Reference 2.9.

This is the breathing rate given for a person offsite during the first eight hours of the accident. This is also the maximum breathing rate given, so it is deemed conservative.

4.2 The number of fuel assemblies in the core, 204, is found in Reference 2.10, Section 3.3.1.

4.2 The rated core thermal power, 2530 MWu is from Reference 2.10.

4.3 The assembly radial peaking factor of 1.8 was used for this analysis. The value of 1.76 is the factor for cycle 11 and can be found in Reference 2.12, but 1.8 was used to bound radial peaking.

4.4 The decay heat fractional energy release 2 days after shutdown, based on Reference 2.13, is less than 0.003. This is less than twice the value used in Reference 2.14. This yields a Linear Heat Generation Rate less than the value given in Reference 2.14. This makes the fuel rod pressure given in Reference 2.14 valid.

4.5 The fuel rod gas pressure is less than 1200 psig based on Reference 2.14. This meets the requirements needed to make the Reference 2.1 assumptions valid. Reference 2.14 can be found as Attachment 8.1. *

(@ consumers Power l'flWllUIM lft l'flOA£5S PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET E-A-NL-91-169-01 Sheet _s_ Rev # --=-0_ _

4.6 The Pool Decontamination Factors are 100 for iodine and 1 for all other nuclides. This is taken from Reference 2.1 based on Reference 2.14.

4.7 The radionuclide activities for one day after reactor shutdown were found in Reference 2.5 and are listed in Table 1. For the iodine isotopes, the values in Table 1 are doubled because Reference 2.5 gives 50% of the value needed.

4.8 The radionuclide half lives are found in Reference 2.6 and are listed in Table 1. The values found in Reference 2.6 were converted to minutes before being input into the table.

4.9 The dose conversion factors for the radionuclides are from Reference 2.8, and are listed in Table 1. The values found were converted to new units before being input into the table.

Table 1 Nuclide Si t112 Dose Conversion (Ci/MW1) (min) Factors See Note 1-131 2.696E+4 11577.6 1073000 1-132 7.204E+2 138 6290 1-133 2.178E+4 1248 181300 1-134 3.040E-4 52.6 1073 1-135 4.778E+3 396.6 31450 Kr-83m 6.039E+O 109.8 .000003 Kr-85m l.489E+2 268.8 .03031 Kr-85 2.999E+2 5638291 .000473 Kr-87 2.295E-2 76.3 .1447 Kr-88 4.452E+ 1 170.4 .369 Xe-131m l.708E+2 17049.6 .001324 Xe-133m l.547E+3 3153.6 .005375 Xe-133 5.171E+4 7552.8 .006259 Xe-135m 6.693E+2 15.36 .07647 Xe-135 7.388E+3 546.6 .04676 Xe-138 4.830E-27 14.13 .1969 NOTE: The D.s>se Conversion Factors are (Rem/Ci) for Iodine and (Rem/sec)/(Ci/m3 ) for all other nuclides.

. ~ Power consumers PALISADES NUCLEAR PLANT E-A-NL-91-169-01 l'llWUIBM ANALYSIS CONTINUATION SHEET

.~#SNOAW Sheet _ 6 _ Rev # __o __

S.O ASSUMPTIONS 5.1 The core is assumed to have run at 102% of rated power, or 2580.6 MWt.

5.2 The FHA occurs inside containment, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown, as per Reference 2.12. This creates the most severe consequences for a fuel handling accident.

5.3 One fuel assembly is completely damaged as a result of the fuel handling accident. This is consistent with Reference 2.1.

5.4 The fuel assembly with the peak inventory is the one damaged, consistent with Reference 2.1 5.5 Consistent with References 2.1 and 2.7, 30% of the Kr-85 found in the gap between the cladding and the fuel is released, along with 12% of the 1-131 and 10% of the remaining iodine, krypton, and xenon.

5.6 The release of radionuclides fails to trip high-radiation alarms and escapes from containment, unfiltered, into the environment. There is a direct flow path through one of the clean waste receiver tanks as long as a containment isolation signal is not generated 5.7 The release is over a two hour period as per Reference 2.1.

6.0 ANALYSIS 6.1 Fuel Activity The first part of this analysis was to determine the activity of each radionuclide released to the reactor cavity water from the pellet-clad gap of the damaged assembly. First, the core inventory of each radionuclide per MWt one day after shutdown was found in Reference 2.5. The value was divided by 204 (number of fuel assemblies in the core) to find the activity per MWt for the one damaged assembly. This was multiplied by the 102% rated power of 2580.6 MWt as well as the _assembly peaking factor of 1.8 to find the activity in the peak assembly one day after shutdown. This value is then corrected for the extra day of decay using the radioactive decay equation. This value, unique to each radionuclide, is multiplied by the fraction of that nuclide in the pellet-clad gap that is released to yield the total nuclide release to the water from the damaged assembly. The equation for this can be written as:

-LN(2) 1440 S- PF tl/2 A.=F( .z r)e

.1 g 204 where

~ = activity released to water from pellet-clad gap, Ci Si = Source term of individual iso.tope, Ci/MWt Fg = fraction of inventory in the pellet-clad gap inventory released P = power of plant at 102% of rated power, MWt Fr = peaking factor of peak assembly t 112 = half life of nuclide, in minutes

PALISADES NUCLEAR PLANT E-A-NL-91-169-01 ANALYSIS CONTINUATION SHEET Sheet 7 Rev # -~o __

6.2 Activity Released Outside Containment The next step in this analysis was to find the activity that is released from the water and into the outside air. This was done using the overall decontamination factor for the water. For all iodine isotopes, Dcff is 100, meaning that one hundredth, or 1% of all iodine released from the pellet clad gap makes it to the outside air. For all other isotopes, D ff is set to 1, meaning t.hat 0

all of the activity released from the pellet clad gap makes it to the outside air. The denotation Q is used to represent the activity of each nuclide released to the outside air.

From there, the dose received due to each isotope was computed. This was done by multiplying the Q value for each isotope by the dose conversion factor (DCFerr)* Then it was multiplied by the Atmospheric Dispersion Factor. This number is based on the distance away from the location, so two values were used: one for the site boundary and one for the low population zone. For the iodine isotopes, the value produced was multiplied by the breathing rate, since the calculated dose from iodine is thyroid dose due to inhalation. This is represented by the following equations:

• H :b=Q DCF w.

_x Q

for iodine isotopes and:

Hthy= (Q DCF i) BR for all other radionuclides where Hlhy = dose to thyroid from inhalation of iodine Ifwb = dose to whole body from xenon and krypton exposure x/Q = atmospheric dispersion factor for location DCF = dose conversion factor for isotope Q = activity released outside containment BR = breathing rate of an individual subject to radionuclide exposure The values of Hlhy were then summed separately for the site boundary (sb) and the low population zone (lpz). The same was done for Hwb*

The values found as maximum doses are:

Hthy(sb) = 41.19 rem Hthy(lpz) = 7.23 rem Hwb(sb) = 0.12 rem Hwb(lpz) = 0.02 rem These values are well within the values given as limits by References 2.2 and 2.4.

PALISADES NUCLEAR PLANT E-A-NL-91-169-01 ANALYSIS CONTINUATION SHEET Sheet _8_ Rev# _ _;o~-

7.0

SUMMARY

This analysis was performed to demonstrate that the radiological consequences of a fuel handling accident are within the limits of References 2.2 and 2.4. The activity in the pellet-dad gap of the peak fuel assembly at the time of the accident was found. Then, the fraction of activity that is would be released is accounted for. The decontaminating ability of the water was then factored in to provide the activity released outside containment of each radionuclide involved. Finally, the dose related to this release was found at the site boundary as well as the low population zone, and the thyroid dose was corrected for the breathing rate of the person subject to the exposure. The exposures calculated for each isotope involved were them summed to find the overall exposures encountered.

The doses calculated at the site boundary were 41.19 rem thyroid and 0.12 rem whole body dose. The doses for the low population zone were calculated to be 7.23 rem thyroid and 0.02 rem whole body.

These are well inside the limits established and are considered acceptable.

@) consumers Pawer PALISADES NUCLEAR PLANT E-A-NL-91-169-01

.M.:MliA#Sl'tlWllUW l'flOADS ANALYSIS CONTINUATION SHEET Sheet _ 9 _ Rev # _ _,a,___

List of Attachments 8.1 Reference 2.14: Letter from H.G. Show (Siemens) to R.J. Gerling dated October 6, 1992.

Subject:

Fuel Internal Rod Pressure at End of Life.

8.2 Technical Review Checklist 8.3 Engineering Analysis Checklist 8.4 Document Review Sheet

Siemens Power Corporation Proprietary SIEMENS October 6, 1992 HGS:327:92 Mr. R. J. Gerling Consumers Power Company Palisades Nuclear Power Plant 2nao Blue Star Memorial Hwy.

Covert, Ml 49043-9530

Dear Bud:

In response to information requested by CPCo, a RODEX2 gas pressure analysis for a fuel rod stored in the Palisades spent fuel storage pool has been performed.

Key input variables for the analysis are the FSAR design basis of 1SOOF for the spent fuel storage pool and an LHGR = 0.05 kW/ft. ANSl/ANS-5.1-1979 indicates that the ratio of the decay heat rate to the heat rate for infinite operation is 0.0028 two weeks following shutdown. Multiplying 0.0028 by the core average LHGR (5.23 kW/ft) gives 0.014 kW/ft; thus, the LHGR of 0.05 kW/ft is conservative. The hot operating end of life internal rod pressure was 2698 psia.

The analysis indicates that gas pressure in the fuel rod stored in the pool will be less than 1170 psia. For reference purposes, the SPC calculation number is E-5059-337-SX.

• If you have any questions, please contact Mr. Jim Hulsman.

Very truly yours,

~H. G. Shaw Contract Administrator tlm c: T. C. Bordine, CPCo Jackson

• • • • M*W*Palisades J. W. Hulsman, SPC S. F. Pierce, CPCo Palisades Siemens Power Corporation Nuclear Division - Headquarters 155 108th Avenue NE, PO Box 90777 Bellevue, WA 98009-0777 Tel: (206) 453-4300 Fax: (206) 453-4446

Pree No 9.11 TEQlllC,\L BEYJEV CHECXLJSI Athchment 5 Revision S EA

• 4-Al!..-1/-/6'1-cl REY. 0 Page 1 of 1 This checklist provides guidance for th* review of 1ngine1ring analyses.

Answer questions Yes or No, or N/A if tn.y do not apply. OocW11nt all conaents on a 3110 Font. Satisfactory resolution of COlllllnts and completion of this checklist 1s noted by th* Technically R1v1ewld signature on the Initiation and Review record block of Fon1 3619.

(Y, ~ N/A)

1. Have the proper input codes, standards and design r_

principles been specified? *

2. Have the input codes, standards and design principles b11n l./

properly appl itd? ---'---

3. Are all inputs and assu.ptions valtd and the basis far V 4.

th11r use docU111nted?

Is V1ndor 1nformatton used as input addressed correctly tn the analysis? V

s. If the analysis 1rgU111nt departs frOll Y1ndor Infon11t1on/Recetm11ndations, is the dep&rturc just1f1catton documented? N(d

. &. Are assU11Pt1ons accurately dtscriblcl and reasonable! t

7. Has* th* use of 1nginHring Judg-t bffft doc..,.tect and Justified? *

• y

8. Are 111 constants, variables and for.alu cornet and properly applied! v 5 et? ,:;*ttJJcAel

9. Have any *1nor (instptfi,cant) 1rron beea 1dtfttif1td1 If yes; Id1nttfy on the 3110 Fol'll UICI Justtf1 their 3r1c&,,.,,A.-

1ns1gntftcanca. l

10. Does anal1111 tnvolv* *lcltftl? If Yes; verifJ tl9e following tnfor111t1* ts accurately repnsetlcl Oii the analysts dr*tnt (CJuttNt doc-t).

• Type of W.1*

• Sia"' Vllct *

• Md9tal Betnt Joined .

• TMclmu* of Material Being Joined

• locatt* of V.lcl(s)

• Appropriate Veld Symbolog vI

n. Has tha ob.jecttve of tha analysis bee *t?

12. Have adlltntstrattve requ11"1111ftts suclt 11 ftUIDlr1ng an4 for11at beta sattsfteclf

Form 3698 9-89 PALISADES NUCLEAR PlANT ENGINEERING ANALYSIS CHECKLIST Affected Revision Items Affected By This EA Yes No Required Identify* Closeout

1. Other EAs 0 }Zf

2. Design Documents Elec E-38 through E-49 0

3. Design Documents Mech J2f M259, M664, M665 0 Pf 4.0 LICENSING DOCUMENTS 4.1 Final Safety Analysis Report (FSAR) 11 0 Yes FSAe 14. \ 4 FC-t'/3L/

4.2 Technical Specifications 0 ~

4.3 Standing Order 54 D tzJ 5.0 PROCEDURES 5.1 Administrative Procedures 0 ~

5.2 Working Procedures D ~

5.3 Tech Spec Surveillance Procedures D ~

6.0 OTHER DOCUMENTS

~

Q-List 0 Plant Drawings 0

~

6.3 Equipment Data Base 0 6.4 Spare Parts (Stock/MMS)* 0 JZI

~

6.5 Fire Protection Program Report 0 (FPPR) 6.6 Design Basis Documents D JZI 6.7 Operating Checklists 0 fZl 6.8 SPCC/PIPP Oil and Hazardous Material Spill Prevention Plan 0 JZf 6.9 EEQ Documents 0 SA' Do any of the following documents need to be generated as a result of this EA:

Yes No

1. Corrective Action Document? 0 ;zf Reference

2. Safety Evaluation? .er D Reference

3. EEQ Evaluation Sheet? 0 0' Reference C Review of this EA Required? 0 0 Completed B y >>  ?!/~ Date 12~1/~

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• Identify Section, No, Drawing, Document, etc.

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• • IM*n***rw NUCLEAR OPERATIONS DEPARTMENT Document Review Sheet Rev1s1on Number Page / of 2.

Item Page and/or Number Section Number Comments Response or Resolution I

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Tee..( .£ev, el(,- f3.4*~-A/t.-1/-/1 'J I~<'~ 2 er/'* ;t Site Boundary Low Population Zone t=24 hrs t=24 hrs (1/min) t=48 Thyroid Whole Bod Thyroid Whole Body Ci/MW Aa (Ci) Lambda Aa (Ci) Q (Ci) (Rem) (Rem) (Rem) (Rem)

Kr-83m 6.039E+OO 137.5 6.313E-03 0.015 0.001549 8.8E-13 1.5E-13 Kr-85m 1.489E+02 3390.5 2.579E-03 82.678 8.267843 0.000038 0.000006 Kr-85 2.999E+02 6828.7 1.230E-07 6827.514 2048.254 0.000150 0.000026 Kr-87 2.295E-02 0.5 9.084E-03 0.000 0.000000 2.4E-12 4.3E-13 Kr-88 4.452E+01 1013.7 4.068E-03 2.896 0.289638 0.000016 0.000002 Xe-131m 1.708E+02 3889.1 4.065E-05 3667.998 366.7997 0.000075 0.000013 Xe-~33m 1.547E+03 35225.2 2.198E-04 25668.112 2566.811 0.002138 0.000375 Xe-133 5.171 E+04 1177436.7 9.177E-05 1031682.524 103168.2 0.100088 0.017563 Xe-135m 6.693E+02 "15240.0 4.513E-02 0.000 9.1E-26 1.1 E-30 1.9E-31 Xe-135 7.388E+03 168224.8 1.268E-03 27095.860 2709.585 0.019638 0.003446 Xe-138 4.830E-27 0.0 4.906E-02 0.000 2.3E-57 7.0E-62 1.2E-62 1-131 2.696E+04 613879.2 5.987E-05 563172.157 675.8065 39.00168 6.844165 1-132 7.204E+02 16403.5 5.023E-03 11.848 0.011847 0.000004 0.000000 1-133 2.178E+04 495930.6 5.554E-04 222885.904 222.8859 2.173409 0.381398 1-134 3.046E-04 0.0 1.318E-02 0.000 4.0E-14 2.3E-18 4.0E-19 1-135 4.778E+03 108795.1 1.748E-03 8778.854 8.778854 0.014849 0.002605 41.18994 0.122146 7.228170 0.021434