ML063100647

Amendment, Regarding Elimination of Main Steam Line Radiation

ML063100647
Person / Time
Site:	Duane Arnold
Issue date:	11/15/2006
From:	Richard Ennis NRC/NRR/ADRO/DORL/LPLIII-1
To:	Vanmiddlesworth G Duane Arnold
K. Feintuch, LPL3-1
Shared Package
ML0631000635	List: ML063100647
References
TAC MC8883
Download: ML063100647 (20)

Text

November 15, 2006 Mr. Gary Van Middlesworth Vice President

Duane Arnold Energy Center

3277 DAEC Road

Palo, Iowa 52324-9785

SUBJECT:

DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT REGARDING ELIMINATION OF MAIN STEAM LINE RADIATION MONITOR

TRIP FUNCTION (TAC NO. MC8883)

Dear Mr. Van Middlesworth:

The U.S. Nuclear Regulatory Commission (NRC or the Commission), has issued the enclosed Amendment No. 261 to Facility Operating License No. DPR-49 for the Duane Arnold Energy

Center (DAEC). This amendment revises Technical Specification (TS) Table 3.3.6.1-1, "Primary

Containment Isolation Instrumentation," to elim inate the Main Steamline Radiation Monitor trip function.

The changes to the DAEC TSs are in response to the application by Nuclear Management Company, LLC (NMC, the former licensee) dated November 14, 2005, as supplemented by your letter of September 1, 2006. Amendment No. 260, issued on January 27, 2006, transferred the

DAEC license from NMC to FPL Energy Duane Ar nold, LLC (FPL Energy). By letter dated February 6, 2006, FPL Energy adopted all previous docketed requests before the NRC for

review and approval.

A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.Sincerely,/RA/Richard B. Ennis, Senior Project Manager Plant Licensing Branch III-1

Division of Licensing Project Management

Office of Nuclear Reactor Regulation Docket No. 50-331

Enclosures:

1. Amendment No. 261 to License No. DPR-49

2. Safety Evaluation cc w/encls: See next page

ML063100647Package: ML063100635TS: ML063250039 OFFICELPL3-1/PMLPL3-1/PMLPL3-1/LAEICB/BCIOLB/BCNAMEKFeintuchREnnisTHarrisAHoweNSalgadoDATE11/6/0611/14/0611/6/0611/13/0611/7/06OFFICESBWB/BCAADB/BCOGC/NLOLPL3-1/BCNAMEGCranstonMKotzalasBPooleLRaghavanDATE11/8/0611/7/0611/14/0611/15/06

cc: Mr. J. A. Stall

Senior Vice President, Nuclear and Chief

Nuclear Officer

Florida Power & Light Company

P. O. Box 14000

Juno Beach, FL 33408-0420 Mr. M. S. Ross Managing Attorney

Florida Power & Light Company

P. O. Box 14000

Juno Beach, FL 33408-0420 Mr. R. E. Helfrich Senior Attorney

Florida Power & Light Company

P. O. Box 14000

Juno Beach, FL 33408-0420 Mr. W. E. Webster Vice President, Nuclear Operations

Florida Power & Light Company

P. O. Box 14000

Juno Beach, FL 33408-0420 John Bjorseth Site Director

Duane Arnold Energy Center

3277 DAEC Road

Palo, IA 52324 Steven R. Catron Manager, Regulatory Affairs

Duane Arnold Energy Center

3277 DAEC Road

Palo, IA 52324 U. S. Nuclear Regulatory Commission Resident Inspector's Office

Rural Route #1

Palo, IA 52324 Regional Administrator, Region III U.S. Nuclear Regulatory Commission

Suite 210 2443 Warrenville Road

Lisle, IL 60532-4351 Mr. M. Warner Vice President, Nuclear Operations Support Florida Power & Light Company

P. O. Box 14000

Juno Beach, FL 33408-0420 Mr. D. A. Curtland Plant Manager

Duane Arnold Energy Center

3277 DAEC Rd.

Palo, IA 52324-9785 Mr. R. S. Kundalkar Vice President, Nuclear Engineering

Florida Power & Light Company

P. O. Box 14000

Juno Beach, FL 33408-0420 Daniel K. McGhee Iowa Department of Public Health

Bureau of Radiological Health

321 East 12th Street

Lucas State Office Building, 5th Floor

Des Moines, IA 50319-0075 Chairman, Linn County Board of Supervisors

930 1st Street SW

Cedar Rapids, IA 52404 October 12, 2006 1The application was submitted by Nuclear Management Company, LLC (NMC, the former licensee). Amendment No. 260, issued on January 27, 2006, transferred license DPR-49 fromNMC to FPL Energy Duane Arnold, LLC (FPL Energy, the current licensee). By letter datedFebruary 6, 2006, FPL Energy adopted all previous docketed r equests before the NRC for review and approval.

FPL ENERGY DUANE ARNOLD, LLC DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 261 License No. DPR-49

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Nuclear Management Company, LLC 1 dated November 14, 2005, as supplemented by letter dated September 1, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in

10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the

public, and (ii) that such activities will be conducted in compliance with the

Commission's regulations;D.The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.2.Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility

Operating License No. DPR-49 is hereby amended to read as follows:

(2)Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 261, are hereby incorporat ed in the license. FPL Energy Duane Arnold, LLC shall operate the facility in accordance with the Technical

Specifications.3.This license amendment is effective as of its date of issuance and shall be implemented within 120 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/L. Raghavan, Chief Plant Licensing Branch III-1

Division of Operating Reactor Licensing

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance:

November 15, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 261 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following page of the Facility Operating License No. DPR-49 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERTPage 3Page 3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The changed area is identified by marginal lines.

REMOVE INSERT3.3-573.3-57 Amendment No. 261 2.B.(2)FPL Energy Duane Arnold, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in

accordance with the limitations for storage and amounts required for reactor

operation, as described in the Updated Final Safety Analysis Report, as

supplemented and amended as of June 1992 and as supplemented by letters

dated March 26, 1993, and November 17, 2000.2.B.(3)FPL Energy Duane Arnold, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and

special nuclear material as sealed neutron sources for reactor startup, sealed

sources for reactor instrumentation and radiation monitoring equipment

calibration, and as fission detectors in amounts as required;2.B.(4)FPL Energy Duane Arnold, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated radioactive apparatus

components;2.B.(5)FPL Energy Duane Arnold, LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear materials

as may be produced by the operation of the facility.C.This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I; Part 20, Section 30.34 of Part

30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of

Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional

conditions specified or incorporated below:

Maximum Power Level2.C.(1)FPL Energy Duane Arnold, LLC is authorized to operate the Duane Arnold Energy Center at steady state reactor core power levels not in excess of 1912

megawatts (thermal).(2)Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 261, are hereby incorporat ed in the license. FPL Energy Duane Arnold, LLC shall operate the facility in accordance with the Technical

Specifications.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 261 TO FACI LITY OPERATING LICENSE NO. DPR-49 FPL ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-33

11.0INTRODUCTION

By application dated November 14, 2005, Nuclear Management Company, LLC (NMC, the former licensee), requested changes to the Technical Specific ations (TSs) for the Duane Arnold Energy Center(DAEC). Amendment No. 260, issued on January 27, 2006, transferred the DAEC license (Facility

Operating License No. DPR-49) from NMC to FPL Energy Duane Arnold, LLC (FPL Energy, the current licensee). By letter dated February 6, 2006, FPL E nergy adopted all previous docketed requests beforethe U.S. Nuclear Regulatory Commission (NRC or the Commission) for review and approval. By letter

dated September 1, 2006, FPL Energy provided a supplement pertaining to the requested license amendment. The supplement provided additional informati on that clarified the application, did not expandthe scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards considerati on determination as published in the Federal Register on August 1, 2006 (71 FR 43533).

The proposed amendment would revise DAEC TS Tabl e 3.3.6.1-1, "Primary Containment Isolation Instrumentation," to eliminate the Main Steam Line R adiation Monitor (MSLRM) trip function. Specifically, TS Table 3.3.6.1-1, Function 1.f, "Main Steamline Radiation - High," would be deleted and the Function 1.g, "Turbine Building Temperature - High," would be re-numbered as Function 1.f. The TS Bases would also be changed (under the TS Bases Control Program specified in TS 5.5.10) to reflect the above changes to Table 3.3.6.1-1.

The licensee's application provided the following background information regarding why the change is being requested:

The purpose of the MSLRM is to generate an isolati on signal on conditions of high radiation in the Main Steam Lines (MSL) that are indicative of a Design Basis Accident (DBA) - Control Rod Drop Accident (CRDA). The isolation signal will cause a trip and isolation of the Mechanical Vacuum Pump (MVP), which is used during plant startup to initially establish a vacuum condition in the

main condenser, and will close the MSL drains. In addition, the reactor coolant sample valves in the Main Recirculation System also receive a signal to close. Originally, the MSLRM also generated a Reacto r Protection System [RPS] trip (SCRAM) and aNuclear Steam Supply Shutoff System (NSSSS) trip (Main Steam Isolation Valve (MSIV) closure).

The [boiling-water reactor] BWR Owners' Gr oup (BWROG) developed a licensing topical report[LTR] (NEDO-31400-A, "Safety Evaluation for Elim inating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor," 1991) that justified the removal of the RPS and NSSSS/MSIV trips on the MSLRM high radiation signal. NRC subsequently approved the BWROG LTR. The associated license

amendment for the DAEC was granted as License Amendment # 182 in 1992.

Because not all BWRs have the MVP trip and sample valve isolations on the MSLRM signal, the BWROG topical report did not address these specific isolations. In addition, the topical report did not include the isolation of the MSL drains. T hus, these isolations were not removed during implementation of Amendment # 182 at the DAEC.

Due to obsolescence issues with the current MSLRM instruments, NMC seeks to eliminate this trip function as a cost-beneficial solution to replacement of the current instruments with those of a

newer design. In addition to the immediate cost savings from not having to replace the existing instruments, there is a long-term cost savings from avoided maintenance costs (both corrective

maintenance and surveillance testing/calibrations).

2.0REGULATORY EVALUATION

As discussed in DAEC Updated Final Safety Anal ysis Report (UFSAR) Section 6.2.4, "ContainmentIsolation System," the safety objective of the primary containment system is to provide the capability in

conjunction with other safeguard features, including the c ontainment isolation system to limit the release of fission products in the event of a postulated DBA so that offsite doses are held to a practical minimum, and do not exceed the values set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.67.The current DAEC design and licensing basis for the MSLRM containment isolation function is discussed in UFSAR Sections 7.3.1.1.8, 11.5.1, and 15.2.4. High radiation in the vicinity of the MSLs could indicate a gross release of fission products from the fuel. On indication of such a release, a containment isolation signal is generated. The isolation signal trips and isolat es the MVP, closes the MSL drains, and closes the reactor coolant sample valves in the main recirc ulation system. The MSLRM containment isolationfunction is currently credited in the CRDA analysis in order to reduce the radiological consequences of the accident. In support of the proposed amendmen t, the licensee has reanalyzed the radiological consequences of a CRDA without taking credit for the MSLRM containment isolation function.

Based on a review of UFSAR Section 3.1; NUREG-0800, "Standard Review Plan" (SRP), Sections 6.2.4, 6.4, 13.2.1, 13.2.2, 13.5.2.1, 15.0.1, 15.4.9, and 18.0; and the licensee's application dated November 14, 2005, the NRC staff identified the following regulatory requirements and guidance as being applicable to the proposed amendment:

!10 CFR 50.67, "Accident source term," insofar as it sets standards for radiological consequences of postulated accidents.

!10 CFR Part 50, Appendix A, General Design Criterion (GDC) 13, "Instrumentation and control,"

insofar as it requires that instrumentation be pr ovided to monitor variables and systems over their anticipated ranges for normal operation, for ant icipated operational occurrences, and for accident conditions as appropriate to assure adequate safety.

!10 CFR Part 50, Appendix A, GDC 19, "Control room," insofar as it requires that the control room be maintained in a safe, habitable condition under a ccident conditions by providing adequate protection against radiation.  !10 CFR Part 50, Appendix A, GDC 20, "Protection system functions," insofar as it requires the protection system to be designed to sense accident c onditions and to initiate the operation of systems and components important to safety.

!10 CFR Part 50, Appendix A, GDC 28, "Reactivity limits," insofar as it requires that reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to

assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor

coolant pressure boundary greater than limited local yi elding nor (2) sufficiently disturb the core, itssupport structures, or other reactor pressure vessel internals to impair significantly the capability to

cool the core.

!10 CFR 50.36, "Technical specifications," insofar as it specifies the regulatory requirements related to the content of the TSs. Specifically, 10 CFR 50.36(c)(2)(ii), which requires that a TS limiting condition for operation (LCO) be established for each item meeting one or more of the following criteria:Criterion 1:Installed instrumentation that is used to detect and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Criterion 2:A process variable, design feature, or oper ating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the

integrity of a fission product barrier. Criterion 3:A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or

presents a challenge to the integrity of a fission product barrier. Criterion 4:A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

!NRC Information Notice (IN) 97-78, "Crediting of Oper ator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Res ponse Times," insofar that it provides guidancerelated to evaluating the acceptability of subs tituting manual actions for automatic actions.

!Regulatory Guide (RG) RG 1.183, "Alternative Radi ological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Regulatory Position 4.4.

!RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants."

3.0TECHNICAL EVALUATION

3.1Operator Performance The NRC staff reviewed the proposed amendment with respect to operator performance. As discussed in the licensee's application, under the current DAEC design if a CRDA were to occur during MVP operation, any resulting fission product release from the reac tor would travel through the open MSIVs to the main condenser where the MVP would exhaust them to the pl ant Offgas Stack, resulting in an elevated release until the MSLRM trip automatically isolates this pathway. Under the proposed amendment, tripping the MVP and terminating the elevated release from the Offgas Stack would be a manual action performed by operators.

In the request, the licensee referenced NRC IN 97-78, and stated that it used this guidance to review the following factors to determine whether substituting manual for automatic actions is appropriate: (1) whether there are sufficient alarms or other indi cations that can be used to diagnose the event; (2) whether plant procedures and training exist to gui de the operator response or actions; and (3) the sensitivity of the timing of these actions (includi ng the consequences of any errors in executing those actions).

Alarms and Other Indications

The licensee stated that there are several alternate i ndications of fuel damage and offsite release. For example, the normal range/post-accident range radiati on monitoring system (KAMAN) alarms in the main control room, and the alarm settings, which are in accordance with the Offsite Dose Assessment Manual (ODAM), are well below radiation levels anticipat ed during any analyzed accident. Backup monitoring is provided by the Offgas Stack radiation monitors. T hese monitors alarm in the control room and cause Engineered Safety Feature (ESF) trips on a high r adiation signal. The licensee stated that during aCRDA, the alarm settings would be exceeded quickly (estimated time approximately 2-5 minutes) and prompt the operator action to trip the MVP.

Plant Procedures, Training, and Operator Actions

Regarding procedures, the alarms mentioned above co rrespond to a section in an Annunciator Response Procedure (ARP). The ARPs direct operators to respond to the alarms, including taking actions in associated Abnormal Operating Procedures (AOPs) and O perating Instructions (OIs). The AOP for the alarms discussed above directs the operator to secure the MVP. Training is provided to operators on ARPs, AOPs, and OIs as part of the annual license re-qualification program. In addition, the licensee

states that, if NRC approval for the request is received, operating procedures (ARPs, AOPs, and OIs) will

be consolidated to reflect the removal of the MSLR Ms and to streamline the instructions for manually securing the MVP upon a confirmed high radiation condi tion on the remaining offgas radiation monitors.

Associated training for operators will be conducted.

As discussed in DAEC UFSAR Section 15.0.4, the plant licensing basis is that for the initial response to an accident (first 10 minutes), all actions shall be automatic and require no decision or manipulation of controls by plant operations personnel. Therefor e, credit for manual actions may be taken after 10 minutes. The licensee's supplement dated September 1, 2006, provided a detailed best estimate time line of the sequence of events and actions beginning with the CRDA event and ending with the operator verifying the MVP has been tripped. The time line assumes that the MVP is tripped after 10 minutes.

Securing the MVP involves turning the associated hand swit ch to the "STOP" position. To verify the MVPtrip, the operator will check the indicator light located directly above the hand switch. Since the controls

for securing the MVP are in the main control room, the operator does not need to enter the plant to

perform the action, and no support personnel are needed.

Based on the actions shown in the time line, and the location of the MVP controls, the NRC staff finds that the assumption of 10 minutes is reasonable.

In addition, the NRC staff finds that the 10 minute ti me frame for manual action is consistent with the licensing basis as stated in UFSAR Section 15.0.4.

The licensee also stated that the existing RG 1.97 in strumentation for monitoring offsite releases andassessing core damage will remain adequate for directing emergency response to a CRDA event.

Therefore, no new or upgraded instrumentation is needed to compensate for the removal of the MSLRMs.

Sensitivity and Timing of Operator Actions

The licensee's submittal dated September 1, 2006, stated that the best estimate time line for accomplishing the assumed operator actions in the dose analysis supporting the amendment request is not considered to constitute "time critical operat or actions," because a sensitivity study performed to support the submittal demonstrated that the "no operator action" case (i.e., worst-case operator error) still

achieves acceptable results in terms of dose rate

s. The NRC staff's evaluation of the radiological consequences associated with the proposed amendment is discussed in safety evaluation (SE) Section

3.2. Conclusion for Operator Performance The NRC staff finds that, consistent with the guidance in IN 97-78, the licensee has given adequate consideration to the alarms and other indications available to the operators to diagnose a CRDA, to the procedure changes and associated training needed for im plementation of the change, to the operator actions related to tripping the MVP, and to sensitiv ity and timing of the operator actions. Based on these considerations, the NRC staff concludes that t he proposed amendment is acceptable with respect to operator performance related to substituting the automatic tripping of the MVP with manual operator

actions. 3.2Radiological ConsequencesThe NRC staff performed independent confirmatory dose calculations, for the affected CRDA event, using the NRC-sponsored radiological consequence computer code, "RADTRAD: Simplified Model for RAD ionuclide T ransport and R emoval A nd D ose Estimation," Version 3.03, as described in NUREG/CR-6604. The RADTRAD code, developed by the Sandi a National Laboratories for the NRC, estimates transport and removal of radionuclides and radiologica l consequence doses at selected receptors. The findings of this SE are based on the description of the licensee's analysis and other supporting information docketed by the licensee.

As discussed above in SE Section 2.0, the MSLRM isol ation function (i.e., tripping of MVP and closing of MSL drains and reactor coolant sample valves) is currently credited in the CRDA analysis in order to reduce the radiological consequences of the acci dent. In support of the proposed amendment, thelicensee has reanalyzed the radiological consequences of a CRDA without taking credit for the MSLRMisolation function. Specifically, the licensee's reanalysis assumes that the MSLRM will not automatically trip the MVP, this will be a manual action by the operators. In addition, the analysis assumes that the MSL

drains and reactor coolant sample valves are not isolated following a CRDA. The impact of the proposed amendment on potential DBA radiological c onsequences is assessed and analyzed below.

Control Rod Drop Accident CRDA is the only DBA event affected by the proposed TS change. The key assumptions of the submittedanalysis are listed in Table 1. The licensee's re-analys is of the CRDA event is consistent with the current licensing basis, except for the following:

!release fractions, currently based on draft DG-1081, are based on the final RG 1.183, Appendix C;

!use of the 5 seconds release time, as suggested in the RG 1.183 Section 3.3, instead of the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> release assumed in the current calculation; and

!performing calculations of the control room (CR) and technical support center (TSC) doses (current licensing basis includes an engineering judgement that the CRDA is not a limiting event for the DAEC).These changes, made to the current licensing basis anal ysis, are consistent with the regulatory guidance stated in the RG 1.183, and are, therefore, acceptable.

The radionuclide core inventory used in this re-analysis was previously reviewed and approved by the NRC staff in License Amendment No. 240 (ADAMS Accession No. ML011660142).

The licensee performed a bounding analysis assuming an operator failure to isolate the MVP until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of the MSLRM signal, rather than isolating the system after the 10 minute time frame allowed for manual operator actions (as discussed in SE Section 3.1). The results, presented in Table 2, indicate that the calculated bounding doses are within the regulatory limits. However, the calculated bounding dose at the exclusion area boundary (EAB) of 7.4 rem of total effective dose equivalent (TEDE) exceeds the suggested limit of 6.3 rem.

The suggested EAB limit of 6.3 rem for fuel rod drop accidents, stated in the RG 1.183, was established in SRP 15.4.9 to represent the NRC staff's guidance to be "well within" (defined as 25 percent of) the 10 CFR 50.67 regulatory limit of 25 rem. Since the calculated bounding EAB dose is based on a conservative assumption (as reviewed and accepted by the NRC staff in License Amendment No. 240),

and since there are additional conservatisms embedded in the licensee's analysis, the NRC staff

considers such a deviation as insignificant from the poi nt of view of the intent of the regulatory guidance, and, therefore, is acceptable.

Atmospheric Dispersion

As presented in the November 14, 2005, license amendment request, the licensee generated new atmospheric dispersion factors (/Q - read as "CHI over Q" values) for elevated releases from the plant stack to the CR and TSC. These new /Q values were generated partially based on the guidance in RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants." The licensee also used existing /Q values for postulated elevated releases to the EAB and low-population zone (LPZ) and ground level releases to the CR, TSC, EAB, and LPZ. The

existing /Q values were previously generated by the licensee in support of DAEC License Amendment No. 240.Meteorological Data

All of the /Q values used in the dose assessment for the current license amendment request were generated using the site meteorological data collected from 1997-1999 which the licensee previously

provided in support of DAEC License Amendment No. 240. Based upon the discussion of atmospheric

dispersion factors in the SE associated with Amendment No. 240, the NRC staff has concluded that the

1997-1999 site meteorological database provides an acc eptable basis for making atmospheric dispersion estimates for use in support of this license amendment request.

Control Room Atmospheric Relative Concentrations

The licensee used previously approved /Q values for the CR and TSC dose assessments for postulated ground level releases from the main condenser. These /Q values are discussed in the SE associated with DAEC License Amendment No. 240 (see ML0116601420 dated July 31, 2001).

The licensee generated new CR /Q values for postulated releases from the plant stack to the CR and the TSC partially based on the guidance in RG 1.194 fo r elevated releases. RG 1.194 states that /Q values for postulated elevated releases to the CR should be based on /Q values generated from both the ARCON96 (NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes" and

PAVAN (NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Bases

Accident Releases of Radioactive Material from Nucl ear Power Stations") computer codes. The resulting ARCON96 and PAVAN /Q values are then combined as a function of time period to determine the effective /Q values to be used in the dose assessments.

In generating the effective /Q values used in the dose assessments, the licensee provided new PAVAN/Q values and used previously generated ARCON96 /Q values approved in DAEC Amendment No. 240.

The meteorological input for the new PAVAN /Q values consisted of 1995-1997 joint frequency distributions compiled from wind data measured at t he 10-meter level on the onsite meteorological tower.

The meteorological input to the previous ARCON96 /Q values consisted of 1995-1997 hourly wind speed and direction data from the 10-meter and 45.7-meter levels on the onsite meteorological tower. Stability

class input to both the PAVAN and ARCON96 /Q values was based on delta-temperature measurements made between the 45.7-meter and 10-meter levels on the onsite meteorological tower. The NRC staff qualitatively reviewed the inputs to the new PAVAN calculations and found them generally consistent with site configuration drawings and NRC st aff practice with two exceptions. When calculating/Q values for an elevated release using PAVAN, RG 1.

194 states that calculations should be made for several distances in each wind direction sector with the objective of identifying the maximum /Q value.

The licensee input the actual distances and directions from the stack to the CR and TSC. This would result in /Q values less than the maximum /Q values due to the relative closeness of the DAEC stack to the CR intake and TSC. The second exception regards input of terrain heights. Use of terrain heights is

not applicable to the CR /Q calculations and would result in an increase in /Q values in this scenario due to a reduction in effective release height.

The NRC staff accepts use of the licensee's 0-2 hour, 2-8 hour, and 8-24 hour /Q values presented in Exhibit A of their November 14, 2005 submittal because staff generated /Q values following the guidance in RG 1.194 and found the licensee's /Q values for these time intervals to be conservative. Although they were not used as part of this license amendment request, staff does not find the /Q values for the 1-4 day and 4-30 day /Q values presented in Exhibit A acceptable for use in other design-basis accident control room analyses. The elevated release control room /Q values should be regenerated by fully implementing the guidance in RG 1.194 or using another methodology found acceptable by the NRC staff before being used in any other design-basis control room dose analysis.

On the basis of the discussion presented above, the review described in the SE associated with Amendment No. 240, and a review of the licensee's use of these /Q values in this license amendment request, the NRC staff has concluded that the CR and TSC /Q values presented in Table 3 are acceptable for use in the DBA assessments described in this SE.

EAB and LPZ Atmospheric Dispersion Factors (/Q values)

The licensee used existing ground level and elevated release /Q values that were accepted by the NRC staff in Licensing Amendment No. 240 to evaluate the impact of the DAEC postulated releases to the EAB

and LPZ. Based on the review described in the SE associated with DAEC License Amendment No. 240

and a review of the licensee's use of these /Q values in this license amendment request, the NRC staff has concluded that the EAB and LPZ /Q values presented in Table 4 are acceptable for use in the design basis accident assessments described in this SE.

Summary of Revised Analysis

The licensee re-evaluated the radiological consequenc es resulting from the postulated CRDA event,affected by the proposed TS changes, and concluded that the radiological consequences at the EAB, LPZ, CR and TSC are within the dose criteria provided in 10 CFR 50.67 and accident dose guidelines specified in SRP 15.0.1. The NRC staff's review has found that the licensee used analysis assumptions and inputs

consistent with applicable regulatory guidance i dentified in Section 2.0 of this SE.

Deviations from the current licensing basis were found acceptable to the NRC staff and are discussed in SE Section 3.2, subsection "

Control Rod Drop Accident

." The NRC staff performed independent calculations using the licensee's assumptions. These NRC staff analyses confirmed the licensee'scalculated results. The NRC staff finds that the EAB, LPZ, CR and TSC doses, estimated by the licensee

for the CRDA, presented in Table 2, meet the applicable dose criteria and are, therefore, acceptable.

Conclusion for Radiological Consequences The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of a postulated CRDA event affected by the proposed TS modification. TheNRC staff finds that the licensee used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Sect ion 2.0, with the exceptions noted for RG 1.194 as discussed in SE Section 3.2, subsection "Control Room Atmospheric Relative Concentrations.

" The NRC staff compared the doses estimated by the licensee to the applicable criteria identified in Section 2.0. The NRC staff also finds, with reasonable assurance, that the licensee's estimates of the EAB, LPZ, CR and TSC doses will comply with these criteria, except for the RG 1.183 EAB dose criterion as

discussed in Section 3.2, subsection "Control Rod Drop Accident."

The NRC staff finds reasonable assurance that, as modified by the proposed TS changes, DAEC will continue to provide sufficient safety margins, with adequate defense-in-depth, to address unanticipated

events and to compensate for uncertainties in accident progression and in analysis assumptions and

parameters. Therefore, the proposed license amendment is acceptable with respect to radiological

consequences.

Table 1Selected parameter values and assumptions for CRDAPower level (102%)1950 MWtMVP rated flow rate1800 cfm Condenser volume55,000 ft 3 Condenser leakage1% volume per day CR building volume155,000 ft 3TSC volume 68,300 ft 3CR intake flow rate3150 cfm TSC unfiltered intake flow rate900 cfmNo. of failed fuel rods 1200Melted fuel percentage0.77% Table 2Revised calculated CRDA doses (TEDE)(rem)Receptor Base Case (10 min Delay)

Bounding Case(24 hrs Delay)Regulatory Limit Accident Dose Criteria (from RG 1.183)EAB2.857.425.06.3LPZ1.293.5225.06.3CR0.480.885.05.0TSC0.540.69N/A5.0 Table 3 DAEC Control Room /Q Values (sec/m 3)Control RoomTechnical Support CenterTime PeriodGround LevelElevatedGround LevelElevated0-2 hrs1.48x10

-3 1.68x10-5 2.14x10-3 1.37x10-52-8 hrs1.27x10

-3 3.75x10-7 1.86x10-3 2.16x10-78-24 hrs5.56x10

-4 1.33x10-7 8.44x10-4 8.00x10-7 Table 4 DAEC EAB and LPZ /Q Values (sec/m 3)Time PeriodReceptorGround LevelElevated0-0.5 hrsEAB5.57x10

-4 7.03x10-5 (fumigation)0.5-2 hrsEAB5.57x10

-4 6.95x10-60-0.5 hrsLPZ1.34x10

-4 3.15x10-5 (fumigation)0.5-2 hrsLPZ1.34x10

-4 6.69x10-62-8 hrsLPZ6.43x10

-5 *3.58x10-6 *8-24 hrsLPZ4.46x10

-5 2.61x10-6* The 0-8 hour LPZ /Q values are being conservatively applied to the 2-8 hour time period. 3.3Deletion of MSLRMs The NRC staff reviewed the proposed amendment regardi ng the impact of the deletion of the MSLRMs. Specifically, the NRC staff evaluated whether the proposed change complied with GDCs 13, 20, and 28, and 10 CFR 50.36.

GDC 13 requires, in part, that instrumentation be pr ovided to monitor variables and systems over their anticipated ranges for normal operation, for antici pated operation occurrences, and for accident conditions as appropriate to assure adequate safety. The licensee's application stated that the primary signal used to assure adequate safety (with respect to a CRDA) is the reactor scram on high neutron flux initiated by

the neutron monitoring system, which is not affected by the subject application. This statement isconsistent with the DAEC accident analysis for a CRDA as discussed in UFSAR Section 15.2.4 which

shows that, for the CRDA sequence of events, the average power range monitor 120 percent power signal scrams the reactor and the scram terminates the accident. UFSAR Section 15.2.4 also indicates that the

MSLRMs trip and isolation function reduces the r adiological consequences of a CRDA. However, based on the discussion above in SE Sections 3.1 and 3.2, r eplacing the automatic actions performed by the MSLRMs with manual operator actions is acceptabl e with respect to maintaining the radiologicalconsequences within regulatory limits. The NRC staff finds that the MSLRMs are not required to assure adequate safety in mitigating a CRDA based on the scram function performed by the neutron monitoringsystem. Therefore, the NRC staff concludes that the proposed deletion of the MSLRMs does not affect DAEC's conformance with the intent of GDC 13.

GDC 20 requires, in part, that the protection sy stem be designed to sense accident conditions and to initiate the operation of systems and components important to safety. The licensee's application stated

that the MSLRMs and their associated trip func tion are not needed to mitigate the consequences of an accident, and thus, are no longer systems and components important to safety. The licensee also

reiterated that the neutron monitoring system trip on high neutron flux is the primary safety actionnecessary to mitigate a CRDA. The NRC staff finds that, for a CRDA, the neutron monitoring system is

sufficient to meet the intent of GDC 20. Therefore, the NRC staff concludes that the proposed deletion of

the MSLRMs does not affect DAEC's conformance with the intent of GDC 20.

GDC 28 requires, in part, that reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local

yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internalsto impair significantly the capability to cool the core. As discussed in UFSAR Section 15.2.4, the initiator

for a CRDA is when a control blade becomes decoupled from its control rod drive mechanism, and sticks inside the core at the fully-inserted position, such that it does not follow its drive when withdrawn from the

core. Later in the startup sequence, when the stuck control rod is at its maximum possible control rod

worth, it breaks free and drops at its maximum velocity , causing a prompt, super critical reactivity event.

The licensee's application stated that design features will be retained that will limit the consequences of a CRDA. Specifically, the mechanical design of the c ontrol blades for General Electric BWRs includes a velocity limiter. This device is located on the bo ttom of the control blade and is designed to provide flow resistance such that, should the blade come uncoupled from its drive mechanism, the blade falls slowly out of the core (i.e., limits rate of reactivity additi on). The licensee's application also stated that the rod worth minimizer and its associated banked position wit hdrawal sequence rod pattern controls, would be retained in the TSs to limit individual control rod worths below those used in the CRDA analysis. In consideration of the design of the control blades , and the rod worth minimizer TSs, the NRC staff concludes that the intent of GDC 28 is satisfied wit h respect to the CRDA without consideration of the trip signals received from the MSLRMs.

As discussed in SE Section 2.0, 10 CFR 50.36(c)(2)(ii), which requires that a TS LCO be established for each item meeting one or more of the following criteria: Criterion 1:Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Criterion 2:A process variable, design feature, or oper ating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the

integrity of a fission product barrier. Criterion 3:A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or

presents a challenge to the integrity of a fission product barrier. Criterion 4:A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

With respect to Criterion 1, the MSLRMs are not used to detect leakage in the reactor coolant pressure boundary. As such, the MSLRMs do not satisfy Criterion 1.

With respect to Criterion 2, the MSLRM trip function is not an initial condition assumed in the CRDA analysis. As such the MSLRMs do not satisfy Criterion 2.

With respect to Criterion 3, as discussed in SE Section 3.2, the MSLRM trip is not necessary to ensure that the radiological consequences of a CRDA remain within regulatory limits. As such, the MSLRMs do not satisfy Criterion 3.

With respect to Criterion 4, the licensee's applicat ion provided a qualitative risk assessment. The licensee stated that various existing TS provisions that prec lude this event are being maintained to ensure that theprobability of a CRDA is maintained at a very low value (e.g., control rod coupling checks in TS 3.1.3.5, stuck rod provision in TS 3.1.3, "slow" control rod s eparation criteria in TS 3.1.4). The licensee also stated that the MSLRM trip does not prevent a CRDA; its sole function is in response to the event after it occurs.

As such, the elimination of the MSLRMs will not increase the risk of core damage from a CRDA event. In

addition, since the elimination of the MSLRM trip function would not result in unacceptable dose consequences, the large early release frequency woul d not be increased. Thus, the licensee concludedthat the overall risk of a CRDA event would remain extremely low after removal of the MSLRM trip

function. The NRC staff concurs with the licensee's assessment that the MSLRM trip function is not

significant to public health and safety. As such, the MSLRMs do not satisfy Criterion 4.

The NRC staff concludes that since none of the four criteria in 10 CFR 50.36(c)(2)(ii) are satisfied, a TS LCO for the MSLRM trip function is not required.3.4Technical Evaluation Conclusion Based on the considerations discussed in SE Sections 3.1 through 3.3, the NRC staff finds the proposed changes to TS Table 3.3.6.1-1 to be acceptable.

The licensee stated that the TS Bases would also be changed (under the TS Bases Control Program specified in TS 5.5.10) to reflect the changes to Table 3.3.6.1-1. The NRC staff considered the proposedTS Bases changes as information only. The NRC staff did not review or make a finding with respect to these changes.

4.0STATE CONSULTATION

In accordance with the Commission's regulations, the Iowa State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0ENVIRONMENTAL CONSIDERATION

SThe amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the

amendment involves no significant increase in am ounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consi deration, and there has been no public comment on suchfinding (71 FR 43533). Accordingly, the amendment meets the eligibility criteria for categorical exclusion

set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or

environmental assessment need be prepared in connection with the issuance of the amendment.

6.0CONCLUSION

The Commission has concluded, based on the considerati ons discussed in this document, that: (1) thereis reasonable assurance that the health and safety of the public will not be endangered by operation in the

proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations,and (3) the issuance of the amendment will not be inimical to the common defense and security or to the

health and safety of the public.

7.0REFERENCES

1.Van Middlesworth, G., Duane Arnold Energy C enter, "Technical Specifications Change Request:

'Elimination of Main Steam Line Radiation Moni tor Trips," letter to U.S. Nuclear Regulatory Commission, November 14, 2005. ADAMS Accession No. ML053260438.2.Van Middlesworth, G., Duane Arnold Energy Cent er, "Disposition of Pending NRC Actions," letter to U.S. Nuclear Regulatory Commission, February 6, 2006. ADAMS Accession No.

ML060530621.3.Van Middlesworth, G., Duane Arnold Energy Center, "Response to Request for Additional Information Related to the Proposed Amendment Requesting Elimination of Main Steamline

Radiation Monitor Trips (TAC NO. MC8883),"

letter to U.S. Nuclear Regulatory Commission, September 1, 2006. ADAMS Accession No. ML062890143.4.General Electric Company, "Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line

Radiation Monitor," NEDO-31400-A, October 1992. 5.U.S. Nuclear Regulatory Commission, "Safety Evaluation by the Office of Nuclear ReactorRegulation Related to Amendment No. 182 to Facility Operating License No. DPR-049, Iowa

Electric Light and Power Company, Central Io wa Power Cooperative, Corn Belt Power Cooperative, Duane Arnold Energy Center, Docket No. 50-331," March 24, 1992.

Principal Contributors: J. Cai B. Parks

A. Drozd

L. Brown

K. Feintuch

R. Ennis Date: November 15, 2006