ML20138M495

Suppl to TS Change Request 192 to Licenses DPR-24 & DPR-27, Modifying TS Section 15.3.3, Eccs,Acs,Air Recirculation Fan Coolers & Containment Spray, to Incorporate Outage Times Similar to NUREG-1431,Rev 1

ML20138M495
Person / Time
Site:	Point Beach
Issue date:	02/13/1997
From:	Dante Johnson WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20138M498	List: ML20138M495 ML20138M509
References
RTR-NUREG-1431 PBL-97-0057, PBL-97-57, NUDOCS 9702250408
Download: ML20138M495 (13)

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. Wisconsin i Electnc

. PONER COMPANY 231 W Mechigart PO Box 2046. Mdwoukee.WI 53201-2046 (414)221-234s PBL 97-0057 10 CFR 50.4 10 CFR 50.90 February 13,1997 U.S. NUCLEAR REGULATORY COMMISSION Document Control Desk Mail Station PI-137 l Washington, DC 20555 l 1

Gentlemen:

DOCKETS 50-266 AND 50-301  !

SUPPLEMENT TO TECHNICAL SPECIFICATIONS I CHANGE REOUEST 192 -

POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 In a letter dated September 30,1996, Wisconsin Electric requested Technical Specifications change  !

request 192. This Technical Specifications change request proposes to modify Technical Specifications section 15.3.3, " Emergency Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan Coolers, and Containment Spray" to incorporate allowed outage times similar to those contained in NUREG-1431, Revision 1, "Westingho'use Owner's Group Improved Standard Technical Specifications,"

and modify the operability requirements for the service water system. The proposed changes to Technical Specifications section 15.3.7, " Auxiliary Electrical Systems," also ref'ect the modified service water i operability requirements. The proposed change to Technical Specifications section 15.5.2," Containment," I modifies the heat removal capacity of the reactor containment air cooler units. Previous supplements to this Technical Specifications change request were provided in a letters dated November 26,1996 and December 12,1996.

This letter also provides supplemental information for Technical Specifications change request 192. j Attachment I contains an analysis of environmental consequences of a Loss of Coohnt Accident (LOCA) and attachment 2 contains additional information and minor corrections to the supplement dated December 12,1996. Also provided with this submittal are three diskettes containing meteorological data. ,

Two diskettes contain meteorological data from Kewaunee Nuclear Plant for the years 1991-1996. One  !

di:ielte contains meteorological data from Point Beach Nuclear Plant for the year 1996.

1 In a letter dated January 16,1997, we stated that we are in the process ofidentifying possible alternatives I that can be used to reduce the control room thyroid dose to s30 rem without the use of KI and that we l would propose a plan for resolving the control room KI issue as a supplement to Technical Specifications change request 192. This information will be provided as a future supplement to this Technical ,

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, Document Control Desk i February 13,1997 Page 2 l

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We have determined that the additional information does not involve a significant hazards consideration,

authorize a significant change in the types or total amounts of any effluent release, or result in any significant increase in individual or cumulative occupational exposure. Therefore, we conclude that the proposed amendments meet the requirements of 10 CFR 51.22(c)(9) and that an environmental impact statement or negative declaration and environmentalimpact appraisal need not be prepared. The original 4

"No Significant Hazards" determinations for operation under the proposed Technical Specifications remain applicable.

Please contact us if you have any questions.

l Sincerely, Dougl s . Johnson Manager-Regulatory Services and Licensing 1

cc: NRC Resident Inspector l

NRC Regional Administrator PSCW Subscribed and sworn before me on l this _)y' ' day of Tdu;m _,1997.

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em taJ thb uy@ublic, State of Wisconsin My commission expires uht/ anno I l

f ATTACHMENT 1 TECHNICAL SPECIFICATIONS CHANGE REOUEST 192 '

ENVIRONMENTAL CONSEOUENCES OF A LOSS OF COOLANT ACCIDENT Introduction A large pipe rupture in the RCS is assumed to occur. As a result of the accident, it is  :

assumed that core damage occurs and iodine and noble gas activity is released to the containment atmosphere. A portion of this activity is released via containment leakage to the outside atmosphere (Reference 1). Also, once recirculation of the Emergency Core  !

Cooling System (ECCS) is established, activity in the sump solution may be released to j the environment by means ofleakage from ECCS equipment outside containment or the failure of an ECCS passive component (e.g. a pump seal or gasket) in the auxiliary .

building (Reference 2). It is not necessary to model the ECCS passive failure release path !

because Point Beach has a filtered ventilation system in the auxiliary building (Reference ;

2). This section describes the assumptions and analyses performed to determine the l amount of radioactivity released and the offsite and control room radiological consequences resulting from these releases.

Input Parameters and Assumptions The analysis of the large break LOCA radiological consequences uses the analytical l methods and assumptions outlined in the Standard Review Plans (References I and 2). j One hundred and two (102) percent of the current power level of 1518.5 MWt (1548.9 1 MWt) is assumed in the analysis. Both the offsite and control room doses are calculated I

based on the conservative assumptions listed in Reference 3.

Containment Leakage i

Following the large break LOCA, 50% of the core iodine activity and 100% of the core )

noble gas activity are assumed to be immediately released to containment when determining doses due to containment leakage. Fifty percent of the iodine released to containment is assumed to instantaneously plate out on containment surfaces. This leaves 25% of the core iodine activity and 100% of the core noble gas activity instantaneously available for leakage from the containment (Reference 3). This iodine is assumed to be 91% elemental,4% methyl and 5% particulate (Reference 3).

The particulates and the elemental iodine are removed from the containment atmosphere by the action of the containment sprays. The organic form ofiodine is not easily removed from the containment atmosphere and is assumed to be removed only by radioactive decay and leakage. The elemental iodine spray coefficient of 20 hr is determined based on the model suggested in Reference 4. Credit is taken for this spray removal until a decontamination factor of 200 in the containment inventory of elemental iodine is reached (Reference 4). The particulate iodine spray coefficient of 6.02 hr is also determined based on the model suggested in Reference 4. Credit is taken for particulate iodine j i

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, removal until containment spray has been terminated. At this time a decontamination factor of 19.8 in the particulate iodine inventory in containment has been reached.

The Technical Specification containment design allowable leak rate of 0.4% by weight of containment air is used for the initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Thereafter the containment leak rate is l assumed to be one-half the design value, or 0.2%/ day (Reference 3)

ECCS Equipment Leakage f

I When ECCS recirculation is established following the LOCA, leakage is assumed to occur from ECCS equipment outside the containment. It is also assumed that 50% of the total core iodine is in the sump water being recirculated (Reference 2). Hence, the ECCS equipment leakage results in the release of a significant amount ofiodine activity to the

outside environment. For this activity release path, no credit is taken for plateout of 1

elemental iodine on containment surfaces or for iodine removal by the atmosphere filtration system in the auxiliary building. The iodine release from this path is j conservatively assumed to be 100% elemental.

The ECCS equipment leaks at a rate of 400 cc/ min. This leak rate is conservatively assumed to continue at this constant rate form the time ECCS recirculation is established

until 30 days following accident initiation. Ten percent of the iodine in the leakage is assumed to become airborne due to flashing (Reference 2).

There is no noble gas activity in the ECCS recirculation water.

Control Room Parameters The doses to personnel in the control room are determined for both of the activity release paths discussed above. The control room ventilation system volume is 65,243 ff, the filtered makeup flow is 4950 cfm with no filtered recirculation, and the unfiltered l inleakage flow is 10 cfm. The control room filter removal efficiencies are 90% elemental, I 90% organic and 99% particulate.

Attached Tables The thyroid dose conversion factors, breathing rates and atmospheric dispersion factors used in the dose calculations are given in Table 1. The core activities used in the dose calculations are given in Table 2. The control room assumptions and parameters are given in Table 3. The major assumptions and parameters used to determine the doses due to containment leakage are given in Table 4 and those for ECCS equipment leakage are given in Table 5.

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.. j Description of Analyses Performed l l

The offsite thyroid and y-body doses, as well as the control room thyroid, gamma body and beta skin doses, are determined using the analytical methods and assumptions outlined in the Standard Review Plans (References 1 and 2). Both the containment leakage and the ECCS continuous leakage activity release pathways are included in this analysis.

Acceptance Criteria The offsite doses must meet the guidelines of 10CFR100, or 300 rem thyroid and 25 rem whole body for the initial 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the accident at the Site Boundary (SB) and for the duration of the accident at the Low Population Zone (LPZ). The dose criteria for control room personnel following the accident are 5 rem y-body, 30 rem thyroid, and 30 rem p-skin (or 75 rem -skin with protective clothing) per Reference 5.

Results The offsite and control room thyroid, y-body, and beta skin doses due to the large break LOCA are given in Table 6.

Conclusions  !

I The offsite thyroid and whole body doses are within the current NRC acceptance criteria for a loss of coolant accident. The control Room whole body and beta skin doses are within the current NRC acceptance criteria for the control room. The control room thyroid dose exceeds the 30 rem limit; however, by the use of potassium iodide pills, the control room thyroid dose would be reduced by approximately 28 rem which is within the 30 rem limit. The use of potassium iodide pills to reduce the thyroid dose to the control room operator by a factor of 10 is based on NRC Safety Evaluation Report dated August 10,1982 for resolution NUREG-0737 item III.D.3.4 - Control Room Habitability.

References

1. NUREG-0800, Standard Review Plan 15.6.5, Appendix A, " Radiological Consequences of a Design Basis Loss-of-Coolant Accident Including Containment Leakage Contribution," Rev.1, July 1981.

2. NUREG-0800, Standard Review Plan 15.6.5, Appendix B, " Radiological  ;

Consequences of a Design Basis Loss-of-Coolant Accident: Leakage from Engineered

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Safety Features Components Outside Containment," Rev.1, July 1981. l

3. Regulatory Guide 1.4, Rev. 2, July 1974, " Assumptions Used for Evaluating the I Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors." l 3

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4. NUREG-0800, Standard Review Plan 6.5.2, " Containment Spray as a Fission Product  ;

Cleanup System," Rev. 2, December 1988.  !

5. NUREG-0800, Standard Review Plan 6.4, " Control Room Habitability System", Rev.

I 2, July 1981. i i

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, TABLE 1 DOSE CONVERSION FACTORS, BREATHING RATES AND ATMOSPHERIC 4

DISPER510N FACTDRS Thyroid Dose Conversion Factors U3 Isotope (rem / curie) i

l131 1.07 E6 ['

l l-132 6.29 E3

l-133 1.81 E5 '

l-134 1.07 E3 i 1135 3.14 E4 4

Time Period Breathing Rate W i (hr) (m3/sec) i 0-8 3.47 E-4 S

8-24 1.75 E-4 24-720 2.32 E-4 Atmospheric Dispersion Factors W

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l Site Boundary i 0-2 hr 5.0 E-4 Low Population Zone j

0-8 hr 3.0 E 5 l J

8 24 hr 1.6 E-5 t i 24-96 hr 4.2 E-6 96-720 hr 8.6 E 7 i Release from Release from Auxiliary l Control Room Containment Building )

0-8 hr 2.1 E-3 1.0 E-3

8-24 hr 1.3 E-3 7.0 E-4 i 24-96 hr 8.3 E 4 3.9 E-4 96-720 hr 3.3 E-4 1.3 E-4 D' ICRP Publication 30 m Regulatory Guide 1.4 W

Wisconsin Electric letter NPL 97-0041 5

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. TABLE 2 CORE ACTIVITIES III l

l Total Core Activity at i Nuclide Shutdown (Cl) '

l-131 4.13 E7 i

1-132 5.92 E7 I I

j 1-133 8.45 E7 l-134 9.30 E7

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I-135 7.89 E7 Kr-85 5.07 E5 Kr-85m 1.13 E7 1

Kr-87 2.16 E7 Kr 88 3.0 E7 Xe 131m 4.41 E5 Xe-133 8.36 E7 Xe-133m 2.63 E6 i i

Xe-135 2.30 E7 Xe-135m 1.60 E7 4

Xe-138 7.04 E7 l

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These core activities are based on a core power level of 1548.9 MWt. The activities

, were updated as a part of the Point Beach fuel upgrade program.

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a TABLE 3 CONTROL ROOM PARAMETERS Volume 65.243 ft3 Unfiltered inleakage 10.0 cfm Total Flow Rate 19800cfm Filtered Makeup 4950cfm Filtered Recirculation o efm Fitter Efficiency

• Elemental 90 %

Organic 90 %

Particulate 99 %

Occupancy Factors 0 - 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1.0 12 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.0 24 - 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 1.0 36 - 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 0.0 48 - 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> 1.0 i 60 - 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 0.0 72 - 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> 1.0 84 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 0.5 4 - 30 days 0.4 i

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4 TABLE 4 ASSUMPTIONS USED FOR LARGE BREAK LOCA DOSE ANALYSIS CONTAINMENT LEAKAGE i

Power (102%) 1549 MWt lodine Chemical Species

Elemental 91 %

Methyl 5%

Particulate 4% .

1 lodine Removal in Containment instantaneous lodine Plateout 50 %

Containment Spray Start delay time 90 seconds .

Injection spray flowrate 1190 cfm Duration of injection spray 1.37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> i Spray removal coefficient I

Elemental 20 hr-1 Particulate 6.02 hr*1 2

i Containment Net Free Volume 1.065E6 ft 3 Sprayed Volume 475,000 ft 3 Unsprayed Volume 590,000 fr a i

Containment Mixing Containment Fan Coolers I Start Delay Time 90 seconds I I

Number of Units 2 l

Flow Rate per Unit 38,500 cfm l

Containment Leak Rate j 0 24 hr O.4% / day I

> 24 hr 0.2%/ day I 8 a

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e TABLE 5 ASSUMPTIONS USED FOR LARGE BREAK LOCA DOSE ANALYSIS ECCS EQUIPMENT LEAKAGE i

Power (102%) 1549 MWt lodine Activity in Recirculation Water 50% Core lodine Activity i

lodine Chemical Species 100% Elemental j

Leakage Rate 400 cc/ min I Leakage Duration From start of recirculation through 30 day duration Time of Recirculation Initiation 20 minutes

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Sump Water Volume 223635 gallons lodine Flashing Fraction to Environment 10 %

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TABLE 6 LARGE BREAK OFFSITE AND CONTROL ROOM DOSES

1. Thyroid Doses Dose (Rem) l i

SB (0-2 Hrl LPZ (0-30 Day) CR (0-30 Davi 4

Containment Leakage 132.7 23.7 196.4 4

ECCS Equip. Leakage 25.2 16.3 84.7 Total 157.9 40.0 281.1

2. y-body Doses Dose (Rem)  ;

SB (0 2 Hr) LP2 (0-30 Davl CR (0 30Ular  !

Containment Leakage (10) 0.72 0.07 0.02 Containment Leakage (NG) 2.52 0.38 0.85 ECCS Equip Leakage 0.11 0.03 0.003 Total 3.35 0.48 0.87

3. CR E-Skin Dose Dose (Rem) 30 Day Containment Leakage (10) 0.14 Containment Leakage (NG) 25.52 ECCS Equip. Leakage 0.03 Total 25.7 10 = lodines NG = Noble Gases 10 i

I A'ITACHMENT 2 TECHNICAL SPECIFICATIONS CHANGE REOUEST 19J ADDITIONAL INFORMATION AND CORRECTIONS TO PREVIOUS SUPPLEMENT CORRECTIONS la the cover letter and the attachment to the December 12,19% submittal, the reference to a previous supplement dated November 13,1996 is incorrect. The actual date of the supplement previous to the December 12,1996 supplement shoulJ be November 26,1996.

Change 2. in the December 12,19% subndttal should have included a change to the basis of this Technical Specification to be consistent with the modification to the proposed change. In particular, the requirement for five operable service water pumps prior to entry into TS 15.3.3.D.2.d. The revised page 15.3.3-10 is attached.

ADDITIONAL INFORMATION The assumed the delay time for containment spray and containment fan coolers in the FSAR section 14.3.4 " Containment Integrity Evaluation" is 60 seconds. A condition report (CR 96-1486) documents the discovery that the delay time for containment spray could be as long as 67 seconds and the delay time for containment fan coolers could be as long as 62 seconds. The recent containment integrity analyses performed (see letter dated November 26,19%) assumed these longer delay times for containment spray and containment fan coolers.

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