ML17319B670

DC Cook Unit 2 Potential Radiological Consequences of Incidents Involving High Exposure Fuel

ML17319B670
Person / Time
Site:	Cook
Issue date:	11/17/1982
From:	Hansen L, Krysinski T, Williamson H SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
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ML17319B669	List: ML17319B668 ML17319B670
References
XN-NF-82-90, NUDOCS 8211290388
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XN-NF-82-90 Issue Date: ll/17/82 D.

C.

COOK UNIT 2 POTENTIAL RADIOLOGICAL CONSEQUENCES OF INCIDENTS INVOLVING HIGH EXPOSURE FUEL prepared by: ~C jy'

/ 2-

ansen, Unit anager BWR Neutronics Reviewed by:

. L. Krys ns i, anager BWR Neutronics Approved by:

'H.

E. Williamson, Manager Neutronics 8 Fuel Manageme Approved by:

tout, anage Licensing and Saf Engineering Approved by:

o er>. anager Fuel Engine'ering 5! Technical Services

/mar E)5(ON NUGAEAR COMPANY,Inc.

PDR ADOCK 0500031h P,.....

PDR

IHPORTANT NOTICE REGARDING CONTENTS At(0 USE OF THIS OOCUHENT PLEASE READ CAREFULLY This technical repor t was derived through research and'development programs sponsored by Exxon Nuclear Company, Inc.

It is being submitted by Exxon Nuclear to the USNRC as part of a 'technical contribution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear-'fabricated reload fuel or other technical services provided by Exxon Nuclear for light water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief.

The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulations.

Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf A.

Hakes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the.

information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or O.

Assumes any liabilities with respect to thn >>s~ nf, nr for damages resulting from the use of, any infonnaLiun apparatus, method, or process disclosed in this document.

XN-NF-82-90 TABLE OF CONTENTS Section

1.0 INTRODUCTION

2.0

SUMMARY

AND CONCLUSION 3.0 EFFECTS OF FUEL EXPOSURE AND REACTOR POWER INCREASES 4.0 RADIATION SOURCES 4.1 FISSION PRODUCT INVENTORIES.

4.2 ISOTOPIC COMPOSITION OF FISSION PRODUCTS 4.3 DOSE PROPORTIONALITY CONSTANTS 5.0 ACCIDENT CONSEQUENCES.

5.1 FUEL HANDLING ACCIDENT 5.1.1 Auxiliary Building 5.1.2 Inside Containment 5.2 LOCKED ROTOR INCIDENT.

5.3 RCC ASSEMBLY EJECTION INCIDENT

~Pa e

2 4

7 7

10 10 12 12 12 13 13 14 5.4 STEAM GENERATOR TUBE AND MAIN STEAM LINE RUPTURES...

15 5.5 LOCA EVENT

6.0 REFERENCES

16 18

I l

5 i

t

XN-NF-82-90 LIST OF TABLES Tabl e

~Pa e

1 Assumptions for Calculating Fission Product Inventories.

19 2

Core and Gap Fission Product Activities 20 3

Fuel Assembly and Gap Fission Product Activities 100 Hours After Shutdown 22 4

Comparison of Computed Fission Product Activities 23 5

Summary of Activities Considered in Accident Evaluations 24 6

Constants Relating Off-Site Doses from Accidents Involving Design Base and Proposed Operations 25 7

Fuel Handling Accident in the Auxiliary Building Two Hour Site Boundary Doses 26 8

Fuel Handling Accident Inside Containment Two Hour Site Boundary Doses 27 9

Off-Site Doses Resulting from a Loss of Coolant Accident 28

1 I

XN-NF-82-90 D.

C.

COOK UNIT 2 POTENTIAL RADIOLOGICAL CONSEQUENCES OF INCIDENTS INVOLVING HIGH EXPOSURE FUEL

1.0 INTRODUCTION

Current operation of D.

C.

Cook Unit 2 is based, in part, on the results of analyses which conservatively demonstrate that the off-site radiological consequences of incidents are within the limits specified in 10 CFR 100.11.

Those analyses are documented in Chapter 14 of the Updated Final Safety Analysis Report (UFSAR) for D.

C.

Cook Units 1

and 2 dated July 1982.

Analyses presented in. the UFSAR include consideration of fission product inventories (radiological source terms) that result from opera-tion of Unit 2 at a power level of 3391 MWt and an end-of-cycle average fuel exposure of approximately 25,000 MWD/MTU.

Since increases in power and fuel exposure are planned for future D.

C.

Cook 2 operating cycles, consideration was given to the effect of those changes on the potential off-site radiological consequences of incidents.

The results of those analyses which support operation at a

reactor power of 3,425 MW with an end-of-cycle average exposure of 30,000 MWD/MTU are pres'ented herein.

I I

XN-NF-82-90 2.0

SUMMARY

AND CONCLUSION Changes in the potential off-site radiological consequences of incidents were evaluated considering planned increases in reactor power and fuel exposure.

Analyses included consideration of the 1) potential for increased fuel failures resulting from accidents/transients, 2) increases in gaseous and volatile fission product inventories within the fuel, 3) changes in the fractional release of gaseous and volatile fission products to the pellet-clad gap and plenum, and 4) differences in the isotopic mixture of fission products available for release from the fuel.

As a result of those analyses, it is concluded that:

o No increase in fuel failures will occur as a result of accidents/

transients due to the proposed increases in reactor power to 3425 MWt and the batch average discharge exposure to 40,000 MWD/MTU.

o A slight increase

(~5K) in the total inventory of gaseous and volatile fission products exists within the core at end-of-cycle conditions which would result in a proportional change in off-site exposures under extreme (maximum hypothetical)

LOCA conditions.

The off-site doses,

however, remain well within the limits specified in 10 CFR 100.

o The inventory of gaseous and volatile fission products which is most readily available for release (i.e., that contained in the pellet-clad gap and plenum) is lower in the ENC 17x17 fuel

XN-NF-82-90 when operating at the proposed increased core power and exposure than in the 15x15 fuel used as a basis for prior analyses which considered operation at the lower core power and average exposure conditions.

o Changes in the fission product isotopic mixtures that result from increases in power and exposure and which are available for release from the fuel are not of radiological significance (i.e.,

no changes occur to negate prior assumptions used for evaluating off-site radiological consequences of accidents).

o No changes occur (e,g.,

in fuel rod pressure, decontamination l

factors, etc.) which would invalidate prior assumptions made in evaluating the off-site radiological consequences of accidents.

The analyses performed and/or discussed herein conservatively demon-strate that the proposed increase in reactor

power, accompanied by the planned increase in batch averaged discharge

exposure, can be accomplished in continued compliance with the limits specified in 10 CFR 100.

XN-NF-82-90 3.0 EFFECTS OF FUEL EXPOSURE AND REACTOR POWER INCREASES The off-site radiological consequences of incidents (i.e., potential radiation doses at the Exclusion Area Boundary and within the Low Population 2one) are a function of the following:

1.

Gaseous and volatile fission product inventories available for release from the fuel (i.e., in the fuel matrix and/or within the fuel rod gap);

2.

The quantity of gaseous and volatile fission products released from the fuel in an incident (i.e., the number of rods damaged and the number of curies within those rods available for release);

3.

The mixture of i'sotopes released to the environment; 4.

The fraction and rate of release of fission products to the environment (e.g.,

containment leak rate, pool decontamination factor, filter efficiencies, etc.);

5.

Meterological conditions at the time of the incident; 6.

Distances from the point of release to the point at which the dose is evaluated; and 7.

Biological factors associated with the exposed population (e.g.,

breathing rate, biological pathways, etc.).

Of the above items, only items 1, 2 and 3* are affected by changes in the reactor power, fuel design, and fuel irradiation history (e.g.,

specific power and total exposure).

For those factors, proportionality The spent fuel pool decontamination factor for iodine is dependent on the fuel rod gap pressure.

However, calculations indicate that the pressure remains below the threshold value of 1200 psig at which the assumed decontamination factor remains conservatively valid (see Regulatory Guide 1.25).

XN-NF-82-90 constants have been developed to relate the thyroid and whole body doses resulting from changes in the'adiation source terms to doses previously determined for the various incidents considered in the DC Cook Unit 2 UFSAR.

In particular, thyroid and whole body doses associated with incidents involving operations at higher power and exposure levels can be expressed as:

02(TH)

= DI(TH) AI(DCF)I, and 2(WB) 1(WB)

EC, where; 0

(

)

=

Thyroid dose resulting from a particular incident involving high exposure fuel and high power operations.

0>(TH)

=

Thyroid dose calculated for DC Cook Unit 2 existing operating conditions.

A

=

The ratio of the total number of curies of iodine available I

for release from fuel operating at increased power and exposure to the number of curies available for release from fuel operating at conditions considered in the UFSAR.

DCF

=

The ratio of the thyroid dose per unit concentration I

(Rem/PCi inhaled) of the iodine isotopic mixture available for release from fuel s operating at increased power and

XN-NF-82-90 exposure to the dose per unit concentration of the iodine isotopic mixture available for release from fuels operating at conditions considered in the UFSAR.

0

=

Whole body external dose resulting from a particular incident involving high exposure fuel and high power operation.

0

=

Whole body external dose calculated for DC Cook Unit 2 existing operating conditions.

The ratio of the average energy per disintegration of the fission products released from fuels operating at high exposure and increased power to the average energy per disintegration of fission products released from fuels operating at conditions considered in the UFSAR.

The ratio of the total number of curies available for release from fuels operating at high exposure and increased power to the total number of curies available for release from fuels operating at conditions considered in the UFSAR.

The equation which relates thyroid doses is precise.

For whole body doses,

however, the relationship assumes that the ratio of B and Y doses remains unchanged when increasing the power and fuel exposure.

While not exact, that assumption should be valid for reasonable changes in the power and exposure of fuels considered herein.

~

li

~

~

~

XN-NF-82-90 4.0 RADIATION SOURCES 4.1 FISSION PRODUCT INVENTORIES Off-site radiological consequences of an incident involving irradiated fuel result from the release of fission product gases to the environs.

Such releases would normally involve only those fission products which are contained in gaseous form (noble gases and volatile fission products) within the fuel rod gap.

In some cases, however,'onsideration is also given to the total inventory of gaseous and volatile fission products to conversatively bound the possible off-site consequences of an incident.

Core and gap fission product inventories, computed for the current DC Cook Unit 2 17xl7 fuel assembly design, are given in Table 14.1-3 of the Unit 2 UFSAR.

Those data were calculated assuming full power (3391 NWt) operation for a period of 650 days.

Similar calcula-tions of core and gap fission product inventories for the DC Cook Unit 1

15x15 fuel assembly are shown in Table 14.A.2-1 in Appendix 14A of the Unit 1 UFSAR.

Comparison of the core and gap fission product inventories given in the UFSAR for the two fuel designs indicates that the total inventory remains virtually unchanged but the percent of the fission product gases released to the gap is lower for the current 17x17 design.

As a consequence of the above observation, it was concluded that the

XN-NF-82-90 off-site radiological consequences of accidents cited in Chapter 14 of the DC Cook Unit 1 UFSAR are also applicable to Unit 2 (see Section

.14.3.5 of the UFSAR for Unit 2).

For the purpose of evaluating the consequences of incidents involving irradiated fuel, consideration must be given to the total core fission product inventories and related gap activities or, as appropriate, to the same values for the worst case fuel assembly.

The total core and worst case fuel assembly activities calculated for OC Cook Unit 1 are given in Tables 14.A.2-1 and 14.A.3-1 in Appendix 14A of the UFSAR.

Examination of the data and assumptions used to obtain the core and gap activities given in Table 14.A.2-1 indicates that the activities are representative of operation at rated power for a period of 650 days (i.e., to a core average exposure of 24,560 NWD/NTU).

Assumptions used to obtain the activities for the "highest rated discharged assembly" given in Table 14.A.3-2 are somewhat more obscure.

Comparison of the iodine-131 and krypton-85 activities given for the "highest rated assembly" with those for the total core indicates that the highest fuel assembly power was 1.37 times the average value and the maximum assembly exposure was

~42,400 NWD/HTU (i.e.,

~ 1.73 times the core average value).

To develop the relationship between off-site accident doses for the current operational limits and those proposed for future operation, it is important to determine the radiological source terms for both

XN-NF-82-90 conditions in a consistent manner.

Consequently, fission product inventories and gap release fractions for the DC Cook Unit 1 15x15 fuel assembly were re-evaluated to establish a basis for comparison with the DC Cook Unit 2 ENC 17x17 fuel assembly designed to operate at an increased power and exposure.

Basic assumptions used for computing the fission product inventories and gap release fractions for the total core and worst case fuel assembly are given in Table 1.

For each case identified'in Table 1, the ORIGEN Code(

was used to calculate the fission product inventories for the assumed operating conditions and exposures.

Gap release fractions for the various isotopes were computed using the RODEX 2 Code with fission gas release models (2)

~

described in the proposed ANS Standard 5.4.

The results of the ORIGEN and RODEX calculations are given in Tables 2 and 3 for the core averaged and worst case fuel assemblies, respectively.

A summary of the total activity of various elements of interest is given in Table 4.

These data support the previous estimate of the exposure (42,400 MWD/MTU) used to derive data given for the "highest rated" fuel assembly.

Also, the data show that the total inventory of fission products of interest within the core and/or worst case fuel assembly increases slightly, but the quantity available for release from the pellet-clad gap decreases for the ENC 17x17 design even at the proposed higher power and exposure conditions.

A similar trend in data is apparent

10 XN-NF-82-90 in Table 5 which presents a summary of source terms needed for evaluating the relative accident consequences associated with the proposed operation at higher power and fuel exposure.

4. 2 ISOTOPIC COMPOSITION OF FISSION PRODUCTS The effects of changes in the isotopic composition of fission

products, released from the fuel or gaps, were examined using the DACRIN code.

In particular, the thyroid and whole body doses due to inhalation of the isotopic mixtures of fission products available for release from the 17x17 and design base fuels were evaluated.

These calculations support conclusions drawn by Scherpelz and Desrosiers relative to the (4) contribution of various isotopes to the whole body, thyroid and other organ doses which would result from similar accidents (see page 198 of Reference 4).

More importantly, there are no isotopic composition changes which result from operation at increased power and exposure which would negate prior assumptions relative to the importance of the thyroid and whole body external doses as the controlling factors in evaluating the consequences of accidents.

4.3 DOSE PROPORTIONALITY CONSTANTS Assuming that the rate and fractions of fission products released from the fuel remain unchanged from those presented in the UFSAR, then the thyroid and whole body doses for fuels other than those considered in the UFSAR are related as shown in Section 3.0.

Values for the constants

XN-NF-82-$0 which relate the dose for high exposure fuel with equivalent doses for the design base fuel have been calculated from the data given in Tables 2

and 3, from DACRIN calculations of thyroid dose per pCi of iodine

inhaled, and from the B

and Y disintegration energies given on the RIBD library Those constants, necessary for the computation of (5) off-site doses due to increased power and exposure, are given in Table 6.

l E

a

12 XN-NF-82-90 5.0 ACCIDENT CONSEQUENCES 5.1 FUEL HANDLING ACCIDENT Fuel handling accidents evaluated in the UFSAR consider accidental damage to a worst case fuel assembly.

Section 14.2.1.1 describes the assumptions and results of analyses of such an accident within the Auxiliary Building and Section 14.2. 1.2 addresses the occurrence of such an accident inside containment.

In each case the accident is assumed to occur 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown.

5.1.1 Auxi 1 iary Bui 1 din To evaluate the consequences of an accident in the Auxiliary Build'ing it was assumed that the gap activities calculated to be present r

in the worst case fuel assembly would be released (see page 14.2.1-9, Section 4.2.1. 1).

Consequently, for the higher power and exposure fuels, the off-site thyroid and whole body doses are proportional to the ratio of gap inventories of the worst case ENC 17x17 fuel assembly to the worst case 15x15 fuel assembly.

Calculated two-hour whole body and thyroid doses for the base case and proposed higher power and fuel exposure are given in Table 7.

As can be seen, potential radiation doses at the site boundary remain only a small fraction of the limiting values specified in 10 CFR 100 for accidents involving fuel used at proposed higher power and fuel exposures.

13 XN-NF-82-90 5.1.2 Inside Containment In considering a fuel handling accident inside containment, assumptions relative to the release of fission products were taken from Regulatory Guide 1.25 (conservative case) and Regulatory Guide 4.2 (realistic case).

Those assumptions are specified in Section 14.2. 1.2 of the UFSAR.

In each case it is assumed that the gap activity available for release from the fuel rods is a fixed fraction of the total inventory of the rod.

Consequently, the off-site dose for high exposure fuel operating at increased reactor power is proportional to the ratio of the total inventory of the isotopes within the worst case fuel assemblies.

(For the realistic case, the fraction of the inventory available for release also decreases since one row of fuel rods in the 17x17 design is a smaller fraction of the total assembly than is one row of rods in the 15x15 design.

That difference is conservatively ignored in calculating the doses for the high exposure fuel operating at an increased reactor power).

Calculated thyroid and whole body doses due to the higher exposure fuel and higher reactor power are given in Table 8.

Doses at the site boundary resulting from a fuel handling accident inside containment remain well

. below the limits specified in 10 CFR 100 for proposed operations at higher reactor power and with increased fuel exposure.

5.2 LOCKED ROTOR INCIDENT An analysis of the locked rotor event is presented in Section 14.2.1.6 of the Unit 2 UFSAR, but releases of fission product activity which result from that incident are not separately addressed.

The analysis

14 XN-NF-82-90 is based on conservative assumptions of the peaking factor at the time of the incident and, as a result, indicates that some fuel failures may occur as a result of the incident.

(The assumed value of F~

= 2.5 is greater than the limiting value given in the Technical Specifications for DC Cook Unit 2).

An analysis of the locked rotor incident was subsequently performed by Exxon Nuclear

'assuming operation of the reactor at the proposed (6) increased power level of 3425 MWt and with the proposed increased total peaking factor (F

) of 2.1 and an F of 1.60.

This analysis demonstrates that an adequate margin in the MDNBR exists such that no fuel failures are expected to occur as a result of a locked rotor incident.

Consequently, I

no further consideration of the potential radiological impact of proposed increases in reactor power and fuel exposure is warranted.

5.3 RCC ASSEMBLY EJECTION INCIDENT Section 14.2.6 of the UFSAR presents an evaluation of the RCC assembly ejection incident.

As for the locked rotor incident, conservative assumptions (greater than technical specification limits) were made relative to peaking factors in the core.

Results of these conservative analyses indicate that less than ten (10) percent of the affected fuel rods enter DNB.

(This is less than 2X of the total core).

The off-site radiological consequences of this event were not expressly evaluated since it is clear that the resulting doses would be only a small fraction

(~2K)of those calculated for the double-ended coolant pipe break (LOCA) event (see pages 14.2.6-12, Section 14.2.6 of the Unit 1 UFSAR).

0 15 XN-NF-82-90 An analysis of the RCC assembly ejection incident has been performed by Exxon Nuclear using procedures developed in the Generic (7) 4

. (s)

Rod Ejection Analysis and assuming operation of the reactor at a

power level of 3425 MW and with an F

value of 2.1.

This analysis demonstrates that the maximum pellet energy deposition is <170 cal/gram.

Consequently, no failures of ENC fuel are expected to occur as a result of an ejected RCC assembly.

Should failures occur in other fuels within the mixed core configuration as is indicated in the UFSAR, they would be

'imited to a small fraction of the total core and would, therefore, result in an off-site release of fission product activity which is only a small fraction of that which would occur under LOCA conditions (i.e.,

off-site doses would be far below the limits given in 10 CFR 100).

5.4 STEAM GENERATOR TUBE AND MAIN STEAM LINE RUPTURES Analyses are presented in Sections 14.2.4 and 14.2.5 which show that no fuel rods will experience transition boiling as a result of a steam generator tube rupture or a main steam line break.

Consequently, since no fuel failures are expected to result from these incidents, the off-site thyroid and whole body dose calculations are based on the assumption of operation with equilibrium fission product activity in the primary and secondary systems determined by assuming that 1X of the fuel cladding is defective.

16 XN-NF-82-90 Similar analyses for ENC 17x17 fuel, operating at the increased (6) reactor power level, support the contention that no fuel failures will result from generator tube ruptures or main steam line breaks.

Since the technical specification limit for coolant activity remains unchanged and no fuel failures occur as a result of the incidents, the estimated thyroid and whole body doses given in Section 14.2.7 of the Unit 1 UFSAR as a function of the primary-to-secondary leak rate remain unchanged.

In addition, since the inventory of iodine available in the fuel rod gaps of the ENC 17x17 design at proposed power and exposure levels is less than for the design base conditions, the effect of iodine spiking behavior will be less than for the cases considered in the UFSAR.

5.5 LOCA EVENT Section 14.3.5 of the Unit 2 UFSAR notes that the environmental consequences of a loss-of-coolant accident for Unit 1 is conservatively valid for Unit 2.

That analysis, presented in Section 14.3.5 of the Unit 1 UFSAR, includes evaluations of the off-site radiological conse-quences of 1) the design base accident, and 2) the maximum hypothetical accident.

The Design Base Accident (DBA) assumes that the entire inventory of volatile fission products contained in the pellet-cladding gap is released.

That inventory, as given in Table 14.3.5-2 and Table 14A.2-1 of the Unit 1 UFSAR is based on an end-of-cycle core aver aged exposure of 24,560 MWD/MTU with a rated reactor power of 3391 MW.

Consequently,

17 XN-NF-82-90 the off-site dose for proposed changes in power and end-of-cycle average fuel exposure are proportional to the change in fission product inventory within the pellet-clad gap.

For the maximum hypothetical

accident, the fission product source is assumed to be a fixed fraction of the total core inventory of volatile fission products.

As for the DBA, the inventory is based on the end-of-cycle core average values.

Consequently, for the hypothetical accident assumptions, the off-site dose resulting from operations at the proposed higher power and end-of-cycle fuel exposure is proportional to the change in the total core inventory of volatile fission products.

Off-site doses which could result from a loss-of-coolant accident involving fuel operating at the proposed reactor power and exposure levels are given in Table 9 and compared with values determined in the UFSAR.

As can be seen, the estimated off-site radiological consequences of a LOCA are within the limits specified in 10 CFR 100.

18 XN-NF-82-90

6.0 REFERENCES

=1)

M. J. Bell, "ORIGEN - The ORNL Isotope Generation and Depletion Code",

Oak Ridge National Laboratory, ORNL-4628 (May 1973).

2)

XN-NF-81-58(P),

"RODEX 2 Fuel Rod Thermal Mechanical

Response

Evaluation Model", Exxon Nuclear Company, Inc., August 1981.

3)

J.

R. Houston, et. al.,

"DACRIN - A Computer Program for Calcu-lating Organ Dose from Acute or Chronic Radionuclide Inhalation",

Battelle Pacific Northwest Laboratories, BNWL-B-389/UC-41 (December

1974, Reissued April 1976).

4)

R. I. Scherpelz and A. E. Desrosiers, "Doses Received While Crossing a Plume of Radioactive Material Released During an Accident at a Nuclear Power Plant", Health Physics, Vol. 43, No.

2 (August 1982),

pp. 187-203.

5)

R. 0. Gumprecht, "Mathematical Basis of Computer Code RIBD",

Douglas United Nuclear, Inc.,

DUN-4136 (June 1968).

6)

XN-NF-82-32(P), "Plant Transient Analysis for the Donald C.

Cook Unit 2 Reactor at 3,425 MWt", Exxon Nuclear

Company, Inc.', April 1982.

7)

XN-NF-82-37, "D. C.

Cook Unit 2, Cycle 4 Safety Analysis Report",

Exxon Nuclear Company, Inc., April 1982.

8)

XN-NF-78-44, "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors",

Exxon Nuclear

Company, Inc., January 1979.

.I

19 XN-NF-82-90 Table 1

Assumptions for Calculating Fission Product Inventories Core Activity ase ase x 5 N

x Worst Case Fuel Assembl ase ase 5x 5 N

7x 7

Rated Power (MWt)

Actual Power (102K of Rated) 3391*

3459 3425**

3494 3391*

3459 3425**

3494 Fuel Exposure (MWO/MTU) 24,560 30,000 42,400 43,000 Radial Peaking Factor (F~H)

Local Peaking Factor Total Peaking Factor N/A 1.0 1.99*

N/A 1.0

]0**

1.55*

1.0 1.99*

1.55*

1.0 10**

Limiting value from technical specifications.

• • Proposed limiting value.

Table 2

CORE AND GAP FISSION PRODUCT ACTIVITIES Isotop'e urres n

Core ractron Of (UI'Ies

~ln Ga Design Base 15xl5 Fuel Assembly EOC Average Fuel Exposure of 24,560 HWD/HTU Rated Reactor Power 3391 MWt

~N'eS I nba r acheron 07 Activit In Ga Core ENC 17x17 Fuel Assembly Design EOC Average Fuel Exposure of 30,000 HWD/HIU Rated Reactor Power 3425.HWt T

1-129 I-131 1-132 1-133 l-134 I-135 Total I

Kr-85 Kr-85m Kr-87 Kr-88 Kr-89 Total Kr Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Xe-137 Xe-138 Total Xe 2.35 9.57 x 108 1'3 108 1.86 x 108 2.12 x 108 1.67 x 10 7.98 x 10 8.35 x 107 2.36 x 107 4.67 x 107 6.83 x 107 8.61 x 10 2.26 x 10 7.00 x 108 5

1.86 x 106 5.02 x 107 3.86 x 107 4.96 x 108 1.81 x 108, 2.03 x 10 6.64 x 10

.5113 x 10 I

.1336 x 10 2

.4114 x 10 2.

.8279 x 10 3

.5214 x 10 2

.4643 x 10

.2123 x 10 2

.2129 x 10 3

.1639 x 10 2

.1722 x 10 3

.2382 x 10

.1312 x 10 2

.6255 x 10 1

.1011 x 10 2

.1498 x 10 3

~ 5259 x 10 3

'322 x 10 3

.3797 x 10 0.12 1.28 x 105 5.64 x 106 1.54 x 105 1.11 x 105 7.75 x 10 4.27 x 10 1.77 x 104 5 02 x 103 7.65 x 105 1.18 x 104 2.05 x 10 2.14 x 10 9.18 x 106 3

1.16 x 104 5.08 x 104 5:78 x 104 2.61 x 104 4.20 x 104 7.71 x 10 1.42 x 10 2.54 1.02 x 108 1.45.x 108 1.96.x 108 2.22 x 108 1.75 x 10 8.40 x 10 8.79 x 107 5

2.39 x 107 4.71 x 107 6.90 x 107 8.66 x 10 2.28 x 10 7.53 x 108 5

1.95 x 106 5.29 x 107 3.44 x 107 5.22 x 108 1.92 x 108 2.16 x 10 6.96 x 10

~ 2317 x 10 2

.5057 x 10 2

.1569 x 10 2

.3142 x 10 3

.1982 x 10 2

.1767 x 10

.9053 x 10 3

.8120 x 10 4

.6238 x 10 3

.6568 x 10 4

.9069 x 10

.4986 x 10 2

,.2387 x 10 2

.3847 x'0 3

.5711 x 10 3

.2003 x 10 4

.8839 x 10 3

.1446 x 10 0.06 5.16 x 105 2.28 x 105 6;16 x 104 4 40 x 105 3.09 x 10 1.71 x 10 7.96 x 104 3

1.94 x 103 2.94 x 104 4 53 x 103 7.85 x 10 8.35 x 10 3.75 x 105 3

4.65 x 104 2.04 x 104 1.96 x 104 1.05 x 104 1.70 x 104 3.12 x 10 5.67 x 10 5 OC I

rr I00 I

LD C)

1 >l>le 2 Coot iwiw.4 Cs-134 Cs-134m Cs-135 Cs-136 Cs-137 Cs-138 Cs-139 Cs-140 Cs-141 Cs-142 Cs-143 1.35 x 106 5.04 x 101 1.66 x 106 4.49 x 106 7.00 x 108 1.71 x 108 1.73.x 108 1.59 x 108 1.15 x 107 9.00 x 107 4.69 x 10

.2935 x 10 2

.2476 x 10 1

.2935 x 10 1

.2574 x 10 1

.2935 x 10 2

.1067 x 10 3

.5789 x 10 3

.1974 x 10'3

.1190 x 10 4

.3688 x 10 4

,.3435 x 10 3.96 1.25 0.49 1.16 2.05 1.82 1.00 3.14 1.37 3.32 1.61 x 104 5

x 10 x 105 5

x 105 x 105 x 104 x 104 x 103 x 103 x 10 1.71 x 106 7

6.43 x 101 1.58 x 106 5.24 x 106 7.63 x 108 1.81 x 108 1.81 x 108 1.67 x 108 1.20 x 107 9.43 x 107 4.89 x 10

.1270 x 10 3

.9437 x 10 1

.1270 x 10 2

.9809 x 10 1

.1270 x 10 3

.4060 x 10 3

.2203 x 10 4

.7510 x 10 4

.4527 x 10 4

.1403 x 10 4

.1306 x 10 2.17 x 103 6.07 x 10

0. 2,0 5.14 x 104 9.69 x 104 7.35 x 104 3.99

x. 104 1.25 x 103 5.43 x 103 1.32 x 102 6.39 x'0 Total Cs Te-125m Te-127 Te-127m Te-129 Te-129m Te-131 Te-131m Te-132 Te-133 Te-133m Te-134 Te-135 Total Te 7.85 x 10 1.84 x 106 7.23 x 106 1.48 x 107 3.60 x 106 2.16 x 107 8.34 x 107 1.43 x 108 1.32 x 107 5.18 x 108 1.43 x 108 1.87 x 107 3.92 x 10 6.97 x 10

.1923

.1727 x 10

.2451

.6035 x [0

.2341

.3626 x 10 1

.3082 x 10 1

.4955 x 10 2

.2568 x 10 2

.5135 x 10 2

.4707 x 10 2

.1029 x 10 1.06 3.54 1.25 3.63 2.17 5.06 3.02 4.41 6.54 1.33 7.34 8.80 4.03 x 106 x 205 x 10 x 105 5

x 105 x 105 x 105 x 106 x 105 x 105 x 105 x 104=

x 10 1.03 x 10 7 8.29 x 10 2.07 x 106 7.96 x 106 1.63 x 107

. 3.88-x 106 2.36 x 107 8.91 x 107 1.52 x 108 1;41 x 107 5.47 x 108 1.51 x 108 1 96 x 107 4.11 x 10 7.39 x 10

.7935 x 10-2

.6539 x 10

.1078

.2296 x 10

.1013

.1382 x 10-1

~ 1172 x lo 1

~ 1890 x 10 3

.9774 x 10 2

.1953 x 10 2

.1790 x 10 3

.3910 x 10 5.05 x 10 1.64 x 104 4

5.21 x 105 1.76 x'104 8.91 x 105 2.39 x 105 1.23 x 105 1.78 x 106 2.66 x 104 5 35 x 105 2.95 x 105 3.51 x 104 1.61 x 10 re 4.25 x 10 OC I

ll I

CO fO I

CD

I >bi>> 3 rUri ASSrHDI T AND GAP rlSSION PRODUCT ACTIVITlrs lnn IIOURS ArlER SIIIITDOWN Worst, Case 15xl5 fuel Assembly EOL Exposure of 42,400'WD/HTU Rated Reactor Power 3391 HWt

'Worst Case ENC 17x17 fuel Assembly EOL Exposure of 43,000 HWD/HIU Rated Reactor Power 3425 HWt Nuclide 1-131 I-132 I-133 l-135 Total I

Kr-85 Total Kr Xe-131m Xe-133 Xe-133m Xe-135 Total.Xe Cs-134 Cs-136 Cs-137 Total Cs Te-125m Te-127 Te-127m Te-129 Te-129m Te-131 Te-131m Te-132 Total Te er ue ssem y

100 ifours After Shutdown

$.89 x 105 4.78 x 104 5.75 x 101 4.26 x 10 1.13 x 10 6.58 x 10 6.58 x 10 6.49 x 106 3

'1.02 x 104 1.58 x 103 1.91 x 10 1.04 x 10 2.01 x 104 4.34 x 104 6.49 x 10 3.09 x 10 2.10 x 104 4.48 x 104 1.45 x 104 3.25 x 104 5.06 x 103 2.26 x 104 1.24 x 105 4.65 x 10 6.24 x 10 rac

>on

.2950

. 2826

.2925

.0657

.2584

.2949

.2882

.2932

.1747

.2891

.3209

.2891

.3204

.2982

.2987

.2881

.3109

.2817

.3576

.3862 Vries

~la Ga 1.74 x 105 1.35 x 104 1.68 x 10 2.80 3.26 x 10 1.70 x 10 1.70 x 103 1.91 x 105 2.94 x 103 4.63 x 102 3.34 x 10

'.01 x 10 5.81 x 104 1.39 x 104 1.88 x 10 9.08 x 10 6.73 x 104 1.34 x 103 4.33 x 103 9.36 x 104 1.57 x 102 6.37 x 103 4.43 x 105 1.80 x 10 2.'29 x 10 er ue ssem y

100 iiours After Shutdown 6.06 x 105 5

4.89 x 104 5.90 x 101 4.36 x 10 1.15 x 10 6.02 x 103 6.02 x 10 6.26 x 106 1.07 x 104 1.64 x 103 1.95' 10 1.10 x 10 1.82 x 104 5

4.12 x 104 5.73 x 10 2.81 x 10 1.83 x 104 4.56 x 104 1.47 x 104.

3.39 x 104 5.29 x 103 2.31 x 104 1.27 x 105 4.77 x 10 6.41 x 10 rac ion Activit In Ga

.2959

. 261.9

.2936

.0601

.2434

.2956

.2806

'.2932

.1601

.2699

.3061

.2699

.3948

.2983

.3689

.2783

.3683

.2601

.3106

.3206 ul les

~ln Ga 1.79 x 105 5

1.28 x 104 1.73 x 10 2.62 3.24 x 10 1.47 x 103 1.47 x 10 1.85 x 105 3.00 x 103 4.81 x 102 3.12 x 10 3.07 x 10 4.91 x 104 1.26 x 104 1.55 x 10 7.72 x iO4 7.22 x

GO4 1.36 x 103 5.42 x 103 9.43 x 104 1.95 x 102 6.01 x 103 3.94 x 105 1.53 x 10 2.06 x 10 Using the AIIS 5.4 fission gas release model results in a calculated internal rod pressure which exceeds. system pressure for the 15x15 fuel when irradiated at average power to 42,400 HWD/HIU.

Gap release fractions, therefore, were calculated at a more realistic case of 33,000 HWD/HTU exposure but applied to the fission product inventories which are consistent with those used in the UFSAR.

Table 4

COMPARISON OF COMPUTED FISSION PRODUCT ACTIVITIES EOC CORE AVERAGED EXPOSURE Core Activity Gap Activity Iodine Krypton Xenon Cesium Tellurium Totals UFSAR 7.86 x 108 2.28 x 108 2.90 x 10 1.30 x 10 R

EN Base Case 7.89 x 108 8

2.26 x 108(

)

6.64 x 108 7.85 x 108 6.98 x 10 3.17 x 10 N

Proposed 8.40 x 108 8

2.28 x 108(

)

6.96 x 108 8.29 x 108 7.39 x 10 3.33 x 109 UFSAR 4.79 x 105 6

7.93 x 106(

)

3.72 x 10 9.30 x 10 6 R

N Base Case 4.27 x 105 6

2.14 x 106(

)

1.42 x 106 1.06 x 107 1.03 x 10 1.73 x 10 7 ORI EN

~Pro osed 1.71 x 104 6

8.35 x 105(

)

5.67 x 105 5.05 x 106 4.25 x 10 7.12 x 10 6 WORST CASE ASSEMBLY AT TIME OF REACTOR SHUTDOWN Assembl Activit Ga Activit Iodine Krypton Xenon Cesium Tellurium Totals 5.60 x 10

(

)

1.48 x 106 1.73 x 10 8.81 x 10 6.55 x 106 6

1.45 x 106(

)

4.89 x 106 6.47 x 106 6.88 x 10 2.62 x 10 7

6.66 x 106 6

1.54 x 106(

)

5.43 x 106 6.51 x 106 6.63 x 10 2.68 x 10 7 2.87 x 103 4

8.27 x 104(

2:22 x 10 5.92 x 104 1.50 x 105 6

1.51 x 105(

)

6.03 x 105 4.51 x 106 1.87 x 10 4.58 x 106 1.49 x 105 6

1.43 x 10 6.16 x 10

(

)

4.10 x 106 5

1.64 x 10 4.30 x 106 (a)

Includes short lived isotopes not included in the UFSAR total.

Considering only those isotopes in the base case as were included in the UFSAR data, the total number of curies calculated by ORIGEN is 2.79 x 108 Ci which compares well with the UFSAR value of 2.90 x 108 Ci.

(b)

Evaluated with an assumed radial peaking factor of 1.37 rather than the limiting value of 1.55 as was assumed in the ORIGEN calculations.

(c)

Includes short lived isotopes which were not included in the UFSAR data since they decay prior to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.

ORIGEN calculated data equivalent to the xenon activity given in the UFSAR results in an activity of 1.75 x 106 Ci which is in excellent agreement with the UFSAR value of 1.73 x 106 Ci.

Table 5

Summary of Activities Considered in Accident Evaluations Core Activity (Curies)

Worst Assembly Activity

-100 Hours After Shutdown (Curies Material Iodine Krypton Xenon Cesium Tellurium Total ore ct>vasty a

ctlvlty 7.98x10 2.26x10 6.64x10 7.85x10 6.98xlO 3.17xlO 8.40xlO8 2.28xlO8 6.96x10 8 8.29x10 8 7.39x10 8 3.33x10 9 4.27x10 6 2.14xl0 5

l. 42x10 6 1.06x10 6 1.03xlO 7

1.73xlO 7 1.71x10 6 8.35x10 4 5.67x10 5 5.05x10 5 4.25xlO 6 7.12x10 6 ase ase

~ro osed Hase ase

~ro osed ase ase 1.13xlO 6 6.58x10 3 1.04xlO 6 3.09x10 5 6.24x10 5 3.llx10

~Pro osed

l. 15x10 6 6.02x10 3 1.10x10 6 2.81x10 5 6.41x10 5 3.18x10 6 ore ctivity a

ct1 V1 ty Hase Case

~Pro osed 3.26x10 3.24xlO 5

5 1.70x10 1.47x10 3

3 3.0lx10 3.07x10 5

9.08x10

'.72x10 2.29xl0 2.06x10 5

5 9.49xl0 9.16xl0 5

5

>C I

I00 C7

25 XN-NF-82-90 Table 6

Constants Relating Off-Site Doses From Accidents Involving Design Base and Proposed Operations Radiation Source Worst Case Assembly (Gap Activity)

Worst Case Assembly (Total Activity)

Core Average Fuel (Total Activity)

Core Average Fuel (Gap Activity)

A*

I 0.993 1.026 1;053 0;400 CCF

• I 1.036 1.003 1.010 1.005 0.995 0.982 0.998 1.000 0.939 1.022 1.050 0.412 Constants are as defined in Section 3.0.

26 XN-NF-82-90 Table 7

Fuel Handling Accident in the Auxiliary Building Two Hour Site Boundary Doses Desi n Base 0 eration*

Pro osed 0 er ation xpecte Case-Rem Whole Body 0.22 Thyroid 0.84 onservative Case-Rem 0.51 2.15 xpecte Case-Rem 0.21 0.86 onservative Case-Rem 0.48 2.21 10 CFR 100 Limit-Rem 25 300 See Table 14.2.1-4 of the Updated Final Safety Analysis Report for DC Cook Units 1 and 2.

27 XN-NF-82-90 Table 8

Fuel Handling Accident Inside Containment Two Hour Site Boundary Doses Desi n Base 0 eration*

Pro osed 0 eration Whole Body Thyroid onservatlve Case-Rem 1.3 82.3 ea lstlc Case-Rem 3.73x10 5.61x10 onservati ve Case-Rem 1'. 3 84.7 Rea istic Case-Rem 3.74xl0 5.77x10 10 CFR 100 Limit-Rem 25 300 See page 14.2.1-13 of Section 14.2.1.2 of the Updated Final Safety Analysis Report for OC Cook Units 1

and 2.

Table 9 Off-Site Ooses Resulting From a Loss of Coolant Accident

~ite onndarR days x = 3218 meters x

= 610 meters Oesi n Base 15x15 Fuel*

~ox P~ou ati~on one

~ice oondar

-2 hours x = 610 meters ow o u ation one days x = 3218 meters ENC 17xl7 Hi h Ex osure Fuel Thyroid Whole Body

~ense Rema

~dose Rems)

Thyroid Whole Body Dose (Rema),

~dose (Rmns Thyroid Whole Body

~0ose (Rema

~Dose Rema Thyroid Whole Body

~dose Rema

~ense Rema 10 CFR 100 Guideline Limit 300 25 300 25 300 25 300 25 Design Basis Accident (lero-Ice Condenser I

Removal Efficiency)

Maximum Hypothetical Accident (2ero Ice Condenser I2 Removal Efficiency)

Oesign Basis Accident 20% Ice Condenser I

Removal Efficiency 4(C Ice Condenser I

Removal Efficiency Maximum Hypothetical

'ccident 201 Ice Condenser I

Ren)oval Efficiency 4iyf, Ice Condenser l2 Removal Efficiency 2.45 154 2.37 2.35 148 146 0.031 7.64 0.031 7.6 3.88 221.8 3.85 3.81 220.4 220 0.052 9.78 0.05-9.7 0.98 164 0.95 0.94 158 155 0.013 8.01 0.0013 8.0 1.56 236 1.55 1.53 235 234 0.021 10.25 0.02 10.2 See Table 14.3.5-7 of the Updated Final Safety Report for OC Cook Units 1

and 2.

XN-NF-82-90 Issue Date:

11/17/82 DISTRIBUTION USNRC/GF OWSLEY (35)

AMERICAN ELECTRIC POWER/HG SHAW (4)

MJ ADES CO BROWN LE HANSEN (2)

TL KRYSINSKI GA SOFER RB STOUT GN WARD HE WILLIAMSON