ML062610440

Issuance of License Amendment 199 Changes to Technical Specifications to Replace the Current Accident Source Term Used in Design Basis Radiological Analyses

ML062610440
Person / Time
Site:	Columbia
Issue date:	11/27/2006
From:	Brian Benney NRC/NRR/ADRO/DORL/LPLIV
To:	Parrish J Energy Northwest
Terao D, NRR/ADRO/DORL/LPL4, 415-1302
Shared Package
ML062610429	List: ML062610440
References
TAC MC4570
Download: ML062610440 (49)

Text

November 27, 2006 Mr. J. V. Parrish Chief Executive Officer Energy Northwest P.O. Box 968 (Mail Drop 1023)

Richland, WA 99352-0968

SUBJECT:

COLUMBIA GENERATING STATION - ISSUANCE OF AMENDMENT RE:

ALTERNATIVE SOURCE TERM (TAC NO. MC4570)

Dear Mr. Parrish:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 199 to Facility Operating License No. NPF-21 for the Columbia Generating Station (CGS). The amendment consists of changes to the Technical Specifications and Final Safety Analysis Report in response to your application dated September 30, 2004, as supplemented by letters dated March 16, September 29, 2005, and March 21, August 7, August 24, and September 11, 2006.

The proposed change will replace the current accident source term used in design-basis radiological analyses with an alternative source term pursuant to Title 10 of the Code of Federal Regulations, Section 50.67 (10 CFR 50.67), "Accident source term." Energy Northwest updated the current loss-of-coolant accident, the control rod drop accident, the fuel-handling accident, the main steam line break analyses, and proposed revisions to several technical specifications with an exception. That exception is the Technical Information Document (TID)-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," which will continue to be used as the radiation dose basis for equipment qualification, and radiation zone maps/shielding calculations.

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

Sincerely,

/RA by D. Terao for/

Brian J. Benney, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397

Enclosures:

1. Amendment No. 199 to NPF-21

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

Pkg ML062610429 (Amendment ML062610440, TS Pg ML063280053)

OFFICE NRR/LPL4/PM NRR/LPL4/LA DSS/D DRA/D OGC - NLO NRR/LPL4/BC NAME BBenney:MFields for LFeizollahi TMartin CHolden DRoth DTerao DATE 11/9/06 11/9/06 11/9/06 11/9/06 11/17/06 11/21/06 ENERGY NORTHWEST DOCKET NO. 50-397 COLUMBIA GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 199 License No. NPF-21

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

A. The application for amendment by Energy Northwest (licensee) dated September 30, 2004, as supplemented by letters dated March 16, and September 29, 2005, and March 21, August 7, August 24, and September 11, 2006, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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; and E. 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 and Final Safety Analysis Report as indicated in this license amendment. Paragraph 2.C.(2) of Facility Operating License No. NPF-21 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. 199 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.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: November 27, 2006

ATTACHMENT TO LICENSE AMENDMENT NO. 199 FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 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. The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE INSERT ii ii iii iii 1.1-2 1.1-2 1.1-3 1.1-3 3.1.7-1, 3.1.7-2, and 3.1.7-3 3.1.7-1, 3.1.7-2, and 3.1.7-3 3.3.6.1-7 3.3.6.1-7 3.3.6.2-4 3.3.6.2-4 3.3.7.1-2 through 3.3.7.1-5 3.3.7.1-2 through 3.3.7.1-4 3.6.1.3-8 through 3.6.1.3-9 3.6.1.3-8 through 3.6.1.3-9 3.6.4.1-1 3.6.4.1-1 3.6.4.1-2 3.6.4.1-2 3.6.4.1-3 --------

3.6.4.2-1 through 3.6.4.2-4 3.6.4.2-1 through 3.6.4.2-3 3.6.4.3-1 3.6.4.3-1 3.6.4.3-2 3.6.4.3-2 3.6.4.3-3 -------

3.7.3-1 through 3.7.3-4 3.7.3-1 through 3.7.3-3 3.7.4-1 3.7.4-1 3.7.4-2 3.7.4-2 3.7.4-3 -------

3.8.2-1 through 3.8.2-4 3.8.2-1 through 3.8.2-3 3.8.5-1 3.8.5-1 3.8.5-2 3.8.5-2 3.8.8-1 3.8.8-1 3.8.8-2 3.8.8-2 3.9.7-1 3.9.7-1

---

3.9.10-1 5.5-7 5.5-7

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 199 TO FACILITY OPERATING LICENSE NO. NPF-21 ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397

1.0 INTRODUCTION

By application dated September 30, 2004 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML042930316), as supplemented by letters dated March 16 and September 29, 2005, and March 21 and September 11, 2006 (ADAMS Accession No. ML052850270, ML050900256, ML060900602 and ML062620329, respectively), Energy Northwest (EN/licensee) requested changes to the Technical Specifications (TSs) (Appendix A to Facility Operating License No. NPF-21) and Final Safety Analysis Report for the Columbia Generating Station (CGS). The supplements dated March 16 and September 29, 2005, and March 21 and September 11, 2006, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published on October 26, 2004 (69 FR 62472), in the Federal Register.

The requested change would replace the current accident source term used in design-basis radiological analyses with an alternative source term (AST) pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, "Accident source term." Energy Northwest updated the current loss-of-coolant accident (LOCA), the control rod drop accident (CRDA), the fuel handling accident (FHA), and the main steam line break (MSLB) analyses. The requested change would revise the licensing and design basis to reflect the application of full scope AST methodology consistent with the guidance provided in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (with the exception that the Technical Information Document, (TID)-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," will continue to be used as the radiation dose basis of equipment qualification (EQ)) and the associated TS changes.

Changes to the TSs are analyzed in this safety evaluation (SE).

The AST methodology allows the licensee to revise the accident source term used in the design-basis radiological consequence analysis. To implement the AST, the requirements for being in MODE 4 requires changes to TS 3.1.7, "Standby Liquid Control System." The request also includes a requirement to maintain the pH of the suppression pool above 7 for a period of 30 days following a LOCA in order to minimize the amount of radioactive iodine released. In addition, the licensee takes credit for drywell spray removal of iodine.

2.0 BACKGROUND

/REGULATORY EVALUATION In December 1999, the U.S. Nuclear Regulatory Commission (NRC) issued a new regulation, 10 CFR 50.67, "Accident Source Term," which provided a mechanism for licensed power reactors to replace the traditional accident source term, used in their design-basis accident (DBA) analyses, with an AST. Regulatory guidance for the implementation of these ASTs is provided in RG 1.183. A licensee seeking to use an AST is required by 10 CFR 50.67 to apply for a license amendment. An evaluation of the consequences of affected DBAs is required to be included with the submittal. ENs application addresses these requirements in proposing to use the AST described in RG 1.183 as the source term in the evaluation of the radiological consequences of the design-basis LOCA, MSLB, CRDA, and FHA at the CGS. As part of the implementation of the AST, the total effective dose equivalent (TEDE) acceptance criteria of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11 and 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 19 as the CGS licensing basis for the DBA LOCA, CRDA, FHA, and the MSLB. Specifically, 10 CFR 50.67(b)(2) states that the NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 sievert (Sv) (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE.

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.

The NRC requirement and guidance documents that are applicable to the NRC staffs review of CGSs license amendment request include:

Part 50 of 10 CFR, Appendix A, GDC 26, requires that each reactor have two independent reactivity control systems of a different design while GDC 29 requires that the reactivity control system be capable of accomplishing its safety function in the event of anticipated operational occurrences.

Implementation of an AST involves re-analyzing DBAs according to 10 CFR 50.67, "Accident source term," and applying for a license amendment under 10 CFR 50.90.

GDC 17, "Electric power systems," of 10 CFR, Part 50, Appendix A requires that nuclear power plants have onsite and offsite electric power systems to permit the functioning of structures, systems, and components (SSCs) that are important to safety. The onsite system is required to have sufficient independence, redundancy, and testability to perform its safety function, assuming a single failure. The offsite power system must be supplied by two physically independent circuits that are designed and located so as to minimize, to the extent practical, the likelihood of their simultaneous failure under an operating and postulated accident, and

environmental conditions. In addition, this criterion requires provisions to minimize the probability of losing electric power from the remaining electric power supplies as a result of loss of power from the unit, the offsite transmission network, or the onsite power supplies.

GDC 18, "Inspection and testing of electric power systems," requires that electric power systems that are important to safety must be designed to permit appropriate periodic inspection and testing.

Section 50.36 of 10 CFR, "Technical specifications," provides the content required in a licensees TSs. Specifically, 10 CFR 50.36(c)(3) requires that the TS include surveillance requirements.

Section 50.49 of 10 CFR, "Environmental qualification of electric equipment important to safety for nuclear power plants." Compliance with this rule requires that the safety-related electrical equipment which are relied upon to remain functional during and following the design-basis events be qualified for an accident (harsh) environment. This provides assurance that the equipment needed in the event of an accident will perform its intended function.

NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," which provides a more realistic source term than the TID-14844 source term. This NUREG provides estimates of accident source term that were more physically based and that could be applied to boiling-water reactors (BWRs) and pressurized-water reactors. These source terms are characterized by the composition and the magnitude of the radioactive material, the chemical and physical properties of the material, and the timing of the release to the containment. Equipment dose calculations performed with this NUREG source term were lower than doses calculated with the TID-14844 source term during the gap release and early in-vessel release phases of core degradation.

RG 1.183 also provides guidance to the licensee of operating power reactors on acceptable applications of MSLBs. This RG states that the licensees may use either the AST or the TID-14844 assumptions for performing the required EQ analyses. It further states that no plant modifications are required to address the impact of the difference in source term characteristics (i.e., AST vs. TID 14844) on EQ doses.

NUREG-0933, Issue 187, "The Potential Impact of Postulated Cesium Concentration on Equipment Qualification." The issue states that the Sandia National Laboratories' report, "Evaluation of Radiological Consequences of Design Basis Accidents at Operating Reactors Using the Revised Source Term," dated September 28, 1998, showed that: (1) for equipment exposed to the containment atmosphere, the TID-14844 source term and the gap and in-vessel releases in the AST produced similar integrated doses, and (2) for equipment exposed to suppression pool water, the integrated doses calculated with the AST remain enveloped by those calculated with TID-14844 for the first 145 days post accident for a BWR, including the 30 percent versus 1 percent release of cesium. It was concluded that there was no clear basis to backfit the requirement to modify the design-basis for EQ to adopt the AST. There would be no discernible risk reduction associated with such a requirement. Longer term equipment operability issues associated with severe fuel damage accidents, (with which the AST is associated) could also be addressed under accident management or plant recovery actions as necessary.

NUREG-0800, "Standard Review Plan," Section 15.7.4, provides guidance to the NRC staff for the review and evaluation of system design features and plant procedures provided for the mitigation of the radiological consequences of postulated FHAs.

Standard Review Plan, Section 15.0.1, "Radiological Consequence Analyses Using Alternate Source Term," issued in July 2000, also provides new guidance on acceptable applications of alternative source terms.

Technical Specification Task Force Traveler (TSTF)-51, Revision 2 approved by the NRC on October 13, 1999, provides for the relaxation of some TS requirements during refueling after a sufficient decay period has occurred.

3.0 TECHNICAL EVALUATION

3.1 Accident Dose Calculations The NRC staff reviewed the technical analyses related to the radiological consequences of DBAs that EN performed in support of this proposed license amendment. EN provided information regarding these analyses in the September 30, 2004, submittal, as supplemented in letters dated on September 29, 2005, and March 21, 2006. The NRC staff met with EN on December 6 through 7, 2005, and held teleconferences on January 5, 2006, January 31, 2006, February 12, 2006, May 9, 2006, and May 25, 2006.

The NRC staff reviewed the assumptions, inputs, and methods used by EN to assess these impacts. The NRC staff did independent calculations to confirm the conservatism of the licensees analyses. However, the findings of this SE are based on the descriptions of the analyses and other supporting information submitted by EN.

In accordance with the guidance in RG 1.183, a licensee is not required to re-analyze all DBAs for the purpose of the application, just those affected by the proposed changes. EN considered the following DBA events, which the NRC staff considers applicable:

• Loss-of-coolant accident

• Fuel handling accident

• Main steam line break

• Control rod drop accident The technical evaluation of these events is described below.

3.1.1 Loss-of-Coolant Accident The objective of analyzing the radiological consequences of a LOCA is to evaluate the design of various plant safety systems. These safety systems are intended to mitigate the postulated release of radioactive materials from the plant to the environment in the event that the emergency core cooling system (ECCS) is not effective in preventing core damage. A LOCA is a failure of the reactor coolant system (RCS) that results in the loss of reactor coolant that, if not mitigated, could result in fuel damage, including a core melt. The primary coolant blows down through the break to the drywell, depressurizing the RCS. As the pressure builds in the drywell, steam and other gases expand into the wetwell. Passing through the suppression pool

water, the steam is condensed, thereby reducing the pressure in the wetwell and drywell. A reactor trip occurs and the ECCS actuates to remove fuel decay heat. Thermodynamic analyses, performed using a spectrum of RCS break sizes, show that the ECCS and other plant safety features are effective in preventing significant fuel damage. Nonetheless, the radiological consequence portion of the LOCA analysis assumes that ECCS is not effective and that substantial fuel damage occurs. Appendix A of RG 1.183 identifies acceptable radiological analysis assumptions for a LOCA. The source term and release pathways related to the LOCA are discussed below.

3.1.1.1 Source Term EN projected the core inventory of fission products using the ORIGEN 2 computer code. The resulting core inventories of dose-significant radionuclides were tabulated in Table 4.4-4 to of Enclosure 1 of the September 30, 2004, submittal. These inventories are based upon an adjusted plant-specific pre-1995 ORIGEN 2 run. The three adjustments were:

(1) a scale factor to bound the power level to 3556 megawatts thermal (Mwt), (2) a correction to increase selected krypton values (based on comparisons to other core inventory tables), and (3) an increase in the activity of longer-lived isotopes. The assumed power level of 3556 Mwt is the licensed power increased by 2 percent to account for measurement uncertainties. The ORIGEN 2 computer code is acceptable to the NRC staff for estimating the core inventory. The core inventory used excluded two cobalt nuclides from the RADTRAD inventory file of 60 nuclides. It also added 8 additional nuclides for a total of 66 nuclides.

Because the list of radionuclides is slightly different from the standard default nuclides in RADTRAD, the NRC staff asked EN to confirm that the most conservative radionuclides were used to determine the source for the CGS shielding studies for the shine doses from external sources to the control room (CR), and inhalation doses both for offsite and CR locations. EN provided this confirmation in the March 21, 2006, supplement.

3.1.1.2 Release Pathways The release to the environment is assumed to occur through the following pathways:

• Design leakage of primary containment atmosphere.

• Design leakage through main steam line isolation valves (MSIVs).

• Design leakage from ECCS piping and components that recirculate suppression pool water outside of the primary containment.

Under the previous TID-14844 source term assumption of instantaneous core damage, the initial blowdown would also include all of the released fission products, a fraction of which would be retained by the suppression pool water. Under the AST, a substantial fraction of the fission product release occurs after the initial blowdown is complete. As such, EN did not credit any reduction in fission products transferred to the wetwell air space by suppression pool scrubbing, assuming instead a well-mixed wetwell air space and drywell after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

EN assumes that a portion of the fission products released from the reactor pressure vessel will be removed by drywell sprays. The sprays are assumed to be initiated at 15 minutes and turned off after 1 day.

3.1.1.2.1 Containment Leakage Pathway The drywell and wetwell are projected to leak at their design leakage of 0.5 percent of their atmospheric contents by weight per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 0.25 percent of their atmospheric contents by weight for the remainder of the 30-day accident duration. Leakage from the drywell and wetwell will collect in the free volume of the secondary containment and be released to the environment via ventilation system exhaust or leakage. Following a LOCA, the standby gas treatment system (SGTS) fans start and draw down the secondary containment to create a negative pressure with reference to the environment. The SGTS exhaust is processed through high-efficiency particulate air filter media before being released to the environment.

EN states that, prior to the completion of the secondary containment drawdown, the containment leakage is assumed to go directly to the environment. After the 20-minute drawdown period, filtration of the leakage is credited; however, no credit is taken for the holdup in secondary containment.

3.1.1.2.2 Secondary Containment Bypass Leakages Two sources of containment leakage that bypass secondary containment are MSIV leakage and miscellaneous leakages. The models for these leakages are discussed below. A new limit of 16 standard cubic feet per hour (scfh) per valve, or 64 scfh for four steam lines, at a test pressure of 39.7 pounds per square foot absolute (psia) (25 pounds per square inch gauge (psig)) is proposed in the TS change submitted with the license amendment request (LAR). Since the TS allowable leakage is assessed in units of scfh, and the steam lines are not at standard conditions of temperature and pressure, EN adjusted the assigned flow rates appropriately for the assumed accident conditions.

Credit was taken for natural deposition within the main steam lines. The main steam lines are seismically qualified up to the turbine stop valves. The main steam line piping between the two condensate filter demineralizers is also credited for natural deposition.

MSIV leakage was reduced by a factor of two at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No credit was taken for the main steam line leakage control system. The operability requirements for this system would be removed as part of the proposed TS changes.

To accommodate a postulated single failure of an MSIV to close, credit for natural deposition was taken for only three of the four steam lines. For the three credited lines, natural deposition was calculated according to AEB-98-03, "Assessment of Radiological Consequences for the Perry Pilot Plant Application Using the Revised (NUREG-1465)

Source Term," dated December 9, 1998. A modified Bixler approach for gaseous iodine removal was used. The Bixler model is taken from NUREG/CR-6605, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," dated April 1998. The Bixler model was modified by adopting the AEB 03 well-mixed flow expression for gaseous iodine removal. Proposed credit for organic iodine removal is discussed below.

The second source of bypass leakage, miscellaneous leakage paths, was assumed to equal the proposed TS limit of 0.04 percent primary containment volume per day at

peak accident pressure. The supporting LOCA analysis was based on this limit for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Consistent with the treatment of MSIV leakage, this leakage value was reduced by a factor of 2 at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The NRC staff finds that the bypass models discussed above are consistent with the broad guidance in Appendix A of RG 1.183 and the previously accepted methodology provided in AEB-98-03 and is, therefore, acceptable to the NRC staff.

3.1.1.2.2.1 Main Steam Deposition of Organic Iodine Calculation, NE-02-04-05, "Columbia Offsite and Control Room Doses for LOCA Using AST and NRC Methods," proposes the use of a "Modified Bixler Model for Organic Iodine Removal." During a December 6 and 7, 2005, meeting on the EN AST submittal, the NRC staff discussed this methodology with EN. The NRC staff has determined from the docket information and the discussion of this docketed information that credit for organic deposition cannot be granted at this time, because the removal mechanisms for the deposition of organic iodine are not well understood.

The NRC staff performed a sensitivity study of the impact of crediting organic iodine deposition in the LOCA main steam line deposition analysis. Based upon this study, the NRC staff determined that the assumption of organic iodine deposition, as credited in this analysis, has very little effect on the overall analysis conclusions. This determination is based on plant-specific parameters and is applicable only to CGS.

Therefore, the NRC staff has reasonable assurance that without this credit the dose acceptance criterion for the LOCA will be met.

3.1.1.2.3 Engineered Safeguards Features (ESF) Leakage During the progression of a LOCA, some fission products released from the fuel will be carried to the suppression pool via spillage from the RCS. Post-LOCA, the suppression pool is a source of water for the ESF systems. Since portions of these systems are located outside the primary containment, potential leakages from these systems are evaluated as a radiation exposure pathway. For the purposes of assessing the consequences of leakage from the ESF systems, EN conservatively assumes that all of the radioiodines released from the fuel are instantaneously moved to the suppression pool. This source term assumption is conservative, in that all of the radioiodine released from the fuel is available for both primary containment atmosphere leakage and the ESF system leakage. EN assumes that 10 percent of the iodine in the ECCS leakage becomes airborne and is available for release as 97 percent elemental and 3 percent organic iodine. The release continues for 30 days. The NRC staff finds these assumptions to be consistent with the guidance of RG 1.183 and, therefore, acceptable.

Two sources of potential ESF leakage were included in the release model. The first is ESF system leakage directly into secondary containment. The current design-basis assumes a value of 1 gallon per minute (gpm). Consistent with RG 1.183, this value was increased by a factor of 2. Leakage was assumed to start at 15 minutes after the event.

The second source of potential ESF leakage is into the condensate storage tanks (CSTs).

During the operation of high-pressure core spray or reactor core isolation cooling systems aligned to the suppression pool, radiological impact of leakage into the CSTs through the CSTs suctions and test returns has been evaluated.

EN determined the dose contribution from the CSTs to the CR and offsite locations to be approximately 1 percent and 2 percent of the total dose, respectively. EN judged that these doses are not significant and did not include them in the total dose from a LOCA. Based upon EN's analysis for the CST pathways, the NRC staff agrees that they are not significant to the conclusion that 10 CFR 50.67(b)(2) regulatory criteria are met for the current analysis.

However, in the EN LOCA calculation, LM-0646, Revision 1, EN states that they conform to RG 1.183, Appendix A, Regulatory Position 5.2. This regulatory position considers the impact of design leakage through valves isolating ESF recirculation systems from tanks vented to the atmosphere. Should circumstances change (for example an increase in the operational or assumed leakage to the CSTs) such that the dose from these pathways either increases or becomes significant to the conclusions derived from the total dose from a postulated LOCA, the analyses should be updated to include the CST dose pathway.

3.1.1.2.2.2 Main Steam Isolation Leakage Control System The original design function of the main steam isolation valve leakage control (MSIVLC) system was to minimize the release of fission products via the main steam lines that could potentially bypass containment and the SGTS after a LOCA. The MSIVLC system performed this function by directing MSIV leakage to the SGTS. This leakage was directed to the SGTS by a blower that served to maintain the pressure in the steam lines negative, with respect to atmosphere.

The routing of this leakage to the SGTS provided for filtration of MSIV leakage and its exhaust via the plant stack.

The MSIV leakage in the AST LOCA dose model is assumed to flow directly to the environment without credit for SGTS filtration. Additionally, the MSIV leakage was analyzed as a release from the turbine generator building exhaust, and this provides a conservative /Q compared to a release via the plant stack. EN is planning to deactivate the MSIVLC system during the implementation of the approved AST LAR.

3.1.1.3 Secondary Containment Drawdown Containment integrity is ensured in part by TS 3.6.4.1, Secondary Containment and its associated surveillance requirements. SR 3.6.4.1.4 demonstrates that the SGT has the capability of drawing down the secondary containment to negative 0.25 inches w.g. in a two minute period of time. Changes which reduce containment integrity would be detected by increases in draw down time in successive testing until ultimately the test would fail.

SR 3.6.4.1.5 demonstrates that the SGT is capable of maintaining a negative pressure in the secondary containment for a period of time at a reduced SGT fan flow rate of 2240 cfm. An increase in inleakage above the fan flow rate would cause the test to fail and indicate that integrity had degraded. The licensee states in its current TS Bases document [t]he internal pressure of the SGT system boundary [secondary containment] is maintained at a negative pressure of 0.25 inches water gauge when the system is in operation, which represents the internal pressure required to ensure zero exfiltration of air from the building using the 95%

meteorological data. SR 3.6.4.1.5 confirms that the SGT can maintain the negative pressure of 0.25 inches w.g. for an extended period with reduced SGT system flow under benign test conditions and provides reasonable assurance that the negative pressure can continue to be maintained under adverse conditions when the SGT system would be operating at its full design flow of 4800 cfm.

The licensee submitted an analysis based on a GOTHIC model of the secondary containment to evaluate the draw down time under accident conditions considering the effects of adverse temperature and wind conditions. The NRC staff did not perform a comprehensive review of this analysis although it noted that the licensees analysis: (1) considered heat transfer from the primary to secondary containment, (2) assumed adiabatic boundary conditions for the surface of the secondary containment structure exposed to the outside environment, (3) considered secondary containment inleakage, (4) considered heat loads generated within the secondary containment, (5) considered fan performance characteristics in evaluating the depressurization of the secondary containment, (6) considered the assumption of loss of offsite power coincident with a LOCA in a manner consistent with SRP 6.2.3 Secondary Containment Functional Design. The analysis predicted a worse case draw down time of 16 minutes. The licensee increased the draw down time to 20 minutes for use in the LOCA analysis.

The draw down time established through the surveillance programs is 2 minutes based on SR 3.6.4.1.4. The NRC staff finds that the use of 20 minutes as the draw down time in the LOCA analysis is acceptable because it considers the potential increase in draw down time due to accident and meteorological conditions and because it is more conservative than that determined through draw down testing.

3.1.1.4 Reactor Building Volume The reactor building has free volume for dilution, but EN effectively did not credit it for holdup.

For modeling purposes, the SGTS was assumed to have a flow rate of 5,000 cubic feet per minute (cfm). Throughout the accident, EN assumed the reactor building volume to be approximately equal to the assumed volumetric flow from the SGTS in 1 minute. Therefore, a small amount of delay is credited by the licensee.

The NRC staff performed an assessment to determine the impact of the small delay and of not modeling the range of SGTS flow rates allowable by TSs. The methodology used by the NRC staff did not credit mixing or holdup, and used design SGTS flow rates with uncertainty. The NRC staff results did not differ from the EN results significantly, therefore, the NRC staff concludes that the results obtained by the licensee are acceptable.

After 20 minutes, the SGTS filter efficiency for all forms of iodine and for particulates is 99 percent. The filter efficiency is reduced to an effective value of 98 percent based on a filter bypass of 50 cfm.

3.1.1.5 Control Room EN evaluated the dose to the operators in the CR. Both CR remote intakes are normally open.

Following a design-basis LOCA, the CR emergency filtration (CREF) system is automatically actuated by a high drywell pressure, low-low reactor water level or high radiation reactor building exhaust. The CR local intake is automatically secured and the CR pressurization

process begins. Both trains of the CREF system receive a start signal and one or both start, depending on whether a single failure of one train was postulated.

The CR volume models the intake of activity from the environment for the purpose of calculating the dose to the CR operators. For the licensing basis case, one CREF train was assumed to fail at time zero, leaving one train operating at 800 cfm. The assumed CREF filter efficiencies were 95 percent for the gaseous iodine species and 99 percent for the particulates.

The unfiltered inleakage for the single-CREF train scenario was 50 cfm. The CR exit flow rate is the sum of filtered and unfiltered incoming flow rates.

From a single-failure perspective, the assumption of a single failure in the CREF system was conservative since this failure was analyzed as occurring simultaneously with the postulated single failure of an MSIV to close. The dose consequences associated with a single failure of an MSIV to close, bound the consequences associated with a single failure of the CREF and the two failures are independent. Nonetheless, for conservatism, the mitigation of the LOCA with credit for only one CREF train is presented as the licensing basis case.

EN evaluated two additional cases. In these cases, both CREF trains were assumed to start as designed. In the first case, EN assumed that the CR operator secured one of the two trains 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the start of the accident. In the second case, EN assumed that both trains operated for the 30-day duration of the accident. The two-train filtered intake flow rate of 1300 cfm and an unfiltered inleakage of 75 cfm were used for these cases. EN determined that the CR dose calculated for both of these scenarios is bounded by the single-train licensing basis case discussed above. Securing a CREF train (when two trains are in operation following a design-basis LOCA) before 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> could increase the dose to the operator. To preclude this undesirable operator action, the appropriate plant procedure(s) will be revised to prohibit the securing of a CREF train within the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of the design-basis LOCA.

The doses calculated in this AST evaluation are based on the limiting combinations of unfiltered leakages and filtered intake flows coupled with conservatively selected /Q s.

3.1.1.6 LOCA Review The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological impacts of the proposed changes. The assumptions and parameters are in Table 1 of this SE. Based upon the information provided by EN, the NRC staff concludes that the licensee used analysis methods and assumptions consistent with the guidance of RG 1.183. The NRC staff compared the radiation doses estimated by the licensee to the 10 CFR 50.67(b)(2) acceptance criteria and to the results estimated by the NRC staff in its confirmatory calculations. The NRC staff finds, with reasonable assurance, that the licensees estimates of the exclusion area boundary (EAB), low-population zone (LPZ), and CR doses for the LOCA will continue to comply with the criteria.

3.1.2 Fuel Handling Accident This accident analysis postulates that a spent fuel assembly is dropped during refueling. Three cases were evaluated. A drop of a fuel assembly in the reactor vessel cavity over the reactor

core, and in the fuel transfer area (between the reactor vessel and the spent fuel pool) or over the spent fuel pool.

The drop of a fuel assembly in the reactor vessel cavity over the reactor core during refueling is the DBA worst case. At this location, the maximum drop (free fall distance) is approximately 34 feet and fuel pin damage is postulated to occur to both the dropped assembly and to some portion of those assemblies impacted in the reactor core.

The extent of damage for both cases is calculated based on the free fall distance and the resulting kinetic energy of the dropped assembly. This drop is postulated to damage 250 pins (based upon a fuel assembly with an 8 x 8 fuel pin array).

The fission product inventory in the core is largely contained in the fuel pellets that are enclosed in the fuel rod clad. However, the volatile constituents of this inventory will migrate from the pellets to the gap between the pellets and the fuel rod cladding. The fission product inventory in the fuel rod gap of the damaged fuel rods is assumed to be instantaneously released because of the accident. Fission products released from the damaged fuel are decontaminated by passage through the pool water, depending on their physical and chemical form. The fission products released from the pool are assumed to be released to the environment without credit for reactor building filtration, holdup or dilution. The CR was modeled without taking credit for automatic system actuation and the no credit is taken for the reactor building or the SGTS.

Therefore, the normal outside air makeup flow of 1100 cfm continues for the duration of the event and no credit is taken for CREF filters.

The NRC staff questioned why EN did not factor the impact of ingress and egress as stated in RG 1.183 into the total unfiltered inleakage into the CR and whether the value of 1100 cfm was appropriate. EN's response, dated March 21, 2006, to NRC Question 19, stated that the 10 cfm for ingress and egress, and increasing the 1100 cfm by 50 percent would not impact the radiological dose results by greater than 0.9 percent. EN also presented the results of a study that calculated the dose at a very large unfiltered inflow rate of 100,000 cfm to demonstrate that the dose results are relatively insensitive to these large values of unfiltered inleakage.

The licensee assumed that 10 cfm for ingress and egress should not be included and that the normal intake value of 1100 cfm is an appropriate assumption for future analysis. The NRC staff did not use this assumption to perform its own calculations, and the overall analysis conclusions are not impacted up to a value of 100,000 cfm. Based upon the results of the EN calculation using 100,000 cfm unfiltered inleakage, the NRC staff has reasonable assurance that the dose acceptance criterion for the FHA is met.

3.1.2.1 Fuel Transfer Area or Spent Fuel Pool Drop EN's evaluation of a second case involving a drop in the fuel transfer area (between the reactor vessel and the spent fuel pool) or over the spent fuel pool. EN stated that the postulated activity released would be substantially lower than the limiting case described above based on the following:

The maximum credible drop height is 17 inches. After a drop height of 17 inches, the kinetic energy available to cause fuel damage is substantially reduced. The number of pins damaged in the design-basis drop would bound the number of pins damaged in a drop elsewhere as the

drop height is significantly greater in the licensing basis case. The TS minimum required water depth over the point of fuel assembly impact is approximately 22 feet, just 1 foot lower than the 23 feet, upon which a decontamination factor (DF) of 200 is based. The difference in water height is approximately 1 percent for normal water level conditions (22 feet, 9 inches) and a maximum difference of approximately 4 percent for the minimum TS water level (22 feet).

The drop height of 17 inches is limited by procedural controls. In accordance with Licensee Controlled Specification (LCS) 1.9.1, the top of active fuel in an assembly must be maintained at least 7 feet, 6 inches below the TS minimum required water level of 22 feet. Based on the comparable water depth available for decontamination and the difference in the postulated drop distances, EN concluded that the consequences of an FHA over the reactor cavity bound those for an FHA over the transfer area or over the spent fuel pool.

In a EN response, dated March 21, 2006, to NRC Question 14, EN provided further clarification of the second case. A rod drop analysis was performed to provide a quantitative assessment of the source term. A drop height of 4 feet was analyzed and the resulting number of rods failing was calculated to be 83 rods. The water height above the release point (assuming the assembly is laying flat) is 22 feet - 5.5 inches (assembly width) = 21 feet, 6.5 inches. Based on the comparable water depth available for decontamination and the difference in the postulated drop distances, EN concluded that the consequences of an FHA over the reactor cavity bound those for an FHA over the transfer area or over the spent fuel pool spent fuel pool.

3.1.2.2 Technical Specifications EN proposed changing TSs 3.6.4.1, "Secondary Containment," 3.6.4.2, "Secondary Containment Isolation Valves," and 3.6.4.3, "Standby Gas Treatment System," by deleting "During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS" from the applicability statements. EN also proposed that footnote (b) be deleted from TS Table 3.3.6.2-1. EN justified the changes by stating that secondary containment is not credited for the mitigation of the FHA. The need to ensure the operability of this system during core fuel handling activities or core alterations is no longer necessary.

The NRC staff requested that EN provide confirmation that the analyzed configuration provided the most bounding atmospheric dispersion factors for all possible release paths (since containment was not credited). In an EN response, dated March 21, 2006, to NRC Question 22, EN provided this confirmation. EN stated that, since the analyzed source is closest to the local intake, no other point on the reactor building produces a higher /Q. Based upon this limiting /Q and the summary that follows, the NRC staff finds these changes acceptable.

3.1.2.3 FHA Review The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological impacts of the proposed changes. The assumptions found acceptable to the NRC staff are presented in Table 1 of this SE. Based upon the information provided by EN, the NRC staff finds that the licensee used analysis methods and assumptions consistent with the guidance of RG 1.183, except were discussed and accepted above. The NRC staff compared the radiation doses estimated by the licensee to the 10 CFR 50.67(b)(2) acceptance criteria and to the results estimated by the NRC staff in its confirmatory calculations. The NRC staff finds,

with reasonable assurance, that the licensees estimates of the EAB, LPZ, and CR doses for the FHA will continue to comply with these criteria.

3.1.3 Main Steam Line Break (MSLB)

The postulated MSLB accident is a double-ended break of one main steam line outside the primary containment. The assumed displacement of the pipe ends permits a maximum blowdown rate. The mass of coolant released is the amount in the steam line and connecting lines at the time of the break, plus the amount passing through the MSIVs prior to closure (6 seconds1). A total of 130,000 pounds mass (lbm) of blowdown (105,000 Ibm of liquid and 25,000 lbm of steam) is released as documented in the current licensing basis. The quantity of blowdown is not affected by the application of the AST methodology to this event.

The release of steam to the environment resulting from the MSLB is assumed to be an instantaneous ground level puff. EN stated that the methodology used to establish the puff transit time and the normalized concentration as a function of distance traveled is consistent with RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants." The initial volume of the puff is established by the amount of steam released by the MSLB and by flashing a portion of the entrained liquid.

The volume of the puff was calculated to be approximately 5.9E4 m3.

3.1.3.1 Source Term The fission product inventory available for release was based on the maximum equilibrium reactor coolant dose equivalent iodine 131 (DEI-131) concentration of 0.2 microcuries per gram (Ci/gm). This is the limit specified in TS Limiting Condition for Operation 3.4.8. In addition to the maximum equilibrium case, RG 1.183 specifies a pre-accident iodine spiking case. To account for iodine spiking, the equilibrium level of DEI-131 was increased by a factor of 20 to achieve a spiking concentration of 4.0 Ci/gm. No fuel damage was postulated for the MSLB.

The activity (in the terms of DEI-131) in the mass of the initial liquid blowdown was assumed to be released to the atmosphere instantaneously, as a ground level release, and no credit was taken for plateout, holdup, or dilution within facility buildings. For example, the DEI-131 total activity release for the iodine spiking case is 4 Ci/gm x 105,000 Ibm (mass of the initial liquid blowdown) x 454 gm/Ibm /1 E6 Ci/Ci =191 Ci.

The NRC staff questioned why EN did not factor in the iodine that would be contained in the 25,000 lbm of steam. EN's response, dated March 21, 2006, to NRC Question 26, stated that neglecting the iodine in the 25,000 lbm of steam was more than compensated by other conservative assumptions. Although the NRC staff believes the steam activity should be added into the total activity released, the NRC staff24 hours 2.3E-4 Breathing rate, control room, m3/s 3.5E-4 Unfiltered inleakage due to ingress and egress, cfm 10 Control room unfiltered infiltration in emergency pressurization mode, cfm2 One train in service 50 Both trains in service 75 Control room normal intake flow, cfm 11003 Control room filtered pressurization minimum flowrate, cfm One train in service 800 Both trains in service 1300 Control room filtered recirculation, cfm none Control room volume, ft3 214,000 Control room intake filter efficiency, %

Particulates 99 Elemental and Organic iodine 95 Noble Gases 0 Control room occupancy factor 0-24 hrs 1.0 1-4 days 0.6 4-30 days 0.4 2

Includes 10 cfm unfiltered inleakage for ingress and egress. The 10 cfm unfiltered inleakage is added to the measured unfiltered inleakage to determine the total unfiltered inleakage.

3 For the FHA and CRDA, emergency pressurization is not credited. The total unfiltered inleakage is assumed to be 100,000 cfm based upon a sensitivity study performed by the licensee. For the MSLB, the dose is taken at the CR intake and is not dependent upon the flow into the CR.

Loss-of-Coolant Accident (LOCA)

Containment Leakage Source Onset of gap release phase, min 2.0 Core release fractions and timing-Containment atmosphere Duration, hrs 0.5 1.5 Noble Gases: 0.05 0.95 Halogens: 0.05 0.25 Alkali Metals: 0.05 0.20 Tellurium: 0.00 0.05 Strontium: 0.00 0.02 Barium: 0.00 0.02 Noble Metals: 0.00 0.0025 Cerium Group: 0.00 0.0005 Lanthanides: 0.00 0.0002 Iodine species distribution Aerosol 0.95 Elemental 0.0485 Organic 0.0015 Primary containment volume, ft3 Drywell 200,540 Suppression pool air space 144,184 Suppression pool water volume 137,262 Containment leakrate, %/day 0- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.5 Greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.25 Standby Gas Treatment System (SGTS) Filter Effective Efficiency, %

Before drawdown (all species) 0 After drawdown (all species except noble gases)4 98 After drawdown (noble gases) 0 Secondary containment volumetric flow rate bypassing SGTS filters, cfm After drawdown 50 Secondary containment bypass, (% primary containment volume/day) 0.04 Secondary containment volume5, ft3 5,000 SGTS drawdown time, min 20 4

The SGTS filter efficiency for all forms of iodine and for particulates is 99 percent. A filter bypass of 50 cfm was also assumed. This reduces the effective filter efficiency to a value of 98 percent.

5 The SGTS was assumed to have a flow rate of 5,000 cfm. The volume is artificially set to SGTS flow rate volume exhausted in 1 minute to represent no mixing or holdup credit in the secondary containment.

Drywell natural deposition Particulate None Elemental None Control room isolation delay, minutes 0 Spray Initiation Time, minutes 15 Aerosol Drywell Spray Removal Rates Time, hr Removal Rate, 1/hr 0 0.00 0.25 6.20 2.44 0.62 24.0 0.00 Main Steam Line Isolation Valves (MSIV) Leakage MSIV technical specification leak rate6 at test pressure of $25 psig, scfh One line 16 Total 64 Normal Steam line (and steam) temperature, EF 544.0 Engineered Safety Feature (ESF) Leakage Iodine species, %

Particulate/aerosol 0 Elemental 97 Organic 3 Iodine flash fraction 0.1 SGTS charcoal filtration Credited ESF estimated leakage into secondary containment, gpm 17 Leakage to CST, gpm 0.24 Atmospheric Dispersion Factors /Q values, sec/m3 Control Room Submittal Table 4.4-2 and 4.4-3 Amended Submittal8 Offsite Submittal Table 4.3-5 6

MSIV leakage is reduced 50 percent after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7 Both the ESF and CST leakage are doubled per Regulatory Guide 1.183. Values not corrected for accident conditions such as the ESF leakage, are taken to be at accident conditions.

8 Table 4.3-4 of the original submittal is amended by the licensees response to the RAI question number 33, dated March 21, 2006.

Fuel Handling Accident Peaking factor 1.7 Fuel rods damaged, rods9 250 Decay period, hrs 24 Pool decontamination factor Iodine 200 Noble Gases 1 Particulate Infinite Fraction of core in gap I-131 0.08 Kr-85 0.10 Other iodines 0.05 Other noble gases 0.05 Alkali Metals10 0.12 Release period, hr 2 Release location SGTS operating SGT stack No SGTS RB vent stack Control Room Emergency Filtration (CREF) initiation Not Credited Control room normal unfiltered intake, cfm 1,100 Water Depth, ft $23 Atmospheric Dispersion Factors /Q values, sec/m3 Submittal Table 4.4-2 and 4.4-3 Main Steam Line Break MSIV closure time, sec 6 Reactor coolant system pressure, psia 1,060 Reactor coolant system temperature, degrees F 552 Reactor coolant activity, Ci/gm dose equivalent I-131 Normal 0.2 Spike 4.0 9

Based upon fuel assembly with an 8x8 fuel pin array, but applied to all fuel types.

10 Cesium and rubidium are present and were considered, but were not included in the calculation because the DF for particulate is assumed to be infinite.

Radioactivity release rate to environment Instantaneous Control Room Occupancy Factor 1 Control room CREF initiation Not Credited Atmospheric Dispersion Factors /Q values, sec/m3 Submittal Table 4.5-1 Control Rod Drop Accident Peaking factor 1.7 Fraction of core Inventory in gap Noble gases 0.1 Iodine 0.1 Other Halogens 0.05 Alkali Metals 0.12 Percentage of core with damaged rod, % 1.8 Percentage of damaged rods that fail, 0.77 Melted fuel release fraction to vessel Noble gases 1.0 Iodine 0.5 Br 0.3 Alkali Metals 0.25 Tellurium Metals 0.05 Ba, Sr 0.02 Noble metals 0.0025 Ce 0.0005 La 0.0002 Fraction of activity released to vessel that enters main condenser Noble gases 1.0 Iodine 0.1 others 0.01 Fraction of activity released from main condenser Noble gases 1.0 Iodine 0.1 others 0.01 Main condenser (plus LP turbine) free volume, ft3 144,000 Release rate from main condenser, %/day 1 Release duration, hours 24 Control room CREF initiation Not Credited Atmospheric Dispersion Factors /Q values, sec/m3 Submittal Table 4.6-1

Columbia Generating Station cc:

Mr. W. Scott Oxenford (Mail Drop PE04) Mr. Dale K. Atkinson (Mail Drop PE08)

Vice President, Technical Services Vice President, Nuclear Generation Energy Northwest Energy Northwest P.O. Box 968 P.O. Box 968 Richland, WA 99352-0968 Richland, WA 99352-0968 Mr. Albert E. Mouncer (Mail Drop PE01) Mr. William A. Horin, Esq.

Vice President, Corporate Services/ Winston & Strawn General Counsel/CFO 1700 K Street, N.W.

Energy Northwest Washington, DC 20006-3817 P.O. Box 968 Richland, WA 99352-0968 Mr. Matt Steuerwalt Executive Policy Division Chairman Office of the Governor Energy Facility Site Evaluation Council P.O. Box 43113 P.O. Box 43172 Olympia, WA 98504-3113 Olympia, WA 98504-3172 Ms. Lynn Albin Mr. Douglas W. Coleman (Mail Drop PE20) Washington State Department of Health Manager, Regulatory Programs P.O. Box 7827 Energy Northwest Olympia, WA 98504-7827 P.O. Box 968 Richland, WA 99352-0968 Technical Services Branch Chief FEMA Region X Mr. Gregory V. Cullen (Mail Drop PE20) 130 228th Street, S.W.

Supervisor, Licensing Bothell, WA 98201-9796 Energy Northwest P.O. Box 968 Ms. Cheryl M. Whitcomb (Mail Drop PE03)

Richland, WA 99352-0968 Vice President, Organizational Performance & Staffing/CKO Regional Administrator, Region IV Energy Northwest U.S. Nuclear Regulatory Commission P.O. Box 968 611 Ryan Plaza Drive, Suite 400 Richland, WA 99352-0968 Arlington, TX 76011-4005 Assistant Director Chairman Nuclear Safety and Energy Siting Division Benton County Board of Commissioners Oregon Department of Energy P.O. Box 190 625 Marion Street, NE Prosser, WA 99350-0190 Salem, OR 97301-3742 Senior Resident Inspector Mr. J. V. Parrish U.S. Nuclear Regulatory Commission Chief Executive Officer P.O. Box 69 Energy Northwest Richland, WA 99352-0069 Mail Drop 1023 P.O. Box 968 Richland, WA 99352-0968 August 2006