ML050410068

From kanterella
Jump to navigation Jump to search

Purpose and Reason of Control Room Air Treatment System Being Required During a Control Pod Drop Accident to Maintain Control Room Doses within 10 CFR 50 Appendix a, GDC 19 Acceptance Criteria
ML050410068
Person / Time
Site: Nine Mile Point  Constellation icon.png
Issue date: 05/01/1998
From:
Niagara Mohawk Power Corp
To:
Office of Nuclear Reactor Regulation
Chawla M, 415-8371, NRR/DLPM
Shared Package
ML050470272 List:
References
-RFPFR, TAC MC5556, YT020050009
Download: ML050410068 (28)
v • d • e


Text

PURPOSE The purpose of this calculation is to determine if the control room air treatment system is required during a control rod drop accident (CRDA) to maintain control room doses within IOCFR50 Appendix A, GDC 19 acceptance criteria.

REASON FOR REVISION 1: The purpose of revision 1 is to incorporate independent reviewers comments, design inputs relating to control room normal ventilation intake flow rate, control room air volume, and to evaluate a ground level release to be consistent with SRP 15.4.9 (loss of off-site power).

METHODOLOGY The UFSAR CRDA analysis assumed a maximum of 1% of the noble gas activity and 0.5% of the halogen activity in a fuel rod are released to the coolant. The SRP assumes 10% of the noble gas and iodines are released to the coolant. The resultant halogen and noble gas radioactivity described in the UFSAR assumed to be released to the coolant as a result of a CRDA is compared to the calculated reactor coolant source term using SRP 15.4.9 methodology. For conservatism, the higher of the two coolant source terms is input to the Stone and Webster computer code DRAGON.

Two cases are analyzed. Case 1 assumes elevated release to be consistent with the UFSAR (REF 1.d) and case 2 assumes ground level release to be consistent with SRP 15.4.9 (loss of offsite power).

DATA/ASSUMPTIONS Nominal Reactor Power is 1850 MWt. A 2 % uncertainty is added. Therefore the Reactor Power used in this calculation is 1887. (REF 1a, 6) 850 fuel rods exceed 170 cal/gm, which is the enthalpy limit for eventual cladding perforation. Since Ul's peak enthalpy will be less than 280 cal/gm, melting does not occur.(REF 2,3, and 4)

Fuel assumptions.(REF 1.b, 1.c, and 7) 532 fuel assemblies.

62 rods/fuel assembly.

32,632 total rods. This number was taken from radiological calculation 1H-009, Control Rod Drop Accident. The total rods were based on the actual fuel type in the Unit 1 core in 1991. Using only 8x8 fuel the total number of rods would be 62 by 532 = 32,984. This would result in an approximate I % decrease in source term (850/32,984 versus 850/32,632). Therefore, the dose contribution using the 1991 total number of rods are conservative and also negligible.

Case 1: elevated release. (REF 1.d). Case 2: ground level release (REF 4). (see METHODOLOGY section).

Release fractions (REF 4)

Amount of fuel gap activity is 10% of total activity in the rod (REF 5)

Fuel rods presumed failed are assumed to have operated at 1.5 times that of the average power level in the core.

All of noble gas and iodine gap activity instantaneously and uniformly mix in reactor coolant in the pressure vessel at the time of the accident.(REF 5) 10% of the iodines and 100% of the noble gases are released from the reactor pressure vessel to the turbine and condenser.

All noble gases in the turbine and condenser are available for release.

90% of the iodines are assumed to be removed by plateout and partitioning in the turbine and the condenser leaving only 10 %

airborne and available for leakage.

The turbine and condensers leak to the atmosphere at 1% / day for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leak terminates.

Main Condenser volume: 5.OOE+04 ft3 (REF 7) - Not needed for this calculation since the DRAGON computer code uses fractions per day times the activity available for release to determine a release rate.

Control Room free volume: 1.31E+05 ft3. (REF 8)

The control room normal ventilation intake flow rate is 2250 cfm +/- 10%. 2250 cfm

+ 10% = 2475 cfm, use 2500 cfm. (REF 14) . An additional 30 cfm in-leakage is to account for 10 cfm inleakage to the control room assumed in accordance with SRP 6.4, section 111.3.d.(2).(ii) and 20 cfm is assumed to account for an unfiltered inleakage from a drain (REF 11).

Credit is not taken for Control Room Air Treatment System initiation to determine if filtration of the intake air is required to meet 10CFR50 Appendix A GDC 19 criteria.

U1 stack 0-2 hour X/Q is 3.12E-04 sec/m3 and the 2-720 hour X/Q is 1.22E-08 sec/m3.

U1 turbine building blowout panel 0-2 hour X/Q is 1.93E-03 seclm3. This value is conservatively used for the postulated 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> release duration. (REF 9)

Radioactivity assumed in the coolant as a result of a postulated CRDA as described in the UFSAR is: Halogens = 5.62E+04 Ci and Noble Gas =

6.64E+04 Ci.(REF 1.d)

Breathing rate of 3.47E-04 m3 / sec is conservatively assumed for the duration (0-720 hours) of the accident. (REF 12) 10CFR50 Appendix A, GDC 19 dose limit of 5 rem whole body or equivalent.

This equates to 30 rem thyroid and 30 rem beta (skin) (REF 13).

CALCULATION 1.0 COOLANT SOURCE TERM BASED ON SRP METHODOLOGY TABLE 2 ISOTOPE Ci/MWt MWt TOTAL CORE ACTIVITY IN ACTIVITY COOLANT Ci Cl 1-131 2.90E+04 1887 5.47E+07 2.14E+05 1-132 4.20E+04 1887 7.93E+07 3.1 OE+05 1-133 4.80E+04 1887 9.06E+07 3.54E+05 1-134 6.20E+04 1887 1.17E+08 4.57E+05 1-135 4.90E+04 1887 9.25E+07 3.61 E+05 TOTAL 1.70E+06 KR-83M 3.00E+03 1887 5.66E+06 2.21 E+04 KR-85M 6.50E+03 1887 1.23E+07 4.79E+04 KR-85 3.OOE+02 1887 5.66E+05 2.21 E+03 KR-87 1.20E+04 1887 2.26E+07 8.85E+04 KR-88 1.70E+04 1887 3.21 E+07 1.25E+05 KR-89 2.OOE+04 1887 3.77E+07 1.47E+05 XE-131 M 1.80E+02 1887 3.40E+05 1.33E+03 XE-133M 2.OOE+02 1887 3.77E+05 1.47E+03 XE-1 33 5.60E+04 1887 1.06E+08 4.13E+05 XE-135M 1.70E+04 1887 3.21 E+07 1.25E+05 XE-1 35 9.80E+03 1887 1.85E+d7 7.23E+04 XE-1 38 4.40E+04 1887 8.30E+07 3.24E+05 TOTAL 1.37E+06 O)GE BWR 6, Technical Description of a single cycle Boiling Water Reactor Nuclear System, January 1, 1974 (Table F.2.3-9) 1850 MWt

  • 1.02 (DATA/ASSUMPTION # 1)

Column Q

  • Column 0 0 Column Q * (850 failed rods / 32,632 total rods)*1.5 peaking factor
  • 0.1 (DATA/ASSUMPTIONS #2, 3, 5.a, 5.b, and 5.c)

As can be seen by comparing the total halogen and noble gas activities in Table 2 Column 4 to those described in the UFSAR (DATA I ASSUMPTIONS #11, the activity in the coolant using SRP methodology is greater than that the UFSAR coolant source term. Therefore the activity in the coolant using SRP methodology will be for this calculation.

2.0 ACTIVITY AVAILABLE IN CONDENSER

As stated in DATAIASUMPTIONS 5.d all the noble gases and 10% of the iodines are released to the turbine and condenser. Therefore, Table 1 Column i) is multiplied by 0.1 for iodines and 1.0 for noble gases to determine activity in the condenser.

TABLE 3 ISOTOPE ACTIVITY TO CONDENSER 1-131 2.14E+04 1-132 3.1OE+04 1-133 3.54E+04 1-134 4.57E+04 1-135 3.61 E+04 KR-83M 2.21 E+04 KR-85M 4.79E+04 KR-85 2.21 E+03 KR-87 8.85E+04 KR-88 1.25E+05 KR-89 1.47E+05 XE-131M 1.33E+03 XE-133M 1.47E+03 XE-133 4.13E+05 XE-135M 1.25E+05 XE-135 7.23E+04 XE-138 3.24E+05 The above condenser activity is input to DRAGON runs # 9014 and 8946 dated 5/16/98. The card inputs to these runs are included in Appendix A.

DRAGON inputs for case 1:

Main Condenser volume: 5.0OE+04 ft3 (DATA/ASSUMPTION # 6)

Main Condenser release rate: 0.01 fractions per day (DATA/ASSUMPTION #5.g)

Unit I Control Room (DATA/ASSUMPTIONS # 7 - 9) volume: 1.31 E+05 ft3 intake rate 0 to 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />s: 2530 cfm filter efficiencies 0 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> = 0 Breathing rate 0 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> is 3.47E-04 m3 /sec (DATA/ASSUMPTION #12) 0-2 hour X/Q: 3.12E-04 sec/m3 2 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> X/Q: 1.22E-08 sec/m3 (DATA/ASSUMPTION #10).

Fraction of iodine inventory available for release is 0.1 Fraction of noble gas inventory available for release is 1.0 DRAGON inputs for case 2 is same as for case I with the exception of 0-24 hour X/Q = 1.93E-3 sec/m3

RESULTS The 0 - 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Unit 1 Control Room Doses are as follows TABLE 4 CRDA TO Ul CONTROL ROOM - ELEVATED RELEASE UNIT I CR DOSES, GDC 19 LIMIT, REM REM THYROID 4.50E-01 30 GAMMA 7.1OE-04 5 BETA 9.21E-03 30 TABLE 5 CRDA TO UI CONTROL ROOM - GROUND RELEASE UNIT I CR DOSES, GDC 19 LIMIT, REM REM THYROID 2.76E+01 30 GAMMA 1.44E-02 5 BETA 3.37E-01 30 CONCLUSIONS The revision to control room volume and control room ventilation intake flow rate had negligible impact on the control room operator doses. The control room doses resulting from both a ground level and elevated release are within the dose guidelines of IOCFR50 Appendix A GDC 19 assuming no control room air treatment system initiation. Therefore, the control room air treatment system is not required to meet the guideline doses.

COMPUTER RUN LOG JOB DATE DESCRIPTION OF RUN 9014 5/16/98 DRAGON (REF 10) CRDA to Ul Control Room-no filters, elevated release 8946 5/16/98 DRAGON (REF 10) Same as above except ground level release.

Card images (2) are given in Appendix A REFERENCES Nine Mile Point I Final Safety Analysis Report Revision 14 Table XV-9 I.B.4.0 XV.C.3.2 XV.C.4.5.1 111.B.2.2 General Electric Standard Application for Reactor Fuel, Licensing Topical Report, NEDE-24011-P-A-13 Class III, August 1996.

Engineering Report for Application of GE1l1 to Nine Mile Point Nuclear Station Unit I Reload 12, GENE-770-31-1292, revision 2, April, 1993 NUREG 0800, Standard Review Plan 15.4.9, Radiological Consequences of Control Rod Drop Accident (BWR).

Regulatory Guide 1.77, Assumptions used for evaluating a control rod ejection accident for pressurized water reactors, Appendix B.1.b, I.c, and 2.c Technical Specifications section 1.14, Amendment 142, page 5 Calculation 1H-009, Control Rod Drop Accident, revision 00 S10-210-HV12, Control Room & Auxiliary Control Room, revision 00, pages 45 and attachment 1-3 and NMPC Drawings: C18810C, sheet 1; C18812C, sheet 1; C18804C, sheet 1.

Letter dated March 19, 1984, from T. E. Lempges (NMPC) to D. B. Vassallo (NRC)

DRAGON computer code, SWEC Number NU-1 15, Version 5, Level 0 NMPC Calculation SI0-CR277.A-U1.210.

SRP 15.6.5, Loss-of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, and Regulatory Guide 1.3, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors.

SRP 6.4, Control Room Habitability System Internal Correspondence, file code M98-014, from T. Mogren to T. Kulczycky dated 5/18/98, "Outside Air Flow Rate for Control Room Ventilation."

APPENDIX A (2 ATTACHED)

CARD IMAGES OF COMPUTER RUNS

From: <kulczyckytonimo.com>

To: WND2.WNP3(DSH)

Date: 5/5/98 7:23am

Subject:

NMP Unit 1 Control Rod Drop Accident

Darl, My original e-mail did not go through, apparently because of the size of the documents. Therefore I will send them to you one at a time.

Attached is one of the 4 calculations for the Unit 1 Control Room Air Treatment system you requested yesterday during our telecon.

Unit 1 Control Rod Drop Accident (See attached file: ulcrda.doc)

The documents are in WORD for Windows version 7.0 format - if unable to read please call. My number is 315-349-1949 or try my pager 1-800-732-4365, pager # 1072.

Ted Kulczycky

TABLE OF CONTENTS page #

Title Page 1 Table of Contents 2 Purpose 3 Background 3 Methodology 3 Data / Assumptions 4 Calculation 5 Results 8 Conclusion 8 Computer Output Log 8 References 9 APPENDIX A - Card Image of Computer Run 10

PURPOSE The purpose of this calculation is to determine if the control room air treatment system is required during a control rod drop accident (CRDA) to maintain control room doses within 10CFR50 Appendix A, GDC 19 acceptance criteria.

BACKGROUND In March, 1984 NMP1 submitted to the staff the results of their control room habitability study in response to NUREG 0737 TMI Task Action Item III.D.3.4, Control Habitability Requirements. The design basis accidents identified in the submittal were the main steam line break and the loss of coolant accidents. The staff required that the licensee use Standard Review Plan (SRP) 6.4, Control Room Habitability System, as one of the documents to determine if the control room habitability acceptance criteria was met. SRP 6.4, section 11.6, acceptance criteria states that "In accordance with GDC 19 (Ref. 3), these doses (5 rem or equivalent) to an individual in the control room should not be exceeded for any postulated design basis accident." Furthermore, SRP 15.4.9, Radiological Consequences of Control Rod Drop Accident (BWR), requires a specific evaluation of the CRDA for the first application involving a particular standardized' design to establish a reference point for comparison of future applications incorporating the design. The Safety Evaluation Report (SER) received from the staff in May, 1984 stated "... The staffs conclusion is that control room operator doses following design basis accidents would be within GDC-19 guidelines.' As stated in the UFSAR section lll.B.1.5, Shielding and Access Control, "The most limiting accidents are the main steam line (MSL) break accident and the loss-of-coolant accident (LOCA) without core spray..." However, supporting documentation could not be found that supported the assumption that the control rod drop and fuel handling (FHA) accident were also evaluated as part of the study. As stated in the PURPOSE section, this calculation is for CRDA only. The FHA has been evaluated in calculation H21C045.

Although it has not been determined if the CRDA was required to be evaluated as part of the Task Action Item. This calculation has been performed to determine if the control room air treatment system would be required to mitigate the radiological consequences in the control room to ensure an individual in the control room will not received greater that 5 rem whole body dose or equivalent for any postulated design basis accident.

METHODOLOGY The resultant halogen and noble gas radioactivity described in the UFSAR assumed to be released to the coolant as a result of a CRDA is compared to a calculated reactor coolant source term based on the number of fuel rods that are assumed to fail as a result of a CRDA. The higher of the two coolant source terms is input to the Stone and Webster computer code DRAGON for conservatism. The method used for release assumptions from the coolant to the environment are consistent with the Standard Review Plan 15.4.9 (REF 4) with the exception that an elevated release is assumed to be consistent with the

UFSAR (REF 1d).

DATA/ASSUMPTIONS Nominal Reactor Power is 1850 MWt. A 2 % uncertainty is added. Therefore the Reactor Power used in this calculation is 1887.

850 fuel rods reach 170 cal/gm, which is th enthalpy limit for eventual cladding perforation. Since Ul's peak enthalpy will be less than 280 cal/gm, melting does not occur.

Fuel assumptions.

532 fuel assemblies.

62 rods/fuel assembly.

32,632 total rods. This number was taken from radiological calculation 1H-009, Control Rod Drop Accident. The total rods were based on the actual fuel type in the Unit 1 core in 1991. Using only 8x8 fuel the total number of rods would be 62 by 532 = 32,984. This would result in an approximate 1 % decrease in source term (850/32,984 versus 850/32,632). Therefore, the dose contribution using the 1991 total number of rods are conservative and also negligible.

Elevated release (see METHODOLOGY section)

Release fractions Amount of fuel gap activity is 10% of total activity in the rod Fuel rods presumed failed are assumed to have operated at 1.5 times that of the average power level in the core.

10 % of the gap activity for noble gas and iodines instantaneously and uniformly mix in reactor coolant in the pressure vessel at the time of the accident. Although Regulatory Guide 1.25 was used to define the gap activity released to the coolant, only 10% of the krypton 85 activity was assumed to be released instead of the 30% stated in RG1.25. Since Kr-85 had a negligible impact on dose. The computer run was not repeated.

10% of the iodines and 100% of the noble gases are released from the pressure vessel to the turbine.

All noble gases in the turbine are available for release.

90% of the iodines are assumed to be removed by plateout and partitioning in the turbine and the condenser leaving only 10 %

airborne and available for leakage.

The turbine and condensers leak to the atmosphere at 1%/day for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leak terminates.

Main Condenser volume: 5.OOE+04 ft3 Control Room free volume: 1.36E+05 ft3. The total volume, 1.69E+05 ft3, is multiplied by 0.8 to account for equipment located in the control room.

envelope (main control room, auxiliary control room, instrument shop, and computer room).

Control Room normal ventilation intake flow rate: 3550 cfm + 30 cfm in-leakage.

This is calculated by taking the maximum flow rate of 16,300 cfm minus the minimum recirculation flow rate of 12,750 cfm (REF 1.e). The additional 30 cfm is to account for 10 cfm inleakage to the control room assumed in accordance with SRP 6.4, section 111.3.d.(2).(ii). An additional 20 cfm is

assumed to account for an unfiltered inleakage.

Credit is not taken for Control Room Air Treatment System initiation.

Ul stack 0-2 hour X/Q is 3.12E-04 sec/m3 and the 2-720 hour X/Q is 1.22E-08 sec/m3.

Radioactivity assumed in the coolant as a result of a postulated CRDA is Halogens = 5.62E+04 Ci and Noble Gas = 6.64E+04 Ci.

Breathing rate of 3.47E-04 m3 / sec for the duration (0-720 hours) of the accident.

10CFR50 Appendix A, GDC 19 dose limit of 5 rem whole body or equivalent.

This equates to 30 rem thyroid and 30 rem beta (skin).

CALCULATION UNIT I COOLANT ACTIVITY TABLE I ISOTOPE Ci/MWt MWt TOTAL ACTIVITY IN ACTIVITY 1 CORE COOLANT FROM RELEASED TO ACTIVITY RODS THE CONDENSER 0

Ci Ci CI 1-131 2.90E+04 1887 5.47E+07 7.09E+03 7.09E+02 1-132 4.20E+04 1887 7.93E+07 1.03E+04 1.03E+03 1-133 4.80E+04 1887 9.06E+07 1.17E+04 1.17E+03 1-134 6.20E+04 1887 1.17E+08 1.51 E+04 1.51E+03 1-135 4.90E+04 1887 9.25E+07 1.20E+04 1.20E+03 TOTAL 4.34E+08 5.62E+04 5.62E+03 KR-83M 3.OOE+03 1887 5.66E+06 1.07E+03 1.07E+03 KR-85M 6.50E+03 1887 1.23E+07 2.33E+03 2.33E+03 KR-85 3.OOE+02 1887 5.66E+05 1.07E+02 1.07E+02 KR-87 1.20E+04 1887 2.26E+07 4.30E+03 4.30E+03 KR-88 1.70E+04 1887 3.21 E+07 6.09E+03 6.09E+03 KR-89 2.00E+04 1887 3.77E+07 7.16E+03 7.16E+03 XE-1 31M 1.80E+02 1887 3.40E+05 6.45E+01 6.45E+01 XE-1 33M 2.O0E+02 1887 3.77E+05 7.16E+01 7.16E+01 XE-133 5.60E+04 1887 1.06E+08 2.01 E+04 2.01 E+04 XE-135M 1.70E+04 1887 3.21 E+07 6.09E+03 6.09E+03 XE-135 9.80E+03 1887 1.85E+07 3.51 E+03 3.51 E+03 XE-138 4.40E+04 1887 8.30E+07 1.58E+04 1.58E+04 TOTALS 3.51 E+08 6.66E+04

(©GE BWR 6, Technical Description of a single cycle Boiling Water Reactor Nuclear System, January 1, 1974 (Table F.2.3-9)

Q 1850* 1.02 (DATNASSUMPTION # 11) 0 Column (D

  • Column © Activity in coolant from rods = (Column 0 / Total Core Activity)
  • Total Coolant Activity Column
  • 0.1 for lodines and 1.0 for Noble Gases

2.0 COOLANT SOURCE TERM ASSUMING 850 FAILED RODS TABLE 2 ISOTOPE CilMWt MWt TOTAL CORE ACTIVITY IN ACTIVITY COOLANT Ci Ci 1-131 2.90E+04 1887 5.47E+07 2.14E+05 1-132 4.20E+04 1887 7.93E+07 3.1 OE+05 1-133 4.80E+04 1887 9.06E+07 3.54E+05 1-134 6.20E+04 1887 1.17E+08 4.57E+05 1-135 4.90E+04 1887 9.25E+07 3.61 E+05 TOTAL 4.34E+08 KR-83M 3.OOE+03 1887 5.66E+06 2.21 E+04 KR-85M 6.50E+03 1887 1.23E+07 4.79E+04 KR-85 3.OOE+02 1887 5.66E+05 2.21 E+03 KR-87 1.20E+04 1887 2.26E+07 8.85E+04 KR-88 1.70E+04 1887 3.21 E+07 1.25E+05 KR-89 2.OOE+04 1887 3.77E+07 1.47E+05 XE-131M 1.80E+02 1887 3.40E+05 1.33E+03 XE-1 33M 2.OOE+02 1887 3.77E+05 1.47E+03 XE-1 33 5.60E+04 1887 1.06E+08 4.13E+05 XE-135M 1.70E+04 1887 3.21 E+07 1.25E+05 XE-135 9.80E+03 1887 1.85E+07 7.23E+04 XE-1 38 4.40E+04 1887 8.30E+07 3.24E+05 3.51 E+08 GE BWR 6, Technical Description of a single cycle Boiling Water Reactor D

Nuclear System, January 1, 1974 (Table F.2.3-9) 1850 MWt

  • 1.02 (DATA/ASSUMPTION # 1)

Column O

  • Column Q Column 0 * (850 failed rods / 32,632 total rods)*1.5 peaking factor
  • 0.1 As can be seen by comparing Table I Column 4 to Table 1 Column 4, the activity assumed in the coolant as a result of 850 failed fuel rods is greater that the activity in the coolant using the UFSAR coolant source term. Therefore the activity from 850 failed fuel rods will be for this calculation.

ACTIVITY AVAILABLE FOR RELEASE As stated in DATA/ASUMPTIONS 5.d all the noble gases and 10% of the iodines are released to the turbine and condenser. Therefore, Table 2 Column is multiplied by 0.1 for iodines and 1.0 for noble gases to

determine activity in the condenser.

TABLE 3 ISOTOPE ACTIVITY TO CONDENSER 1-131 2.14E+04 1-132 3.1 OE+04 1-133 3.54E+04 1-134 4.57E+04 1-135 3.61 E+04 KR-83M 2.21 E+04 KR-85M 4.79E+04 KR-85 2.21 E+03 KR-87 8.85E+04 KR-88 1.25E+05 KR-89 1.47E+05 XE-131 M 1.33E+03 XE-133M 1.47E+03 XE-133 4.13E+05 XE-135M 1.25E+05 XE-135 1.25E+05 XE-138 3.24E+05 The above condenser activity is input to DRAGON run # 5341 dated 4/29/98. The card input to this run is included in Appendix A.

DRAGON inputs:

Main Condenser volume: 5.OOE+04 ft3 (DATA/ASSUMPTION # 6)

Main Condenser release rate: 0.01 fractions per day (DATAIASSUMPTION #5.g)

Unit 1 Control Room (DATAIASSUMPTIONS # 7 - 9) volume: 1.36 E+05 ft3 intake rate 0 to 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />s: 3580 cfm filter efficiencies 0 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> = 0 Breathing rate 0 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> is 3.47E-04 m3 /sec (DATA/ASSUMPTION #12) 0-2 hour X/Q: 3.12E-04 sec/m3 2 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> X/Q: 1.22E-08 sec/m3 (DATAIASSUMPTION #10)

RESULTS The 0 - 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Unit I Control Room Doses are as follows TABLE 4 CRDA TO U1 CONTROL ROOM

UNIT I CR DOSES, GDC 19 LIMIT, REM REM THYROID 4.52E-01 30 GAMMA 7.65E-04 5 BETA 9.68E-03 30 CONCLUSIONS The control room doses are within the dose guidelines of 10CFR50 Appendix A GDC 19 assuming no control room air treatment system initiation. Therefore, the control room air treatment system is not required to meet the guideline doses.

COMPUTER RUN LOG JOB# DATE DESCRIPTION OF RUN 5341 4/29/98 DRAGON (REF 10) CRDA to Ul Control Room-no filters Card image is given in Appendix A

REFERENCES Nine Mile Point I Final Safety Analysis Report Revision 14 Table XV-9 1.6.4.0 XV.C.3.2 XV.C.4.5.1 III.B.2.2 General Electric Standard Application for Reactor Fuel, Licensing Topical Report, NEDE-24011-P-A-13 Class 1II,August 1996.

Engineering Report for Application of GEI 1 to Nine Mile Point Nuclear Station Unit 1 Reload 12, GENE-770-31-1292, revision 2, April, 1993 NUREG 0800, Standard Review Plan 15.4.9, Radiological Consequences of Control Rod Drop Accident (BWR).

Regulatory Guide 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility For A Boiling and Pressurized Water Reactor.

Technical Specifications section 1.14, Amendment 142, page 5 Calculation 1 H-009, Control Rod Drop Accident, revision 00 S10-210-HV12, Control Room & Auxiliary Control Room, revision 00 Letter dated March 19, 1984, from T. E. Lempges (NMPC) to D. B. Vassallo (NRC)

DRAGON computer code, SWEC Number NU-1 15, Version 5, Level 0 NMPC Calculation S10-CR277.A-Ul.210.

APPENDIX A ( 1 ATTACHED)

CARD IMAGE OF COMPUTER RUNS

MAY-05-98 TUE 07:59 AM P.01 I

NINE blILE POINT NUCLEAR STATION Unit (1, 2or o=8oth): I Discipline: ANALYSIS rim Calcuaion No.

CONTROL ROD DROP ACCIDENT (CRDA) IMPACT ON CONTROL H21 C046 ROOM HASITABILTIY (Sub)syslem(s) l Eulding l Floor Ele. Index No.

N/A TB 277 NINA.

rT. M.KURTZ*'

Checkw(s) I Approver(s) .. *

  • A.C. I1OISAN I T. G. KULCZYK Design/Config Prep'd Rev Description Change No. By Date Chk Date PP Date 00 ORIGINAL NA 5/t /98 794 5///?

. ,___.____.___ I -

Computer Output/Microfilm Filed Separately (Yes I No I NA): Yes Safety Class (SR I NSR I Qxx): SR

  • Superseded Document(s): NONE Document Cross Reference(s) - For additional references see page(s) :9 Ref . Doc No Document No. Type Index Sheet Rev

. J$SEEPAGE9 _____

General Reterence(s):

See page 9 Remarks:

Calculation to detrmlne Unit I Control Room doses from a design basis Unit I Control Rod Drop Accident (CRDA). This calculation determines that no Unit I Control Room Air Trestment System Initiation is required to reduce doses below GDC 19 limits.

Confirmation R1quired (Yes I No) : No Final Issue Status File Location I Operations Acceptance See Page(s): _ .1 .F l (APP I FI0 I .VOI):,- APP (Calc_ Hold):

Calc Required (Ye;INo): No Evaluation Number(s): SE 95-010 Component ID(s) (As shown In MEL):

Copy of Applicability Review Attached (Yes I NIR)?N/R NONE KeyWords:DOSE, GDC 19, CONTROL ROOM HA8ITABILITY, CRDA

P.u2 IIAY-O5-98 TUE U(:bq RM

  • VNIAGARA

{ U MOHAWK *i.,.. 'Ni TlNUA Tl(b3 SHEET - Page2 NUCLEAR ENGINEER1ING T:i6'jL ,6

.i. d.- *. (Net ) l Nine Mies Point Nuclear Station Unit: _1_I_ . Disposition: _NIA OrEnteor/Date Chedcer/Data culation No. . Revsion T. M. Kurt I4129/98 A: C. Moisan I 511198 H21C046 00.:

Ref.

TABLE OF CONTENTS paae #

Title Page 1 Table of Contents 2 Purpose 3 Background .3

-Methodology 3 Data / Assumptions 4 Calculation 5 Results 8 Conclusion 8 Computer Output Log B References 9 APPENDIX A - Card Image of Computer Run 10 I I_ _ -. .. _X.

,- ,s, ^ no BA_ ^ Ers FORMAT SNEW-LJ%-Vb, KrtV. UZ JrI4

P.03 1tY-lUb-t3 WUE Uf:bU firlf VNIAGARA Y..~IV NOMOHAWK NUCLEAR ENGINEERING lIcNIIA~iNONTINUATION SEET I-R

-Nine Mile Point Nuclear Station Unit: _1_ Disposition: N1A OfgrbIID8ate lCheckefDate l Cakaiaoi No. ReviscI T. M.Kurtz/_4/29198 A..C. Moisan/ 51/98 H21CO4.0 PURPOSE The purpose of this calculation Is to determine if the control room air treatment system is required during a control rod drop accident (CRDA) to maintain control room doses within 1DCFR50 Appendix A, GDC 19 acceptance criteria.

BACKGROUND In March, 1984 NMPI submitted to the staff the results of their control room habitability study in response to NUREG 0737 TMI Task Action Item IIl.D.3.4, Control Habitability Requirements. The design basis accidents identified in the submittal were the main steam line break and the loss of coolant accidents. The staff required that the licensee.use Stiridard Review Plan (SRP) 6.4, Control Room Habitability System, as one of the documents to determine if the control room habitability acceptance criteria was met. SRP 6.4, section 11.6; acceptance criteria states that Fin accordance with GDC 19 (Ref. 3), these doses (5rem or equivalent) to an individual in the control room should not be exceeded for any postulated design basis accident' Furthermore, SRP 15.4.9, Radiological Consequences of Control Rod Drop Accident (BWR). requires a specific evaluation of the CRDA for the first application involving a particular standardized design to establish a reference point for comparison of future applications incorporating the design. The Safety Evaluation Report (SER) received from the staff in May, 1984 stated ¶... The staffs conclusion is that control room operator doses follow'ing design basis accidents would be within GDC-19 guidelines.' As stated in the UFSAR section lll.B.1.5, Shielding and Access Control, 'The most limiting accidents are the main steam line (MSL) break accident and the loss-of-coolant accident (LOCA) without core spray...' However, supporting documentation could not be found that supported the assumption that the contiol rod drop and fuel handling (FHA) accident.were also evaluated as part Pf the study. As stated in the PURPOSE section, this calculation Is for CRDA only. The FHA has been evaluated in calculation H121C045.

Although it has not been determined if the CRDA was required to be evaluated as partof the Task Action Item. This calculation has been performed to determine if the control room air treatrnent system would be required to mitigate the radiological-consequences in the *control room to ensure an individual in the control room will not received greater that S rem whole body dose or.equivalent for any postulated design basis accident.

METHODOLOGY The resultant halogen and noble bas radioactivity described in the UFSAR assumed to be released to the coolant as a result of a CRDA is compared to a calculated reactor coolant source term based on the number of fuel rods that are assumed to fail as a result of a CRDA.

The higher of the two coolant source terms is input to the Stone and Webster computer code DRAGON for conservatism. The method used for release assumptions from the coolant to the environment are consistent with the Stahdard Review Plan 15.4.9 (REF 4) with the exception that an elevated release is assumed to be consistent with the UFSAR (REF 1.d).

FORMAT # NEP-DES-OB, Rev. 02 (F02)

MAY-05-98 TUE 08:00 AM P.04 Nine.Mile Point Nuclear Station Unit: I Disosition: NIA OJsnaporIDe {Checker/Date Calcubaion o. -- rat T. M.Kurtz_4/29198 A. C. Moisan/5/1198 H21C046 00 Ref, DATNASSUMPTIONS I E4 1. Nominal Reactor Power is 1850 MWt. A 2 % uncertainty is added. Therefore the Reactor Power used in this calculation Is 1887.

'2.7 85p fuel rods reach 170 cal/gm, which is th enthalpy limit for eventual cladding

J perforation. Since UI's peak enthalpy will be less than 280 cal/gm, melting does not occur. - g 1.6 ' 3. Fuel assumptions.
a. 532 fuel assemblies.

7 b. 62 rods/fuel assembly.

c. 32,632 total rods. This number was taken from radiological calculation 1H-009, Control Rod Drop Accident. The total rods were based on the actual fuel type in the Unit 1 core In 1991. Using only 8x8 fuel the total number of rods would be 62 by 532 = 32,984. This would result in an approximate 1 % decrease in source term (850132,984 versus 850/32,632). Therefore, the dose contribution using thei1991 total number of rods are.conservative and also negligible.
4. Elevated release (see METHODOLOGY section).

L *5. Release fractions

a. Amount of fuel gap activity is 10% of total activity in the rod
b. Fuel rods presumed failed are assumed to have operated at 1.5 times that of the average power level in the core.
c. 10 % of the gap activity for noble gas and iodines instantaneously and uniformly mix In reactor coolant in the pressure vessel at the time of the accident. Although Regulatory Guide 1.25 was used to define the gap activity released to the coolant, only 10% of the krypton 85 activity was assumed to be released instead of the 30% stated in RG1.25. Since Kr-85 had a negligible impact on dose. The computer run was not repeated.

d.- 10% of the iodines and 100% of the noble gases are released from the pressure vessel to the turbine.

e. All n6ble gases in the turbine are available for release.
f. 90% of the iodines are assumed to be removed by plateout and p'artitioning in the turbine and the condenser leaving only 10 Va airborne and available for leakage.
g. The turbine and condensers leak to the atmosphere at 1%,D/day for a period.of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leak terminates.

-7 6. Main Condenser volume: 5.OOE+04 ft?

7. Control Room free volume: 1.36E+05 ft'. The total volume, 1.69E405 ft?, is multiplied by 0.8 to account for equipment located in the control room envelope (main control room, auxiliary control room, instrument shop, and computer room).

1- c8. Control Room normal ventilation intake flow rate: 3550 cfm + 30 cfm in-leakage. This is calculated by taking the maximum flow rate of 16,300 cfm minus the minimum recirculation flow rate of 12,750 cfm (REF 1.e). The additional 30 cfm is to account for 10 cfm in!eakabe to the control room assumed in accordarice with SRP 6.4, section 111.3.d.(2).(ii). An additional 20 cfm is assumed to account for an unfiltered inleakage.

FORMAT

  • NEP-DES-08, Rev. 02 (F02)

MAY-05-98 TUE 08:01 AM P.05 IN,UCL, VNIAGARA l AMOHAWK ENGINEERING l j

iON CONTINUATON SHE j . - .,

Page 5 (Next---)

Nine Mile Point Nudear Station Unit _I_ Disposition: N/A 0,ist,. Il ChnkerMte Calmbao Nto. Revwcn T.--. Kurtz/ 4/29/98 A. C. Moisan /51/98 121d046 00

';'1

9. Credit is not taken for Control Room Air Treatment System initiation.

.- 10. U1 stack 0-2 hourX/Q is 3.12E-04 sec/m 3 and the 2-720 hourX/Q is 1.22E-08 secIM3 .

.N 11. Radioactivity assumed in the coolant as a result of a postulated CRDA i's Halogens = 5.62E+04 Ci and Noble Gas = 6.64E+04 Ci.

I)o 12. Breathing rate of 3.47E-04 m3 / sec for the duration (0-720 hours) of the accident.

13. 10CFR50 Appendix A, GDC 19 dose limit of 5 rem whole body or equivalent. This equates to 30 rem thyroid and 30 rem beta (skin). .- -

CALCULATION 1.0 UNIT I COOLANT ACTIVITY TABLE I

. ISOTOPE CVMWt MWt TOTAL ACTVITY IN ACTIVITY. .

0 CORE CCOOLANT FROM RELEASED TO -

ACTIVITY RODS THE CONDENSER 0

Ci Ci Ci 1-131 2.90E+04 1887 5.47E+07 7.09E+03 7.09E+02 1-132 *4.20E+04 1887 7.93E+07 1.03E+04 1.03E+03-1-133 4.80E+04 1887 9.06E+07 1.17E+04 1.17E+03 1-13.4 6.20E+04 1887 1.17E+05 1.51 E104 t1.51 E+03 1-135 4.90E+04 1887 9.25E+07 1.20E+04 1.20E+03 TOTAL 4.34E+08 5.62E+04 5.62E+03.

KR-83M 3.00E+03 1887 5.66E+06 1.07E+03 1.07E+03 KR-85M 6.50E+03 1887 I1.23E+07 2.33E+03 2.33E+03 KR-85 3.OOE+02 1887 6.66E+05 1.07E+02 1.07E+02.

KR-87 1.20E+04 1887 2.26E+07 4.30E+03 4.30E+03 KR-88 1.70E+04 1887 3.21E+07 B.09E+03 .6.09E+03 KR-89 2.OOE+04 1887 3.77E+07 7.16E+03 7.16E+03 XE-131M 1.80E+02 1887 3.40E+05 6.45E+01 6.45E+01 XE-133M 2.00E+02 1887 3.77E+05 7.16E+01 7.16E+01 XE-133 5.60E+04 1887 1.06E+08 2.01E+04 2.01 E+04 XE-1 35M 1.70E+04 1887 3.21 E+07 6.09E+03 6.09E+03 XE-i35 9.80E+03

  • 18B7 1.85E+07 3.51 E+03 3.51 E+03 XE-138 4.40E+04 1887 8.30E+07 1.5BE+04 1.58E+04 TOTALS 3.51 E+08 6.66E+04 Q GE EWR 6, Technical Description of a single cycle Boiling Water Reactor Nuclear Sy' item, January 1, 1974 (Table F.2.3.9) 6 1850 - 1.02 (DATAJASSUMPTION # 11)

O Column O)Columrt i t) Activity in coolant from rods = (C6lumn 0 / Total Core Activity) 'Total Coolant Activity 0 Column (4) 1 0.1 for lodines and 1.0 for Noble Gases FORMAT # NEP-DES.08,'Rev. 02 IF02)

r. uo

'VNIAGARA LdOHWK  :'L~ULTIN.NTINU~ATION-SHEE-T-l..,'

Pg 4

NUCLEAR ENrIwEERING C w *-. (Next-v Nine Mile Point Nuclear Sfation Unit _1_ Disposition: _NIA (OigratorI/Dke T. M. Kurtz 14129198 Ctlw/Date A. C. Moisanl 51198 Calculation No.

H21CD46

  • Revision 00 kRef_ -4

-2.0 COOLANT SOURCE TERM ASSUMING 850 FAILED RODS TABLE 2 ISOTOPE CiJMWt MWt TOTAL CORE ACTIVITY IN Di ACTIVITY COOLANT Ci Ci 1-131 2.90E+04 1887 5.47E+07 2.14E+05 1-132 4.20E-+04 1887 7.93E+07 3.10E+05 1-133 4.80E404 1887 9.06E*07 3.54E*OS 1-134 6.20E+04 1887 1.17E+08 4.57E+05 1-135 4.90E+04 1887 9.25E+07 3.61E+05 TOTAL 4.34E+08 KR-83M 3.00E+03 1887 5.66E+06 2.21ED+04 KR-85M 6.50E+03 1887 1.23E+07 4.79E+04 KR-85 3.OOEt02 1887 5.66E+05 2.21E+03:

KR-87 1.20E+04 1887 2.26E+07 8.85E+04 KR-88 1.70E+04 1887 3.21E+07 1.25E+05 KR-89 2.OOE+04 1887 3.77£+07 1.47E+05 XE-131M 1.80E+02 1887 3.40E+05 1.33E+03 XE-133M 2.00E+02 1887 3.77E+05 1.47E+03 XE-133 5.60E+04 1887 1;06E+08 4.13E+05 XE-135M 1.70E+04 1887 3.21E+07 1.25E+05 XE-135 9.80E+03 1887 1 85E+07 7.23E+04 XE-138 4.40E+04 1887 8.3012+07 3.24E+05 3.51 E-08 OGE. BWR 6, Technical Description of a single cycle Boiling Water Reactor Nuclear System, January 1, 1974 (Table F.2.3-9)

C) 1850 MWt

  • 1.02 (DATA/ASSUMPTION # 1)

- Column G' Column @

6 Column O (850 failed rods / 32,632 total rods)*1.5 peaking factor' 0.1 Asican be seen by comparing Table 1 Column 4 to Table I.Column 4, the activity assumed in the coolant as a result of 850 failed fuel rods is greater that the activity in the coolant using the UFSAR coolant source term. Therefore the activity from 850 failed fuel rods will be for this calculation.

FORMAT X NEP-DES-08, Rev. 02 (F02)

Ifini-UO-00 IUC UO-Ui Mf P.07 NNIAGARA HWCALCUWLATION'Cb~IO kM n-~m U INSH TPage 7 NUCLEAR ENGINEERING I , i, *"l , (Nex Nine Mile Point Nuclear Station Unit: 1_- . Disposition: N/A.

6Oanator/Dwe CheckedDate CRlcuaon No. Revision

.T. M. Kurtz ! 4/29/98 A. C. Moisan /5/1/98 H21C046 l00 Ref, 3.0 ACTIVITY AVAILABLE FOR RELEASE As stated in DATAIASUMPTIONS 5.d all the noble gases and 10% of the iodines are released to the turbine and condenser. Therefore, Table 2 Column Q0 is multiplied by 0.1.for iodines and 1.0 for noble gases to determine activity in the condenser..

TABLE3.

ISOTOPE ACTIVITY TO. CONDENSER 1-131 2.14E+04 1-132 3.1OE+04 1-133 3.54E+04 1-134 4.57E+04 1-135 3.61 E+04 KR-83M 2.21 E+04 KR-85M 4.79E+04 KR-B5 2.21 E+03 KR-87 8.85E+04 KR-88 1.25E+05 KR-89 1.47E+05 XE-131M 1.33E+03 XE-133M 1.47E+03 XE-133 4.13E+05 XE-135M 1.25E+05 XE-135 1 .25E+05 XE-138 3.24E+05 The above condenser activity is input to DRAGON run # 5341 dated 4129/98. The card Input to this run is included in Appendix A.

DRAGON inputs:

Main Condenser volume: 5.00E+04 fl3 (DATAIASSUMPTION # 6)

Main Condenser release rate: 0.01 fractions per day (DATAIASSUMPTION #5.g)

Unit I Control Room (DATA/ASSUMPTIONS # 7 - 9)

. volume: 1.36 E+05 ft

  • intake rate 0 to 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />s: 3580 cfm
  • filter efficiencies 0- 72b hours = 0 Breathing rate 0 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> is 3.47E-04 m3 Isec (DATA/ASSUMPTION #12) 0-2 hourXIQ: 3.12E-04 sec/r 3 2 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> X/Q: 1,22E-08 sec/mr3 (DATA/ASSUMPTION '#10)

-I FORMAT # NEP-DES-0B,.Rev. 02 (FOZ)

rIMMI-us-co J.UL u'uC nfl a . UW I DNIAGARA NN=IN'O

[NUCLEA:RENGINEERING NL;1(Ndl ~--

TNUAI $.H'EET:-'".-

Page 8 N x.

Nine Mile Point Nuclear Station Unit: __I_ Disposition: _N/A OngriaeorltrOai CherIDste *CakulaUnNo.. ReNisin T. M. Kurtz 14/29198 A. C. Moisan 1001/98 H21C046 0 Ref.

.

  • RESULTS The 0 - 720 hour0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Unit I Control Room Doses are as follows TABLE 4 CRDA TO UI CONTROL ROOM U

UNIT I CR DOSES, . GDC 19 LIMIT, REM REM THYROID 4.52E-0I 30 GAMMA 7.65E-04 5 BETA 9.68E-03 30 CONCLUSIONS The control room doses are within the.dose guidelines of 10CFR5O Appendix A GDC 19 assuming no control room air treatment system initiation. Therefore, the control room air treatment system Is not required to meet the guideline doses.

COMPUTER RUN LOG JOB # DATE DESCRIPTION OF RUN

  • 5341 4/29198 DRAGON (REF 10) CRDA to Ul Control Room-no filters Card image is given in Appendix A FORMAT # NEP-DES-08, Rev. 02 (F02)

D no MAY-05-98 TUE B0:02 AM 1

lNUCLEAR ENGINEERING %ET:~lPr VNNIAGARA MOHAWK

,.:I:.

.LC~

.ELA!

JIL&LJA1 UI~ wS EPage 9

'Nine Mile PointNuciearStaton Unit I Disposition: WIA OrtInorDate Checkef/Dats Calcuiaion No. . Revison

.T.M. Kurtz 4129198 A. C. Moisan 1511/98 H21C046 00 Ref.

. REFERENCES

1. Nine Mile Point 1 Final Safety Analysis Report Revision 14
a. Table XV-9
b. 1.8.4.0 -
c. XV.C.3.2
d. XV.C.4.5.1
e. II .B.2.2
2. General Eleciric Standard Application for Reactor Fuel, Licensing Topical Report, NEDE-24011-P-A-13 Class lIl, August 1996.
3. Engineering Report for Application of GE1 1 to Nine Mile Point Nuclear Station Unit 1 Reload 12, GENE-770-31-1292. revision 2, April, 1993
4. NUREG 0800, Standard Review Plan 15.4.9, Radiological Consequences of Control Rod.

Drop Accident (BWR).

5. Regulatory Guide 1.25, Assumptions Used for Evaluating the Potential Radiological

. Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility For

. .A Boiling and Pressurized Water Reactor.

6. Technical Specifications section 1.14,.Amendment 142, page 5
7. Calculation 1H-009, Control.Rod Drop Accident, revision 00
8. SIO210-HV12, Control Room & Auxiliary Control Room, revision 00
9. Letter dated March 19, 1984, from T. E. Lempges (NMPC) to'D. S. Vassallo (NRC)

. 10. DRAGON computer code, SWEC Number NU-115, Version 5, Level O

11. NMPC Calculation S10-CR277.A-U1.210.

FORMAT# NEP-DES-08, Rev. 02 (F02)

tlAY-05-98 TUE OB02 AMl P.10 V NIAGARA l 2 Sk OLA i (NxIO Ru4mOA K i .. A LC T,ICNCONTINUAlO ET M4SH 2... Page 10 NUCLEAR ENGINEERINM..

Nine Mile Pbint Nuclear Sfationi Unit: __1_ Disposition: N/A Odgbnaie . , CheckerJDate Caculation No. -Revision T M.Kurtz/ 4/29/98 A. C. Mooson / 5/1198 H21C046 00 Ref, APPENDIX A ( 1 ATTACHED)

CARD IMAGE OF COMPUTER RUNS I . . -- -- -..__.

FORMAT # NEP-DES-08, Rev. 02 (FOZ)

.**.** 12noc- AU -- -- Nut,'o.-v:.1gs..Lvon-- 4J0o46 -- W*&- A I It I

  • tA:or.

CD, CD

  • 6 * *
  • a * * & CARA) f4rGj--.0r INP.UT SU.4lrTI7D 10 0nA4oN w

' I *' *4 4 * -3 M

CAR*3 c0LUMPN% 1 2 3 a 4 l s 7 n CA"o CULUNFII _r, 1234567USN C n C4 UZ56 S7190 IP1 s67eon t2aA5p090A1234567?22PS234567490 CAR') No.

I **NMPI CADA QRCRFATCO USING sAP 14001:1 AND 050 OR .12b3Z PINS PErF

  • 3 fall 1 1011 0 I 0 0 0 .10 0 1*

4AtHNCONTC-NS CAR ' s.004 0 '1 n 0 0*01 .0 0 0 CO4TROL RONM4 1. 36#;5 0 0 o 0 1.0 :0 0 0*00tO Z. .1444 3.1014 1. 494 4.5r74 ,J. I+4 0.00+0 0.0010 0*00+0 0. 00.o o.ooeo z.*214 4*7'?4 2.21 -I IA. f 1.25 FS 7 1.4FI5 to .Z343 1.471.1 4.l395 I.Z5*i r.2304 0.00o0o3.245 0 I a I I 0 3q3no. 35d0.3. 12-4 *I I 2.0

'I 1 a 0 0 0 3sA0. 35'0* 1*22-9 .I I 44.R to .3 a 0 a 0 0 3Iso. I500. 1.2-0 'I I 720.0 173456ft740 12345679901734S67801o234S6789ntP 34567n9soIZ34y67890IP3456s09U IZ34557ftsn VI-C'%

Retrieved from "https://kanterella.com/w/index.php?title=ML050410068&oldid=2596356"