ML040720726

3/11/04 Callaway Plant, Unit #1 - Issuance of Amd. 159 Main Feedwater/Auxiliary Feedwater Modification and Steam Generator Tube Rupture Re-Analysis (Tacs. MB9875/MB9876)

ML040720726
Person / Time
Site:	Callaway
Issue date:	03/11/2004
From:	Donohew J NRC/NRR/DLPM/LPD4
To:	Randolph G Union Electric Co
Donohew J N, NRR/DLPM,415-1307
Shared Package
ML040720789	List: ML040720726 ML040750310
References
TAC MB9875, TAC MB9876
Download: ML040720726 (31)

Text

March 11, 2004 Mr. Garry L. Randolph Vice President and Chief Nuclear Officer Union Electric Company Post Office Box 620 Fulton, MO 65251

SUBJECT:

CALLAWAY PLANT, UNIT 1 - ISSUANCE OF AMENDMENT RE: MAIN FEEDWATER/AUXILIARY FEEDWATER MODIFICATION AND STEAM GENERATOR TUBE RUPTURE RE-ANALYSIS (TAC NOS. MB9875 AND MB9876)

Dear Mr. Randolph:

The Commission has issued the enclosed Amendment No. 159 to Facility Operating License No. NPF-30 for the Callaway Plant, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated June 27, 2003 (ULNRC-04592), as supplemented by letter dated December 12, 2003 (ULNRC-04928).

The amendment (1) revises the definition of dose equivalent radioiodine 131 (I-131), and (2) increases the maximum allowed closure time of each main feedwater isolation valve (MFIV) from 5 seconds to 15 seconds. A plant modification would replace the electro-hydraulic MFIV actuators with system-medium actuators to improve MFIV reliability and reduce maintenance requirements, and the MFIV stroke time would be increased. A plant modification would also replace swing check valves in each auxiliary feedwater (AFW) motor-driven pump discharge line with an automatic recirculation control check valve to reduce the potential for vibration and increase AFW flow margin. The NRC also approves the re-analysis of the steam generator tube rupture with overfill accident submitted in the application.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.

Sincerely,

/RA/

Jack Donohew, Senior Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosures:

1. Amendment No. 159 to NPF-30

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

March 11, 2004 Mr. Garry L. Randolph Vice President and Chief Nuclear Officer Union Electric Company Post Office Box 620 Fulton, MO 65251

SUBJECT:

CALLAWAY PLANT, UNIT 1 - ISSUANCE OF AMENDMENT RE: MAIN FEEDWATER/AUXILIARY FEEDWATER MODIFICATION AND STEAM GENERATOR TUBE RUPTURE RE-ANALYSIS (TAC NOS. MB9875 AND MB9876)

Dear Mr. Randolph:

The Commission has issued the enclosed Amendment No. 159 to Facility Operating License No. NPF-30 for the Callaway Plant, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated June 27, 2003, (ULNRC-04592), as supplemented by letter dated December 12, 2003 (ULNRC-04928).

The amendment (1) revises the definition of dose equivalent radioiodine 131 (I-131), and (2) increases the maximum allowed closure time of each main feedwater isolation valve (MFIV) from 5 seconds to 15 seconds. A plant modification would replace the electro-hydraulic MFIV actuators with system-medium actuators to improve MFIV reliability and reduce maintenance requirements, and the MFIV stroke time would be increased. A plant modification would also replace swing check valves in each auxiliary feedwater (AFW) motor-driven pump discharge line with an automatic recirculation control check valve to reduce the potential for vibration and increase AFW flow margin. The NRC also approves the re-analysis of the steam generator tube rupture with overfill accident submitted in the application.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions next biweekly Federal Register notice.

Sincerely,

/RA/

Jack Donohew, Senior Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-483 DISTRIBUTION:

PUBLIC GHill (2)

Enclosures:

1. Amendment No. 159 to NPF-30 PDIV-2 Reading TBoyce

2. Safety Evaluation RidsNrrDlpmPdiv (HBerkow) RDennig RidsNrrPMJDonohew DSolorio cc w/encls: See next page RidsNrrLaEPeyton JUhle RidsOgcRp DTrimble RidsACRSACNWMailCenter RidsRegion4MailCenter (D. Graves)

• See memos with the concurrence date from the branches to Projects TS: ML040750310 NRR-100 PKG.: ML040720789 ACCESSION NO.: ML040720726 NRR-058 OFFICE PDIV-2/PM PDIV-2/LA IROB/SC SRXB/SC SPSB/SC SPLB/SC OGC PDIV-2/SC NAME JDonohew EPeyton DTrimble* JUhle* RDennig* DSolorio* MBupp RGramm for SDembek DATE 3/1/04 2/27/04 01/27/04 02/05/04 01/20/04 01/06/04 3/9/04 3/10/04 DOCUMENT NAME: C:\ORPCheckout\FileNET\ML040720726.wpd OFFICIAL RECORD COPY

UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 159 License No. NPF-30

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Union Electric Company (UE, the licensee) dated June 27, 2003, as supplemented by letter dated December 12, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions 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 Commissions 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; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions 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. NPF-30 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 159 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Also, by Amendment No. 159, the license is amended to authorize revision of the Final Safety Analysis Report (FSAR), as set forth in the application for amendment by Union Electric Company dated June 27, 2003 and supplement dated December 12, 2003.

Union Electric Company shall update the FSAR to reflect the reanalysis of the steam generator tube rupture accident with overfill, as described in the licensees letters dated June 27 and December 12, 2003, and the NRC staffs safety evaluation dated March 11, 2004, in accordance with 10 CFR 50.71(e).

3. This amendment is effective as of its date of issuance, and shall be implemented prior to the entry into Mode 3 in the restart of the Callaway Plant from Refueling Outage (RO) 13, which is scheduled for April 2004.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA by Robert A. Gramm for/

Stephen Dembek, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 11, 2004

ATTACHMENT TO LICENSE AMENDMENT NO. 159 FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

REMOVE INSERT 1.1-2 1.1-2 3.7-8 3.7-8

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 159 TO FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By application dated June 27, 2003, as supplemented by letter dated December 12, 2003, Union Electric Company (the licensee) requested changes to the Technical Specifications (TSs, Appendix A to Facility Operating License No. NPF-30) for the Callaway Plant, Unit 1 (Callaway).

The amendment would revise the TSs in two parts. It would (1) revise the definition of dose equivalent radioiodine 131 (I-131) by adding the phrase "or those derived from the data provided in International Commission on Radiological Protection Publication 30, Limits for Intakes of Radionuclides by Workers, 1979" to the current definition, and (2) increase the maximum allowed closure time of each main feedwater isolation valve (MFIV) from 5 seconds to 15 seconds in Surveillance Requirement (SR) 3.7.3.1. A plant modification would replace the electro-hydraulic MFIV actuators with system-medium actuators to improve MFIV reliability and reduce maintenance requirements. The MFIV stroke time would be increased.

In the application, the licensee also described a plant modification to replace swing check valves in each auxiliary feedwater (AFW) motor-driven pump discharge line with an automatic recirculation control (ARC) check valve to reduce the potential for vibration and increase AFW flow margin, and a re-analysis of the steam generator tube rupture (SGTR) with overfill accident. This plant modification and the re-analysis does not require any changes to any requirements in the TSs.

The licensees description of the proposed changes, technical analysis, and regulatory analysis in support of its proposed license amendment are given in Sections 2.0, 4.0 and 5.2, respectively, of the licensees application. The detailed evaluation below will support the conclusion that: (1) there is 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 Commissions 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.

The additional information provided in the supplemental letter dated December 12, 2003, does not expand the scope of the application as noticed and does not change the staffs original proposed no significant hazards consideration determination published in the Federal Register on July 22, 2003 (68 FR 43394).

2.0 REGULATORY EVALUATION

In the main feedwater (MFW) system, the modification would replace the electro-hydraulic actuators currently installed on the MFIVs with system-medium actuators. In the AFW system, the modification would replace the existing swing check valve in each AFW motor-driven pump discharge line with an ARC check valve. The SGTR with overfill accident is re-analyzed because of the plant modifications.

The regulatory requirements for which the staff reviewed the application against are the following:

 General Design Criteria (GDC) 16, "Containment design," requires that the containment and associated systems be designed to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded as long as postulated accident conditions require. The design pressure of the Callaway containment is 60 psig and a design temperature is 320EF (FSAR Table 6.2.1-2). For Callaway, the associated systems referred to in GDC 16 are the containment spray system and the containment fan coolers.

 GDC 34, "Residual Heat Removal." The single failure criteria of GDC 34 requires that the safety functions of the feedwater systems can be performed assuming a single active component failure coincident with the loss of offsite power. GDC 34 also requires the feedwater system through suitable interconnections to provide a path for the addition of AFW for reactor cooldown under emergency shutdown conditions.

 GDC 38, "Containment heat removal," requires that these systems rapidly reduce the containment pressure and temperature following any loss-of-coolant accident (LOCA) and maintain them at acceptably low levels. This is verified by the accident analyses.

 GDC 50, "Containment design basis," requires that the containment be designed so that the containment structure can accommodate, without exceeding its design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA.

 The accident dose guidelines in 10 CFR 100.11, as supplemented by accident-specific criteria in Standard Review Plan (SRP) Section 15.6.3, "Radiological Consequences of Steam Generator Tube Rupture (PWR)," and GDC 19, "Control Room," as supplemented by SRP Section 6.4, "Control Room Habitability Systems."

In its application, the licensee stated that although the iodine spiking methodology of Regulatory Guide (RG) 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," had been approved by the NRC for use by other licensees, the use of a source term methodology at Callaway required a license amendment request pursuant to 10 CFR 50.67. The NRC staff does not agree with this conclusion. The pertinent requirement is 10 CFR 50.67(b) which states in part: "A licensee

who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under §50.90." The term "source term" is defined in 10 CFR 50.2 to be:

Source term refers to the magnitude and mix of the radionuclides released from the fuel, expressed as fractions of the fission product inventory in the fuel, as well as their physical and chemical form, and the timing of their release.

The iodine spike model addresses the phenomenon in which the concentration of iodine isotopes in the reactor coolant increase following a plant transient. This phenomenon is typically understood to be a leeching of iodine salts that were deposited on the inside surface of the fuel rod cladding. The iodine spiking methodology does not change the magnitude and mix of radionuclides, as defined, since the iodine spike is not expressed as fractions of the fission product inventory in the fuel. The staffs intent in publishing the 10 CFR 50.67 rule is explained in the Statements of Consideration at 64 Federal Register 71990, dated December 23, 1999 (see especially staffs response to public comment No. 7 on page 71995). The staff has concluded that the licensees proposed change to the iodine spiking methodology is not subject to the requirements of 10 CFR 50.67 and, therefore, this amendment request was not reviewed against these requirements.

The licensee also requested NRC approval to use RG 1.195 for other licensing basis dose applications. The staff believes that the authority for changes in analysis methodology is available in the Commissions regulations at 10 CFR 50.59, and the supporting guidance in RG 1.187, "Guidance for Implementation of 10 CFR 50.59 Changes, Tests, and Experiments."

Advance blanket approval for the use of RG 1.195 is unwarranted. The staff bases this decision upon the availability of the change process of 10 CFR 50.59, the wide scope of RG 1.195, its interfaces with RG 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors," and guide provisions that are applicable on a case-by-case basis only. This is addressed in Section 3.2 of this safety evaluation.

3.0 TECHNICAL EVALUATION

The licensee has proposed the following changes to the TSs:

1. Revise the definition of dose equivalent radioiodine 131 (I-131) in TS Section 1.1 by adding the phrase "or those derived from the data provided in International Commission on Radiological Protection Publication 30, 'Limits for Intakes of Radionuclides by Workers,' 1979" to the current definition.

2. Increase the maximum allowed closure time of each MFIV from 5 seconds to 15 seconds in SR 3.7.3.1, which requires the verification that the closure time of each MFIV is less than or equal to the specified time in seconds. The proposed specified closure time is 15 seconds. The frequency specified for performing SR 3.3.7.1 not being changed.

The amendment addresses the impact of (1) plant modifications of components in the MFW and AFW systems, and (2) a re-analysis of the SGTR with overfill accident. The plant modifications will replace the electro-hydraulic MFIV actuators with system-medium actuators to improve MFIV reliability and reduce maintenance requirements. The MFIV stroke time would be increased which will reduce the magnitude of the feedwater system pressure transients.

Because this modification will result in a longer closure time than required by SR 3.7.3.1, the licensee has proposed to increase the time to 15 seconds.

Another plant modification would replace swing check valves in each AFW motor-driven pump discharge line with an ARC check valve to reduce the potential for vibration and increase AFW flow margin to the steam generators. The licensee stated that this modification was found acceptable under the provisions of 10 CFR 50.59. It did not require any changes to the TSs.

The two modifications discussed above will be done by the licensee in the next refueling outage (RO) for Callaway, RO 13, which is scheduled for April 2004.

The re-analysis of the SGTR with overfill accident will change the license, and the licensing basis for Callaway, in that (1) the definition of dose equivalent I-131 in TS Section 1.1 had to be revised for the dose consequences of the accident to be found acceptable, and (2) a condition on the license is imposed through this amendment that requires the licensee to incorporate the re-analysis of SGTR with overfill, which is in the licensees letters dated June 27 and December 12, 2003, letters, into the Final Safety Analysis Report (FSAR) during the next licensing document regulatory update. The licensee stated, in Attachment 7 of its application, that this is within six months after the end of RO 13.

The technical evaluation of the amendment is broken down into the reviews of (1) the plant modifications and (2) the SGTR with overfill re-analysis. These reviews are described below.

In the NRC staffs review of the application, requests for additional information (RAIs) were sent to the licensee. The licensee provided draft responses to the questions and discussed these responses with the staff in the meeting of November 12, 2003. The draft responses were provided to the staff before the meeting. The summary of the meeting was issued on December 4, 2003, and the staffs RAIs and the licensees draft responses were attached to the meeting summary and placed on the Callaway docket (Accession No. ML033240668). The licensee then provided its formal responses to the staffs RAIs, with any revisions due to the discussions held in the November 12, 2003, meeting, in its letter dated December 12, 2003.

3.1 Plant Modifications The plant modifications are the replacement of (1) the electro-hydraulic MFIV actuators with system-medium actuators, and (2) the swing check valves in each AFW motor-driven pump discharge line with an ARC check valve. The modifications were reviewed by the NRC staff in the following areas:

 Plant systems,

 Containment systems, and

 Radiological dose consequences.

These areas are addressed below.

3.1.1 Plant Systems 3.1.1.1 MFIV Actuator Replacement The licensee is proposing replacement of the existing MFIV actuators in order to correct problems that the licensee has had with them. In its application, the licensee explained that the existing actuators have a poor maintenance history which includes leaks that have resulted in loss of generation capacity, operational delays, and increased personnel exposure to a hazardous material associated with the electro-hydraulic actuators.

The licensee stated that, to accommodate the new actuators, it is proposed that the existing valve bonnets and yokes be removed and replaced with new bonnets. This component will house the new system-medium (process fluid) actuated design. The new actuators are steam pistons with the piston shaft attached directly to the valve stem. They are operated by process fluid to close the valve and either process fluid or instrument air to open the valve. The safety function is to close the valve. Pressure from process fluid acts upon the actuator piston to position the valve to the desired position. The process fluid is directed to the actuator piston chamber by solenoid valves. Each MFIV has a redundant train of solenoid valves, i.e., two trains that are each powered by a safety-related source. Each MFIV has three solenoid valves per train with a total of six solenoids per valve.

The staff requested additional information that was submitted in the licensees letter dated December 12, 2003. Specifically, the staff inquired as to whether the licensee had performed a failure analysis of the design and if sufficient process fluid pressure would exist to actuate the MFIVs closed in the event of an MFW line break. The closed MFIV is needed to act as a pressure boundary for AFW injection and is also relied upon in containment pressurization analysis for a main steamline break (MSLB) accident inside containment.

The licensees response, in the letter dated December 12, 2003, contained a detailed failure analysis for the MFIV actuator replacement and an analysis of the ability of the MFIVs to close coincident with an MFW line break. The analysis evaluated the single train and common mode outcome of all possible failure cases. The staff reviewed the licensees failure analysis and found it to be thorough. In each case it was demonstrated that the single failure did not prohibit the actuator from performing its safety function. Where appropriate, the licensee confirmed the performance of the components through testing and discussion with the valve manufacturer.

Accordingly, the staff finds the licensees single failure analysis to be thorough and that it substantiates the licensing basis of the actuators delineated in GDC 34.

The staff also reviewed the licensees response to the question as to whether sufficient process fluid pressure would be sustained in order to actuate the MFIVs closed in the event of a MFW line break. The licensees response indicated that following the assembly of each MFIV system-medium actuator, a hot functional test was performed using Callaways spare MFIV body. The testing showed that the MFIV would close within a time frame that would not impact the plants primary side heatup analyses, with as little as 0 psig of secondary side system pressure. The staff also met with the licensee in a meeting held on November 12, 2003, to

discuss the proposed modification. In this meeting, the staff inquired as to how the valves could actuate closed with no system pressure. The licensee indicated that the dead weight of the vertically mounted actuator piston and stem, which are massive components, would actuate the valve closed as long as the lower piston chamber was vented. The lower piston chamber is vented to the condenser via redundant pathways. These pathways were evaluated in the licensees failure analysis described above and found to meet single failure criteria. Therefore, the NRC staff concludes that the actuators are capable of closing the MFIVs in the event of system depressurization due to a MFW line break.

3.1.1.2 AFW Pump Discharge Check Valve Replacement The AFW system employs two motor-driven pumps (MDAFWPs) and one steam turbine driven pump (SDAFWP). Each MDAFWP discharges through a check valve and a locked-open isolation valve to feed two steam generators. The existing pump discharge check valves are swing style check valves. The licensee stated that plant experience with these valves indicates they have contributed to high vibration on both trains of MDAFWP piping systems in instances where leaking MDAFWP discharge flow control valves have been shown to be the direct causal effect of the vibration. Also, the licensee stated that in certain AFW system lineups, use of the discharge swing style check valves contributes to system hydraulic instability. As part of the corrective action for this instability, the licensee reworked the discharge flow control valves to eliminate leakage and the direct cause of vibration. Further corrective action to minimize piping vibration and system instability is to replace the existing swing check valve in each MDAFWP discharge line with an ARC valve.

The design of the ARC valve is such that all flow sensing, bypass pressure reduction, reverse flow protection and modulating recirculating flow is performed in the integral three port valve.

The staff discussed operation of the ARC valves with the licensee during the meeting at NRC headquarters on November 12, 2003. The ARC valves will modulate until main process flow demand exceeds minimum recirculation flow. As the process flow demand increases beyond minimum flow, the bypass port will close and recirculation flow will decrease. This feature effectively increases the AFW flow margin to the steam generators.

In response to the RAI from the NRC staff, the licensee considered failure analysis for the modification to replace the MDAFWP discharge check valves with ARC valves. The only change reported between the existing system with a swing style check valve plus a recirculation line orifice, and the replacement ARC valve, is the interaction between the two performed by the armature in the replacement ARC valve. The licensee stated that in the unlikely event that the armature that connects recirculation flow control and the lifting check disc were to fail (break), the result would be that the recirculation line of the ARC valve would fail open. When the valve fails to the open position, the system effectively returns to the current configuration installed at Callaway, with flow through the recirculation line being continuous, but restricted by the bypass pressure reducer orifice of the ARC valve. Based on this, the NRC staff agrees with the licensees assessment that the ARC valves meet single failure criteria in accordance with GDC 34, and are a suitable replacement for the existing swing check valves.

3.1.2 Containment Systems 3.1.2.1 Containment Technical Findings In its application, the licensee is proposing two design changes to Callway. The first change is the replacement of electro-hydraulic actuators on the MFIVs with system-medium (feedwater) actuators. This change increases the stroke time of the MFIVs from five to fifteen seconds.

The consequence of this change is an additional mass of feedwater to the steam generator associated with the ruptured steam line and hence the potential to increase the containment pressure.

The second design change is the replacement of the existing swing check valves in each MDAFWP discharge line with an ARC valve. The new ARC valves maintain minimum flow requirements for the respective pump, but as pump flow increases, the circulation line of the ARC automatically closes. The flow that is no longer recirculated increases the flow to the steam generators, in particular, the steam generator with the ruptured main steam line. This increased flow also has the potential to overpressurize the containment and possibly to raise the containment temperature.

In order to verify that these changes will not result in exceeding the containment design pressure and temperature, the licensee has evaluated the effect of these proposed changes on the LOCA and MSLB accident mass and energy release analyses and reanalyzed the containment response to a postulated design basis MSLB inside containment. The MFW line break inside containment was also considered by the NRC staff, and it is addressed in Section 3.1.2.6.

3.1.2.2 Large Break LOCA Mass and Energy Release Analysis The licensee states that the short-term and long-term LOCA mass and energy release analyses do not credit nor model MFIV valve closure. An instantaneous isolation is assumed. Such a closure would occur on a safety injection Phase A signal (see TS Table 3.3-3). A faster isolation results in increased heating of the mass discharged to the containment for a cold leg pump suction or pump discharge break. Therefore, the NRC staff concludes that the proposed increased stroke time of 15 seconds for MFIV valve closure is bounded by the assumptions of the current analysis.

The large break LOCA analysis models the AFW flow. The increase in AFW flow due to the ARC valves will cool the secondary side (steam generator water) and reduce the amount of energy released. Therefore, the NRC staff concludes that the licensees current analysis is bounding with respect to the replacement of the current AFW swing check valves with the ARC valves.

3.1.2.3 Large Break LOCA Containment Analysis Since the current mass and energy analysis for the large break LOCA within containment bounds the effects of the proposed changes, the NRC staff concludes that no revised containment analysis by the licensee for the large-break LOCA is necessary.

3.1.2.4 Small Break LOCA Analysis The small-break LOCA analysis only credits the turbine-driven auxiliary feedwater pump (TDAFWP) that provides AFW to all four steam generators. Because the AFW flow is not affected by the type of valve in each pump discharge line for the small-break LOCA, the NRC staff concludes that the ARC valve modification does not affect this analysis.

3.1.2.5 MSLB Mass and Energy Release Analysis 3.1.2.5.1 MSLB Mass and Energy Release Outside Containment In its application, the licensee states that for the MSLB outside containment the analysis assumes MFW isolation coincident with the reactor trip with no delays associated with instrumentation or valve stroke time. This assumption is conservative for environmental qualification considerations since this results in more limiting main steam line tunnel pressures and temperatures. For dose analysis, the timing of MFW isolation is not important since added feedwater to the steam generator would not increase the released dose.

3.1.2.5.2 MSLB Mass and Energy Release Inside Containment The licensee has used the methods in WCAP-88221 to determine the mass and energy release added to the containment as a consequence of a postulated MSLB inside containment. WCAP 8822 was approved by the NRC in its letter dated May 27, 1986. Using these methods, the mass and energy release is not dependent on the timing of MFIV closure but the increased mass due to the increased closure time is accounted for. This is explained in the topical report and in the licensees December 12, 2003, letter. WCAP-8822 is used routinely by licensees with Westinghouse-designed reactors, such as Callaway, and these applications have been approved by the NRC.

WCAP-8822 states that the methods described therein apply to Westinghouse Model D steam generators, and, "with minor alterations," to Model 51 steam generators. Callaways steam generators are Model F. The licensee explained in the December 12, 2003, letter that the reanalyses did not involve any change to the current calculation methods. The Callaway FSAR references methods applicable to Model F steam generators based on WCAP 8822.

Since the calculation methods in WCAP-8822 for the mass and energy release inside containment from the MSLB have been previously approved by the NRC and are applicable to Callaway, the NRC staff concludes that these methods are acceptable for the proposed technical specification changes.

3.1.2.5.3 MSLB Inside Containment - Containment Analysis Increased mass and energy release to the containment due to the increased stroke time of the 1

Land, R.E., "Mass and Energy Releases Following a Steam Line Rupture," WCAP-8822 (Proprietary) and WCAP 8860 (Non-Proprietary), September 1976

MFIVs and the installation of AFW ARC valves will result in an increase in the peak containment pressure and temperature. Using the calculated increase in mass and energy, calculated as discussed in Section 3.1.3.1 of this safety evaluation, the licensee performed a calculation to verify that the containment design pressure and design temperature limits are not exceeded. The licensee used the CONTEMPT LT/28 computer code2 for this analysis.

CONTEMPT LT/28 was originally developed for the NRC and has been used by other licensees for licensing calculations which have been accepted by the NRC. The licensees December 12, 2003 letter states that the assumptions used for this analysis are the same as those in the original Final Safety Analysis Report (FSAR) analysis performed by Bechtel with the following two exceptions: (1) the assumption in the proposed analysis of an 8 percent revaporization rate, and (2) the isolation time for the AFW system.

Eight Percent Revaporization Rate The 8 percent revaporization rate accounts for the reentrainment of liquid into the containment atmosphere which tends to lower the temperature of the containment atmosphere. The use of 8 percent revaporization is consistent with the guidance in NUREG-0588, Revision 1,3 and is acceptable for use in the analysis of the MSLB inside containment. The licensees December 12, 2003 letter states that the 8 percent revaporization was first incorporated into the containment analysis in Amendment No. 35 dated March 30, 1988, which approved the increase in the licensed thermal power to the current 3565 MW thermal. The licensee's application for the power uprating included the 8 percent revaporization rate.4 Isolation Time for AFW System The original analysis of the MSLB inside containment assumed that the AFW flow to the steam generator with the ruptured steam line is isolated at 30 minutes. The revised analysis assumes that the isolation time for the AFW flow is 10 minutes. This 10 minute isolation time is already included in the Callaway licensing basis in that FSAR Section 10.4.9.2.3 states Low pump discharge pressure alarms will alert the operator to a secondary side break. The operator will then determine which loop is broken by observing high AFW flow, using the control room flow indication, and close the appropriate discharge control valve. This can be accomplished within 10 minutes after pump start.

The operator would also have other indications of which loop was broken.

2 NUREG/CR-0255, "CONTEMPT-LT/028-A A Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident," March 1979.

3 NUREG-0588, Revision 1, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," US Nuclear Regulatory Commission, July 1981.

4 Letter from Donald F. Schnell, Vice President, Union Electric, to USNRC, March 31, 1987.

The original containment MSLB analysis for Callaway is described in FSAR Section 6.2.1.4.3, "Containment Response Analysis." The analysis was performed using the COPATTA computer code. FSAR Section 6.2.1.4.3 states that the steam line break accident was reanalyzed in June 2000 to assess the impact of reduced containment cooler performance. The reanalysis was done with the CONTEMPT-LT/028 computer code. It was determined that the original FSAR COPATTA analysis was bounding. The licensee states that the containment analysis done to support the proposed changes of this license amendment request are also bounded by the original Bechtel pressure-temperature results.

Other Assumptions The input used in the analysis is given in FSAR Table 6.2.1-5. The licensees December 12, 2003 letter states that the containment volume used in the analysis is that in FSAR Table 6.2-5.

It is 4 percent less than the nominal value and, therefore, conservative. The Uchida heat transfer correlation is used in the analysis to calculate heat transfer from the containment atmosphere to heat sinks inside containment. The heat sinks included in the analysis are those in FSAR Table 6.2.1-4. The area of these heat sinks has been reduced by either construction tolerances or a 10 percent factor for those items that did not have known tolerances. The steam discharged into the containment is assumed to contain no entrained moisture. The NRC staff agrees that these assumptions are conservative.

3.1.2.6 MFW Line Break Inside Containment-Containment Analysis The licensee also addressed the MFW line break inside containment. FSAR Section 6.2.1.4.4 states that the effects of a postulated MFW line break inside containment on the containment are not as severe as the postulated MSLB inside containment because the discharged break effluent during a feedwater line break is at a lower specific enthalpy, that is, it contains less energy. Because of this, the licensee concluded that it is not necessary to analyze the MFW line break for this application and did not submit such an analysis. The NRC staff agrees with this conclusion.

3.1.2.7 Conclusions As discussed above, the NRC staff concludes that the only break inside containment that needs to be re-analyzed for the proposed plant modifications is the MSLB inside containment. For this line break, the assumptions used by the licensee in the containment analysis are conservative, and the containment analysis presented by the licensee shows that the containment design pressure and temperature are not exceeded, and that the calculated values from the re-analysis are bounded by the values currently reported in the FSAR (i.e., the containment peak temperature and pressure in the current FSAR are not being changed by this amendment).

Based on this, the NRC staff further concludes that the amendment will not in any way cause the plant to not meet GDC 16, 38, and 50. Therefore, the NRC staff concludes that Callaway will continue to meet these GDC.

3.1.3 Radiological Dose Technical Evaluation The radiological dose technical evaluation considers only those impacts that affect assumptions or inputs in the radiological consequence analysis of a design basis accident (DBA).

The licensee considered the impact of the replacement of the MFIV actuator and the associated proposed change in closure time requirements, the impact of the replacement of the AFW pump discharge check valve replacement, and the proposed re-definition of dose equivalent I-131, on the previously approved safety analyses. These evaluations were summarized in Sections 3.0 and 4.0 of the licensees application. As a result of these evaluations, the licensee concluded that only the radiological consequences of an SGTR with overfill event warranted a re-analysis. This conclusion was based on the following reasons: (1) the MFIV isolation timing was not explicitly incorporated in many DBA analyses, (2) with the exception of the SGTR with overfill, the increased AFW flow was a benefit in the DBA analyses, and (3) the steam mass release parameters used in the analyses of the loss of AC power, locked reactor coolant pump, MSLB, and rod control cluster assembly ejection were determined by a methodology that is independent of MFIV timing and AFW flow (being conservatively based on the total amount of heat needed to be removed from the plant to reach cold shutdown). The NRC staff reviewed the impact evaluations and, based on its experience with previous reviews and the above three reasons, agreed with the licensees conclusion that only the re-analysis of the radiological consequences of an SGTR with overfill event is required for the proposed amendment.

The NRC staff reviewed the description of the technical analysis of the radiological consequences of SGTR with overfill performed by the licensee in support of its proposed license amendment. This is provided in Sections 3.0 and 4.0 of Attachment 2 to the application and in the licensees December 12, 2003, response to the RAI. The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess these impacts and performed independent calculations to confirm the conservatism of the licensees analyses. However, the NRC staffs findings in this safety evaluation are based on the information provided by the licensee in its application and supplemental letter.

3.1.3.1 Effect on SGTR Radiological Consequences on Proposed SR 3.7.3.1 The proposed change to SR 3.7.3.1, to increase the MFIV stroke time acceptance criterion from five seconds to 15 seconds, was included in the re-analysis of the radiological consequences of an SGTR with overfill accident because this accident had been found to be the only DBA radiological consequence analysis significantly affected by this proposed change. This is addressed in Section 3.2 of this safety evaluation. In Section 3.2.1, the NRC staff concluded that the licensees re-evaluation of the SGTR with overfill used conservative assumptions and input to the re-analyses, and used conservative methods to perform the calculations, and that the radiological consequences of the re-analyzed SGTR with overfill accident are acceptable, and meet the dose guidelines in 10 CFR 100.11, as supplemented by SRP Section 15.6.3, and requirements in GDC 19, as supplemented by SRP Section 6.4.

3.1.3.2 Control Room Habitability The licensee did not discuss the impact of the proposed changes on the control room habitability in its application. In response to the NRC staffs RAI, the licensee provided a qualitative engineering judgement that the control room dose consequences from an SGTR with overfill would not exceed those consequences previously postulated for the LOCA and, since the consequences of the LOCA were previously found acceptable, it would be reasonable to assume that the consequences of an SGTR with overfill would also be found acceptable.

Also, the proposed modifications in the amendment do not alter either the control room or the control room design needed to meet the habitability requirements in GDC 19.

The licensee stated that the control room isolation would occur upon receipt of a safety injection (SI) signal. It considered several offsite dose consequence cases to evaluate the sensitivity of these results to the timing of an SI actuation. The maximum offsite dose occurred for the case involving SI actuation at the start of the event (T=0) and this was the timing assumed in the SGTR analysis evaluated herein. It was also determined that SI actuation could be delayed to approximately six minutes. However, since the onset of radioactivity release occurs at approximately 11 minutes, the control room would be isolated by SI actuation prior to the start of the release. It should be noted that the NRC staffs confirmatory analysis assumed that the control room realignment would not occur for 30 minutes and this timing is based on the need for manual operator action as discussed in Section 15A.3.1 of the Callaway FSAR.

Regulatory Information Summary (RIS) 2001-19, "Deficiencies in the Documentation of Design Basis Radiological Analyses Submitted in Conjunction with License Amendment Requests,"

was issued by the NRC to inform licensees of inadequacies in licensee's documentation of DBAs radiological analyses in license amendment applications. The NRC recommendations in the RIS are not requirements and no specific action or written response to the RIS by licensees was required. However, items 7a through 7d of RIS 2001-19 addressed considerations of radiation monitors to isolate the control room that affect the validity of control room habitability evaluations. Licensees were requested to address these items. Union Electric Company provided its response to these items in Section III (page 16 of 44) of the attachment to its letter dated December 12, 2003.

Items 7a through 7d addressed considerations that affect the validity of control room habitability evaluations based on the response to a different accident. In the responses to these items, the licensee stated that the control room isolation is initiated prior to the release of radioactivity.

The NRC staff concludes from its review of the licensee's December 12, 2003, response that the licensee has adequately addressed items 7a through of 7d of RIS 2001-19.

On June 12, 2003, the NRC issued Generic Letter (GL) 2003-01, "Control Room Habitability."

This GL identifies the NRC staff's concerns regarding the reliability of current surveillance testing to identify and quantify control room inleakage, and requests licensees to confirm the most limiting unfiltered inleakage into their control room envelope. The GL was issued shortly before the licensee's application was submitted and did not address any part of the GL. On August 11, 2003, the licensee submitted its "60-day" response to this GL, which identified several actions committed to by the licensee, including the development and performance of inleakage testing no later than September 30, 2004. In this response, the licensee stated that it

had performed a control room habitability assessment and, in the absence of confirmatory testing, the conclusion of the assessment is that the control room habitability systems are designed, constructed, configured, operated, and maintained consistent with the plants design and licensing basis.

Although the staff has reviewed the licensees response, as well as those received from other licensees, the staff action, if any, in response to the licensees' responses, including Callaway's, to the GL has not been decided at this time. Nonetheless, the NRC staff has determined that there is reasonable assurance that the Callaway control room will be habitable during an SGTR with overfill and that this amendment may be approved prior to the staffs review of Callaway's final response to the GL. The staff is basing this determination on (1) the Callaway engineering evaluation that the LOCA consequences for control room habitability are more limiting than the other postulated accidents at Callaway, (2) the staffs confirming analysis, (3) the relative magnitude of the infiltration assumed in the Callawway analyses, (4) the control room habitability assessment performed for Callaway, (5) the programmatic elements identified in the Callaway 60-day response to GL 2003-01, and (6) the design of the Callaway control room and control building. Also, the proposed modifications in the amendment do not alter the control room design for control room habitability. It should be noted that the NRC staffs approval of this amendment does not relieve the Callaway licensee of addressing the information requests in GL 2003-01 and does not imply that the NRC staff would necessarily find the analysis in this amendment acceptable as a response to information request 1(a) in GL 2003-01.

3.1.4 Conclusions for Plant Modifications The licensee proposed the following: (1) the replacement of electro-hydraulic actuators with system-medium actuators for the MFIVs, and (2) the replacement of the existing swing check valve in each AFW motor driven pump discharge line with an ARC check valve. These changes affect the containment and the radiological consequences of the SGTR with overfill accident. The licensee has evaluated these changes and concluded that the changes will not affect the safety of the Callaway. Based on the above evaluation, the NRC staff concludes the following for the proposed modifications:

 Based on the evaluation in Section 3.1.1, the NRC staff concludes that the new equipment will not cause any safety problems, will correct problems that the licensee has had with the existing equipment, and that the regulatory requirement of GDC 34 regarding single failure criteria continues to be met.

 Based on the evaluation in Section 3.1.2, the NRC staff concludes that there is reasonable assurance that the containment design pressure and temperature will not be exceeded during any DBA and that Callaway will continue to meet GDC 16, 38, and 50.

 Based on the evaluation in Section 3.1.3, the NRC staff concludes that (1) the radiological dose consequences of the re-analyzed SGTR with overfill are acceptable and meet the dose guidelines in 10 CFR 100.11, as supplemented by SRP Section 15.6.3, and the requirements in GDC 19, as supplemented by SRP Section 6.4, and (2) the existing control room habitability design at Callaway is not being changed by the amendment and GL 2003-01 will be addressed separate from this amendment.

3.2 SGTR With Overfill Accident Re-analysis In support of the plant modifications addressed in Section 3.1 above, the licensee re-analyzed the SGTR with overfill accident because this accident was affected by the modifications. The revised analysis uses the thyroid dose conversion factors (DCFs) in International Commission on Radiological Protection Publication 30 (ICRP 30) and the iodine spiking model in RG 1.195.

The NRC staffs review of the use of the ICRP DCFs and iodine spiking model is given in Section 3.2.2.1.

This re-analysis was also performed to respond to a condition reported in Licensee Event Report (LER) 03-003-00, "Incorrect Steam Generator Tube Rupture Analysis Contained in FSAR," dated May 9, 2003. This amendment will also revise the FSAR by incorporating the description and evaluation of an SGTR with overfill event. This analysis incorporated revised DCFs, a revised iodine spiking model, revised operator response times, and other revised inputs and assumptions consistent with current plant configuration and operation.

The revised iodine spiking model used was that described in RG 1.195. In addition to requesting use of this iodine spiking model for the revised SGTR analysis, the licensee also requested approval to use RG 1.195 for other licensing basis dose applications. In its December 12, 2003, RAI response, the licensee reduced the scope of this request to only the use of the iodine spiking factor of 335 and the ICRP-30 DCFs on a forward-fitting bases for all of the Callaway FSAR Chapter 15 radiological consequence analyses.

The SGTR event causes direct release of radioactive material contained in the primary coolant to the environment through the ruptured steam generator tube and safety or atmospheric relief valves. Reactor protection and engineered safety features are actuated to mitigate the accident and restrict the offsite dose within the guidelines of the 10 CFR Part 100 guidelines. The NRC staffs review covers postulated initial core and plant conditions, the method of thermal and hydraulic analysis, sequence of events assuming with and without offsite power available, assumed response of reactor system components, functional and operational characteristics of the reactor protection system, required operator actions consistent with the plant emergency operating procedures (EOPs), and the results of the accident analysis. A single failure of a mitigating system is assumed for this event. Acceptance criteria for this event are: (1) the radiological consequences of the SGTR event are within the guidelines of the 10 CFR Part 100 limits, and (2) there is no overfill of the steam generator during the mitigation of this event which could cause unacceptable radiological consequences or potential failure of the main steam system. Specific review criteria are contained in SRP Section 15.6.3.

3.2.1 Thermal Hydraulic Analysis The thermal-hydraulic portion of the licensees re-analysis of the SGTR accident with overfill uses the same RETRAN computer code as that for the previous analysis at Callaway Plant.

The use of the RETRAN model for performing the SGTR analysis was previously reviewed and accepted by the NRC staff. The major changes to the input parameters and assumptions in the thermal hydraulic portion of the re-analysis of the SGTR accident with overfill are discussed below.

 The re-analysis of the SGTR with overfill event assumes a slightly lower reactor coolant system (RCS) temperature of 578.4EF from its previous value of 583.4EF and a lower steam generator steam pressure of 908 psia from its previous value of 939 psia. These changes will result in higher calculated break flow for a more conservative dose assessment.

 The re-analysis assumes a higher RCS flow of 382,640 gpm from its previous value of 374360 gpm. This change will result in a higher early stage break flow for a more conservative dose assessment.

 The re-analysis assumes a MFIV isolation delay time of 17 seconds from its previous value of 5 seconds. This change will result in more limiting conditions regarding SG overfill and leads to a more conservative dose assessment. This assumption is consistent with the licensees proposed change to SR 3.7.3.1 to increase the required stroke time to 15 seconds from the current stroke time of 5 seconds.

 The re-analysis assumes a higher decay heat model which will increase the calculated break flow for a more conservative dose assessment.

 The re-analysis assumes a higher AFW flow to the failed steam generator which will provide more limiting conditions regarding steam generator overfill and leads to a more conservative dose assessment. The revised AFW flow is consistent with the flow generated from the AFW with the licensees proposed change of the ARC at the discharge of each motor-driven AFW pump.

 The re-analysis assumes that when AFW is actuated, its flow ramps up to full flow conditions between 5 seconds and 30 seconds. This is a more realistic prediction for the initial AFW flow than the previous analysis which assumes a 30-second delay to full AFW flow.

 The re-analysis assumes an operator action time to isolate AFW flow to the failed steam generator of 20 minutes in lieu of the previous value of 16 minutes, an operator action time to initiate RCS cooldown of 30 minutes in lieu of the previous value of 24 minutes, and an operator action time to complete RCS depressurization of 40 minutes in lieu of the previous value of 35 minutes. The longer times to perform these operator actions will result in more limiting conditions regarding steam generator overfill and result in a more conservative dose assessment. The operator time is addressed in Section 3.2.2.3.

As discussed above, the NRC staff concludes that the assumptions used in this thermal-hydraulic analysis are conservative. The licensee has presented the changes to Section 15.6.3.2 of the FSAR to reflect the above discussed changes in an attachment to its application.

The NRC staff has reviewed the revised FSAR section and finds that it is consistent with the revised analysis.

The results of the licensees re-analysis of the SGTR accident with overfill concluded that a steam generator overfill during this event will occur which will cause water filled in the main

steam line up to the main steam stop valve and the radioactive water will relieve to the environment through the main steam safety valves. However, the licensees mechanical analysis has confirmed that the main steam line will maintain its integrity when it is filled up with water. The radiological consequences of this event with steam generator overfill is assessed in Section 3.2.2.

The NRC staff has reviewed the licensees re-analysis of the SGTR accident with overfill and concludes that the licensees analysis has correctly and conservatively accounted for operation of the plant at the current power level with proposed modifications in the MFIV and AFW system, and that the thermal-hydraulic analysis was performed using acceptable analytical methods and approved computer codes. The NRC staff further concluded that the assumptions used in this thermal-hydraulic analysis are conservative. Therefore, the NRC staff finds that the licensees thermal-hydraulic portion of the re-analysis of the SGTR accident with overfill is acceptable.

3.2.2 SGTR With Overfill Radiological Consequences The SGTR with overfill is a DBA and the following is a description of the event. The SGTR with overfill is a complete severance of a single tube in one of the steam generators resulting in the transfer of RCS water to the ruptured steam generator. The primary-to-secondary break flow through the ruptured tube following the event results in radioactive contamination of the secondary system. A reactor trip occurs, safety injection actuates, and a loss-of-offsite power (LOOP) occurs concurrently with the reactor trip. At least one of the AFW pumps starts to inject water into the steam generators. The AFW flow control valve for the ruptured steam generator is assumed to fail fully open resulting in excessive flow to the steam generator. The water level in the ruptured steam generator increases and fills the main steamline up to the MSIV. This is the overfill of the steam generator. A main steam safety valve (MSSV) on the ruptured steam generator opens at about 20 minutes to relieve pressure and contaminated water is released to the environment. Operators take actions in accordance with EOPs to depressurize the primary and to terminate safety injection. When equilibrium is reached between the primary and secondary pressures at about one hour, the break flow into the ruptured steam generator will cease. The pressure and the water level in the ruptured steam generator will decrease and the MSSV will close. Since the MSSV was not designed to relieve water, the licensee conservatively assumes that the MSSV remains partially open once the pressure is reduced providing a release path for contaminated steam.

Operator actions, which are evaluated in detail in Section 3.2.2.3, to depressurize the primary system and to cool down the plant following termination of the break flow involve the intentional opening of atmospheric dump valves (ADVs) on the unaffected steam generators. This action is necessary since the main condenser is assumed to be unavailable due to the LOOP. The licensee conservatively assumes that there is primary-to-secondary leakage into the unaffected steam generators and that the steam releases from the unaffected steam generators would be contaminated by this leakage. The steam releases from the ruptured and unaffected steam generators continue until the residual heat removal (RHR) system can be used to complete the cooldown of the reactor at approximately six hours into the event.

The licensees analysis conservatively assumes that the RCS primary-to-secondary leakage, the initial RCS specific activity, and the initial steam generator specific activity are at the maximum values allowed by the TSs. The licensee considered two cases of iodine spiking. In the first case, an iodine spike is initiated by the event, resulting in the release of radioiodine from the fuel at 335 times the normal appearance rate (see Section 3.2.2.1 below). For the second case, it is assumed that an iodine spike had occurred prior to the event and that the RCS iodine specific activity is 60 uCi/gm dose equivalent I-131. Again, this is the maximum concentration allowed by the TSs.

The assumptions used by the NRC staff in its confirmatory calculation are presented in Table 1 attached to this safety evaluation. The licensees analysis was performed using dynamic thermo-hydraulic inputs obtained from the RETRAN computer code. The NRC staffs confirming calculation was based on a low resolution extrapolation of graphs of thermo-hydraulic data provided in the licensee's application. The calculations confirmed the conservatism of the licensee's analyses, including input assumptions and methodology. The exclusion area boundary (EAB) and low population zone (LPZ) doses calculated by the licensee for the SGTR with overfill, and given in pages 18 and 19 of Attachment 2 to the licensee's application, are the following:

Exposure Location Iodine Spike Dose (REM) Regulatory Limit*

(REM)

Thyroid EAB Pre-accident 46.2 300 Accident initiated 13.4 30 LPZ Pre-accident 4.71 300 Accident initiated 1.43 30 Whole Body EAB Pre-accident 0.362 25 Accident initiated 0.396 2.5 LPZ Pre-accident 0.0385 25 Accident initiated 0.0424 2.5 Note:

• 10 CFR 100.11 guidelines of 100 rem thyroid and 25 rem whole body, except that for the postulated accident with the accident initiated iodine spike the calculated doses should not exceed a small fraction (i.e., 10 percent) of the previous doses, or 30 rem thyroid and 2.5 rem whole body.

The doses calculated for the SGTR with overfill are for (1) the postulated accident with an assumed pre-accident iodine spike in the reactor coolant, and (2) the postulated accident with the equilibrium iodine concentration for continued full power operation in combination with an assumed accident-initiated iodine spike. Comparing these separate doses to the regulatory limits given above, the NRC staff concludes that the calculated doses for the SGTR with overfill do not exceed the 10 CFR Part 100 guidelines, as supplemented by SRP Section 15.3.6.

3.2.2.1 Use of ICRP DCFs and Iodine Spike Multiplier of 335 The licensees intent in requesting approval of the use of RG 1.195 at Callaway was to gain NRC approval for the use of the ICRP-30 DCFs and the 335 iodine spiking factor, on a forward-fitting basis, for all of the Callaway FSAR Chapter 15 accident radiological consequence analyses. In this regard:

 The NRC staff has previously accepted the use of an iodine spike multiplier of 335 for SGTRs for other licensees. This factor does not apply to any other DBA accident. In reviewing the SGTR with overfill analysis described by the licensee in this amendment request, the NRC staff determined that this value was appropriate for the analysis of SGTR with overfill, and that it would also be acceptable for the other DBA SGTR analyses in the Callaway licensing basis.

 In RIS 2001-19, the NRC staff identified the thyroid DCFs based on ICRP-30 to be an acceptable change in methodology that does not warrant prior review. This conclusion is consistent with the 10 CFR 50.59 implementation guidance in that the NRC staff has accepted the ICRP-30 DCFs for other licensees and that there is nothing inherently specific to a site, plant, licensee, or accident sequence regarding these factors. As such, the NRC staff concludes that the approval sought by the licensee for use of the ICRP-30 DCFs is granted.

Therefore, as stated above, the NRC staff concludes that the use of the iodine spike multiplier of 335 for the STGR accident and the use of ICRP-30 thyroid DCFs at Callaway is acceptable.

3.2.2.2 Control Room Habitability and GL 2003-01 This is discussed in Section 3.1.3.2. In that section, the NRC staff concluded that there is reasonable assurance that the Callaway control room will be habitable during an SGTR with overfill and that this amendment may be approved prior to the staffs review of the licensee's final response to GL 2003-01 on control room habitability.

3.2.2.3 Human Factors Assessment This assessment of the licensee's application involved a review of the operator actions identified by the licensee in its re-analysis of the SGTR with overfill accident. The licensee provided information regarding operator manual actions for responding to an SGTR with overfill accident in its application and its letter dated December 12, 2003. Operator actions were also discussed in the meeting held on November 12, 2003. The following is a summary of the licensees responses and the NRC staffs conclusions regarding the area of operator performance during the accident.

Operator Actions The licensee stated that the re-analysis of the SGTR overfill sequence involves a revised set of operator action times. The licensee conducted exercises on the plant referenced simulator during Spring 2003 to establish the new action time requirements, and personnel from both the

licensee and Westinghouse reviewed operator action times used by other Westinghouse plants in order to ensure that the new times were reasonable. The licensee also indicated that sufficient indications, controls, alarms, and procedures would be provided to enable operators to carry out functions successfully. Additionally, the licensee stated that all operating crews, including both on-shift and staff crews, have demonstrated that they are capable of satisfying the new set of action times and provided the staff with the actual response times demonstrated during simulator exercises by the crews. All response times were within the acceptable values for the accident. Based on the demonstrated crew response times from the simulator provided by the licensee, the NRC staff finds the revised operator times to be sufficient for successful completion of the required tasks during the accident, and that they should not have an adverse effect on operator performance or the safe operation of the facility.

Emergency Operating Procedures The licensee indicated that the revised operator times related to the re-analysis of the SGTR with overfill have been incorporated into the associated EOPs. Following the initiation of the accident sequence, EOP Procedure E-0, "Reactor Trip or Safety Injection," will be used initially, and time-critical diagnostics steps are necessary for the transition from E-0 to Procedure E-3, "Steam Generator Tube Rupture." The licensee provided a description of these procedural diagnostic methods. Procedure E-3 has been revised to support the SGTR with overfill accident re-analysis to include operator action times with a more aggressive response to the event, and the licensee provided a description of these revisions. The licensee stated that all changes were validated on the simulator and that all licensed operators have received training on these changes. Again, the licensee indicated that all crews have been able to demonstrate the capability to meet the new time values in the diagnostic methods specified by E-0 and E-3.

Based on this, the NRC staff concludes that the EOPs for the SGTR with overfill accident are acceptable.

Operator Training and Control Room Simulator The licensee indicated that training on the SGTR with overfill accident sequence and its related EOPs is provided by their licensed operator training program, and all licensed operators have been trained on the updated E-3 procedure through both classroom and simulator training. The licensee stated that the training program used all critical attributes from FSAR Chapter 15, "SGTR Overfill Analysis," in the development of the simulator training exercises and provided a list of examples of such attributes. In addition, during simulator training, the licensees safety analysis engineers observed all exercises to verify that simulator results and operator actions were consistent with FSAR Chapter 15 results. Based on this, the NRC staff finds the licensees training and simulator training for the SGTR with overfill to be acceptable.

Conclusion The NRC staff has reviewed the licensees proposed revised operator action times for response to a SGTR with overfill accident for the licensee's re-analysis of the accident. The staff concludes that the licensee has adequately considered the impact of the proposed revisions and has provided reasonable assurance for demonstrating that the revised operator actions times are sufficient for operators to successfully perform required tasks in the accident. Based

on this, the NRC staff concludes that the revisions in operator action times associated with re-analysis of the SGTR with overfill accident are acceptable.

3.2.2.4 Conclusion As discussed above, the NRC staff concludes that the use of ICRP 30 DCFs and the iodine spiking multiplier of 335 in calculating dose consequences for the SGTR with overfill accident is acceptable, that the revisions in operator action times associated with re-analysis of the SGTR with overfill accident are acceptable, and that the licensees analyses of the accident is conservative. Based on this and the dose consequences calculated by the licensee for the SGTR with overfill accident, which are given above in Section 3.2.1, the staff concludes that the dose consequences for the SGTR with overfill accident meet the dose guidelines in 10 CFR 100.11, as supplemented by SRP Section 15.6.3, and are, therefore, acceptable.

3.2.2.5 Effect on Radiological Consequences on Proposed TS Changes 3.2.2.5.1 TS 1.1, "Definitions" The licensee proposed to revise the definition of dose equivalent iodine -131 to allow the use of thyroid DCFs based on ICRP 30. This redefinition is consistent with the re-analyzed SGTR with overfill and would facilitate the use of the revised DCFs in other DBA radiological consequence analyses. As stated in Section 3.2.2, the NRC staff has concluded that the use of the thyroid DCFs derived from ICRP-30 to be acceptable for use in DBA radiological consequence analyses in plant-specific amendment requests. Based on this, the NRC staff concludes that the proposed change to TS 1.1, which is to allow the licensee to use the new DCFs in DBA radiological consequence analyses, is acceptable.

3.2.1.5.2 SR 3.7.3.1 The licensee proposed to increase the MFIV stroke time acceptance criterion from 5 seconds to 15 seconds. As discussed in Section 3.2.1, this proposed change was incorporated into the re-analysis of the radiological consequences of an SGTR with overfill accident because this accident had been found to be the only DBA radiological consequence analysis significantly affected by this proposed change. In Section 3.2.2.4, the NRC staff concludes that the radiological consequences of an SGTR with overfill accident are acceptable. This included the NRC staff accepting the use of the ICRP 30 DCFs and the iodine spiking factor of 335 in the calculation of the radiological consequences of the accident, which are addressed in Section 3.2.2.1. Based on this, the NRC staff concludes that the proposed change to SR 3.7.3.1 is acceptable.

3.3 Conclusion as to the Amendment Meeting Regulatory Requirements In Sections 3.1.4 and 3.2.1.4, the proposed changes to the plant and to the TSs were reviewed against the regulatory requirements given in Section 2.0. In these sections, the NRC staff concluded that all the regulatory requirements in Section 2.0 were being met. In addition, the staff has concluded that the proposed changes to the TSs in the amendment were the results of the proposed changes to the plant and to the analysis of the SGTR with overfill accident

which were approved in Sections 3.1.4 and 3.2.1.4. Based on this, the NRC staff concludes that the proposed amendment to the TSs in the licensees application date June 27, 2003, is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Missouri State official was notified of the proposed issuance of the amendment. The State official did not offer any comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes a surveillance requirement. The NRC staff has determined that the amendment involves no significant increase in the amounts, 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 consideration and there has been no public comment on such finding (68 FR 43394). 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.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is 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 Commissions 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.

Attachment:

Table 1 - SGTR With Overfill Analysis Assumptions Principal Contributors: J. Cai C. Liang R. Eckenrode R. Lobel J. Golla Date: March 11, 2004

Callaway Plant, Unit 1 cc:

Professional Nuclear Consulting, Inc. Mr. Rick A. Muench 19041 Raines Drive President and Chief Executive Officer Derwood, MD 20855 Wolf Creek Nuclear Operating Corporation P.O. Box 411 John ONeill, Esq. Burlington, KA 66839 Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W. Mr. Dan I. Bolef, President Washington, D.C. 20037 Kay Drey, Representative Board of Directors Coalition for the Mr. Mark A. Reidmeyer, Regional Environment Regulatory Affairs Supervisor 6267 Delmar Boulevard Regulatory Affairs University City, MO 63130 AmerenUE P.O. Box 620 Mr. Lee Fritz, Presiding Commissioner Fulton, MO 65251 Callaway County Court House 10 East Fifth Street U.S. Nuclear Regulatory Commission Fulton, MO 65151 Resident Inspector Office 8201 NRC Road Mr. David E. Shafer Steedman, MO 65077-1302 Superintendent, Licensing Regulatory Affairs Mr. Chris Younie AmerenUE Manager, Quality Assurance P.O. Box 66149, MC 470 AmerenUE St. Louis, MO 63166-6149 P.O. Box 620 Fulton, MO 65251 Mr. Keith D. Young Manager, Regulatory Affairs Manager - Electric Department AmerenUE Missouri Public Service Commission P.O. Box 620 301 W. High Fulton, MO 65251 P.O. Box 360 Jefferson City, MO 65102 Mr. Scott Clardy, Director Section for Environmental Public Health Regional Administrator, Region IV P.O. Box 570 U.S. Nuclear Regulatory Commission Jefferson City, MO 65102-0570 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Certrec Corporation 4200 South Hulen, Suite 630 Mr. Ronald A. Kucera Fort Worth, TX 76109 Deputy Director for Public Policy Department of Natural Resources P.O. Box 176 Jefferson City, Missouri 65102

TABLE 1 SGTR WITH OVERFILL ANALYSIS ASSUMPTIONS Reactor power, MWt 3636 (includes 2% uncertainty)

RCS* mass, lbm Application**, Figure 15.6-3.2.d Initial iodine RCS specific activity, µCi/gm dose equivalent I-131 1.0 Initial noble gas RCS specific activity % failed fuel 1.0 Initial secondary specific activity as fraction of primary specific activity 0.1 RCS to secondary leak rate, gal/min 1.0 Dose conversion factors ICRP 30 3

Offsite breathing rate, m /sec 0-8 hours 3.47E-4 8-24 hours 1.75E-4 24-720 hours 2.32E-4 Iodine spike appearance rate parameters Filtration efficiency fraction 1.0 Letdown flow, gpm 140 RCS initial activity, uCi/gm d.e. I-131 1.0 RCS leakage, gpm 1.0 Control Building parameters Mixing volume, ft3 1.50E5 Filter Intake, cfm 0-30 minutes (30 minutes-720 hours) 900 (450)

Unfiltered inleakage, cfm 300 Filter efficiency, all forms of iodine, % 95 Control Room parameters Volume, cfm 100,000 Filtered flow from control building, cfm 440 Unfiltered flow from control building, cfm, 0-30 min (30 min to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />) 440 (0)

Unfiltered inleakage, cfm 10 Filtered recirculation, cfm 1360 Filter efficiency, all forms of iodine, % 95 Control room breathing rate, m3/sec 3.47E-4 Control room occupancy factors 0-24 hours 1.0 1-4 days 0.6 4-30 days 0.4 Limiting control room /Q (includes occupancy factors), sec/m3 0-8 hrs 7.2E-4 8-24 hrs 5.3E-4 1-4 days 1.7E-4 Offsite /Q, sec/m3 EAB: 0-2 hr 1.5E-4 LPZ: 0-8 hr 1.5E-5 8-24 hr 1.0E-5 24-96 hr 4.6E-6 96-720 hr 1.5E-6 Pre-incident iodine spike activity 60.0 µCi/gm dose equivalent I-131 Co-incident spike multiplier 335 Iodine spike duration, hrs 8 Event timing, sec Reactor trip 0 LOOP 0 SI signal 0 Ruptured SG water relief begins 1149 Break flow stops 3623 Ruptured SG relieves steam 4560 RHR cut in conditions reached 21,600 Steam generator mass @, lbm Application, Figure 15.6-3.2.d Break flow to affected SG, lbm/s Application, Figure 15.6-3.2.i Break flow flash fraction Application, Figure 15.6-3.2e Iodine partitioning in ruptured SG Flashed fraction 1.0 Unflashed fraction 0.01 From released overfill water 0.5 Steam release from ruptured SG, lbm/s Application, Figure 15.6-3.2.g Primary to secondary leakage to unaffected SG, lbm 3,456 (based on 1 gpm at 55 lbm/ft3 for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />)

Steam release from unaffected SGs, lbm/s Application, Figure 15.6-3.2.h Iodine partitioning in unaffected SG Flashed fraction 1.0 Unflashed fraction 0.01

• Acronyms:

EAB = exclusion area boundary LOOP = loss of offsite power LPZ = low population zone RCS = reactor coolant system RHR = residual heat removal SG = steam generator SI = safety injection

/Q = wind dispersion from source (sec/meters3)

• • Application = licensee's license application date June 27, 2003