Text

CFR 50.59 Summary Report for Period Ending October 6, 2008
	ML082900148
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/13/2008
From:	James D; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	2CAN100802
	Download: ML082900148 (53)
	v • d • e

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-4710 Dale E. James Acting Director Nuclear Safety Assurance 2CAN100802 October 13, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

10 CFR 50.59 Summary Report Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

Dear Sir or Madam:

In accordance with 10 CFR 50.59(d)(2), enclosed is the Arkansas Nuclear One - Unit 2 (ANO-2) 10 CFR 50.59 summary report for the time period ending October 6, 2008. This report contains a brief description of changes in procedures, the facility as described in the ANO-2 Safety Analysis Report (SAR), changes in the ANO-2 Technical Requirements Manual (TRM),

and changes in the ANO-2 Technical Specification (TS) Bases for which a safety evaluation was conducted. The report also contains a description of tests and experiments conducted, if any, which were not described in the SAR, and other changes to the SAR for which a safety evaluation was conducted. A copy of each safety evaluation, both ANO-2 specific and those evaluations that may be common between ANO-2 and ANO - Unit 1 (ANO-1), is included in the attachment to this submittal.

If you have any questions or require additional information, please contact Dale James at 479-858-4619.

Sincerely, Dale E. James DEJ/dbb

Attachment:

ANO-2 10 CFR 50.59 Summary Report

2CAN100802 Page 2 of 2 cc: Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS O-7 D1 Washington, DC 20555-0001

Attachment to 2CAN100802 ANO-2 10 CFR 50.59 Summary Report

ANO-2 10 CFR 50.59 Summary Report 50.59 # Initiating Document Summary 2007-010 EC-1640 ANO-2 TSP Replacement With NaTB 2007-011 OP-2202.002 & Operator Action Providing Margin Associated OP-2202.009 With Possible Vortexing in the RWT 2007-012 EC-704 Installation of HPSI Pressurization System (HPS) 2008-001 CALC-ANO2-NE-08-00001 ANO-2 Cycle 20 Reload Report 2008-002 CALC-ANO2-NE-08-00002 ANO-2 Cycle 20 COLR 2008-003 CR-ANO-2-2008-1078 Revise SAR/TRM Regarding Non-Essential Boration Systems 2008-004 CR-ANO-C-2007-0266 Creation of New TRMs to Capture Appendix R Requirements 2008-007 EC-2244 Containment Building Sump GSI-191 Compliance Acronyms CR Condition Report CALC Calculation EC Engineering Change SAR Safety Analysis Report NE Nuclear Engineering TRM Technical Requirements Manual HPSI High Pressure System Injection COLR Core Operating Limits Report ANO-2 Arkansas Nuclear One, Unit 2 RWT Refueling Water Tank TSP Trisodium Phosphate NaTB Sodium Tetraborate OP Operations Procedure

ANO 50.59 Evaluation Number 2007-010

10 CFR 50.59 EVALUATION FORM Sheet 1 of 6 I. OVERVIEW / SIGNATURES1 Facility: ANO-2 Evaluation # / Rev. #:07-010 / 0 Proposed Change / Document: ANO-2 TSP REPLACEMENT WITH NaTB / EC-1640 Description of Change: This change will replace the existing containment sump pH buffer agent Trisodium Phosphate (TSP) with Sodium Tetraborate (NaTB). NRC Generic Safety Issue GSI-191 resolution has been determined that the postulated large break loss-of-coolant-accident (LBLOCA) blowdown generates debris including calcium silicate (cal-sil) insulation particles that become suspended and dissolved in the sump recirculating fluid. The chemical reaction of the existing TSP with dissolved cal-sil produces chemical precipitates including calcium phosphate and sodium aluminum silicate. These precipitates in containment sump fluid can impede the Containment Spray System (CSS) and Emergency Core Cooling System (ECCS) functions, performance and reliability. Precipitates recirculating in the sump fluid and contacting CSS and ECCS flow paths increase hydraulic head losses and thereby degrade CSS and ECCS performance and reliability. Flow may also be impeded through the reactor building sump strainer due to adherence and clogging of particulates on the screens. Precipitates must be minimized so as to not create failure mechanisms of SSCs. Compared to TSP the replacement pH buffer NaTB will improve the characteristics of the recirculating sump fluid during accident conditions.

Is the validity of this Evaluation dependent on any other change? Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

ANO-2 will submit a License Amendment Request (LAR) for a TS Amendment to revise Containment Systems TS 3/4.6.2.2 including the Limiting Condition for Operation and surveillance requirements.

Approval of the TS Amendment prior to Return to Service of the ECCS and CSS systems following implementation of this change will be tracked via Action Request 00016045-18.

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: Bruce A. Burke / ORIGINAL SIGNED BY BRUCE A. BURKE / DP Eng / DE-MCS / 08-13-2007 Name (print) / Signature / Company / Department / Date Reviewer: Jerry W. Howell / ORIGINAL SIGNED BY JERRY W. HOWELL / EOI / DE-Projects / 08-13-2007 Name (print) / Signature / Company / Department / Date OSRC: Curt Bregar / ORIGINAL SIGNED BY CURT BREGAR / 08-15-2007 Chairmans Name (print) / Signature / Date OSRC Meeting # 07-027 1

Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 2 of 6 II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of Yes evaluation ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question No

8. If No, answer all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS:

Design Basis Accidents (DBAs) previously evaluated for ANO-2 that potentially could be impacted by this change include the large break loss-of-coolant-accident (LBLOCA), small break loss-of-coolant-accident (SBLOCA), the containment design basis accident (CDBA), the main steam line break (MSLB) accident, and the main feedwater line break (FWLB). This is predicated on the fact that these accidents assume ECCS cooling function and/or CSS containment cooling function and/or CSS containment cleanup function. The loss of shutdown cooling event can also utilize CSS operation with manual realignment to provide RCS core cooling during accident conditions.

The pH buffer does not initiate any of these accident scenarios. The pH buffer is a passive chemical additive stored on the containment floor that dissolves upon exposure to the post-accident sump fluid. This dissolution process will not change with the upgraded buffer. Therefore, the frequency of occurrence of any of these accidents previously analyzed in the ANO-2 SAR is independent of the selection of pH buffer.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of Yes a structure, system, or component important to safety previously evaluated in the No UFSAR?

BASIS:

The ANO-2 structures, systems and components (SSCs) important to safety that are affected by this design change include the reactor building (or containment), reactor coolant system (RCS), the high pressure safety injection (HPSI) system, the containment spray system (CSS), the sump pH buffer containers (2T-194A/B/C), the sump strainer, and the sump fluid pH buffer (presently trisodium phosphate, commonly referred to as TSP). The applicable safety functions include cooling of the containment and reactor core, scrubbing and retention of fission products from the containment atmosphere, and minimizing corrosion potential of components within the containment including structural members. The function of the sump pH buffer is to improve environmental conditions within containment to ensure safety functions of these SSCs are sustained.

Chemical precipitates have been identified as adversely impacting ECCS and CSS safety functions.

Precipitates recirculating in the sump fluid and contacting CSS and ECCS flow paths increase hydraulic head losses and thereby degrade CSS and ECCS performance and reliability. Flow may also be impeded through the reactor building sump strainer due to adherence and clogging of particulates on the screens. Sump fluid flows must satisfy the CSS and ECCS design bases to ensure that their safety system functions are satisfied.

Effects to equipment qualifications for electrical equipment and instrumentation considered the change to post-LOCA condition. Adverse effects to elastomer materials were also determined to be negligible for the buffer change. Changes to post-LOCA hydrogen generation rates resulting from the buffer upgrade was within the conservative design basis limitations. Generation of hydrogen from SSC materials in containment exposed to boric acid solution during refueling water tank (RWT) drawdown will not be changed from that with the TSP buffering agent.

(continued)

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 3 of 6 Corrosive reaction with aluminum and zinc metals was decreased with the lower pH range.

Oxidation of zinc metals generates hydrogen and contributes to adverse post-LOCA conditions in the containment atmosphere, but hydrogen generation was bounded by the current design basis. Post-LOCA oxidation of aluminum components in containment supports formation of sodium aluminum silicate precipitate in the sump fluid. NaTB reduced the quantity of this precipitate slightly Corrosion of carbon steels and low alloy steels in the pH range of 7 to 8 will result in a corrosion rate of about 10 mils per year as compared to a corrosion rate of about 1 mil per year in the pH range of 9 to 10. However, most safety function components in containment composed of carbon steel or low alloy carbon steel have a protective coating that inhibits oxidation and corrosion. An exception are the containment structural members, but the increase in corrosion rate has been judged to not adversely impact the mitigating safety function for the 30 day duration of the DBA.

The purpose of this design change is to minimize the potential of chemical precipitates that can impede containment sump fluid flows. The upgrade to a pH buffer that has been determined to be superior to that presently used will sustain those safety functions. Analysis demonstrates that the NaTB buffering agent ensures a post-LOCA containment sump fluid pH 7. Determination of the sump fluid with NaTB predict a long term sump fluid pH range of 7 to 8.

No deleterious effects of the improved pH buffer have been observed or realized that could generate adverse effects or malfunctions of the affected SSCs. Replacing TSP with NaTB, which achieves the same pH buffering requirements, will not increase the probability of a LOCA. Therefore, it is unlikely that this change will increase the likelihood of occurrence of any malfunction of an SSC important to safety.

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No BASIS:

Design basis accidents previously evaluated assumed ECCS cooling function and/or CSS containment cooling function and/or CSS containment cleanup function.

The postulated LBLOCA, and SBLOCA require CSS cleanup function. CSS cleanup function is not required for the postulated CDBA, MSLB, or the FWLB accident scenarios. The postulated LBLOCA requires both CSS cleanup and cooling function. The effect of the upgraded pH buffer on the CSS containment cleanup function (i.e., scrubbing and retention of fission products) was considered as negligible. Scrubbing and retention effectiveness of the spray water is dependent on factors that were not adversely impacted by this design change. The particular buffering agent used to attain the pH 7 is not regarded as a significant factor. Consistent with Standard Review Plan Section 6.5.2, ANO-2 will maintain a post-recirculation pH 7 in order to credit iodine scrubbing by containment spray. CSS spray and retention effectiveness are dependent on several variables including pH, temperature, drop size, drop velocity, and containment volume. CSS cleanup function design basis continues to be bounded by the LBLOCA analysis. As with the previous buffering agents, NaTB will maintain pH 7 in the recirculation water following the postulated LBLOCA and SBLOCA. No adverse impact to the CSS cleanup function or effectiveness are expected as a result of this buffer upgrade.

The postulated CDBA, MSLB, and FWLB accident scenarios are dependent on the CSS cooling function rather than its cleanup function. LBLOCA also requires CSS cooling function. The effect of the upgraded pH buffer on the CSS containment cooling function was considered to be negligible.

The CDBA scenario analyzes the worst case conditions to which the containment integrity function can be compromised. The CSS cooling function design basis continues to be bounded by the CDBA analysis.

NaTB results in a slightly decreased sump fluid density as compared to that with TSP but will still be less than the density with the original sodium hydroxide, NaOH. The effects of density will not challenge the performance or power requirements of CSS and ECCS pumps, motors, heat exchangers, instrumentation, valves, or other system components. The specific gravity comparison was 1.0033 for TSP versus 1.0027 for NaTB, and 1.0066 for NaOH. Affects to instrumentation accuracies was determined to be negligible. (continued)

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 4 of 6 The reduction in precipitate species and quantities minimizes challenges to the safety system performance and reliability. In particular, the potential for sump strainer clogging has been significantly reduced as a result of eliminating the sodium phosphate precipitate and reduction of the quantity of sodium aluminum silicate precipitate. Upgrade of pH buffering agent to NaTB will enhance the reliability of ECCS and CSS systems to perform their safety functions.

The boron solubility with NaTB is slightly decreased as compared to that with TSP, but boron solubility will still be greater than boron solubility with the NaOH pH buffer. In addition, NaTB will provide increased reactivity margin as a result of the additional boron contained in NaTB as compared to TSP. Deterministic evaluations support that long term cooling requirements are satisfied in accordance with 10 CFR 50.46 Criterion 5. Cooling of the reactor core would not be impeded by the change in buffering agent. No adverse impact to the cooling functions or capacities of CSS or ECCS are expected as a result of this buffer upgrade.

In conclusion, doses resulting from challenging the containment cleanup function will not be adversely impacted by this design change to upgrade to the sodium tetraborate buffer for the containment sump fluid. The doses resulting from challenging the containment integrity will not be adversely impacted by this design change to upgrade to the sodium tetraborate buffer for the containment sump fluid.

4. Result in more than a minimal increase in the consequences of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No BASIS:

The estimated dose consequences as a result of a malfunction of an SSC important to safety will not change for accidents that have previously been evaluated. The ANO-2 SSCs important to safety that are affected by this design change include the reactor containment building, RCS, HPSI, CSS, and the containment sump components including the pH buffer containers, the pH buffer, and the sump strainer. The sump components are located on the reactor containment building floor but do not serve any function during normal plant operation. The consequences of using this improved buffer will not result in challenging the safety function of the containment structure, RCS, HPSI, CSS, or any related components.

Sump safety function supports ECCS and CSS operation during the postulated accident conditions mentioned previously. These components perform their safety function when the lower containment volume becomes flooded. The pH buffer then dissolves into the sump fluid by dissolution. The upgraded buffer NaTB was selected to reduce an unfavorable effect of the existing pH buffer TSP.

The intent of this buffer upgrade was to eliminate formation of precipitates that can form in the sump fluid. Formation of precipitates can reduce or block ECCS and CSS flow passages and compromise the dependability of ECCS and CSS operation.

Retention of airborne iodine from the containment atmosphere that has been scrubbed by CSS spray will be maintained with NaTB, thereby minimizing the potential for a radiological release from the containment during accident conditions. This change will not increase the consequences from the accident due to containment atmosphere leakage.

Safety function of affected SSCs will be more dependable with this change that will utilize a superior pH buffering agent. Chemical interactions of the proposed buffer on existing SSC materials in the post-accident environment were considered and determined to be negligible. Although the long term sump fluid pH is reduced slightly below that with TSP, the effects of corrosion of carbon steel and low carbon alloy steels were determined to be negligible for the period that SSCs would be exposed to that condition. ECCS and CSS systems and components will continue to perform their safety functions as designed. There were no other changes to the design or operation of the plant that will affect system safety functions other than sump fluid pH control as a result of upgrading the pH buffering agent.

Therefore, this change in pH buffering agent will not increase the consequence of a malfunction of an SSC important to safety.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 5 of 6

5. Create a possibility for an accident of a different type than any previously evaluated in Yes the UFSAR? No BASIS:

The change has no adverse effect on the safety function of any SSC. This design change will not challenge the dependability, performance or integrity of any safety related SSC. No new accident scenario, accident initiator, or SSC failure mechanism will be introduced as a result of this design change.

The change of pH buffer affects solubility of the boric acid. Whereas the solubility of boric acid is increased when using NaTB as compared to using sodium hydroxide (NaOH), the boric acid solubility with NaTB decreases as compared to the existing pH buffer TSP. Although the margin in boron precipitation in the core during long term core cooling is decreased compared to that with TSP, the margin is increased compared to that with the original pH buffer NaOH. Effects of boric acid solubility on post-LOCA long term core cooling were evaluated. Calculations determined that long term cooling requirements were satisfied. The potential of boron precipitation to occur in the reactor core post-LOCA was considered to be bounded by the current design basis. No other new types of precipitates were observed as part of buffer testing.

SSCs designed to mitigate accident consequences that were previously evaluated remain capable of fulfilling those intended safety functions with this design change. The change from trisodium phosphate to sodium tetraborate has no impact on environmentally qualified equipment. Chemical interactions of the proposed buffer on existing SSC materials in the post-accident environment were evaluated and determined to be negligible. Although slightly increased corrosion rates of carbon steel and low carbon alloy steels materials would be expected from using NaTB as compared to TSP, the effects would be negligible for the period that the SSCs would be exposed to that condition.

Therefore, the change does not create the possibility of a new or different kind of accident from any previously evaluated.

6. Create a possibility for a malfunction of a structure, system, or component important to Yes safety with a different result than any previously evaluated in the UFSAR? No BASIS:

NaTB has been evaluated for compatibility with existing SSCs that perform safety functions. No new failure mechanism was found for this buffering agent to initiate a malfunction of that SSC. Testing and analysis has proven NaTB to be superior as a pH buffer for the sump fluid. ECCS, CSS, and containment sump safety functions are improved with the use of NaTB. The sump strainer has a reduced likelihood of plugging with precipitates formed as a result of the interaction of the pH buffer.

When used as the sump fluid pH buffer, NaTB eliminate formation of the calcium phosphate precipitates that form with calcium silicate dissolved in the sump fluid. No new type of precipitates was observed during testing activities.

Therefore, no different results exist for failures of an SSC important to safety than those previously evaluated.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 6 of 6

7. Result in a design basis limit for a fission product barrier as described in the UFSAR Yes being exceeded or altered? No BASIS:

NaTB is compatible with SSCs that provide a safety function. NaTB will not adversely affect fission product barriers such as concrete, steel, and fuel cladding. NaTB is compatible with zirconium fuel claddings. NaTB in the sump fluid will control the sump fluid pH and thereby provide sufficient retention of radioactive iodine in solution and minimize the potential of an off-site release during accident conditions. Challenges to containment integrity will be minimized with the CSS cooling function being improved as a result of this change. Integrity of the reactor coolant system will not be compromised by the change in buffer. Upgrade of pH buffering agent to NaTB will improve ECCS reliability.

CSS cooling function will be improved as a result of the pH buffer revision. This will maintain the dependability of the CSS safety function to cool the containment atmosphere and thereby decrease the potential of challenging the containment integrity.

Retention of airborne iodine from the containment atmosphere that has been scrubbed by CSS spray will be maintained with NaTB, thereby minimizing the potential for a radiological release from the containment during accident conditions. This change will not increase the consequences from the accident due to containment atmosphere leakage.

Therefore, no design basis limit for a fission product barrier will be exceeded or altered.

8. Result in a departure from a method of evaluation described in the UFSAR used in Yes establishing the design bases or in the safety analyses? No BASIS:

The change in the chemical characteristics of the buffering agent does not introduce new methods in mitigating the consequences of the postulated accident scenarios previously evaluated or in determining the design bases of the systems.

Therefore, the change does not alter or depart from existing methodology for establishing the design basis or as used in the safety analyses.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 3

ANO 50.59 Evaluation Number 2007-011

10 CFR 50.59 EVALUATION FORM Sheet 1 of 5 I. OVERVIEW / SIGNATURES2 Facility: ANO-2 Evaluation # / Rev. #: 07-011 Proposed Change / Document: 2202.003, Loss of Coolant Accident, and 2202.009, Functional Recovery Description of Change:

This change adds steps to manually initiate Recirculation Actuation Signal (RAS) when level is between 13%

and 14%. An NRC inspector has asked for empirical data illustrating that no vortexing occurs during the automatic transfer of Emergency Core Cooling System (ECCS) and Containment Spray suction from the Refueling Water Tank (RWT) to the Containment Sump upon receipt of a Recirculation Actuation Signal (RAS) following a Loss of Coolant Accident (LOCA). Although ANO design basis provides reasonable assurance that vortexing does not occur or that any vortexing would be insignificant with regard to system function, no empirical data exists. ANO has subsequently contracted resources to perform actual vortex testing in a mock-up facility.

The manual actions to be credited will act to ensure manual RAS initiation above the current automatic setpoint of 6% +/- 0.5% RWT level in order to provide additional margin to the RWT level in which vortexing may begin.

This additional margin will also act to offset the amount of air entrainment even if it is discovered that some vortexing may be evident assuming automatic RAS initiation at current setpoint.

Vortexing can lead to air entrainment and subsequent voiding of downstream components. Although pumps and other system components can withstand some amount of air entrainment with little or no impact on maintaining safety function, voiding could progress to the point where pumps or systems can become air-bound or otherwise be rendered incapable of performing their intended safety function. In relation to the RWT, excessive vortexing could bring into question the operability of ECCS and/or Containment Spray systems, both of which are required to be capable of automatic transfer to the Containment Sump following RAS initiation per Technical Specifications (TSs). The proposed procedure changes do not affect the ability of this automatic transfer and, on the surface, it would appear that TS compliance is maintained. However, if sufficient air can be entrained such that the systems are rendered incapable of performing their safety function, then the automatic transfer capability does not meet the operability intent of TSs and TS 3.0.3 would be entered due to loss of two trains of TS-required equipment.

Vortexing does not directly relate to system operability or inoperability. Systems are designed to function normally or within functional margins with substantial amounts of air entrainment. 10% voiding has generically been found acceptable for ECCS systems, although it has been shown at some facilities that up to 25% voiding can occur before system function is significantly challenged.

Vortex eliminators come in various designs. Some industry experience has indicated that crucifix-type vortex eliminators may not be completely effective. ANO-2 has this type device in the RWT. However, the ANO-2 design includes a cover plate, which has a significant affect on aiding in vortex elimination. Therefore, based on current design, high confidence of ECCS system operability is maintained. As discussed above, even if vendor test results indicate vortexing may indeed commence prior to completion of the transfer of ECCS/Spray suction from the RWT to the Containment Sump, the amount of air entrained should not approach levels that bring into question system operability.

Nevertheless, Entergy believes it is prudent to establish additional margin above and beyond that required for minimum operability. Therefore, until testing can be completed and further evaluations, including calculations, can be adequately performed, the Emergency Operating Procedures (EOPs) are being revised to have a dedicated operator manually initiate RAS prior to the automatic setpoint being reached.

2 Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 2 of 5 Should the manual operator action be established, certain assumptions are considered. First, it is assumed that RWT level will be maintained at a higher limit than the 91.7% required by TSs. For example, the 3% margin-to-RAS desired (equating to 9% RWT level) will require operator action to be initiated above this 9% level to provide time for appropriate operator identification of the initiation criteria and performance of manual initiation.

The 9% value provides significant margin to the existing 6% setpoint while maintaining ample margin to the total RWT inventory needed to be transferred to the Reactor Coolant System (RCS) to mitigate LOCA events.

Conservatively, whatever manual initiation setpoint would be chosen to ensure actuation by 9% RWT level, the total RWT volume would be increase by an equal amount, further ensuring that injection resources are maintained. Therefore, the first assumption is that RWT level will be maintained at 97% (current level is above 97%).

A simulator scenario has been completed based on worst case operator response time conditions, i.e., a large break LOCA where RWT level would be decreasing most rapidly. The difference between the TS RWT volume and the aforementioned administrative limit of 97% is approximately 5%. Adding this difference to the desired 9% RAS initiation point, the operator was ordered to begin manual RAS initiation at 14% RWT level, but prior to 13% RWT level. During the scenario, the time to approach 14% RWT level was very long, prompting an temporary increase in the size of the simulated RCS break to much greater than that assumed in the analysis.

The break size was reduced to that expected prior to nearing the 14% RWT level desired. This indicates that there is ample time, even during a large break LOCA, to dedicate an operator to the task of monitoring RWT level and being prepared to manually initiate RAS.

Between 14% and 13% RWT level, the operator began routine initiation of RAS, which requires depressing two buttons on one control room panel and then another two buttons on another control room panel. This evolution took about 30 seconds. Based on the time to perform the actuation and the time it took for RAS-related components to fully reposition, the RWT stabilized at just below 9% level. Although it is unlikely operator action would be delayed based on plant control being well established at this time in a LOCA event, RWT level should not be much less than 9% even if the operator response time was increased to 90 seconds. This is because the majority of the RWT level is lost during the period of actual transition to the Containment Sump. The transition is relatively long (1.5 to 2 minutes) and establishes a direct gravity-drain path from the RWT to the Containment Sump. The amount of RWT level decrease based on the operator response time from setpoint to actual initiation is, therefore, relatively insignificant. In any case, it is apparent that initiation can easily be achieved by the operator such that substantial additional inventory is retained in the RWT as compared to the final level achieved if RAS initiation did not begin until 6% RWT level.

A high pressure makeup system is currently installed to maintain Safety Injection Tank (SIT) inventory and to minimize the possibility of void buildup with the ECCS under Temporary Alteration TA-04-02-002. The TA assumes an RWT inventory of 95% is maintained. This system may remain in service based on crediting reduced instrument uncertainty. The calculation of record has 2.34% random uncertainty allowance to bound maintaining initial RWT level via the indicator and automatic actuation via the bistable. Given Operations logs every shift, if the same indicator is used for maintain initial RWT level as well as initiating manual actuation, then there is essentially no instrument uncertainty applicable and most of the 2.34% instrument uncertainty allowance is then margin. This off sets the TA postulated maximum leak of 5,256 gallons, which equates to 1.1% tank (RWT) level. Additional conservatism is available from the inventory released following RAS initiation up to the time of RWT outlet valve closure.

Based on the above discussion, there is reasonable assurance that the proposed manual operator action can be readily achieved without significant adverse impact to safety analysis injection assumptions or ECCS/Spray train operabilities. To further assess the use of operator aid to gain the additional margin, procedure EN-OP-104, Operability Determinations, and NRC Inspection Manual Part 9900 were referred to with regard to crediting operator action. This assessment is included below. Specific requirements are listed to the left and compliance with the specified requirement is indicated to the right of each. The majority of the basis for compliance with each listed item is included above. The 50.59 evaluation below covers other aspects.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 3 of 5 Ability to recognize criteria (RWT level of 14%) Satisfied Indication and equipment necessary easily accessible (all in control room) Satisfied Complications such as interlocks, auto-resets, etc. (none) Satisfied Repositioning following completion (none) Satisfied Action timing Satisfied Manning requirements Satisfied Procedure guidance Satisfied Training to verify procedure Satisfied Access timing Satisfied Personnel training Planned Based on the long period of time demonstrated on the simulator to reach 14% RWT level during a large break LOCA, manning can be achieved either by the normal shift rotation or time may be available to establish additional staff. Regardless of the method employed, the procedure is being revised to require the operator to be dedicated at the time the RAS pre-trip setpoint is reached (40% RWT level). In addition, the procedure ensures the verification checklist for RAS components be completed immediately following RAS initiation.

Training on the new steps will be performed at the beginning of each crew taking the shift for the 1st time following procedure implementation.

Guidance relating to operator action that has not been pre-approved by the NRC does not permit reliance on operator action for those automatic actions that protect a TS Safety Limit. The two Safety Limits related to RCS inventory are LPD and DNBR. The TSs require protection against exceeding these limits in Modes 1 and 2.

RAS initiation does not normally occur until Mode 4 (and always below Mode 2). In addition, the TS contains Limiting Safety System Settings (LSSS) that are required to ensure Safety Limits will not be exceeded. These are Reactor Protection System (RPS) setpoints (although Emergency Safeguard Features (ESF) actuations may occur at the same setting) and refer the user to the RPS TS (3.3.1.1) if any of the settings are negatively impacted. RAS is an ESF function and its initiator (RWT level) is not included in the LSSS listing. Therefore, establishing operator action for manual RAS initiation does not impact the TS Safety Limits for Modes 1 and 2.

When addressing manual operator action, the NRC is most concerned with ensuring the licensee develops adequate responses to Questions 3 and 5 below, in addition to a concern that a significant reduction in the margin to safety may take place. For the changes proposed, the margin to safety is significantly increased from the perspective that action will be taken conservatively to establish long term cooling while avoiding potential air entrainment concerns. The risk of the operator failing to perform the action will not prevent automatic actuation as currently assumed in the license basis. The risk of the operator taking action prior to reaching the 14% RWT level criteria is minimal, as nearly all control room actions are provided a peer check by a separate individual, prior to the action taking place. Therefore, there is no significant reduction in the margin to safety. The response to Questions 3 and 5 are described below.

In summary, this evaluation provides acceptable justification for the use of operator action to establish long term RCS cooling by commencing manual initiating RAS between an RWT level of 13% and 14%. The manual action will not prevent safety functions from being met or render any TS-related equipment inoperable, and acts to provide additional margin to eliminate concerns of air entrainment into the ECCS and Containment Spray trains as a result of potential vortexing in the RWT.

NOTE: The upper RWT level limit for manually initiating RAS (i.e. 14%) is supported by CALC-93-EQ-2001-03 margins and other Operations administrative actions for maintaining additional RWT inventory (i.e.

RWT Level 97%).

References:

1. NRC Inspection Manual Part 9900

2. EN-OP-104, Operability Determinations

3. ANSI/ANS 58.5-1994, Time Response Design Criteria for Safety-Related Operator Actions

4. NRC Information Notice (IN) 97-78

5. NRC Draft Guide DG-1052, Time Response Design Criteria for Safety-Related Operator Actions EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 4 of 5 Is the validity of this Evaluation dependent on any other change? Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: David Bice / ORIGINAL SIGNED BY DAVID BICE / EOI / Licensing / 10-05-07 Name (print) / Signature / Company / Department / Date Reviewer: Rhouis E. Allen / ORIGINAL SIGNED BY RHOUIS ERIC ALLEN / EOI / Design Eng / 10-05-07 Name (print) / Signature / Company / Department / Date OSRC: Curt A. Bregar / ORIGINAL SIGNED BY CURT A. BREGAR / 10-05-07 Chairmans Name (print) / Signature / Date 07-032 OSRC Meeting #

II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8. If No, answer No all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS: The proposed change to perform manual initiaiton of RAS has no impact on accident scenarios.

The initiation of RAS, whether manual or automatic, is a mitigating function, required to support long term cooling of the RCS. Therefore, the change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the SAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No BASIS: Manual operator action to initiate RAS has no impact on RAS-related components because the action only requires depressing initiation pushbuttons in the control room. The actual reposition of RAS-initiated components continues to happen automatically. As discussed above, failure of the operator to perform this manual action has no significant affect the margin to safety and early performance of the action (prior to reaching 14% RWT level) is not considered credible. Therefore, the proposed change does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the SAR.

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No BASIS: Refer to response to Question 2 and 5. Because the probability of a malfunction occurring is not affected, the proposed change continues to provide the same mitigation capability as assumed in the SAR and has no impact on the consequences of an accident. Therefore, the proposed change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the SAR.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 5 of 5

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the UFSAR? No BASIS: Refer to response to Question 2 and 5. Because the probability of a malfunction occurring is not affected, the proposed change does not result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the SAR.

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes UFSAR? No BASIS: The only new event that could develop would be a result of the operator performing the manual actuation early. Early performance could result in less than desired inventory being delivered to the RCS prior to long term recirculation. However, it is expected that the operator error, if assumed, would occur just above the 14% criteria based on the RWT level. Therefore, it is unlikely that significant impact to assumed inventories in the RCS will result and it is not expected that the required Containment Sump volume to support long term recirculation would be significantly impacted. As discussed above, operator error is not considered credible since there is ample time to establish the action criteria and because normal control room operations require a peer check prior to performing actions. In this event, the dedicated operator would, at a minimum, be required to obtain permission from the Control Room Supervisor prior to performing the action. In addition, the operator error would more directly impact the consequences of an accident previously evaluated and it would generally not be appropriate to assume such as an accident initiator. Therefore, the proposed action does not create a possibility for an accident of a different type than any previously evaluated in the SAR.

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the UFSAR? No BASIS: Refer to response to Question 2 and 5. Because the probability of a malfunction occurring is not affected, the proposed change does not create a possibility for a malfunction of a structure, system, or component important to safety with a different result than any previously evaluated in the SAR.

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being Yes exceeded or altered? No BASIS: The manual action acts to enhance mitigation strategies (such as spray, injection, and long term cooling), by decreasing the potential of air entrainment due to RWT vortexing into the ECCS or Containment Spray trains. The containment structure, fuel cladding, and fuel pellet remain within the assumptions of the analysis, given the proposed action. The integrity of the RCS pressure boundary is not affected by the proposed change. Therefore, the proposed change does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing Yes the design bases or in the safety analyses? No BASIS: Manual initiation of an ESF feature has no bearing on any method of evaluation described in the SAR. The proposed change has been reviewed with respect to ensuring assumptions for required injected inventory, Containment Spray requirements, and long term recirculation requirements are maintained, including the assumption that at least one train of ECCS and Containment Spray will remain functional during the injection and recirculation phases, as required. The method used to verify needed inventories and flow paths are not affected. Therefore, the proposed change does not result in a departure from a method of evaluation described in the SAR used in establishing the design bases or in the safety analyses.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 3

ANO 50.59 Evaluation Number 2007-012

10 CFR 50.59 EVALUATION FORM Sheet 1 of 6 I. OVERVIEW / SIGNATURES3 Facility: ANO-2 Evaluation # / Rev. #: 07-012 / 0 Proposed Change / Document: EC-704 Description of Change: Installation of a new HPSI Pressurization System (HPS)

Is the validity of this Evaluation dependent on any other change? Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: Daniel H. Williams / ORIGINAL SIGNED BY DAN WILLIAMS / EOI / Fuels & Analysis / 11-15-07 Name (print) / Signature / Company / Department / Date Reviewer: Tim Woodson/ ORIGINAL SIGNED BY TIM WOODSON / EOI / SYE / 11-15-07 Name (print) / Signature / Company / Department / Date OSRC: Curt Bregar/ ORIGINAL SIGNED BY CURT BREGAR / 11-15-07 Chairmans Name (print) / Signature / Date 07-036 OSRC Meeting #

II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8. If No, answer No all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS:

The sphere of influence of this change is limited to systems used to mitigate accidents. Because the pressurization pump can only take suction from the RWT, there is no potential to cause a boron dilution event. There is sufficient air delivery capacity in the IA system to drive the pneumatic pumps without adversely affecting any other IA supported systems or increasing the failure rate of the IA system itself.

There is no potential relationship to the initiation of any of the other SAR chapter 15 events.

3 Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 2 of 6

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No BASIS:

HPSI Header Interface The HPSI header interface creates four new potential flow diversion paths per header for HPSI injection flow. Critical flow through a 3/8 inch diameter path at 1400 psia and 100°F is around 185 gpm. With the time to start of post-LOCA recirculation being as long as 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (see CALC-93-E-0021-05) more than 25,000 gallons of injection fluid could end up in the Auxiliary building instead of the Containment if one of these paths is open. This could lead to potential NPSH problems with the HPSI and Containment Spray pumps in both trains after recirculation and to excessive off site doses from the release into the atmosphere of recirculation fluid. Each HPSI header has two safety related check valves in series that isolate the HPSI system from the HPS at each HPSI header. These check valves are in parallel with a drain line that is isolated by two manual globe valves in series and the two parallel paths are tied together both on the HPSI header side of all four valves and between the first and second of the two sets of series valves. The isolations are both hydraulic and seismic to non-seismic class breaks.

TO #1 HPSI HEADER M-2232 SH 1 V

(D-4) 2HPS-1011 3/8 INSTR TUBING

2 HPSI HEADER M-2232 SH 1 (C-4) 3/8 INSTR TUBING LSPPR 2T-21 TANK ROOM 3/8 INSTR TUBING 2DCB 2DCB 2CCB 2CCB 3/8 INSTR 2HPS-39 2HPS-38 TUBING 2DCB-29-1/2 2HPS-33 2HPS-34 2CCB 3/8 INSTR 2CCB-77-1/2 TUBING 3/8 2DCB 2DCB 2HPS-37 2HPS-36 INSTR TUBING 2HPS-31 2CCB 2HPS-32 3/8 INSTR TUBING STR TUBING FD FD EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 3 of 6 These check valves have a safety function to close and these manual globe valves have a safety function to remain closed and intact. The check valve failure to close rate from PRA-A2-01-003S04, Rev. 0, Appendix 1, is 1.00E-3. The common cause failure multiplier shown there is 5.00E-2. Therefore, the probability of one check valve pair failing to close on demand is 5.10E-5. The probability of a manual valve spuriously operating is provided in PRA-ES-001-01, Rev. 1, Table 5, and is 1.7E-6 per hour. The upstream manual globe valves (2HPS-34 and 2HPS-39) could be open and not discovered until the next quarterly surveillance. Using the average time until the next surveillance these valves each have a probability of 1.90E-3 of being open. If either of the downstream manual globe valves (2HPS-32 and 2HPS-37) were open, flow from the HPS pump through the valve would be approximately 2.75 GPM (max flow into atmospheric pressure). Operator rounds/log taking, etc. occurs on an average of every six hours.

These valves are hard piped to the floor drain so it is possible that the operator could miss seeing the flow.

No RWT level alarms would actuate and the rate of RWT level decrease would be too small to detect (it is about 200,000 gallons per percent level in the RWT). However, the pump would be at full speed (40 cycles or 80 strokes per minute) which makes quite a bit of noise and the header pressure would fall since the flow from the pump is being diverted out the open valve, so it is likely Ops would be investigating the situation and would certainly find an open valve within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If HPS-32 or 2HPS-37 were open and the check valves were not working, the results would probably be the same as with the check valves working except header pressure would fall more rapidly. It is unlikely that Ops would initially check the position of the drain valves (although if they were checking the position of the valve for some reason, they would detect the out of position condition by verifying the movement of the stem handle), but they would investigate why the HPSI header pressure has fallen and why the HPS pump is screaming. In summary, the open drain valve would certainly be discovered within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Therefore, the downstream manual globe valves have a probability of 2.04E-5 of being open and the probability both valves in a set of manual globe valves being open is 3.88E-8. The probability that the right combination of check valve and manual valve would provide a path past the boundary must also be considered. The probability of the downstream manual globe valve being open and the check valve nearest the HPSI header failing to close on demand is 2.04E-8. The probability of the upstream manual globe valve being open and the check valve farthest from the HPSI header failing to close on demand is 1.9E-6. Since there are two HPSI headers involved, the total increase in probability of a flow diversion path from either header is 1.06E-4. However, even though the HPS system is non-seismic, the probability of it not being intact, i.e. of flow actually being lost through the check valve farthest from the HPSI header, is small enough to eliminate this path from further non-deterministic consideration. Therefore, only the paths through the downstream drain valve need be considered and the probability of one of those paths being open is 1.18E-7. This is a minimal increase.

HPS Pumping There is no increase in the probability of introducing air into the HPSI header because the 2P-252 pumps are only capable of compressing air in the water lines to 65 psig. HPSI full flow tests during outages with the closure head removed have demonstrated that the header pressure does not drop below 280 psig.

Also note that the pumps discharge check valves were purposefully located away from the pump to prevent air from being pumped into the HPSI headers.

Pumping additional flow into the #2 HPSI header cold leg injection line after aligning for hot leg injection may slightly redistribute flow between the hot leg and cold leg but cannot cause a net reduction of flow in either injection path. Any redistribution between the two paths will be too small to be a confusion factor for the operators.

HPS suction from 2T-3 RWT A break in the non-seismic portion of the HPS could cause a loss of 2T-3 RWT inventory. Administrative limits exist to maintain the RWT level to a level in excess of the safety analysis credited RWT inventory level. These limits are set to protect the values in the bases for Technical Specification 3/4.5.4. The minimum indicated level value in the Technical Specification Bases provides a margin to the Technical Specification required available volume. This margin is adequate to cover level and actuation uncertainties and potential inventory losses through the non-seismic HPS. The maximum flow through this break is 22 GPM (see ER-ANO-2000-3275-003) via gravity flow. This 22 GPM would not reduce the 2T-3 inventory below that required to support accident mitigation functions if the break occurs during a LOCA. The existing RWT level alarm setpoints and the responses to those alarms assure this. The slow drain rate EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 4 of 6 through a broken line will also permit adequate time for the detection of the drop in RWT level through normal monitoring and isolation of the potential leak point with seismically qualified, safety related manual valves.

HELB CALC-06-E-0004-01 analyzed the HPSI header lines for HELB analysis after changing the normal pressure to approximately 600 psig in the HPSI headers. Prior to ER-ANO-2000-3275-003, the U2 SAR stated that the HPSI headers will be above 275 psig no more than 2% of the time. A licensing document change request has already been processed to reflect that the HPSI headers have been analyzed for HELB considerations and found acceptable.

Flooding The new HPS will be installed in the U2 auxiliary building on elevation 335 in Rooms 2054, 2055, and 2085, which are primarily the 2T-21 tank room, the lower south piping penetration room, and hallways areas on this elevation. The maximum potential leakage from a tubing break and gravity drain of the 2T-3 RWT will not damage other equipment important to safety at this or lower elevations. The floor drains in these areas are capable of removing the water from any potential leaks. Flooding effects were also evaluated and found to be acceptable in CALC-06-0004-01.

Other Considerations All wetted materials installed by this EC in the HPS are consistent with design requirements for use with borated water. Most materials are stainless steel or elastomers acceptable for use with borated water.

Where applicable, piping specification SPEC-ANO-M-2555 was used for designing piping and tubing fittings with appropriate tubing details.

By reducing the chance of gas formation in the HPSI system without frequent system venting operations, the likelihood of HPSI system malfunction is slightly reduced.

The discharge pressure capability of the pressurization pump is well below that of the HPSI pumps.

Therefore, operation of the pressurization pump cannot jeopardize the integrity of the HPSI header pressure boundary.

The location of the HPS equipment has been selected such that necessary operator access to equipment important to safety is not impeded.

The temporary pump discharge tubing has an ID of approximately 0.25. This will not allow the check valve ball to leave the pump should it fail. The ball check spring is designed for the conditions that will be experienced and any debris from failure that can pass through the tubing would be small enough such that a failure of HPSI/SIT components will not occur. The injection point for any potential debris would be downstream of the HPSI pumps.

There is sufficient air delivery capacity in the IA system to drive the pneumatic pumps without adversely affecting any other IA supported systems or increasing the failure rate of the IA system itself.

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No BASIS:

Engineering Report 97-R-2002-01 calculates the potential leakage of ECCS recirculation fluid for use in dose analyses. Installation of the HPS adds new potential leakage points. Engineering Report 97-R-2002-01 has been revised to reflect the installation of the HPS. The revision shows that the total potential leakage including the installation of the HPS, 2034 cc/min, is still less than the 2060 cc/min assumed in the LOCA dose analysis for the exclusion area boundary, the low population zone and the control room (ANO-2 SAR §15.1.13.4.1).

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 5 of 6

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the UFSAR? No BASIS:

The consequences of the malfunction of the new HPSI header pressure boundary valves is the same as those of other such HPSI header pressure boundary valves. The RWT sample line valves that are now closed will be open, creating the possibility of a malfunction in their ability to close that previously did not exist. However, the administratively controlled excess RWT level eliminates any consequences of a malfunction in their ability to close. Similarly, the consequence of a malfunction in the open HPSI header interface valves is eliminated by the the new valves (check valves and drain valves) in the extended HPSI header pressure boundary.

The consequences of the new HELB (see the response to question 2 above) are no greater than for HELBs currently considered. From NES-13 and Calc-84-EQ-0001-01, the limiting HELB events for the areas of interest (Rooms 2007, 2009, 2055, 2054 and 2084) is the steam generator blow down line break (SGBD) or the Main Feedwater line break (MFW). The MFW HELB bounds the results from a SGBD for rooms 2007 and 2009. Therefore, only the SGBD event will be discussed. ANO-2 SAR §3.6.4 describes the dynamic analyses of high energy line breaks. The results of those evaluations are affected by input assumptions regarding the enthalpy and pressure of the fluids in the piping in which the breaks are postulated. For the SGBD event, the initial pressure and enthalpy are 885 psig and 535 Btu/lb, respectively. The HPSI header and supply tubing will be pressurized to 635 psig. At saturated liquid conditions, the enthalpy will be 481 Btu/lb. The saturation temperature at this pressure is 495F and since the fluid is from the RWT, the temperature of this fluid will not approach these conditions. Therefore, both the pressure and enthalpy of the fluid in the header and tubing is bound by the initial conditions of the SGBD event. The temperature in these rooms following a SGBD or MFW break is higher than the temperature of the water from the RWT and all rooms reach 100% humidity. Therefore, a HELB of the tubing or the HPSI discharge header would be less severe than the current limiting event. Spray effects from the HPSI header or the HPS could only potentially affect the pumps in the HPSI train in which that header is located and that train of the LPSI and Containment Spray systems. Neither the HPSI nor the portion of the Containment Spray systems that could be sprayed upon are necessary to achieve cold shutdown. Although the LPSI system is necessary to achieve cold shutdown, a break in the HPSI header would not result in protective action. Therefore, a loss of redundancy in the LPSI system from such a break is permissible. Spray from the HPSI header or HPS could also impinge on valves in both trains of the shutdown cooling system. These valves can be manually operated after termination of any spray from these lines in order to achieve cold shutdown. Spray effects were also evaluated and found to be acceptable in CALC-06-0004-01.

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes UFSAR? No BASIS:

The new HPS has no potential to initiate an accident of any kind, previously evaluated or not. See the response to question 1 above.

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the UFSAR? No BASIS:

Draining of the RWT inventory to an inventory sink other than the RCS would be considered to be a malfunction of the RWT that would have a different result than any previously evaluated in the SAR since the RWT has no backup and is not assumed to fail in any existing evaluation. However, the administratively controlled excess in the RWT level assures that any such draining (due to the sample line valves now being open to a non seismic flow path) will not lead to less inventory available for safety injection and containment spray than currently credited in the safety analysis. Changes to the RWT level alarm response procedure assure that a leak in the HPS system would not only be isolated before the TS level limit is reached but while adequate excess is still available to offset any wastage through the HPS that might occur during a LOCA. In the event of a leak in the HPS system, operations would have 10.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months to respond to the RWT Low Level alarm at 95% level before the second alarm occurs at 92%.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 6 of 6 Any potential leakage of recirculation fluid from the HPSI headers through failed or leaking components in the HPS has no different result than potential leakage from the HPSI headers themselves or other attachment to those headers. Any increase in likelihood of such leakage has been shown above to be minimal.

Therefore, there is no different result than any previously evaluated in the SAR.

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being Yes exceeded or altered? No BASIS:

The installation of the new HPS cannot prevent the HPSI system from providing its design basis protection for the fuel cladding. This change has no affect on the design basis limits themselves for any fission product barrier.

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing Yes the design bases or in the safety analyses? No BASIS:

No change in any methods of evaluation described in the SAR is involved with the installation of the HPS.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 3

ANO 50.59 Evaluation Number 2008-001

10 CFR 50.59 EVALUATION FORM Sheet 1 of 7 I. OVERVIEW / SIGNATURES4 Facility: ANO-2 Evaluation # / Rev. #: 08-001 Proposed Change / Document: CALC-ANO2-NE-08-00001,ANO-2 Cycle 20 Reload Analysis Report Description of Change:

This engineering report documents the evaluation of the design and performance of the ANO-2 Cycle 20 Reload core. There are 88 fresh assemblies in the Cycle 20 reload core.

Implementation of Next Generation Fuel Design in Cycle 20.

Is the validity of this Evaluation dependent on any other change? Yes No If YES, list the required changes/submittals. The changes covered by this 50.59 Review cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

The validity of this review is dependent upon the following items:

ANO-2 Cycle 20 Core Operating Limits Report (COLR)

ANO-2 Cycle 20 COLSS and CPC setpoint installation Contingencies listed in Tables 6-10 and 7-3 of the Cycle 20 reload report. (EC-6376 and EC-6378)

The Cycle 20 reload report and its associated 10CFR50.59 review is needed prior to the onload of the Cycle 20 core. The items listed above can be completed after the Cycle 20 core is loaded.

The Cycle 20 COLR does not impose any operating limits on the core until Mode 5 is reached. ANO-2 currently has a procedural constraint on the transition into Mode 5 until the COLR for that cycle has been issued by Licensing to the NRC.

Prior to the reactor trip breakers being closed, the COLSS and CPC constants need to be revised to incorporate the Cycle 20 setpoint analysis. This is due to the fact that the CPCs will provide reactor protection once the logarithmic power level - high bypass is removed.

As noted in the reload report, the contingencies need to be verified prior to the Cycle 20 startup. This is being tracked by EC-6376 and EC-6378.

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: James H. Willoughby / ORIGINAL SIGNED BY JIMMY WILLOUGHBY / ESI / Fuels & Analysis / 3-20-08 Name (print) / Signature / Company / Department / Date Reviewer: Daniel W. Fouts / ORIGINAL SIGNED BY DAN FOUTS / EOI / Fuels & Analysis / 3-20-08 Name (print) / Signature / Company / Department / Date OSRC: Curt Bregar / ORIGINAL SIGNED BY CURT BREGAR / 3-20-08 Chairmans Name (print) / Signature / Date 08-008 OSRC Meeting #

4 Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 2 of 7 II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8. If No, answer No all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No Basis:

The Cycle 20 fuel design changes have been explicitly incorporated in the neutronics models. The Cycle 20 Physics Assessment Checklist (PAC) was evaluated based on the Cycle 20 specific core performance.

Parameters that were not confirmed to be bounded by assumptions employed in the analysis of record (AOR) were evaluated with a Cycle 20 specific analysis or plant operation will be restricted to meet the parameters by incorporation into the COLR or setpoints.

The Batch Z fuel assembly design has changed for Cycle 20. The changes are:

Use of Optimized ZIRLO material for fuel rod cladding and all grids except the top and GUARDIAN grids

Reduced fuel pellet and fuel rod cladding diameters

An Inconel top grid, new mid-grids with I-Springs and intermediate flow mixing grids and mixing vanes on some mid-grids

Use of Stress-Relief Annealed (SRA) ZIRLO material for guide tubes

A new manufacturing method for the grid cage that requires some changes to the lower end fittings, guide tubes, and upper end fitting

Use of bulge joints to connect grid assemblies and guide tubes (vs. welding)

Attaching the lower Guardian grid to the lower end fitting with inserts (vs. welding)

An anti-rotation joint between guide tubes and the upper nozzle The frequency of a fuel handling accident will not be increased. The Cycle 20 assemblies have an equivalent structural cage as that previously used at ANO-2 and will be capable of withstanding the expected handling loads. These assemblies are compatible with the fuel handling equipment. The portion of the top nozzle that interfaces with the refuel and spent fuel machine are unchanged. The manner of handling the new fuel assemblies will be unchanged. The external envelope of the new fuel is the same as previous designs. The grid strap elevation remains unchanged with the exception of the two additional intermediate flow mixing grids. The intermediate flow mixing grids are slightly recessed to prevent contact with the fuel rods in previous designs. The mass of the new assemby design is equal to or slightly lower than the previous designs. The new design is analyzed to be co-resident with the existing fuel designs.

The frequency of CEA misoperation is not increased. The dimensions and positions of the CEA guide tubes are unchanged compared to the assemblies used in the previous cycles. Also, any dimensional changes due to irradiation, such as assembly bow, will not be increased with the changes made in the guide tube material.

The Cycle 20 core configuration does not require any changes to plant equipment or modes of operation.

The initiators to accidents previously evaluated in the FSAR are not affected and the probability of an accident is not increased due to the Cycle 20 fuel design.

Plant operating conditions remain unchanged such that the accident initiators remain unaffected due to the Cycle 20 reload core. Therefore, the frequency of occurrence of an accident is not increased due to this reload.

No changes to plant equipment or modes of operation are required for Cycle 20 due to the reload core.

As discussed previously, there are no impacts to any of the accident initiators due to the fuel design.

Therefore the proposed changes will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 3 of 7

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No Basis:

The Cycle 20 reload core has similar reactivity performance as Cycle 19 (i.e., the increase in reactivity during the first approximately 135 EFPD of the cycle). The critical boron levels, reactivity coefficients, and power distributions are consistent with cycle-to-cycle variations and the expected variation in reactivity behavior early in the cycle. The introduction of the Cycle 20 reload fuel will not require equipment important to safety to be operated in a different manner or with a higher duty. Therefore, the probability of a malfunction of a structure, system or component important to safety is not increased due to the introduction of the Cycle 20 core.

The changes in the grid cage assembly design are considered in the Cycle 20 reload analyses. The Batch Z assemblies are dimensionally and structurally equivalent to previous fuel designs. No changes in the assumptions concerning structure, systems or components availability or failure modes are made.

Therefore the changes to the design of the reload assemblies do not increase the likelihood of occurrence of a malfunction of a structure, system or component previously in the FSAR.

The continued use of past operating characteristics and parameters that are bounded by current safety analyses maintains the plant response to abnormalities or accident conditions within the parameters used as design bases for engineered safety features.

The use of Westinghouse Topical Reports WCAP-16500-P-A and WCAP-12610-P-A to implement NGF does not include any changes concerning the availability of any structure, system or component important to safety. No new failure modes were assumed in any analysis. Therefore, the probability of a malfunction of a structure, system or component important to safety is not increased due to the change in the cladding rupture strain model.

No changes in the assumptions concerning structure, system or component availability or failure modes are made as a result of the proposed changes. The structure, system and components important to safety will be maintained and operated within the design and licensing bases for ANO-2. The proposed changes do not impact the function of the supporting equipment that is important to safety. Based upon the above discussion, the Cycle 20 reload core will not result in more than a minimal increase in the likelihood of a structure, system or component important to safety previously evaluated in the FSAR malfunctioning.

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No Basis:

The Cycle 20 fuel design changes have been explicitly incorporated in the neutronics models. The Cycle 20 PAC was evaluated based on the Cycle 20 specific core performance. Parameters that were not confirmed to be bounded by assumptions employed in the AORs were evaluated with Cycle 20 specific analysis. These evaluations either determined that results of the AORs remained bounding or plant operation will be restricted by the COLR or setpoints such that the results of the AORs remain bounding.

The mechanical design of the Batch Z reload fuel assemblies are structurally equivalent to the assemblies that have been utilized at ANO-2 previously. Fuel performance bounding Analyses of Record were performed on the new design and there is no change in the safety analysis or mechanical analysis related Fuel Performance results for Cycle 20.

The consequences of events for which the AORs bound the Cycle 20 fuel design changes are, by definition, not increased.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 4 of 7 The new ECCS Performance Analyses concluded that the ANO-2 Cycle 20 operation conforms to the acceptance criteria given in 10CFR50.46. The calculation of the post-LOCA dose consequences is based on a set of maximum hypothetical core damage assumptions that far exceed the worst case damage estimates of the ECCS analyses. The assumptions used in the post-LOCA dose consequences were not changed due to implementation of NGF fuel and associated use of Westinghouse Topical Reports WCAP-16500-P-A and WCAP-12610-P-A.

With the exception of a CEA withdrawal from subcritcal conditions, a CEA withdrawal from hot zero power and a CEA ejection event, the Cycle 20 core does not impact any non-LOCA safety analyses input. For the CEA withdrawal events, conservative input data was identified to offset the Cycle 20 exceptions such that the analyses-of-record (AOR) remain valid. The CEA ejection analysis was rerun utilizing new physics parameters that are bounded by the AOR physics limits and no fuel failure was predicted to occur.

Therefore, the radiological dose consequences due to a CEA ejection that were determined for Cycle 16 remain bounding for Cycle 20.

No changes in the radiological release rate / duration, or new release mechanism were postulated and no impact to any of the radiation release barriers occurred due to the Cycle 20 reload core. Based on this, the Cycle 20 core and its operation will not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR.

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the UFSAR? No Basis:

The Cycle 20 fuel design changes have been explicitly incorporated in the neutronics models. The Cycle 20 PAC was evaluated based on the Cycle 20 specific core performance. Parameters that were not confirmed to be bounded by assumptions employed in the AORs were evaluated to be acceptable by Cycle 20 specific analysis or plant operation will be restricted by the COLR or setpoints such that the AORs remain bounding.

The changes in the design of the reload fuel assemblies are considered in the Cycle 20 reload analyses.

The Batch Z assemblies are dimensionally and structurally equivalent to previous reload batches. The fuel performance analysis of the Cycle 20 core has been evaluated using NRC approved codes and all design criteria were confirmed to be met. No changes in the assumptions concerning equipment availability or failure modes were made.

The Cycle 20 fuel design places no greater reliance on any important to safety structure, system or component than has been assumed in the AORs. The AORs in turn have been demonstrated to remain bounding or will remain bounding through application of the COLR limits. Since no greater reliance is placed on any structure, system or component to perform a safety related function, the consequences of malfunction have not increased.

The introduction of NGF in Cycle 20 does not change any assumption concerning equipment availability or failure modes, and does not require new equipment or a change in operating procedures, other than setpoints. No greater reliance is placed on a specific system, structure, or component to perform its design function as a result of this change. Thus, the transient response of the plant to abnormalities and accident scenarios analyzed in the FSAR will not be altered by the implementation of the Cycle 20 core design and as such, all associated accident initiators and any single-failure equipment malfunction postulations remain valid with respect to their impact upon the accident analyses. Specifically, no change will occur in the radiological release rate/duration, no new release mechanisms can be postulated, and no impact will occur to any radiation release barriers.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 5 of 7 The changes introduced as part of the Cycle 20 reload core do not place any greater reliance on a specific structure, system or component to perform a safety function than what has been previously evaluated. The specific changes associated with the cores design, evaluation and operation do not change the manner in which plant systems are operated, and do not change equipment availability, performance or failure modes. Therefore, the Cycle 20 reload core will not result in more than a minimal increase in the consequences of a malfunction of a structure, system or component important to safety previously evaluated in the FSAR.

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes UFSAR? No Basis:

The Cycle 20 reload core has similar reactivity performance as the previous cycle (the increase in reactivity during the first approximately 135 EFPD of the cycle). No fuel or core design changes were made that would introduce any new accident initiators. The critical boron levels, reactivity coefficients, and power distributions are consistent with cycle-to-cycle variations and the expected variation in reactivity behavior early in the cycle. The current fuel storage criticality analysis has been confirmed to be applicable to Batch Z reload fuel. The current storage and handling criteria preclude criticality during normal and postulated events, and this will continue to be the case.

The fuel performance of the fuel designs present in the Cycle 20 reload core was evaluated using NRC approved codes. All design criteria were confirmed to be met. The maximum rod burnup projected for Cycle 20 is less than the limit listed in the ANO-2 license. The fuel performance analysis demonstrated that no change will occur in the radiological release rate / duration, no new release mechanisms can be postulated, and no impact will occur to any of the radiation release barriers.

The introduction of the Cycle 20 reload fuel will not require equipment to be operated in a different manner or with a higher duty. No new failure modes that could lead to an accident of a different type were determined. No initiator to any of the accidents was impacted. Therefore the introduction of the Cycle 20 core does not create the possibility of an accident of a different type.

The use of Westinghouse Topical Reports WCAP-16500-P-A and WCAP-12610-P-A to implement NGF does not introduce any new operating conditions, plant configurations or failure modes that could lead to an accident of a different type than any previously evaluated in the FSAR. The safety analyses do not require the use of new equipment or existing equipment to be operated in a different manner. No initiators are impacted by use of NGF.

The Cycle 20 reload core and associated analyses did not introduce any new structure, system or component that would introduce a new accident initiator or failure mechanism that has not already been considered in the FSAR. There are no new system interactions or relationships that are created by the Cycle 20 reload core. The possibility for an accident of a different type than any previously evaluated in the FSAR will not be created by the introduction of the changes associated with the Cycle 20 reload core.

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the UFSAR? No Basis:

The Cycle 20 reload core has similar reactivity performance as the previous cycle (the increase in reactivity during the first approximately 135 EFPD of the cycle). The critical boron levels, reactivity coefficients, and power distributions are consistent with cycle-to-cycle variations and the expected variation in reactivity behavior early in the cycle. The introduction of the Cycle 20 reload fuel will not require structure, system or component important to safety to be operated in a different manner or with a higher duty.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 6 of 7 The Cycle 20 core design does not modify the design or operation of structures, systems, or components important to safety beyond the fuel itself. Structures, systems and components important to safety will function in the same manner with the reload core as with the Cycle 19 reload core. The changes in core characteristics do not change any parameter that would affect the function of structures, systems or components important to safety. There are no new methods of failure associated with any of the changes associated with the Cycle 20 core.

The FSAR considers the RCS activity that would result from a limited number of fuel failures existing at all times. Regardless of the fuel failure mechanisms, which are not explicitly addressed in the FSAR, the changes introduced by the Cycle 20 core will not lead to a malfunction that is not bounded by the existing assumption on coolant activity. Fuel performance analyses have indicated that the fuel rod performance is bounded by the results of the current AOR. The introduction of the Batch Z fuel assemblies does not create the possibility of an accident of a different type.

Based on the above, the introduction of the Cycle 20 core does not create the possibility of a malfunction of a structure, system or component important to safety with a different result than any previously evaluated in the FSAR.

Structures, systems or components will not be operated in a manner different from how they currently operate due to the use of Westinghouse Topical Reports WCAP-16500-P-A and WCAP-12610-P-A to implement NGF. There are no changes to the failure modes of any structures, systems, or components.

New accident initiators are not assumed and no new equipment is required. This change will not create a possibility for a malfunction of a structure, system or component important to safety with a different result than any previously evaluated in the FSAR.

The Cycle 20 reload core design, analysis or operations do not modify the design or operation of any structure, system or component important to safety beyond the fuel itself. The structures, systems, and components important to safety will function the same as the Cycle 19 reload core. The changes in core characteristics do not change any parameter that would affect the function of structures, systems or components important to safety. Therefore, the Cycle 20 core will not create the possibility for a malfunction of a structure, system or component important to safety that has a different result than those already evaluated in the FSAR.

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being Yes exceeded or altered? No Basis:

The Cycle 20 fuel design changes have been explicitly incorporated in the neutronics models. The Cycle 20 PAC was evaluated based on the Cycle 20 specific core performance. Parameters that were not confirmed to be bounded by assumptions employed in the AORs were evaluated with Cycle 20 specific analysis, or plant operation will be restricted by the COLR or setpoints such that the AORs remain bounding. The protection criteria assumed in the AORs has not changed.

The current fuel storage criticality analysis has been confirmed to be applicable to Batch Z reload fuel. The criticality analysis demonstrated that fuel stored consistent with the Technical Specification limits will maintain a k-effective of 0.95 in the spent fuel racks, containment temporary storage racks and fuel carrier under all conditions. The analysis of the new fuel vault demonstrates that fuel will maintain a k-effective below 0.95 during normal conditions and below 0.98 under optimum moderation conditions. Since these analyses remain applicable to the Batch Z fuel, the design basis limits for the fission product barriers are not exceeded or altered.

The fuel performance of the fuel designs has been evaluated using NRC approved codes and all design criteria were confirmed to be met.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 7 of 7 The Cycle 20 ECCS Performance Analysis, utilizing Westinghouse Topical Reports WCAP-16500-P-A and WCAP-12610-P-A, demonstrated that the results of the LBLOCA, SBLOCA and long term cooling are acceptable. Coolable geometry and long term cooling were both maintained. The change in analysis did not change the limit on allowable clad oxidation.

Analyses have been performed to demonstrate compliance with the design basis limits for the three fission product barriers. The Cycle 20 reload core design and operation is predicted to be conservative with respect to the design basis limits for these barriers. The Cycle 20 reload core changes do not result in a FSAR described design basis limit for a fission product barrier being exceeded or altered.

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing Yes the design bases or in the safety analyses? No Basis:

The Cycle 20 core neutronics analysis was performed using the Westinghouse analysis methods. The applicability of these methods for ZrB2 poisoned fuel has been previously evaluated and approved by the NRC.

The Cycle 20 reload analyses were performed utilizing Westinghouse Topical Reports WCAP-16500-P-A, WCAP-12610-P-A, WCAP-16523-P, CENPD-387-P-A, and CENPD-132. These topical reports provide the analytical methods associated with implementation of NGF and are currently not described in the ANO-2 FSAR. However, a request for approval to use these topical reports for ANO-2 NGF safety analyses was submitted to the NRC via 2CAN070701, dated July 31, 2007. In addition, letter 2CAN070702, dated July 31, 2007, submitted for NRC review and approval the results of the Cycle 20 ECCS Performance Analysis that uses WCAP-16500-P-A. The FSAR will be revised appropriately due to this reload and receipt of the NRC Safety Evaluation Reports associated with these two submittals.

No other new methodologies were introduced during the evaluation and assessments of the Cycle 20 reload core. For implementation of the NGF fuel new analytical methods were performed and as stated above have been sent to the NRC for approval. The FSAR will be revised accordingly upon receipt of the NRC SERs. All other evaluations and assessments continued to use existing FSAR described methodologies for neutronics, fuel performance, transient and accident analyses.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 3

ANO 50.59 Evaluation Number 2008-002

10 CFR 50.59 EVALUATION FORM Sheet 1 of 4 I. OVERVIEW / SIGNATURES5 Facility: ANO-2 Evaluation # / Rev. #: 08-002 Proposed Change / Document: CALC-ANO2-NE-08-00002, ANO-2 Core Operating Limits Report for Cycle 20 Description of Change:

Update the Core Operating Limits Report (COLR) for Cycle 20 operations. Changes from the Cycle 19 COLR consist of:

The cycle specific middle and end of cycle positive MTC limits (safety analysis credited values) have been recalculated for Cycle 20.

LHR limit with COLSS OOS reduced to 13.7 kW/ft.

References 21 through 25 have been added to Section IV.

COLSS Out of Service (OOS) limit lines in Figures 4 and 5 were revised to remove the adjustment made in Cycle 19 that accounted for crediting the ABB-NV CHF correlation for CPC DNBR calculations. This credit is not being taken in Cycle 20.

Is the validity of this Evaluation dependent on any other change? Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Requires NRC approval of 2CAN070701 (Reload Methodologies).

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: Daniel W. Fouts / ORIGINAL SIGNED BY DAN FOUTS / ESI / Fuels & Nuclear Analysis / 3-21-08 Name (print) / Signature / Company / Department / Date Reviewer: Todd A. Erskine / ORIGINAL SIGNED BY TODD ERSKINE / EOI / Reactor Engineering / 3-24-08 Name (print) / Signature / Company / Department / Date OSRC: Curt Bregar / ORIGINAL SIGNED BY CURT BREGAR / 3-27-08 Chairmans Name (print) / Signature / Date 08-010 OSRC Meeting #

5 Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 2 of 4 II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8. If No, answer No all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS:

The cycle-specific limits presented in the Cycle 20 Core Operating Limits Report (COLR) will ensure that ANO-2 is operated during Cycle 20 in a manner that is consistent with the assumptions used in the safety analyses for this cycle. The appropriate actions required if these limits are violated are in the ANO-2 Technical Specifications and are not being changed. The COLR serves to protect the initial conditions assumed by the accident analyses and has no impact on the initiating event of any accident previously evaluated in the FSAR.

No changes to plant equipment or operating procedures are required for Cycle 20 due to the COLR. As discussed above there are no impacts to any of the accident initiators due to the changes in the operating limits. Therefore, the proposed changes will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No BASIS:

The Cycle 20 COLR only affects the operational limits and ensures that the Cycle 20 core is operated in a manner that is consistent with the assumptions used in the analyses for this core design. No changes in the assumptions concerning the structure, system or component important to safety availability or failure modes were made in the Cycle 20 reload core design. The structures, systems and components will be maintained and operated within the design and licensing basis for ANO-2. The revised COLR does not impact the functions, duty or failure modes of the supporting equipment. Based on the above discussion, the Cycle 20 operational limits presented in the cycle-specific COLR will not result in more than a minimal increase in the likelihood of a structure, system or component important to safety malfunctioning.

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No BASIS:

The operational limits in the COLR ensure that the unit is operated in Cycle 20 in a manner that is consistent with the assumptions used in the safety analyses for Cycle 20. The required actions, if these limits are violated, are in the ANO-2 Technical Specifications and are not being changed. Changing the limit for Linear Heat Rate and DNBR margin operating limit based on CPCS with COLSS out of service does not result in a change in the evaluated consequences of accidents. Compliance with the revised COLR preserves the initial conditions of the accident analyses and ensures the evaluated consequences of accidents remain bounding. The analyses for the reload have been performed with NRC approved methodologies to ensure the Specified Acceptable Fuel Design Limits (SAFDLs) will not be violated and the dose consequences are bounded by the results of the licensing basis analyses. Therefore, the consequences of an accident previously evaluated in the FSAR will not be increased.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 3 of 4

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the UFSAR? No BASIS:

The Cycle 20 COLR did not require any changes to the assumptions concerning structure, system or component availability or failure modes. This document does not involve any changes in the structures, systems or components. In addition, the COLR does not impact the overall function or duty of the structures, systems or components important to safety. These operational limits will not result in a change to the evaluated consequences of the accidents, which also include consideration of all relevant structure, system and component malfunctions. Therefore, the consequences of a malfunction of a structure, system or component important to safety will not be increased.

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes UFSAR? No BASIS:

The operational limits provided in the COLR will ensure that ANO-2 is operated during Cycle 20 in a manner that is consistent with the assumptions used in the Cycle 20 safety analyses. The COLR does not create an additional failure mode than what has already been analyzed. No initiators to any of the accidents are impacted by this document. No new operating conditions or plant configurations are created that could lead to an accident of a different type than any previously evaluated in the FSAR. Based on the above, the possibility of an accident of a different type than any previously evaluated in the FSAR will not be created.

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the UFSAR? No BASIS:

No changes in the failure modes of the structures, systems or components important to safety are assumed in the Cycle 20 COLR. No new operating conditions or plant configurations are created that could lead to a malfunction of structures, systems, or components of a different type than any previously evaluated in the FSAR. Therefore, the possibility of a malfunction of a structure, system or component important to safety with a different result than previously evaluated in the FSAR will not be created by this document.

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being Yes exceeded or altered? No BASIS:

The Cycle 20 reload safety analyses were performed to demonstrate compliance with the design basis limits for the three fission product barriers. The revised COLR does not alter these design basis limits.

The operating limits presented in the Cycle 20 COLR will ensure that the Cycle 20 core is operated in a manner that is consistent with the Cycle 20 safety analyses assumptions and in a manner that ensures the design basis limits for the fission product barriers are not exceeded. Based on this, the COLR for Cycle 20 does not result in a FSAR described design basis limit for a fission product barrier being altered or exceeded.

EN-LI-101-ATT-9.1, Rev. 3

10 CFR 50.59 EVALUATION FORM Sheet 4 of 4

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing Yes the design bases or in the safety analyses? No BASIS:

ANO-2 Technical Specification 6.6.5 lists the NRC approved methodologies that are to be used to determine the core operating limits that are presented in the COLR.Section IV of the COLR provides details on which revisions and supplements of the Technical Specification listed methodologies are used to demonstrate compliance with the safety analyses and Technical Specification limits. The listing is being modified for ANO-2 Cycle 20 and a license amendment request (2CAN070701) has been submitted to the NRC seeking approval for use of new topical reports in the determination of core operating limits. The Cycle 20 limits were developed using these additional methodologies. Therefore, NRC approval of that submittal is required prior to implementation and use of the ANO-2 Cycle 20 COLR.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 3

ANO 50.59 Evaluation Number 2008-003

10 CFR 50.59 EVALUATION FORM Sheet 1 of 4 I. OVERVIEW / SIGNATURES6 Facility: ANO-2 Evaluation # / Rev. #: 08-003 Proposed Change / Document: Revise TRM and SAR requirements for non-essential Boration Systems Description of Change:

The NRC approved relocation of the charging pumps and boration systems (boric acid pumps, tanks, valves, and attending components) to the TRM via TS Amendment 229, dated January 31, 2001. The basis for this relocation was that the SSCs were not credited in the safety analysis for accident mitigation and, therefore, did not meet the criteria of 10 CFR 50.36 for inclusion in the TSs. The charging pumps and boration systems are important in that they provide the ability to makeup for RCS inventory losses and reactivity control during normal power and shutdown operations. The SSCs can also provide a makeup source following an accident, although they are not required in this respect (the HPSI pumps are credited as the ultimate means of makeup and boration for accidents, and can be used to support plant cooldown). Based on these facts, requiring two redundant (separate flow paths and electrical supply trains) charging pumps or systems to be operable in Modes 1-4 is no longer necessary. The SAR and TRM are being updated to clarify that redundancy for these SSCs is not required. The TRM will still require two of each related SSC. The changes will permit, for example, the two required charging pumps to be powered from a single train. Additionally, the changes will no longer require both a normal and emergency power source to maintain equipment operable.

In describing the ANO-2 application to GDC 33, SAR Section 3.1.4 states:

Reactor coolant makeup during normal operation is provided by the CVCS which includes three positive displacement charging pumps rated at 44 gpm each. The design incorporates a high degree of functional reliability by provision of redundant components (three pumps), and an alternate path for charging. The charging pumps can be powered from either onsite or off-site power sources, including the emergency diesel generators. The system is described in Section 9.3. It is not the function of the CVCS to provide protection against small breaks; this safety function is provided by the safety injection system. The CVCS does have the capability of replacing the flow loss to the containment for leaks in the reactor coolant piping up to 0.30-inch equivalent diameter with only one charging pump available. However, loss of this CVCS capability in no way compromises the safety of the reactor plant.

The proposed changes do not make the above SAR excerpt untrue. System design is not being changed.

SAR Section 8.3.1.1.8.3 states:

The design criterion governing the assignment of extra redundant loads (the third high pressure safety injection pump, the third service water pump, and the third charging pump and associated valves) is to ensure the availability of one of each type pump in each ESF channel under all conditions. Prior to rendering any one of the three pumps inoperable, administrative controls require that the remaining two pumps be connected to the redundant buses.

Although the above SAR excerpt is intended to discuss design related to EDG loading, the wording could be interpreted to imply that two redundant charging pumps must be available when removing the swing (3rd) pump from service. These statements are being revised because 1) the SAR is not normally the document containing detail of actions taken when equipment is removed from service and, 2) the current TRM permits two charging pumps to be removed simultaneously. The proposed change to this SAR section is attached.

(continued) 6 Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 2 of 4 ULD-2-SYS-33 (CVCS) designates the boration and CVCS (includes charging pumps) as having a regulatory/safety significant function. The description provided for this function is:

1. Functions listed within this category are less significant than the safety related functions but are important because of regulatory significance beyond the design bases requirements. Functions required for Technical Specification "operability" of non-Q equipment will fall in this category. Functions of systems which backup the safety related functions of other systems might be included in this category.

2 These functions should include only those non-safety related functions for which a specific regulatory commitment exists. Note that although the SAR frequently describes system operational features, the inclusion of those features does not, in itself, constitute a regulatory commitment. Any system function described in the SAR (or any other formal NRC correspondence) for any purpose other than normal operation should be considered for inclusion here. Provide an appropriate documentation reference.

The proposed changes will not prevent the subject boration systems and charging pumps from continuing to be included in this functional category. No changes to the ULD are being proposed. These SSCs will continue to be maintained in a manner commensurate with their importance to safety.

The TRM bases speak of boration systems, but do not differentiate between flow paths, pumps, etc. The bases are being revised to clarify the normal use of the boration systems (including the charging pumps) and provide discussion of operability requirements, especially with regard to charging pumps. A markup of the proposed TRM changes is also attached.

Although discussed elsewhere, no other SAR sections or LBDs are impacted by the proposed changes. In addition, the changes are not associated with any test or experiment either described or not described in the SAR.

Is the validity of this Evaluation dependent on any other change? Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: David Bice / ORIGINAL SIGNED BY DAVID BICE / EOI / Licensing / 05-08-08 Name (print) / Signature / Company / Department / Date Reviewer: Robert Clark / ORIGINAL SIGNED BY ROBERT CLARK / EOI / Licensing / 05-08-08 Name (print) / Signature / Company / Department / Date OSRC: Bob Eichenberger / ORIGINAL SIGNED BY J.R. EICHENBERGER / 05-15-08 Chairmans Name (print) / Signature / Date 08-015 OSRC Meeting #

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 3 of 4 II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8. If No, answer No all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS:

The operation of the non-essential boration SSCs is not associated with any accident initiator, other than a dilution accident. The system must be in service to impact dilution accident assumptions. Because the changes only impact the probability that the system would not be available, the changes can have no impact on the boron dilution accident. Therefore, the proposed changes do not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the SAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No BASIS:

A malfunction of the non-essential boration SSCs, whether due to loss of supporting equipment (such as electrical power sources) or other equipment failure is not impacted by the proposed changes. The likelihood of any such malfunction remains unchanged. By only requiring either a normal or emergency power source be available to these SSCs, the likelihood of the loss of the SSC is increased if no backup (normal or emergency) power source is available when the connected power source is lost. Nevertheless, the probability of a malfunction of either power source is not impacted by the proposed changes. In addition, these SSCs are not important from the aspect of preventing an accident or credited in accident mitigation. Therefore, the proposed changes do not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the SAR.

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No BASIS:

The accident analysis credits the HPSI pumps as the accident mitigation source for both long term reactivity control and short and long term inventory control of the RCS. The changes impact the potential unavailability of the subject SSCs and cannot impact the boron dilution accident assumptions, since the system(s) must be in operation for a dilution to occur. The changes to the non-essential borations SSCs do not impact the accident analysis or any system credited in the accident analysis. Because these SSCs are not relied upon for mitigation, the proposed changes do not result in more than a minimal increase in the consequences of an accident previously evaluated in the SAR.

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the UFSAR? No BASIS:

A portion of the proposed changes would permit two required SSCs (for example, two charging pumps) to be aligned to a single electrical power train. This change could result in an increased probability of SSC loss upon a loss of offsite power concurrent with a single failure of the EDG to which the SSC is aligned.

In addition, the change permitting these SSCs to be considered operable without both a normal and emergency power source could lead to the same scenario. However, the loss of these SSCs does not result in an immediate automatic unit trip, nor would their loss result in the ability to mitigate any accident or prevent plant cooldown during non-accident events. Operations procedures for loss of charging act to offset the immediate effects of the loss of these SSCs in that time is available to re-establish a source of (continued)

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 4 of 4 normal makeup prior to forcing a shutdown of the unit (due to decreasing Pressurizer level cause by RCP controlled bleedoff losses). If not restored in a timely manner, plant procedures provide for a shutdown of the reactor and depressurization to permit the use of the credited source (HPSI) for RCS makeup. In addition, it is not likely the non-essential boration SSCs will be operated in a configuration which removes redundancy over the long term and, the probability of a loss of offsite power event in conjunction with the loss of the specific EDG aligned to the operating boration SSC (smart failure) is minimal. Therefore, the proposed changes do result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety previously evaluated in the SAR.

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes UFSAR? No BASIS:

The SAR accounts for the loss of the subject SSCs. Such a loss is unrelated to accident initiators.

Therefore, the proposed changes do not result in the possibility for an accident of a different type than any previously evaluated in the SAR.

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the UFSAR? No BASIS:

The SAR accounts for the loss of the subject SSCs. These SSCs are not credited in the accident analysis and not assumed to be available for accident mitigation. Therefore, the proposed changes do not create the possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the SAR.

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being Yes exceeded or altered? No BASIS:

As previously discussed, the subject SSCs are not credited in the accident analysis. Because they are not relied upon for accident prevention or accident mitigation, the loss of the subject SSCs cannot challenge a fission product barrier. Therefore, the proposed changes do not result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered.

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing Yes the design bases or in the safety analyses? No BASIS:

The proposed changes are unrelated to any method of evaluation with regard to the accident analysis.

The changes act to only define operability alignment requirements for the subject SSCs. Therefore, the proposed changes do not result in a departure from a method of evaluation described in the SAR used in established the design bases or in the safety analyses.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 4

ANO 50.59 Evaluation Number 2008-004

10 CFR 50.59 EVALUATION FORM Sheet 1 of 4 I. OVERVIEW / SIGNATURES7 Facility: ANO - Common Evaluation # / Rev. #: 08-004 Proposed Change / Document:

New fire-related Technical Requirement Manual (TRM) operability, recovery actions, and testing requirements are being developed to replace the existing requirements of both units Safety Analysis Report (SAR)

Section 9D and OP 1000.152.

Description of Change:

In developing the subject TRMs, information was largely relocated from the SAR and/or OP 1000.152. The majority of this activity involved editorial or more restrictive changes as discussed in the accompanying PAD.

Those changes that are considered less restrictive were assumed to potentially have an adverse impact on a design function or method of controlling/evaluating a design function, and are, therefore, discussed in this 50.59 evaluation.

The shutdown action associated with inoperable fire suppression systems was removed and replaced with an action to initiate a Condition Report (CR) to determine impact on continued operation of the unit. During relocation of any Technical Specification (TS) to the TRM, shutdown statements are always removed (this is an industry standard), since it has been determined that the subject equipment being relocated does not meet the 10 CFR 50.36 requirements for inclusion in the TSs. Due to the obvious lower safety significance of the relocated component, it is unreasonable to assume a shutdown is required when a subject component is lost.

Therefore, shutdown actions are replaced with an action to initiate a CR. In so doing, an assessment is made to determine the impact of continued plant operation while TRM-related equipment remains out of service. The conclusions of the assessment may be based on risk insights (qualitative or quantitative or both), other required equipment simultaneously removed from service, alternate means of achieving the support function, compensatory measures to offset the effect of the out of service equipment, etc. On the other hand, the assessment may conclude that a shutdown is appropriate based on accrued risk or the aggregate safety significance of several required components being out of service simultaneously.

In addition to the above, the following test frequencies have been extended:

1. Testing of reactor building fire detection from 6 months to 18 months.

The RB environment and radiological conditions may not be supportive of RB equipment maintenance and test activities between refueling-related intervals. Increasing the interval from 6 to 18 months does not have an impact on RB fire detector operability based on past performance and because they are normally only tested during refueling shutdowns due to the very infrequent occurrence of plant shutdowns for reasons other than refueling operations. In addition, testing activities routinely illustrate continued operability and are not directly associated with component failures. This guidance is consistent with the Standard Technical Specifications (STS) (Reference LCO 3.0.3 Bases for missed surveillances).

2. The fire suppression system bases is modified to state that the backup system must, at a minimum, meet the capacity of the single largest safety-related fire area for which it supports.

The capacity requirement of the backup fire suppression system has not been firmly concluded in the past.

Without such guidance, historically practices have led to a conservative assumption that the backup system should be equal to the full capacity of the installed fire pumps. However, it is unreasonable to assume that a fire will occur in all safety related areas simultaneously during the rare occasions when both installed fire pumps are inoperable. Ensuring that the backup capability to be relied upon can supply the single largest 7

Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 2 of 4 safety-related fire area is sufficient to minimize risk and plant impact until at least one installed pump is installed. Because this has not previously been defined, the change is not necessarily adverse.

Nevertheless, this change is conservatively included in this 50.59 evaluation.

Ensuring the appropriate backup capability may require isolation of non-vital fire water sections and establishing compensatory measures for isolated areas until one of the two high pressure fire water pumps is returned to service. Because at least one pump must be returned to service in a reasonable period of time, the capacity of the backup system described in the bases is sufficient to ensure a fire in any safety-related area can be extinguished via the backup system during this period.

3. Reactor building hose station inspections from 31 days (during shutdown) to 18 months.

The Frequency of visual inspections of fire hose stations located within the RB is changed to 18 months to coincide with expected accessibility of the RB. This current 31-day inspection is not applicable to RB stations during operation, but would become applicable if a shutdown occurred. Because the units rarely experience a shutdown between refueling outages and because the RB is closed during plant operation, there is no need to inspect these stations more frequently than 18 months. This change has no significant impact on system function, given access to the RB is not normally permitted between refueling outages and, therefore, the opportunity for necessary equipment to be removed from the hose stations is minimized.

In addition, testing activities routinely illustrate continued operability and are not directly associated with component failures. This guidance is consistent with the STS.

Note that with the guidance of the Maintenance Rule, equipment failures are trended where appropriate and corrective action taken when a rising trend is detected. Given this information, the proposed testing frequencies listed above can be reduced if failure trends were rising. However, given the passive nature of the fire system, it is not expected that the increased test intervals will have any adverse impact on continued system functionality.

Fire-related systems, structures, and components (SSCs) are not credited in the accident analysis; therefore, their application to the eight 50.59 questions below are limited. However, these SSCs are assumed to be normally available to support a safe shutdown of the unit during a fire event. The provision to consider other options in lieu of a unit shutdown during fire pump inoperabilities will permit appropriate actions that continue to support fire mitigation and minimize plant risk (possibly avoiding shutdown/transient risks). Likewise, requiring a backup system equivalent to the largest single safety-related load will continue to support the intend function of the fire system (safe shutdown during a fire event can still be accomplished). Increasing the test interval for certain fire-SSCs is not expected to impact equipment availability or, subsequently, the ability to safely shutdown the unit during a fire event. Therefore, the changes described above will not result in a significant adverse impact with regard to meeting the intended function of the fire system.

Is the validity of this Evaluation dependent on any other change? Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: David Bice / (see LAR 2007-0204) / EOI / Licensing / 07-08-08 Name (print) / Signature / Company / Department / Date Reviewer: Stephenie Pyle / (see LAR 2007-0204) / EOI / Licensing / 07-08-08 Name (print) / Signature / Company / Department / Date OSRC: John N. Miller / (see LAR 2007-0204) / 07-10-08 Chairmans Name (print) / Signature / Date 08-020 OSRC Meeting #

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 3 of 4 II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8. If No, answer No all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS:

Fire SSCs are intended to detect and mitigate the effects of a fire. Absent a fire, however, the fire SSCs are unrelated to those systems and components credited to prevent or mitigate an accident. A fire, or the loss of any fire SSC, is not considered an accident initiator. Likewise, the established capacity of a backup suppression system is unrelated to accident initiators. The removal of the shutdown requirement does not prevent a unit shutdown, but simply provides an opportunity to establish other corrective action (compensatory measures, risk assessment, establishing fire watches, etc.) to be considered in lieu of a shutdown.

Increasing the testing frequency of certain fire SSCs does not have a direct impact on system availability.

Affected SSC failure rates are monitored in accordance with the guidance of the Maintenance Rule or station programs. In addition, the fire SSCs affected are passive in nature and, subsequently, are unlikely to fail a required test. Finally, fire SSCs are not considered accident initiators.

Therefore, the proposed change does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the SAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No BASIS:

Increasing test frequencies do not result in the loss of any fire SSC. Likewise, the aforementioned changes to fire-related actions or definitions do not impact the probability of a fire which could damage a safety-related component. As discussed above, compensatory measures or other actions designated in the TRM or station procedures are sufficient to ensure fire events will continue to be avoided/mitigated and plant risk minimized. Therefore, the proposed change does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the SAR.

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No BASIS:

The accident analysis does not credit fire SSCs to mitigate the consequences of an accident, nor is a fire assumed to occur concurrent with an accident previously evaluated in the SAR. The TRM and other fire-related procedures and instructions provide ample guidance with regard to appropriate action to compensate for inoperable fire SSCs. As stated in response to Question 1, the action to perform a shutdown is not precluded, but other actions are simply offered that may be implemented such that a shutdown is not necessary. These considerations are sufficient to offset fire-risks and are not required to prevent an accident or credited in accident mitigation. Likewise, the frequency in which fire SSCs are tested has no relation to accident mitigation and are not credited in the accident analysis. Because these SSCs are not relied upon for accident mitigation, the proposed change does not result in more than a minimal increase in the consequences of an accident previously evaluated in the SAR.

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 4 of 4

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the UFSAR? No BASIS:

As discussed previously, fire SSCs are not credited in the SAR, nor is a malfunction of a fire SSC related to the accident analysis. The effect of a malfunction of any safety-related SSC required to support the accident analysis is not impacted by the proposed change. Therefore, the proposed change does result in more than a minimal increase in the consequences of a malfunction of a SSC important to safety previously evaluated in the SAR.

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes UFSAR? No BASIS:

The proposed change does not result in a physical change to the plant or plant SSCs. The modified action, definition of backup system capacity, and the change in test frequencies do not introduce a new accident type. Additionally, the TRM and plant procedures provide instruction for the loss of the subject SSCs to ensure equipment is restored in an acceptable time period to minimize any impact on the fire-related safe shutdown assumptions. The proposed change is only related to a fire event, which is already considered in the design basis. Therefore, the proposed change does not result in the possibility for an accident of a different type than any previously evaluated in the SAR.

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the UFSAR? No BASIS:

The impact of a fire at the station is already considered in the analysis for safe shutdown capability. No change to the design, operation, or function of any SSC (fire-related or otherwise) is proposed. Therefore, the proposed changes do not create the possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the SAR.

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being Yes exceeded or altered? No BASIS:

As previously discussed, the subject SSCs are not credited in the accident analysis. The proposed change is not expected to result in the loss of any fire-related SSC. Furthermore, because these SSCs are not relied upon for accident prevention or accident mitigation, the sole loss of the subject SSCs cannot challenge a fission product barrier. Therefore, the proposed changes do not result in a design basis limit for a fission product barrier as described in the SAR being exceeded or altered.

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing Yes the design bases or in the safety analyses? No BASIS:

The proposed change is unrelated to any method of evaluation with regard to the accident analysis. The changes act to only modify action requirements, establish a capability definition, and modify the test frequencies for the subject SSCs. The assumptions for evaluating safe shutdown acceptance criteria for a fire event is not impacted by the proposed change. Increasing the test intervals does not change the manner in which fire SSCs are operated or the method of testing. Therefore, the proposed changes do not result in a departure from a method of evaluation described in the SAR used in establishing the design bases or in the safety analyses.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 4

ANO 50.59 Evaluation Number 2008-007

10 CFR 50.59 EVALUATION FORM Sheet 1 of 5 I. OVERVIEW / SIGNATURES8 Facility: ANO Unit 2 Evaluation # / Rev. #: 08-007 / 0 Proposed Change / Document: EC 2244 Description of Change:

NRC issued GSI-191, Assessment of Debris Accumulation on PWR Sump Performance, with the objective to ensure that post-accident debris blockage will not impede or prevent operation of the Emergency Core Cooling System (ECCS) and Containment Spray System (CSS) in recirculation mode at Pressurized Water Reactors (PWR) during Loss of Coolant Accidents (LOCA) or other High Energy Line Break (HELB) accidents for which containment sump recirculation is required. Following completion of its technical assessment of GSI-191, the NRC issued Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Recirculation during Design-Basis Accidents at Pressurized Water Reactors on June 9, 2003. In developing this bulletin, the NRC staff recognized that it may be necessary for addressees to undertake complex plant-specific evaluations, and that the methodology needed to perform these evaluations was not currently available.

GL 04-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors was issued by the NRC on September 13, 2004. The purpose of the letter was to request performance of an evaluation of debris blockage on ECCS and CSS system recirculation function and take actions, as appropriate, to ensure system function.

In order to provide a consistent, industry-wide approach to performing the requested evaluations per GL 04-02, the Nuclear Energy Institute (NEI) issued guidance report NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, December 2004, Volume 1. The NRC Staff reviewed this guidance report and issued a Safety Evaluation Report (SER) on NEI 04-07 document titled Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 04-02, Volume 2. These documents provide comprehensive guidance with regards to issues related to GSI-191.

The resolution of GSI-191 will be documented in ANOs response to GL 04-02.

Engineering Change (EC) 2244 documents the design bases change associated with compliance to GSI-191 requirements. The following issues related to GSI-191 have been addressed within the scope of this EC:

1. Break Selections - Identify the break size and location that present the greatest challenge to post accident sump performance. Break locations were evaluated in both SG cavities at various locations in the RCS to find the break that generates the highest combination of fiber debris and particulate debris.

This ensures that the analysis is bounding and presents the greatest challenge to post-accident sump performance.

2. Debris Generation/Zone of Influence (ZOI) - Determine for each debris type (i.e. various insulation and coating materials) the zone within which the break jet forces would be sufficient to damage materials and create debris. A spherical zone of influence is established in pipe diameters as determined by jet impingement tests, which may be conducted by industry groups or for site specific conditions. Based on the debris ZOI, and modeling of the debris locations the amount of debris generated by various postulated breaks can be determined.

3. Latent Debris - Latent debris represents dust and dirt that is found in containment and would be washed into the basement by containment spray. NRC approved industry guidance documents (NEI 04-07) provide the basis for the distribution of latent debris between fine fibers and particulate material. Plant walkdown observations were performed at numerous locations throughout containment on various surface types to determine a representative mass of latent debris for use in strainer head loss tests and downstream effects analysis.

8 Signatures may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 2 of 5

4. Debris Transport - Estimate the fraction of debris that would be transported from debris sources within containment to the containment sump strainers based on modeling of flow velocities in the containment basement combined with data established by testing performed for ANO. The erosion and transport testing consider the erosion rates of pieces of calcium silicate insulation and the transport rate that occurs for fine particulate of this material at limiting velocities applicable to the ANO-2 basement during post-LOCA recirculation. The ANO-2 containment sump strainer qualification has been performed with a bounding case that conservatively assumes 100% transport of all other detrimental debris types (i.e.

fiber, coatings, and latent debris) potentially generated following a LOCA, with a conservatively derived erosion and transport rate applied only to the calcium silicate insulation.

5. Head Loss and Vortexing - Strainer head loss was established via scaled testing of strainer modules with debris loading. The test methods addressed various NRC comments from observation of industry strainer testing to conservatively deal with details such as debris preparation, addition sequence, termination criteria and data analysis. Analysis was also performed to establish that vortexing through the strainer or internally within the strainer would not occur within the limits established by the analysis.

6. Net Positive Suction Head (NPSH) - Calculate the NPSH margin for the High Pressure Safety Injection (HPSI) and CSS pumps during the recirculation mode that would exist during a LOCA. The systems are designed to perform their required function assuming a single failure. For conservatism, all trains of HPSI and CSS that draw suction from the containment sump are assumed to be operational during recirculation for purposes of evaluating containment sump performance and NPSH adequacy, since this condition causes maximum flow and associated maximum head loss across the strainer. The most limiting NPSH condition occurs at saturated water conditions, which is conservatively modeled as occurring at the saturation temperature for the minimum containment initial pressure. NPSH margin is evaluated for sub-cooled temperatures below this temperature to determine the temperature at which NPSH is no longer the limiting parameter for strainer head loss (i.e. when containment sump strainer structural design limit is more restrictive than available NPSH margin). Credit is not taken for increased containment pressure associated with the LOCA when determining NPSH margin.

7. Sump Structural Analysis - Verify the structural adequacy of the containment sump strainer including seismic loads and loads due to differential seismic movement, temperature and hydrodynamic water masses. Buoyancy force was considered for submerged strainers. The structural analysis calculation for the replacement sump strainer was revised by this EC to evaluate the impact of additional supports added to the strainer plenum assembly.

8. Upstream Effects - The flow paths upstream of the sump were reviewed for possible additional holdup of inventory due to the new debris types and loading and determined that no additional water holdup needed to be accounted for in the containment minimum level calculation.

9. Down-Stream Effects (DSE), Components and Systems - Components in the sump recirculation flow path are evaluated to ensure debris generated by a LOCA that can pass through the strainers 0.0625 inch openings will not result in blockage or unacceptable wear. The methodology for performing this analysis was developed by the industry owners group and approved by the NRC.

10. DSE, Fuel and Vessel - Evaluate the effects that debris carried downstream of the sump strainer and into the reactor vessel has on core cooling. This analysis includes the potential impact of chemical effects on the fuel assemblies from possible deposits and the potential for flow blockage due to debris buildup at the end of the fuel assemblies. This analysis was performed by a fuel vendor in accordance with guidance established by owners group documents.

11. Chemical Effects - To resolve the formation of adverse calcium phosphate precipitants, the Trisodium Phosphate (TSP) buffer has been replaced with Sodium Tetra-Borate (NaTB) under EC1641. EC1640 provided the design bases for this change. Site specific autoclave testing has been completed to show that the formation of aluminum precipitants does not occur until the containment sump temperatures are subcooled. The effects associated with the precipitation of aluminum compounds are considered for strainer head loss and reactor fuel assemblies. The strainer head loss test included the addition of aluminum precipitate compounds in combination with the other debris types and showed that head loss with these chemical effects materials remained within the allowable head loss. The determination of the types and quantities of chemical effects material was performed in accordance with industry owners group guidance that has been reviewed and approved by the NRC. The chemical precipitates are also evaluated by the DSE fuel analysis noted previously.

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 3 of 5 This Evaluation is therefore limited in scope to evaluating changes between existing design parameters and a new array of NRC approved design parameters deemed critical to successful operation of ECCS and CSS during post accident sump recirculation.

Is the validity of this Evaluation dependent on any other change? Yes No If Yes, list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change Yes No require prior NRC approval?

Preparer: Don Graham / See IAS for signature and date / DP Engineering. / ANO-MCS Name (print) / Signature / Company / Department / Date Reviewer: Chris Davenport / See IAS for signature and date / DP Engineering. / ANO-MCS Name (print) / Signature / Company / Department / Date OSRC: John R. Eichenberger / See IAS for Signature and Date Chairmans Name (print) / Signature / Date OSRC-08-025 OSRC Meeting #

II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If Yes, Questions 1 - 7 are not applicable; answer only Question 8. If No, answer No all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS: N/A

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the UFSAR? No BASIS: N/A

3. Result in more than a minimal increase in the consequences of an accident previously Yes evaluated in the UFSAR? No BASIS: N/A

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the UFSAR? No BASIS: N/A EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 4 of 5

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes UFSAR? No BASIS: N/A

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the UFSAR? No BASIS: N/A

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being Yes exceeded or altered? No BASIS: N/A

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing Yes the design bases or in the safety analyses? No*

This was a departure from a method of evaluation described in the UFSAR. However, the methodology is pre-approved by the NRC. See Basis.

BASIS:

Regulatory Guide (RG) 1.82 is currently referenced in the U2 SAR related to the determination of HPSI and CSS pump NPSH margins during recirculation mode following a LOCA. These references to RG 1.82 are being changed to GSI-191 and GL 04-02 criteria, and is essentially the method of evaluation that is being changed.

GSI-191 was initiated by the NRC for utilities to determine if the transport and accumulation of debris in containment following a LOCA will impede the operation of the ECCS and CSS in commercial operating pressurized water reactors (PWR). In September 2004 the NRC issued Generic Letter 2004-02 (GL 04-02), Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (0CNA090401) The letter required all PWRs operating in the United States to evaluate an array of design parameters deemed critical to successful operation of ECCS and CSS during post accident containment sump recirculation. Based on these evaluations, plants were required to inform the NRC of any modification required to satisfy the evaluation. In order to provide a consistent, industry-wide approach to performing the requested evaluations per GL 04-02, the Nuclear Energy Institute (NEI) issued guidance report NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, December 2004, Volume 1. The NRC Staff reviewed this guidance report and issued an SER on NEI 04-07 document titled Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 04-02, Volume 2. These documents provide comprehensive guidance with regards to issues related to GSI-191.

Extensive testing and analysis to quantify the effects of post-LOCA debris on ECCS and CSS operation while in containment sump recirculation following a LOCA have been completed. In addition to analytical efforts, physical design modifications have also been implemented to gain design margin in critical areas.

These modifications were addressed and evaluated under separate EC packages. This evaluation EC shows that all design margins related to post-LOCA containment sump recirculation remain satisfied.

Changes to the Safety Analysis Report (SAR) are identified in the attached License Basis Document Change Request (LBDCR) for EC 2244. The marked changes pertain to the utilization of the new parameters (methodology) associated with resolution of GSI-191 and GL 04-02. The updated design basis for the containment sump analysis now cover evaluations pertaining to debris quantities generated from selected limiting breaks, chemical effects associated with precipitation of aluminum compounds, strainer head loss due to debris and chemical precipitates, down stream effects on components in the containment sump recirculation flow path and down stream effects on reactor vessel internals, including fuel assemblies for potential detrimental effects. Additional areas as identified in the Description of Change section of the Evaluation related to the new methodology were also addressed as part of the overall containment sump analysis.

EN-LI-101-ATT-9.1, Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 5 of 5 The current NPSH calculation (CALC-91-E-0116-01 R5) evaluated NPSH at two static temperatures and the containment sump strainer head loss evaluation assumed a fixed debris head loss based on the original design basis for the system. Reanalysis of the NPSH calculation (CALC-91-E-0116-01 R6) evaluated the NPSH margin as a function of temperature which varies with time starting with the Recirculation Actuation Signal (RAS) through 30 days post LOCA. The NPSH calculation credits the pre-accident minimum air pressure in containment in determination of available NPSH as the sump temperature is sub-cooled. The limiting NPSH condition remains at the minimum saturated temperature condition, corresponding to the minimum containment pressure. The results of the NPSH calculations have changed. The changes made reflect the new criteria and results showing that the available NPSH exceeds the required NPSH for HPSI and CSS pumps with a clean strainer (no debris present). The clean strainer NPSH margin is compared to the strainer head loss testing results to show that acceptable margin remains.

The new evaluation areas and general methodology utilized for analyses has been approved for use by the NRC and all PWR plants have been instructed to comply with GL 04-02. In addition, the method of evaluation now used is more conservative than the existing method as described in the UFSAR.

Therefore additional NRC approval is not required for this activity beyond their standard review of the compliance submittal documents associated with GL 04-02.

If any of the above questions is checked Yes, obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 4

2CAN100802, CFR 50.59 Summary Report for Period Ending October 6, 2008

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