ML17292B534
| ML17292B534 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/15/1999 |
| From: | Richard I, Rickard I ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-PT21-99 LD-99-002, LD-99-2, NUDOCS 9901220123 | |
| Download: ML17292B534 (89) | |
Text
January 15, 1SS9 LD-99-002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Report of a Oefect Pursuant to 1Q CFR 21 Concerning Incorrect Modeling of a BWR Lower Plenum Volume in BISON The purpose of this letter is'to notify the Nuclear Regulatory Commission of a defect under 10 CFR 21, "Reporting of Defects and Noncompliance." The defect concerns the Operating Limit for the Minimum Critical Power Ratio (MCPR) in Boiling Water Reactors (BWRs) analyzed using the BISON fast transient analysis code. Specifically, the defect involves incorrect modeling of the reactor vessel lower plenum volume in the BISON code, which could lead to the establishment of non-conservative MCPR Operating Limits in plant technical specifications.
The Enclosure summarizes the evaluation performed by ABB Combustion Engineering (ABB-CE). If you have any questions, please feel free to contact me or Virgil Paggen of my staff at (860) 2854700, Very truly yours, COMBUSTION ENGINEERING, INC.
lan . ickard, Director uclear Licensing
Enclosure:
As stated cc: M. A. Barnoski (ABB-CE)
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ABB Combustion Engineering Nuclear Power Comhusion Engineer(ng, Inc. P.O. Box 500 Te(ephone (880) 888-1911 2000 Day Hih Rd. Fax (880) 285-5203 Windsor, CT 0609&0500
Enclosure to LD-99-002 Page2of2 (vi) In the case of a basic component which contains a defect orfails to comply, the number and location ofall such components in use at, supplied for, or being supplied for one or more facilities or activities subj ect to the regulations in this part:
The BISON computer code is a proprietary ABB Atom AB program used by ABB-CE for performing BWR safety analyses for U.S. customers. The defect applies only to reload fuel assemblies currently in operation at WNP-2.
(vii) The corrective action which has been, is being, or willbe taken; the name oftheindividual or organization responsible for the action; and the length oftime that has been or willbe taken to complete the actiont An evaluation of the defect has been performed by ABB-CE. The results of that evaluation show that there is no safety problem at WNP-2 for operation of the previous Cycles 12 and 13 and its current Cycle 14 operating state (Normal Scram Speed, Recirculation Pump Trip operable, and Turbine Bypass operable) prior to 5,000 MWd/MTU. That is, the established MCPR Operating Limits currently in the plant technical specifications for this operating state provide adequate protection. However, certain MCPR Operating Limits for other operating states prior to 5,000 MWd/MTU and all operating states after 5,000 MWd/MTU for WNP-2 Cycle 14 must be increased to accommodate the defect. The utility currently projects that WNP-2 will reach 5.000 MWd/MTU in early 1999. ABB-CE has notified the Washington Public Power Supply System of the necessary changes to the WNP-2 MCPR Operating Limits.
The BISON code model for WNP-2 has been revised to correct the error. All ABB-CE BWR fuel users have been notified of this condition.
(viii) Any advice related to the defect orfailure to comply about the facility, activity, or basic component that has been, is being, or willbe given to purchasers or licensees:
The Washington Public Power Supply System has been notified of the defect and has been provided with revised MCPR Operating Limits for WNP-2.
TOTAL P.84
Enclosure to LD-99<02 Pageos of 2 ABB Combustion Engineering Nuclear Power 10 CFR 21 Re ort of a Defect or Failure to Com I The following information is provided pursuant to the requirements set forth in 10 CFR 21.21(c)(4):
(l) liame and address ofthe individuals informing the Commiss1on:
lan C. Rickard, Director Nuclear Licensing Combustion Engineering, Inc.
2000 Day Hill Road .
Windsor, CT 06095-0500 (ii) Identification ofthe facility, the activity, or the basic component supplied for such facility or such activity within the United States which fails to comply or contains a defect:
The activity for which this report is being filed is the establishment of non-conservative MCPR Operating Limits for the Washington Public Power Supply System Nuclear Project Unit 2 (WNP-2) nuclear power plant during Cycles 12, 13, and 14 operation..
(iii) Identification of the fir constructing the facility or supplying the basic component which fails to comply or contains a defect:
Combustion Engineering, Inc.
2000 Day Hill Road Windsor, CT 06095-0500 (iv) iVature ofthe defect orfailure to comply and the safety hazard which is created or could be created by such defect or failure to comply:
The defect identified involves the incorrect modeling of the reactor vessel lower plenum volume in the BISON code. The defect caused the BISON code model to have a lower plenum volume approximately twice its proper size. This incorrect modeling has the potential to lead to calculation of non-conservative MCPR Operating Limits in some situations.
(v) The date on which the information ofsuch defect or failure to comply was obtained:
ABB-CE determined that a defect in the BISON code existed on January 14. 1999.
02/16/1999 U,S. Nuc/ egu/atory Commission Operations Center Report P~ae I X/
General Information or Other (PAR) Event ¹ 35271 Rep Org: ABB COMBUSTION ENGINEERING Notification Date / Time: 01/15/1999 14:15 (EST)
Supplier: ABB COMBUSTION ENGINEERING Event Date / Time: 01/15/1999 14:15 (EST)
Last llodiflcation: 01/15/1999 Region: 1 Docket ¹:
City: WINDSOR Agreement State: No County: License ¹:
State: CT NRC Notified by: IAN RICKARD Notifications: BLAIR SPITZBERG R4 MQ Ops Officer: DOUG WEAVER VERN HODGE NRR Emergency Class: NON EMERGENCY 10 CFR Section:
21.21 UNSPECIFIED PARAGRAPH PART 21 NOTIFICATION RELATED TO MINIMUMCRITICAL POWER RATIO (MCPR)
"The purpose of this letter is to notify the Nuclear Regulatory Commission of a defect under 10 CFR 21, 'Reporting of Defects and Noncompliance.'he defect concerns the Operating Limit for the Minimum Critical Power Ratio (MCPR) in Boiling Water Reactors (BWRs) analyzed using the BISON fast transient analysis code. Specifically, the defect involves incorrect modeling of the reactor vessel tower plenum volume in the BISON code, which could lead to the establishment of non-conservative MCPR Operating Limits in plant technical specifications.
"The Enclosure summarizes the evaluation performed by ABB Combustion Engineering (ABB-CE). If you have any questions, please feel free to contact me [Ian C. RickardJ or Virgil Paggen of my staff."
'I This event effects WNP-2 current operating cycle (Cycle 14).
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REGULA Y INFORMATION DISTRIBUTIO SYSTEM (RIDS)
ACCESSION NBRP9901220123 DOC.DATE: 99/01/15 NOTARIZED: NO DOCKET ¹ FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION RICKARD,I.C. ABB Combustion Engineering Nuclear'uel (formerly Combustio RICHARD,I.C. ABB Atom, Inc. (formerly ASEA Atom, Inc.)
RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)
SUBJECT:
Part 21 rept re incorrect modeling of BWR lower plenum vol in BISON.Decfect applies only to reload fuel assemblies currently in operation at WNP-2.BISON code model for WNP-2 has been revised to correct error.
DISTRIBUTION CODE: IE19T COPIES RECEIVED:LTR ENCL SIZE:
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NOTES:
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llew, O LFQO January 15, 1999 LD-99-002 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Report of a Defect Pursuant to 10 CFR 21 Concerning Incorrect Modeling of a BWR Lower Plenum Volume in BISON The purpose of this letter is to notify the Nuclear Regulatory Commission of a defect under 10 CFR 21, "Reporting of Defects and Noncompliance." The defect concerns the Operating Limit for the Minimum Critical Power Ratio (MCPR) in Boiling Water Reactors (BWRs) analyzed using the BISON fast transient analysis code. Specifically, the defect involves incorrect modeling of the reactor vessel lower plenum volume in the BISON code, which could lead to the establishment of non-conservative MCPR Operating Limits in plant technical specifications.
The Enclosure summarizes the evaluation performed by ABB Combustion Engineering (ABB-CE). If you have any questions, please feel free to contact me or Virgil Paggen of my staff at (860) 285<700.
Very truly yours, COMBUSTION ENGINEERING, INC.
lan . ickard, Director uclear Licensing
Enclosure:
As stated cc: M. A. Barnoski (ABB-CE)
'V ABB Combustion Engineering Nuclear Power Cornbusdon Engineering, Inc. P.O. Box 500 Telephone (860) 688-1911 2000 Day Hill Rd. Fax(860) 285-5203 Mrindsor, CT 060954500
'7901220123 990115 PDR ADQCK 05000397 S ,
Enclosure to LD-99-002 Page 1 of 2 ABB Combustion Engineering Nuclear Power 10 CFR21 Re ortof a Defector Failure to Com I The following information is provided pursuant to the requirements set forth in 10 CFR 21.21(c)(4):
(i) Name and address ofthe individuals informing the Commission:
lan C. Rickard, Director Nuclear Licensing Combustion Engineering, Inc.
2000 Day Hill Road Windsor, CT 06095-0500 (ii) Identification ofthe facility, the activity, or the basic component supplied for suclt facility or suclt activity within the United States which fails to comply or contains a defect:
The activity for which this report is being filed is the establishment of non-conservative MCPR Operating Limits for the Washington Public Power Supply System Nuclear Project Unit 2 (WNP-2) nuclear power plant during Cycles 12, 13, and 14 operation.
(iii) Identification oftlte firm constructing the facility or supplying the basic component which fails to comply or contains a defect:
Combustion Engineering, Inc.
2000 Day Hill Road Windsor, CT 06095-0500 (iv) Nature ofthe defect or failure to comply and tlte safety hazard whiclt is created or could be created by such defect or failure to comply:
The defect identified involves the incorrect modeling of the reactor vessel lower plenum volume in the BISON code. The defect caused the BISON code model to have a lower plenum volume approximately twice its proper size. This incorrect modeling has the potential to lead to calculation of non-conservative MCPR Operating Limits in some situations.
(v) The date on which tlteinformation ofsuch defect or failure to comply was obtained:
ABB-CE determined that a defect in the BISON code existed on January 14, 1999.
Enclosure to LD-99-002 Page2of2 (vi), In the case ofa basic component wltich contains a defect or fails to comply, the number and location ofall such components in use at, supplied for, or being supplied for one or more facilities or activities subject to the regulations in this part:
The BISON computer code is a proprietary ABB Atom AB program used by ABB-CE for performing BWR safety analyses for U.S. customers. The defect applies only to reload fuel assemblies currently in operation at WNP-2.
(vii) The corrective action wlticlt has been, is being, or willbe taken; the name ofthe individual or organization responsible for the action; and the length oftime tltat has been or willbe taken to complete the action:
An evaluation of the defect has been performed by ABB-CE. The results of that evaluation show that there is no safety problem at WNP-2 for operation of the previous Cycles 12 and 13 and its current Cycle 14 operating state (Normal Scram Speed, Recirculation Pump Trip operable, and Turbine Bypass operable) prior to 5,000 MWd/MTU. That is, the established MCPR Operating Limits currently in the plant technical specifications for this operating state provide adequate protection. However, certain MCPR Operating Limits for other operating states prior to 5,000 MWd/MTU and all operating states after 5,000 MWd/MTU for WNP-2 Cycle 14 must be increased to accommodate the defect. The utility currently projects that WNP-2 will reach 5,000 MWd/MTU in early 1999. ABB-CE has notified the Washington Public Power Supply System of the necessary changes to the WNP-2 MCPR Operating Limits.
The BISON code model for WNP-2 has been revised to correct the error. All ABB-CE BWR fuel users have been notified of this condition.
(viii) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or willbe given to purchasers or licensees:
The Washington Public Power Supply System has been notified of the defect and has been provided with revised MCPR Operating Limits for WNP-2.
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Item: ADAMS Document Library: ML ADAMS"HQNTAD01 ID: 003686666'1
Subject:
SUMMARY
OF JANUARY 27, 2000 MEETING WITH ENERGY NORTHWEST REGARDING THE PROPOSED SECONDARY CONTAINMENT/STANDBYGAS TREATMENT SUBMITTA L.
Body:
ADAMS DISTRIBUTION NOTIFICATION.
Electronic Recipients can RIGHT CLICK and OPEN the first Attachment to View the Document in ADAMS. The Document may also be viewed by searching for Accession Number ML003686666.
DF01 - Direct Flow Distribution: 50 Docket (PDR Avail)
Docket: 05000397 Page 1
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lt NUCLEAR REGULATORY COMMISSION n~g e be Jll]lit WASH iNGTON, O.C. 20555-0001 February 15, 2000
<conn~( in 4 gO genre'ICENSEE:
Energy Northwest Rle. ~~
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FACILITY: WNP-2
SUBJECT:
SUMMARY
OF MEETING WITH ENERGY NORTHWEST REGARDING THE PROPOSED SECONDARY CONTAINMENT/STANDBYGAS TREATMENT SUBMITTAL On January 27, 2000, the Nuclear Regulatory Commission (NRC) staff met with representatives of Energy Northwest to discuss the upcoming secondary containment/standby gas treatment license amendment submittal. The proposed changes would revise the following:
SR 3.6.1.3.10 increases the secondary containment bypass leakage from .74 standard cubic feet per hour (scfh) to .028 percent/day approximately 9.4 scfh.
SR 3.6.4.1.1 change the requirement to verify secondary containment pressure is less than .25 inch vacuum water gage to less than 0 inch vacuum water gage.
SR 3.6.4.1.4 increase the secondary containment drawdown time from 2 minutes to 20 minutes.
TS 5.5.7 increase the standby gas treatment system flow rate to 5000 cubic feet per minute (cfm).
Energy Northwest withdrew a similar amendment request in July 1999, when the staff identified a calculation error in determining containment release concentration. Due to the extensive review required of the previous submittal, both the staff and the licensee felt that a meeting to discuss the upcoming submittal would be beneficial in shortening the review and avoiding unnecessary requests for additional information.
In order to improve efficiency the same NRC staff members who were involved in the final review for the first submittal were at the meeting and will review the second submittal.
Enclosure 1 is a list of the meeting participants. Enclosure 2 is a copy of the slides presented by Energy Northwest.
At the outset of the meeting, Mr. John Arbuckle of Energy Northwest presented an overview of the submittal. The emphasis was on how this submittal has changed from the previous submittal. This submittal includes meteorological data and atmospheric dispersion calculations (X/Q calculations). Leta Brown, of the NRC staff, suggested that it would be useful if the X/Q calculations and an electronic version of the meteorological data were provided for staff review.
Mr. David Studley of Scientech discussed the X/Q calculations including the four release points and the control room intakes. The discussion also covered the application of ARGON 96 Code and the assumptions that were made. The description of the site configuration and of the Oo @~~It" ~48KFiX gfjP"
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control room intakes were useful to the staff in understanding the assumptions that were made.
The staff expressed a concern with the vent option used in the ARCON 96 Code that was used to calculate some of the control room atmospheric dispersion factors. It was suggested that an acceptable solution would be to recalculate the vent run cases as a ground release option using ARGON 96.
'r.
Bruce Boyum of Energy Northwest discussed the accidents that were analyzed specifically, main steam line break accident, fuel handling accident, control rod drop accident and loss-of-coolant accident (LOCA). Mr. Boyum stated that the LOCA was the bounding accident.
Mr. Mark Blumberg,'of the NRC staff, stated that it would be useful to include the calculations for the LOCA analysis and include sufficient informatiori on other accidents so that it can be determined that the LOCA is the bounding accident.
Mr. Studley described the Axident Code, which is the dose analysis code used in the submittal.
The Axident Code models the transport of radioactivity to the environment and to the control room. The major assumptions and the reasons for them were discussed.
Mr. Boyum also discussed the release pathways including the use of the Gothic Code to justify the 40 percent mixing assumed in secondary containment. Mr. Richard Lobel, of the NRC staff, said the proposed submittal should include a description of the derivation and use of the flow equation which is the basis for Figure 4 of Attachment 2 to the licensee's October 15, 1996, submittal. In addition, the staff may request input used in reactor building pressure drawdown calculations so that the staff may perform independent calculations. A final decision has not been made.
Mr. Boyum then discussed control room air flows and unfiltered control room in-leakage.
Mr. Blumberg stated that licensees have had to verify their unfiltered in-leakage assumptions.
The ASTM E741, "Standard Test Methods for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution," test method is acceptable to the staff as a verification test for unfiltered in-leakage. The licensee could describe their design and propose an alternative test method that would have to be reviewed and approved by the staff.
The NRC staff felt that Energy Northwest did a good job explaining their submittal and that they were receptive to suggestions from the staff.
Jack Gushing, Project Manager, Section 2 Project Directorate IV 8 Decomissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation DISTRIBUTION:
Docket No. 50-397 Co 'Hard E-Mail Central File JZwolinski/SBlack
Enclosures:
1 ~ List of Meeting Participants PUBLIC S Richards
- 2. Energy Nothwest Slides PDIV-2 R/F DLange, EDO JCushing E Peyton cc w/encls: See next page OGC RLobel ACRS MBlumberg LBrown LSmith, RIV DOCUMENT NAME: GAPDIV-2EWNP2iMTS12700.
r cewe c I U( e, I I u x "f~ = /t//r.'/'=~ '/
OFFICE NAME DATE PDIV-2/PM Jcushing:Icc C PDIV-2/LA EPeyton cB I I "f fOo PDIV-2/SC SDembe ~6hz FFICIAL RECORD COPY Qg~
0 control room intakes were useful to the staff in understanding the assumptions that were made.
The staff expressed a concern with the vent option used in the ARCON 96 Code that was used to calculate some of the control room atmospheric dispersion factors. It was suggested that an acceptable solution would be to recalculate the vent run cases as a ground release option using ARGON 96.
Mr. Bruce Boyum of Energy Northwest discussed the accidents that were analyzed specifically, main steam line break accident, fuel handling accident, control rod drop accident and loss-of-coolant accident (LOCA). Mr. Boyum stated that the LOCA was the bounding accident.
Mr. Mark Blumberg, of the NRC staff, stated that it,would be useful to include the calculations for the LOCA analysis and include sufficient information on other accidents so that it can be determined that the LOCA is the bounding accident.
Mr. Studley described the Axident Code, which is the dose analysis code used in the submittal.
The Axident Code models the transport of radioactivity to the environment and to the control room. The major assumptions and the reasons for them were discussed.
Mr. Boyum also discussed the release pathways including the use of the Gothic Code to justify the 40 percent mixing assumed in secondary containment. Mr. Richard Lobel, of the NRC staff, said the proposed submittal should include a description of the derivation and use of the flow equation which is the basis for Figure 4 of Attachment 2 to the licensee's October 15, 1996, submittal. In addition, the staff may request input used in reactor building pressure drawdown calculations so that the staff may perform independent calculations. A final decision has not been made.
Mr. Boyum then discussed control room air flows and unfiltered control room in-leakage.
Mr. Blumberg stated that licensees have had to verify their unfiltered in-leakage assumptions.
The ASTM E741, "Standard Test Methods for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution," test method is acceptable to the staff as a verification test for unfiltered in-leakage. The licensee could describe their design and propose an alternative test method that would have to be reviewed and approved by the staff.
The NRC staff felt that Energy Northwest did a good job explaining their submittal and that they were receptive to suggestions from the staff.
Jack Gushing, Project Manager, Section 2 Project Directorate IV 8 Decomissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 1. List of Meeting Participants
- 2. Licensee's Slides cc w/encls: See next page
y l
C
WNP-2 CC:
Mr. Greg O. Smith (Mail Drop 927M) Thomas C. Poindexter, Esq.
Vice President, Generation Winston & Strawn Energy Northwest 1400 L Street, N.W.
P. O. Box 968 Washington, DC 20005-3502 Richland, Washington 99352-0968 Mr. Bob Nichols Mr. Albert E. Mouncer (Mail Drop 1396) Executive Policy Division Chief Counsel Office of the Governor Energy Northwest P.O. Box 43113 P.O. Box 968 Olympia, Washington 98504-3113 Richland, Washington 99352-0968 Mr. J. V. Parrish Ms. Deborah J. Ross, Chairman Chief Executive Officer Energy Facility Site Evaluation Council Energy Northwest P.'. Box 43172 P.O. Box 968 (Mail Drop 1023)
Olympia, Washington 98504-3172 Richland, WA 99352-0968 Mr. D. W. Coleman (Mail Drop PE20)
Manager, Regulatory Affairs Energy Northwest P.O. Box 968 Richland, Washington 99352-0968 Mr. Paul Inserra (Mail Drop PE20)
Manager, Licensing Energy Northwest P.O. Box 968 Richland, Washington 99352-0968 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission
, Harris Tower 8 Pavilion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 Chairman Benton County Board of Commissioners P.O. Box 69 Prosser, Washington 99350-0190 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 69 Richland, Washington 99352-0069 Mr. Rodney L. Webring (Mail Drop PE08)
Vice President, Operations Support/PIO Energy Northwest P. O. Box 968 Richland, Washington 99352-0968
ENERGY NORTHWEST MEETING PARTICIPANTS JANUARY 27 2000 ENERGY NORTHWEST Douglas Coleman Bruce Boyum John Bekahazi Linda Woolsley John Arbuckle SCIENTECH-NUS David Studley NRC Jack Gushing Steve Dembek Mark Blumberg Richard Lobel Leta Brown Enclosure 1
ENERGY NORTHWEST Secondary Containment/SGT Submittal Presentation A
Introduction (John Arbuckle)
Meteorological Data And X/Q Calculations (John Arbuckle and Dave Studley)
Dose Analysis And Results (Bruce Boyum and Dave Studley)
Summary (Bruce Boyum)
Enclosure 2
INTRODUCTION Initial Problem Submittal History Analysis Problem/TS Retraction/JCO-FAO Impact Comparison of Key SGT Parameters Technical Specification Changes Submittal Content
INITIALPROBLEM Under Certain Post-Accident Meteorological Conditions, WNP-2 could not Develop 0.25-inch Negative Differential Pressure Within 120 Seconds Therefore, a Revised Design Basis and Dose Analysis was Provided. JCO (FAO) Prepared and Submitted to Staff.
SU BMITTALHISTORY
~ October 1996 Technical Specification. Amendment Request December 1997 June 1999 Formally Responded to Three RAls
ANALYSIS PROBLEM Analysis Problem/TS Retraction/JCO-FAO Impact Jul 1999 Withdrew Technical Amendment Request Discovery of a Nonconservative Error in Determining Containment Release Concentration During Resolution of Proposed RAI 4 No Impact on JCO-FAO, Current Design Basis, Technical Specifications or Recent Analyses
COMPARISON OF KEY SGT PARAMETERS Original Current Proposed Ke Parameter Desi n Desi n JCO-FAO Desi n Drawdown Time 2 minutes 10 minutes 20 minutes SC Leakage 2240 cfm 1475 cfm 2240 cfm SGT Flow 4457 cfm 5385 5850 cfm 5000 cfm
l
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TECHNICAL SP ECIF ICATIONS Proposed Technical Specification Changes SR 3.6.1.3.10 Increase Secondary Containment Bypass Leakage from 0.74 scfh to(0.028%/day) <~<
9.4 scfh)
SR 3.6.4.1.1 Change the Surveillance Requirement to Verify Every 24 Hours that the Pressure Within Secondary Containment is (0 inch (vs 0.25 inch) c~ -'-',,
of Vacuum Water Gauge SR 3.6.4.1 4 Increase Secondary Containment Drawdown Time from 120 seconds to 20 minutes 5.5.7.2.A Increase Standby Gas Treatment System Flow Rate from 4457 cfm to 5000 cfm
SU BMITTALCONTENT Detailed History Supersedes Previous Submiitals Responses to RAls Incorporated Design Basis Meteorology and X/Q Valuey>, ,rii. " 'it Accidents Analyzed - LOCA in Detail, FHA, MSLB and CRDA GOTHIC Model and Benchmarking Efforts Discussion of Standby Gas Treatment System Evaluation of Significant Hazards Environmental Considerations Evaluation
~
Marked-Up and Typed Technical Specifications Marked-Up Technical Specification Bases - For Info Only
I METEOROLOGICAL DATA AND XlQ CALCULATIONS Description of Met Tower, Terrain, and Instrumentation Data Used (6yrs)
ARCON96
- 1. Description of Release Points and Control Room lntakes
- 2. Application of ARCON96 at WNP-2
- 3. Comparison With JCO-FAO and Power Uprate
DESCRIPTION OF MET TOWER, TERRAIN, AND INSTRUMENTATION Meteorological Tower Consists of a 240-ft Structure with a 5-ft Extension Mast The Tower is Triangular in Shape and of Open Lattice Construction to Minimize Tower Interference with Meteorological Measurements Wind Speed and Direction is Monitored by Separate Channels at the 33-ft and 245-ft Elevations A Single Channel Provides Air Temperature Difference Between 33-ft and 245-ft Elevations
DESCRIPTION OF MET TOWER, TERRAIN, AND INSTRUMENTATION Siting of Instrumentation with Respect to Meteorological Tower and Surrounding Vegetation is Very Good The Base of the Tower Maintained as Natural Vegetation Area Around the Tower is Open Terrain with no Natural or Man-Made Obstructions to Impact Data Being Collected
DATA USED (6YRS)
Site-specific Meteorological (Temperature 8 Wind Speed) Data were used for a Six-year Span from January 1, 1984, to January 1, 1990 A Corresponding Calculation Was Performed and a Curve was Generated which Encompasses a Minimum of 96.1 Percent of all WNP-2 Weather Conditions Data Checked for Reasonableness The Curve Excludes Approximately Four Percent as a Conservative Approximation of 95'/0/5'/o
l l F
DATA USED (6YRS)
Winter Cases Yielded Longer Drawdown Times than Summer Cases Limiting Case Very Conservative Atmospheric Temperature of 0'F, no Wind, Standby Service Water System Spray Pond Temperature 77'F, Division 2 Electrical Power (i.e., Division 1 Not in Operation), One Train of Standby Gas Treatment System in Operation, and 50% Room Cooler Efficiency Case Resulted in a Drawdown Time of 711 Seconds (11.85 Minutes)
J C,
DATA USED (6YRS)
Zero Wind Speed and Zero Degrees Temperature. used Because Temperature Impact on Differential Pressure is Prominent Factor, due to the Higher Differential Temperature Between Inside and Outside Temperatures
( 9
X/Q CALCULATIONS .r V
y4waa ARCON96
- 1. Description of Release Points and Control Room intakes
- 2. Application of ARCON96 at WNP-2
- 3. Comparison with JCO-FAO and Power Uprate
X/Q CALCULATIONS ARCON96 ARCQN96 used to Determine X/Q values for Three Control Room lntakes and Four Release Points Utilized the Same Time Period/plant Specific Data as Used in the Drawdown Analysis CR Intakes Local and Two Remote Intakes
e
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l::.-~
NQ CALCULATIONS Release Points Considered .
Turbine Building Walls Release from Turbine Building Walls to be Used with Events such as CRDAs Reactor Building Walls Release from Reactor Building Walls to be Used During Drawdown Period and Secondary Bypass Leakage Reactor Building Roof line (Stack) Release Vent Release from Reactor Building Roof Used for SGTS Releases Reactor Building (King Kong) Doors - Release from Reactor Building Grade Area
X/Q CALCULATIONS - ARCON96-SITE CONFIGURATION SEE NEXT SLIDE
X/Q Calculations - ARCON96 - Site Configuration Intake ¹1 1589W, 12294 KKDoors Reactor Bldg Roofline 42 fl MSL 1176W, 11778N 1243.5W, 11918N 671 ft MSL 441ft MSL Tttto North 12300 ftN Sito North 12200 57S MSL 12100 Turbine Bldg Wall Sources Turbine Bldg, ht 12000 CRIntake L 1467W, 11883N 528 ft MSL 457 66ShtSL 11900 MS 545 tSL L Reactor Roof Elevation Bldg.
11800 506 542 MSL 474 MSL Reactor Bldg 11700 Wall Sources Ptaat Grado
. 441 MSL Intake ¹2 11600 1025W, 11565 442 ft MSL 11500 1600 1500 1400 1300 200 1100 1000ft W
X/Q CALCULATIONS Application of ARGON96 Analyzed each of the four Release Paths for each of the Three Intakes (i.e., 12 Scenarios)
Analyzed all 12 Scenarios for all 6 Years The Maximum Value of the 6 Years was Chosen for each Time Period (i.e., not just the Maximum Year but the Maximum Value each Time Step)
Local Intake not Used for LOCA due to the Presence of an Automatic Isolation Signal
X/Q CALCULATIONS A.
Application of ARCON96 Between 0 to 3 Hours, no Credit for Operator Action -psed Average of the two Remote lntakes ~> --"~ "
From 3 Hours to the 30 Days, Credit for Operator Action Used the .
Lower of the two X/Qs Calculated for the Remote Intakes~@',z -q",
Very Conservative Treatment With the Presence of two Remote
' """c" Intakes, the Plant will be able to Switch to the Upwind Intake and in Effect Preclude the Introduction of Activity During the Accident
X/Q CALCULATIONS SEE NEXT SLIDE
I 1 NQ Calculations ARCON96 Results and Comparison with Original CR NQs for UFSAR NQ Calculated Used for UFSAR Time Period Ground Release s/m'Q with ARGON 96 s/m'o Analysis Ratio of ARGON 96 Value UFSAR Value 0-2hr 9.94E-5 2.17E-4 45.81%
2-3hr 9.91E-5 5.43E-5 182.50%
3-8hr 7.16E-5 4.49E-5 159.47%
8-24 hr 4.37E-5 3.55E-5 123.10%
1-4da s 2.35E-5 1.67E-5 140.72%
4-30da s ~
1.56E-5 1.67E-5 93 41 SGTS Release 0-2 hr 2.54E-4 3.77E-4 67.37%
2-3 hr 'j.70E-4 9.43E-5 180.28%
3-8 hr 8.15E-5 7.80E-5 104.49%
8-24 hr 3.20E-5 6.17E-5 51.86%
1-4da s 2.26E-5 2.90E-5 77 93%
4-30da s 1.90E-5 2.90E-5 65.52%
DOSE ANALYSIS AND RESULTS Accidents Analyzed for Radiological Consequences AXIDENTCode
- 1. How Applied
- 2. Where used Before Summary of LOCA Major Assumptions Changes from Original Design and Previous Submittal LOCA Results
ACCIDENTS ANALYZED Main Steam Line Break (MSLB)
Fuel Handling Accident (FHA)
Control Rod Drop Accident (CRDA)
Loss Of Coolant Accident (LOCA)
1 1
~f AXIDENT CODE Dose Analysis Code Radiological Consequences of the Spectrum of Design Basis Accidents were Analyzed using the SCIENTECH-NUS AXIDENT Code Code Description - AXIDENT Models the Transport of Radioactivity to the Environment and to the Control Room. This .
Code Includes the time Dependent Effects of Containment Sprays, Recirculation, Purge and Intake Filters, Atmospheric Dispersion, Natural Decay, etc. The Code is Based on the Explicit Solution of the Integrated Activity in a Receptor Volume.
AXIDENT CODE Industry Experience/Usage of the AXIDENTCode Developed in the early 70's to Support the Licensing and Licensing Reviews of both U.S. and international Commercial Nuclear Power Plants. Developed to fill the Void in Codes Available to Assess the Emerging Issue at the time Control Room Habitability.
General lndust Usa e
~ Used to Support Licensing Submittals in the 70's Used for a Number of Plants in Support of the Post-TMI Action Item
~ Used Throughout the 80's and 90's to Resolve Control Room Habitability and Accident Analyses Issues
AXIDENT CODE SCIENTECH Experience Successfully Benchmarked the Previous Results Generated by other Codes which Compute the Integrated/exact Solution (i.e.,
Bechtel's LOCADOSE and S&L's PostDBA)
Over the Years, Thousands of Cases have been run on the AXIDENTCode. In all Cases, the Results have Trended as Expected. The Results have also Consistently been in Close Agreement with the Steady State Murphy/Campe Equations.
S I
AXIDENT CODE Recent Calculations Submitted to the NRC Include:
Control Room and Offsite Dose Calculations to Support the FPC Crystal River Restart CRDA and MSLB Analyses for the CPLL Brunswick Station's Power U prate Effort Reanalysis Effort for the Cooper Station Pending Review
I I I <<e
~ I
' ~
~
i ~
~ ~
I I
'Y/////
~ ~ ~ ~ I I I I r ~ I e
C CONTROL ROOM AIR FLOWS Air intake 1100 cfm 95%
Filter Unit:
eff Elemental and Organic Control Room (CR)
CR Total Volume, 2.0E5 ft'xit Flow.
~1110.55 cfm 99% eff. Particulate Unfiltered Infiltration 8c in~ass/egress 10.55 c6n
MAJOR ASSUMPTIONS )~
AXIDENT Code used for Dose Analysis Source Term Release Per TID-14844 (102% Power Level) 100 /0 Noble Gases
- 25% Halogens (50% ESF Leakage) 91% Elemental 5 % Particulate 4 % Organic Dose Conversion Factors in Accordance with ICRP 30
MAJOR ASSUMPTIONS Instantaneous Mixing in Primary Containment Release from Containment of .5'/oiday Total
.32/o/day Containment Leakage
- .18'/o/day MSIV Leakage No Suppression Pool Scrubbing Credited
MAJOR ASSUMPTIONS SGT Filter Efficiency
-- 99% Efficient for Halogens 0% Efficient for Noble Gases
. SGT. Flow:
5000 cfm Single Train 50 cfm Bypass Leakage Control Room Filter Efficiency 95% Efficient for Elemental and Organic iodine 99% Efficient for Particulate iodine
I 7
MAJOR ASSUMPTIONS Secondary Containment Drawdown Time of 20 Minutes During Drawdown Time:
No SGT Filtration Credited Secondary Containment Leakage at a Rate of 1 Volume/day 40% Mixing in Secondary Containment for Entire Scenario
MAJOR ASSUMPTIONS gl
~P ESF Leakage Into Secondary Containment:
-- 1 gpm
- 10% Flashing Fraction
- 50% Core Iodine Source Term Control Room Unfiltered Inleakage:
10 scfm Ingress/egress
.55 scfm Infiltration Secondary Containment Bypass Leakage of .028%/day (9.4 scfh)
New X/Q Values
DOSE ANALYSIS METHODOLOGY MAJOR CHANGES
' ~
CURRENT PREVIOUS PROPOSED ITEM DESIGN SUBMITTAL DESIGN Drawdown Time 5 Min 20 Min 20 Min Sec. Ctmt Mixing None 40% Vol. 40% Vol.
ESF Leakage None 1 GPM.
SGT Filter Bypass 14 cfm 14 cfm 50 cfm Bypass Leakage .00209 054*' .028 cj~ r4,
(% Day)
LOCA ANALYSIS RESULTS CALC. . LlMIT
~REM ~REM CR Whole Body 0.4 5 GR Thyroid 28.1 30 GR Beta 6.8 30 EAB Whole Body 3.7 25 EAB Thyroid 56.6 300 LPZ Whole Body 3.4 25 LPZ Thyroid 131 300
4 E
E LOCA DOSE BY PATH RELEASE PATH CONTROL ROOM LPZ THYROID W. BODY THYROID W. BODY Sec. Ctmt Bypass 21 .02 1Q1 0.5 Sec. Ctmt Leakage .02 8E-5 0.2 4E-3 Sec. Ctmt SGT Rel. 4.2 .16 1.2 ESF.Leakage Sec.Ctmt 0.3 1E-5 1.2 2E-3 MSIV Leakage 3.0 .18 13 1.7 TOTAL 28.1 04 131 34
CONTAINMENT RELEASE IMPACT
(.04%1day Sec Ctmt Bypass Leakage)
RELEASE PATH CR THYROID DOSE (REM)
.32'/o/DAY .5'/o/DAY Sec Ctmt Bypass Sec Ctmt Leakage .02 .03 Sec Ctmt SGT Rel. 4.2 ESF Leak (Sec Ctmt) 0.3 0.3 MSIV Leakage 3.0 3.0 TOTAL 36.9 39.2
t ~
,0 4 I 0
~
I
EFFECT OF RELEASE ELEVATION DESCRIPTION DOSE REM Ground Release = 100%
Roofline Release = 0/0 39.2 Ground Release = 60'/0 Roofline Release = 40'/0 39.6 Ground Release = 50'/0 Roofline Release = 50'/0 39.7 Ground Release = 0'/0 Roof line Release = 100% 39.2 Based on Thyroid Dose in Control Room Assuming .5'/0/day Containment Leakage and .04%/day Secondary Containment Bypass
EFFECT OF SGT FLOWRATE DESCRIPTION DOSE (REM)
THYROID W.BODY 4K cfm for 30 Days 37.3 .357 5K cfm for 30 Days 36.9 .369 10K cfm for 30 Days A09 (20 Min Drawdn) 10K cfm for 30 Days 37.3 .409 (10 Min Drawdn) 10K cfm for 1 Hr Then 4K .362 cfm 30 Days Based On .32%/day Ctmt Leakage and .04%/day Sec Ctmt Bypass Leakage
SUMMARY
Dose Analysis Meets 10CFR50 and 10CFR100 Limits No Hardware Changes Necessary Beyond Those Completed in-Support of the JCO-FAO Tech Spec Submittal to be Made. in February FSAR Changes will be Implemented Following Approval of Tech Spec Submittal FSAR Changes Include:
-. Accident Analysis and Doses Control Room Habitability Analysis
~ Secondary Containment Description
- Description of SGTS and REA JCO-FAO will be Closed Following Approval of Tech Spec Submittal
le'g.
q RC~O UNITED STATES C~
+4 Cy 4V NUCLEAR REGULATORY COMMISSION 0
0 REGION IV Cy 611 RYAN PLAZA DRIVE, SUITE 400
+n ~O ARLINGTON,TEXAS 76011-8064 NOV l 5 I999 Mr. J. V. Parrish (Mail Drop 1023)
Chief Executive Officer Energy Northwest P.O. Box 968 Richland, Washington 99352-0968
SUBJECT:
INVITATIONTO DISCUSSION OF PILOT PROGRAM RESULTS
Dear Mr. Parrtsh:
On May 30, 1999, we initiated a pilot of the risk-informed baseline inspection program at Cooper Nuclear Station and Fort Calhoun Station. The pilot program is scheduled to end November 27. On December 1, 1999, the NRC will meet with Omaha Public Power District and Nebraska Public Power District to discuss their perceptions and lessons learned gathered during the pilot program. You will find the details of the meeting documented in the enclosed meeting notice.
Although the meeting is between the NRC and members of the pilot plant staffs, we will be discussing topics that may be of interest to your organization. This meeting will be open to public observation. Following the meeting, the discussion will be opened for comments and questions from observers. We would entertain and welcome your participation at that time.
Should you have any questions on this inspection plan, please contact Charles Marschall at
~ (817) 860-8185 or David Loveless at (817) 860-81 61.
Sincerely, gals'@zPdK Charles S. Marschall, Chief Project'Branch C Division of Reactor Projects Docket No.: 50-397 License No.: NPF-21
Enclosure:
As Stated
Energy Northwest cc w/enclosure:
Chairman Energy Facility Site Evaluation Council P.O. Box 43172 Olympia, Washington 98504-3172 Rodney L. Webring (Mail Drop PE08)
Vice President, Operations Support/PIO Energy Northwest P.O. Box 968 Richland, Washington 99352-0968 Greg O. Smith (Mail Drop 927M)
Vice President, Generation Energy Northwest P.O. Box 968 Richland, Washington 99352-0968 D. W. Coleman (Mail Drop PE20)
Manager, Regulatory Affairs Energy Northwest P.O. Box 968 Richland, Washington 99352-0968 Albert E. Mouncer (Mail Drop 1396)
General Counsel Energy Northwest P.O. Box 968 Richland, Washington 99352-0968 Paul Inserra (Mail Drop PE20)
Manager, Licensing .
Energy Northwest P.O. Box 968 Richland, Washington 99352-0968 Perry D. Robinson, Esq.
Winston 8 Strawn 1400 L Street, N.W.
Washington, D.C. 20005-3502 Bob Nichols State Liaison Officer Executive Policy Division Office of the Governor P.O. Box 43113 Olympia, Washington 98504-3113
Energy Northwest bcc to DCD (IE45) bcc distrib. by RIV w enclosure:
Branch Chief (DRP/E)
Project Engineer (DRP/E) bcc distrib. w/o enclosure:
Regional Administrator Resident Inspector DRP Director RIV File DRS Director RITS Coordinator Branch Chief (DRP/TSS)
DOCUMENT NAME: R:QWN2>mn12-1.wpd Torecelvecopyofdocurnent, Indicateinbox:"C" =Cop withoutenclosures "E" =Cop withenclosures "N" ~Nocopy RIV'C:DRP/C CSMarschall;vlh 11/1 5/99 OFFICIAL RECORD COPY
%+8 Recu C~
+ ~ ~o UNITED STATES Cy A
A 0
NUCLEAR REGULATORY COMMISSION Ol c REGION IV Cy 611 RYAN PLAZA DRIVE, SUITE 400
+ gO ARLINGTON, TEXAS 76011-8064
~>>*++
November 15, 1999 NOTICE OF LICENSEE MEETING Name of Licensee: Nebraska Public Power District Omaha Public Power District Name of Facility: Cooper Nuclear Station Fort Calhoun Station Docket: 50-285 50-298 Date and Time December 1, 1999 of Meeting: 1 -3 p.m. (CDT)
Location of Meeting: Fort Calhoun Station Auditorium Hwy. 75 - North of Fort Calhoun Fort Calhoun, Nebraska Purpose of Meeting: Discussion of the licensee's perceptions and lessons learned from the pilot inspection program NRC Attendees: K. Brockman, Director, Division of Reactor Projects C. Marschall, Chief, Branch C, Division of Reactor Projects A. Madison, Transition Task Force Leader, Office of Nuclear Reactor'Regulation W. Jones', Senior Reactor Analyst, Division of Reactor Safety L. Wharton, FCS Project Manager, Office of Nuclear Reactor Regulation L. Burkhart, CNS Project Manager, Office of Nuclear Reactor Regulation D. Loveless, Senior Project Engineer, Branch C, Division of Reactor Projects Licensee Attendees: W. Gary Gates, Vice President, Fort Calhoun Station S. Gambhir, Division Manager, Nuclear Operations J. Chase,.Division Manager, Nuclear Assessments M. Tesar, Division, Manager, Nuclear Support Services R. Phelps, Division Manager, Nuclear Engineering J. Solymossy, Manager, Fort Calhoun Station M. Frans, Manager, Nuclear Licensing R. Jaworski, Manager, Revised Reactor Oversight Project B. Hansher, Supervisor, Station, Licensing
Nebraska Public Power District Omaha Public Power District H. Hackerott, Supervisor, Systems Analysis J. McDonatd, Plant Manager, Cooper Nuclear Station P. Caudill, General Manager, Technical Services G. Smith, Nuclear Projects Manager J. Sumpter, Licensing Supervisor D. Robinson, QA Assessment Manager R. Wachowiak, Risk Management Supervisor NOTE:
(1) This meeting is open to attendance by members of the general public.
(2) NRC personnel, not listed above, that desire to attend this meeting should notify C. S.
Marschall at 817/860-8185 by COB on November 19, 1999.
(3) A comment session will be conducted at the end of the meeting for participation of all Region IV operating reactor licensees.
~n Approved By:
Charles S Marschall, Chief
~
Project Branch C Division of Reactor Projects CC:
G. R. Horn, Senior Vice President of Energy Supply Nebraska Public Power District 1414 15th Street Columbus, Nebraska 68601 John R. McPhail, General Counsel Nebraska Public Power District P.O. Box 499 Columbus, Nebraska 68602-0499 B. L. Houston, Nuclear Licensing and Safety Manager Nebraska Public Power District P.O. Box 98 Brownville, Nebraska 68321
Nebraska Public Power District Omaha Public Power District Dr. William D. Leech Manager - Nuclear MidAmerican Energy-907 Walnut Street P.O. Box 657 Des Moines, Iowa 50303-0657 Ron Stoddard Lincoln Electric System 1040 0 Street P.O. Box 80869 Lincoln, Nebraska 68501-0869 Michael J. Linder, Director Nebraska Department of Environmental Quality P.O. Box 98922 Lincoln, Nebraska 68509-8922 Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 1824 N Street Auburn, Nebraska 68305 Cheryl K. Rogers, Program Manager Nebraska Health and Human Services System Division of Public Health Assurance Consumer Services Section 301, Centennial Mall, South P.O. Box 95007 Lincoln, Nebraska 68509-5007 Ronald A. Kucera, Director of Intergovernmental Cooperation Department of Natural Resources P.O. Box 176 Jefferson City, Missouri 65102 Jerry Uhlmann, Director State Emergency Management Agency P:O. Box 116 Jefferson City, Missouri 65101
Nebraska Public Power District .Omaha Public Power District Vick L. Cooper, Chief Radiation Control Program, RCP Kansas Department of Health and Environment Bureau of Air and Radiation Forbes Field Building 283 Topeka, Kansas 66620 Mark T. Frans, Manager Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.
~
P.O. Box 399 Hwy. 75 - North of Fort Calhoun Fort Calhoun, Nebraska 68023-0399 James W. Chase, Division Manager Nuclear Assessments Fort Calhoun Station P.O. Box 399 Fort Calhoun, Nebraska 68023 J. M. Solymossy, Manager - Fort Calhoun Station Omaha Public Power District Fort Calhoun Station FC-1-1 Plant P.O. Box 399 Hwy. 75 - North of Fort Calhoun Fort Calhoun, Nebraska 68023 Perry D. Robinson, Esq.
Winston 8 Strawn 1400 L. Street, N.W.
Washington, D.C. 20005-3502 Chairman Washington County Board of Supervisors Washington County Courthouse P.O. Box 466 Blair, Nebraska 68008 Cheryl K. Rogers, Program Manager Nebraska Health and-Human Services System Division of Public Health Assurance Consumer Services Section 301 Centennial Mall, South P.O. Box 95007 Lincoln, Nebraska 68509-5007
0 C Nebraska Public Power District Omaha Public Power District bcc:
DMB (IE45)
OEDO RIV Coordinator (16E15)
ORA File RIV File E-Mail To:
( ) NRC Attendees PMNS Mtg Announcement Coordinator PAB DEDR FJM A/D/NRR BWS Acting, ADT/NRR BAB2 Acting ADP/NRR LJB L. Burkhart, Project Manager, NRR LRW R. Wharton, Project Manager, NRR OEMAIL D/OE WCW W. Walker, Senior Resident Inspector JAC J. Clark, Senior Resident Inspector CAH C. Hackney, RSLO GFS G. Sanborn, EO BWH B. Henderson, PAO CJG
'AC1, RA Secretaries LAT, DLF,LJB1 DRP Division CLG, LMB, NLH DRS Division CMS, JAK,NSL DNMS Division PDP Dean Papa LPL P. Longdo DOCUMENT NAME: R:QCNS<CN12-1'MN.DRP and R:QFCS>FC12-1 MN.DRP To receive co of document, Indicate In box: "C" = Co without enclosures "E" = Co with enclosures N" = No co RI V:DR P/C C PAO C RSLO CSMarsch all:vlh BWHenderson CAHackne 11/1 5/99 11/15/99 11/1 5/99 OFFICIAL RECORD COPY
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Subject:
Summary of Significant Meeting with Licensee (Part 50)
Body:
PDR ADOCK 05000397 P Docket: 05000285, Notes: N/A Docket: 05000298, Notes: N/A Docket: 05000397, Notes: N/A Page 1