ULNRC-05734, Transmittal of 10 CFR 50.59 Summary Report

Transmittal of 10 CFR 50.59 Summary Report

ML103000160
Person / Time
Site:	Callaway
Issue date:	10/26/2010
From:	Maglio S AmerenUE, Union Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-05734
Download: ML103000160 (19)

Text

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WAmeren Callaway Plant MISSOURI

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October 26,2010 ULNRC-05734 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.59(d)(2)

Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

FACILITY OPERATING LICENSE NPF-30 10 CFR 50.59 Summary Report In accordance with 10 CFR 50.59(d)(2), this letter transmits a report which summarizes the evaluations performed pursuant to 10 CFR 50.59(c)(1) for changes, tests, and experiments approved and implemented for activities at Callaway Plant. This report covers all 10 CFR 50.59 evaluations that were implemented from January 1,2006 to June 30, 2010.

Based on the noted reporting period, it has been greater than 24 months since Callaway submitted its previous 10 CFR 50.59 Summary Report. In addition, this report includes one evaluation that was inadvertently omitted from a previous report. These reporting deficiencies are being addressed under Callaway's corrective action program. An extensive review was completed to ensure no other evaluations or reports were omitted or overlooked.

This letter does not contain new commitments. If there are any questions, please contact Mr. Tom Elwood at (314) 225-1905.

Sincerely, AtNL\ A.. ~2 Scott A. Maglio Regulatory Affairs anager KRA Enclosure

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• PO Box 620 Fulton, MO 65251 AmerenMissouri.com

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• ULNRC-05734 October 26,2010 Page 2 cc: Mr. Elmo E. Collins, Jr.

Regional Administrator U.S. Nuclear Regulatory Commission Region IV 612 E . Lamar Blvd., Suite 400 Arlington, TX 76011-4125 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Mohan C. Thadani Senior Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8G14 Washington, DC 20555-2738 Mr. James Polickoski (2 copies)

Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8G14 Washington, DC 20555-2738

ULNRC-05734 October 26, 2010 Page 3 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 4200 South Hulen, Suite 422 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed. )

Electronic distribution for the following can be made via Other Situations ULNRC Distribution:

A. C. Heflin F. M. Diya L. S. Sandbothe C. O. Reasoner III S. A. Maglio S. L. Gallagher T. L. Woodward (NSRB)

T. B. Elwood K. R. Austgen Ms. Diane M. Hooper (WCNOC)

Mr. Tim Hope (Luminant Power)

Mr. Ron Barnes (APS)

Mr. Tom Baldwin (PG&E)

Mr. Wayne Harrison (STPNOC)

Ms. Linda Conklin (SCE)

Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP)

Missouri Public Service Commission

DOCKET NO. 50-483 Enclosure to ULNRC-05734 UNION ELECTRIC COMPANY CALLAWAY PLANT 10 CFR 50.59

SUMMARY

REPORT JANUARY 2006 ---- JUNE 2010

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 EXECUTIVE

SUMMARY

In accordance with 10 CPR 50.59(d)(2), a summary report has been prepared which provides summaries of the 10 CPR 50.59 evaluations of changes, tests, and experiments approved and implemented for activities at Callaway Plant.

This report covers all 10 CPR 50.59 evaluations that were implemented from January 1,2006 to June 30, 2010. During this period there were 11 changes implemented that required a 10 CPR 50.59 evaluation. Additionally, this report includes one 10 CPR 50.59 evaluation that was inadvertently omitted from a previous report.

Page I of IS

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 REFERENCE/ABBREVIATION KEY CAR Callaway Action Request (a Callaway corrective action document)

COMS Cold Overpressure Mitigation System DLM Diffusion Layer Model ECCS Emergency Core Cooling System EFPY Effective Full Power Years EPRI Electric Power Research Institute ESFS Engineered Safety Feature System ESW Essential Service Water FSARCN FSAR Change Notice LOCA Loss of Coolant Accident MCB Main Control Board MP Modification Package MSIV Main Steam Isolation Valve MSLB Main Steam Line Break NIS Nuclear Information Systems NPSH Net Positive Suction Head ORNL Oak Ridge National Laboratory POD Prompt Operability Determination PORV Power Operated Relief Valve PRA Probabilistic Risk Analysis PTLR Pressure Temperature Limit Report RCS Reactor Coolant System RFR Request for Resolution RHR Residual Heat Removal RSG Replacement Steam Generator RTP Rated Thermal Power RWST Refueling Water Storage Tank SSC Systems, Structures, or Components TIS Technical Specification(s)

TSBCN TIS Bases Change Notice UHS Ultimate Heat Sink Page 2 of IS

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 RFR 23374A (pre-dates creation of the Callaway 50.59 Evaluation Log)

Change in Analysis Methodology for Replacement Steam Generators Activity

Description:

This evaluation addresses a change in analysis methodology implemented for the Replacement Steam Generator (RSG) project. The RSG Containment analyses were performed using newer methodologies (GOTHIC) than those currently described in FSAR Section 6.2 (COPATTA and CONTEMPT).

The original containment analyses for Callaway were perfomed by Bechtel using their proprietary COPATTA computer code. The COPATTA analyses were documented in Bechtel calculations M-YY-43, M-YY-44, and M-98-001 for LOCA and Main Steam Line Break (MSLB) and formed the bases for Callaway FSAR Sections 6.2 and 3.11(B). Callaway then developed containment analysis capabilities using the CONTEMPT computer code, and assumed in-house responsibility for the Licensing Bases containment analyses. As described in the FSAR, Licensing Bases containment analyses, using CONTEMPT to update the COPATTA results, include the work performed to support the allowance for containment cooler degradation and the slower stroke times associated with the new Main Feedwater Isolation Valves.

The new Callaway GOTHIC containment evaluation model was constructed based on the recent Kewaunee GOTHIC containment evaluation model. The Kewaunee GOTHIC containment evaluation model was reviewed and approved by the NRC. NRC approval of the Kewaunee GOTHIC containment evaluation model is documented in NRC Letter from Anthony C.

McMurtray (NRC) to Thomas Coutu (NMC), Enclosure 2, Safety Evaluation, dated September 29, 2003 and in NRC Letter from John G. Lamb (NRC) to Thomas Coutu (NMC), Enclosure 2, Safety Evaluation, dated February 27,2004. Both the Kewaunee and Callaway containment evaluation models use GOTHIC code version 7.1 patch 1 with the DLM heat and mass transfer option.

Summary of Evaluation:

Based on the prior NRC approval of GOTHIC, which is being applied to Callaway in accordance with the NRC conditions for approval of GOTHIC, this change may be implemented without (additional) prior NRC approval.

Note: A summary of this evaluation should have been included in the 10 CFR 50.59 summary report sent on May 18, 2006.

Page 3 of 15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 GOTHIC 7.2a (Callaway 50.59 Evaluation Log No. 06-02)

GOTHIC 7.2a Software Documentation Package Activity

Description:

Current Containment/Main Steam Tunnel temperature/pressure analyses performed for FSAR Chapters 3.B and 6.2 were completed by Westinghouse using version 7.1pl of the GOTHIC code. However, future analyses will be performed by Callaway personnel on site, and GOTHIC version 7.1 pI is no longer distributed by EPRI. GOTHIC version 7.2 replaced version 7.1 pI, but version 7.2a will be considered since most of the changes were corrections of errors from version 7.2. A software documentation package has been prepared to enable the use of GOTHIC version 7.2a per procedure EDP-ZZ-040 11, "Nuclear Engineering Analytical Software Controls." This 10 CFR 50.59 Evaluation is being performed as part of the 10 CFR 50.59 review of the Software Documentation Package.

Summary of Evaluation:

Evaluation question 8, "Does the proposed activity result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses?" is applicable to this change. GOTHIC 7.2a calculates some results that are conservative and some that are non-conservative with respect to results from version 7.1 P 1. A review of the change in these results indicates that the results are conservative or essentially the same as those from GOTHIC version 7.1 pI and thus do not represent a departure from a method of evaluation described in the FSAR and do not require prior NRC approval. Other limitations on the use of GOTHIC remain consistent with NRC approval of the use of GOTHIC in previous applications. In addition, user-controlled enhancements which could impact the results in GOTHIC 7.2a will not be used for Callaway calculations.

MP 00-1009B (Callaway 50.59 Evaluation Log No. 06-04)

MSIV Actuator Replacement Activity

Description:

Modification MP 00-1009B will replace the existing electro-hydraulic Main Steam Isolation Valve (MSIV) actuators, ABHVOOll, ABHVOOI4, ABHV0017 and ABHV0020 with system-medium actuators to improve reliability. With the new actuators using system media (i.e., main steam line pressure) the stroke time of the MSIV s will be greater at lower pressure as compared to the stroke time established for the current electro-hydraulic actuators.

The new actuator will retain two redundant safety trains to actuate the valve. The safety trains will consist of a set of solenoids directing main steam from the main steam line to the upper piston chamber in the actuator to stroke the valve closed. Due to the solenoid and train design, the failure mode of the valve is changing from "as-is" to "fail closed" during a loss of all electrical power.

Page 4 of 15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-OS734 Summary of Evaluation:

The evaluation concludes that changing the electro-hydraulic MSIV actuators to system-medium actuators does not require NRC approval. The primary changes resulting from this modification are the change in MSIV failure mode from "as-is" to "closed" during a loss of all power and having the stroke time of the actuator increase with decreasing system pressure. The change in MSIV failure mode will introduce only one Condition II fault during the life of the plant and the MSIV modification does not impact any of the initiating mechanisms or conditions for events found in the FSAR and thus will not result in a more than minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR.

A PRA evaluation was performed (at the component/function level) and it was determined that the potential impact of the new MSIV actuators on the probability of malfunction does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety.

The stroke time change is bounded by existing accident analyses, and since offsite and control room accident doses are not sensitive to MSIV closure time, the MSIV modification does not result in an increase in the consequences of an accident previously evaluated in the FSAR.

Although the change in MSIV failure mode very slightly increases the potential Condition II event (i.e., inadvertent MSIV closure), by definition such an event does not propagate to cause more serious faults and is not an event with dose consequences, which means this modification does not introduce the posibility of a change in the consequences of a malfunction of an SSC important to safety. Rapid depressurization scenarios that could increase stroke times were reviewed and determined to be bounded by Mode 1 and 2 cases. The new MSIV s will retain the ability to perform their safety function (i.e., close), will fail closed and have redundant trains which will not introduce the possibility for a malfunction of an SSC important to safety with a different result than previously evaluated.

With regard to MSIV performance and assumptions in the safety analyses, the slower MSIV stroke times resulting from the MSIV modification would not cause the fuel cladding, RCS integrity, or FSAR containment pressure limits to be challenged, therefore analyzed MSIV performance (with the MSIVs modified as proposed) does not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered.

There is no change to any FSAR described methodology.

This activity can be implemented without obtaining a License Amendment based on the responses to the eight questions of 10 CFR SO.S9(c)(2). Changes to the MSIV stroke time do not require a change to the Technical Specifications since the MSIV stroke time was moved to the Technical Specification Bases by Amendment 172.

Page 5 of 15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 This evaluation also covers changes to the FSAR and the T/S Bases (FSARCN 04-058 and TSBCN 07-002).

ORIGEN 2.2 (Callaway 50.59 Evaluation Log No. 06-05)

ORIGEN 2.2 Software Documentation Package Activity

Description:

For each fuel cycle, the fission product inventory is calculated using the computer code ORIGEN and compared to the product inventory reported/described in the FSAR to ensure that it is acceptable (or determine if changes are needed). The ORIGEN 2 computer code was used to calculate the fission product inventories currently reported/described in the FSAR (FSAR 15.0.9). ORNL has issued revision 2.2 of ORIGEN. This revision corrects a problem in previous revisions in calculating the fission rate of the minor actinides. This 50.59 Evaluation will approve the use of revision 2.2 at Callaway in calculating the cycle-specific fission product inventories.

Summary of Evaluation:

Revision 2.2 of ORIGEN increases the calculated fission product inventories, which is a conservative change. Since this is a conservative change, it is not considered a departure from a method of evaluation described in the FSAR.

MP 06-0146 (Callaway 50.59 Evaluation Log No. 06-06)

Cold Overpressure Mitigation System Setpoint Changes Activity

Description:

MP 06-0146 is revising the Callaway Pressure Temperature Limit Report (PTLR) and associated Cold Overpressure Mitigation System (COMS) setpoints to be applicable for 28 and 35 EFPY, in addition to resolving issues documented in Callaway's corrective action program. The PTLR and associated COMS setpoints will be calculated by Westinghouse using the methodology from WCAP-14040-A Rev. 4. The methodology currently listed in the plant Technical Specifications is from WCAP-14040-A rev. 2. This 10 CFR 50.59 evaluation addresses only the change in the methodology from Revision 2 to Revision 4 ofWCAP-14040-A. The change required to the Technical Specifications due to the listing ofWCAP-14040-A Rev. 2 in Section 5.6.6.b.2 will be covered seperately and submitted to the NRC under license amendment request OL-1274 (AmerenUE letters ULNRC-05315 and ULNRC-05346).

Summary of Evaluation:

Adopting the methodology from WCAP-14040-A Rev. 4 is an activity that only affects a "method of evaluation." Therefore, only question 8 is required to be addressed in this evaluation. The methodology described in WCAP-14040-A Rev. 4 does not represent a Page 60f15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 "departure from a method of evaluation that would require NRC approval" since the NRC has already approved this methodology. NRC approval is documented in the Safety Evaluation attached to the NRC letter from Herbert N. Berkow to Gordon Bishoff dated February 27,2004.

Apart from this modification, a license amendment is being pursued to remove the specific revision number from WCAP-14040-A in the Technical Specifications in accordance with TSTF-419. Therefore, this methodology may not be implemented until license amendment request OL-1274 is approved by the NRC.

MP 06-0029 (Callaway 50.59 Evaluation Log No. 07-01)

Modify RHR Recirculation Sump Level Sensor Activity

Description:

Modification MP 06-0029 will remove the lower portion of the RHR Recirculation Sump level instrumentation. The new RHR Recirculation Sump strainers that will be installed in accordance with MP 06-0003 will create interference with the level float assembly that is currently installed in each of the sumps. Therefore, the lower portion of the sump level indication system, which is installed in the sump, will be removed, and the existing level indicators will monitor level using only the upper portion of the level sensors. The result will be that this modification will change the range of the RHR Recirculation sump level indication such that it will only indicate from the top of the curb around the sump (six inches above the containment floor) to 72 inches above the containment floor. Currently, the range of the level indication extends from six inches above the bottom of the recirculation sump to 72 inches above the containment floor.

As shown in FSAR Table 6.2.2-6 & 6A, the containment water level at the time that initiation of ECCS switchover begins is at a minimum of 200 1'-6." This is 12 inches above the top of the curb that surrounds the recirculation sump. With the water level at this elevation, both the recirculation sump level instruments and the normal sump level instruments will be monitoring containment water level, albeit with different reference points. It therefore makes no difference which level indication is used in order to perform the sump level monitoring function that is described in section 6.3.2.2 of the FSAR. So, the change will have no adverse impact on the functions described therein.

As stated in the FSAR, the RHR Recirculation sump level instrumentation is not a Type A variable and is not used for event identification. It is not credited as the Containment Sump Water Level indication required by Table 2 of Reg. Guide 1.97 rev. 2 (described as Items B.3.2

& C.2.3 in FSAR Table 7A-3). As shown in Data Sheet 6.2 of Table 7A-3 of the FSAR, it is the Normal Sump Water Level indication that is credited for satisfying the requirements of Reg.

Guide 1.97. Therefore, the changes to the RHR Recirculation Sump level instrumentation will not adversely impact the function of Containment Sump Water Level indication that is required by Reg. Guide 1.97.

Page 70f15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 In conjunction with this Plant Modification FSAR Sections 6.3.2.2 and 7.5.2.1.2 are being changed to clarify that it is the containment "nonnal" sump level indication which is being credited as the indication which provides the assurance that adequate NPSH exists for operation in the recirculation mode. This is in agreement with FSAR Table 7A-3 sht. 29 which credits the nonnal sump level indication as the instruments that are required for function detection and long-tenn surveillance. (This is the tenninology that is used in Reg. Guide 1.97 rev. 2.)

Summary of Evaluation:

Since the RHR Recirculation Sump level instrumentation perfonns indicating and alanning functions only and has no controlling functions, it cannot directly initiate any accident. This modification does not change these basic indicating and alanning functions, it only reduces the range over which the sump level is monitored. This change therefore has no impact on the frequency of any accident or the likelilhood of a malfunction of an SSC important to safety.

Since the level sensors are (and will continue to be) located within the curb that surrounds the sump, this instrumentation will not respond until water level in containment is at least six inches above the floor of containment (top of the curb). For this reason, this indication is not the indication that is credited to satisfy the "Containment Sump Water Level" indication requirements of Reg. Guide 1.97. The containment nonnal sump level instrumentation is the indication that is credited to satisfy these requirements.

The RHR Recirculation Sump level instrumentation will, however, provide operators with an indication of containment water level during an accident that is redundant to the nonnal sump level instrumentation. Since the modified instrumentation will not respond until containment water level is at least six inches above the containment floor as the existing system does, there is no change to the way this instrumentation functions. Therefore, this change has no impact on the consequences of an accident or the consequences of a malfunction of any SSC important to safety.

Since no new functions or failure modes are being created as a result of this change, this change does not create the possibility of either a new type of accident or a malfunction of an SSC important to safety with a different result than any previously evaluated in the FSAR. Also, since this instrumentation has no controlling functions, it cannot affect any system parameters such as RCS or Containment temperature or pressure. Therefore, it has no impact on any design basis limit for a fission product barrier as described in the FSAR.

Based on the above, a License Amendment is not required prior to implementing this change.

Page 8 of15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 FSARCN 07-006 (Callaway 50.59 Evaluation Log No. 07-02)

FSAR Revisions Associated with Cycle 15 Till Coastdown Activity

Description:

Tavg Coastdown is a technique to extend bum-up at the end of a cycle by taking advantage of the large negative moderator temperature coefficient at the end of an operating cycle. Callaway plant intends to implement Tavg Coastdown at the end of cycles as necessary. FSARCN 07-006 will add a description of the Tavg Coastdown strategy into the Callaway Plant FSAR Chapters 7.7.1.8.1 and 15.0.2.2. Appropriate plant procedure changes will be made as well.

The Tavg Coastdown will be implemented by a Temporary Modification (TM) or Procedurally Controlled Temporary Modification (PCTM) in accordance with plant procedures. If a PCTM is used, it will be controlled by plant procedures. The plant modifications tracked as a TM or PCTM are:

1. The one-time adjustment to the steam dump load rejection controller (AB-TC-0500A) gain and the steam dump valve trip open setpoints (AB-TB-0500B and AB-TB-0500C) prior to initiating the Tavg Coastdown (per plant procedures), and

2. The periodic adjustment to the Tref signal (ACTY0505A) for the Rod Control system during the T avg Coastdown (per plant procedures).

These adjustments have been analyzed as part of reference 1 and are acceptable for conditions during the Tavg Coastdown and during the downpower to start the subsequent refueling outage.

The steam dump load rejection controller and the affected Tref signal are not required during refueling outages while the turbine is out of service. Therefore, the applicable mode restraint specified by this 50.59 evaluation is Mode 2 ascending (to ensure restoration prior to placing the turbine back in service). Plant procedures will administratively ensure that, at a minimum, this mode restraint is implemented.

This 50.59 Evaluation covers the Tavg Coastdown strategy as presented in the Tavg Coastdown Engineering Report (reference 1), including the procedural changes required for implementation of the Tavg Coastdown strategy, as well as incorporation of the Tavg Coastdown into the FSAR.

References:

1) Tavg Coastdown Engineering Report, Attachment to SCP-07 -17

2) WCAP-16140, Replacement Steam Generator Program NSSS Report

3) CAR 200606856 Action 19.1 Summary of Evaluation:

Page 9 of 15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 The strategy determined for the Callaway Plant Tavg Coastdown has been evaluated against the seven applicable 50.59 evaluation criteria. Evaluations in reference 1 have shown that the following do not require adjustment for Tavg Coastdown:

• The T' and T" settings for the OP~T & OT~T trips

• The current pressurizer level program

• RCS Flow setpoints However, in order to ensure the pressurizer PORVs are not challenged on a Turbine Trip without Reactor Trip from P-9 setpoint, the load rejection controller gain and the steam dump valve trip open setpoints require adjustment prior to initiation of the T avg coastdown.

As delineated in Appendix A to reference 1, the following guidelines for Tavg Coastdown implementation are included to ensure the analyses which demonstrate acceptability for this 50.59 evaluation remain valid:

1. Maximum rate of reduction in Tavg of 2°F/day.

2. Minimum average Tavg of 579 .2°F while at full power.

3. Minimum extrapolated full power average T avg of 570. 7°F (equivalent to 569.95°F at 94.5% reactor power). Extrapolation is performed using reference hot no-load temperature of 557°F.

4. Minimum Steam Pressure of 867 psia while at full power.

5. Limiting range for Feedwater Temperature of 390-446°F.

6. Any loop-specific full power ~T is not to exceed 1.0°F below that loop's ~To. Maintain through renormalizinglrescaling the ~To as appropriate.

7. The secondary side calorimetric power measurement is not to exceed the NIS channels' indicated power level by more than +2% Rated Thermal Power (RTP). Maintain through NIS adjustment as appropriate.

Callaway procedures will be revised to ensure compliance with the guidelines given above.

Therefore, the validity of the evaluations and analyses in references 1, 2 and 3 will be ensured.

Based upon the evaluations provided in reference 1, all acceptance criteria were met for the accidents/transients that were required to be analyzed. Maintaining operation of the plant at a reduced Tavg within the guidelines given above will ensure that adequate operational margins remain. Therefore, implementation of T avg Coastdown as described in this evaluation and in reference 1 will not result in the increase of probability or consequences of accidents or malfunction of equipment evaluated in the FSAR. In addition, Tavg Coastdown will not result in an accident of a different type than described in the FSAR or the possibility of a malfunction of an SSC with a different result than what is described in the FSAR. Since the evaluation for impact on applicable accident analyses, as performed for this modification, determined that the applicable acceptance criteria continue to be met (thus preserving intended margins to fuel safety limits), Tavg Coastdown will not cause a design basis limit of a fission product barrier to be exceeded or altered.

Page 10 of15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-20 10 Enclosure to ULNRC-05734 The results of the evaluations performed for the seven applicable 50.59 evaluation questions show that this activity may be implemented without prior NRC approval.

FSARCN 07-037 (Callaway 50.59 Evaluation Log No. 07-03)

Revision of Description of Regulatory Guide 1.52 Position Related to ANSI Standard Used for the Design, Construction, and Testing of Air Heaters Activity

Description:

This 10 CFR 50.59 Evaluation is completed for FSAR Change Notice 07-037. This FSAR change revises Callaway's commitment to Regulatory Guide 1.52 as documented in FSAR Table 9.4-2. Only a portion of the commitment to Regulatory Guide 1.52 is being modified by this FSAR change.

Specifically, on Sheet 7 of Table 9.4-2, Regulatory Guide 1.52 Position 3.b specifies that air heaters should be designed, constructed, and tested in accordance with the requirements of Section 5.5 of ANSI N509-1976. Callaway's FSAR currently states that the Control Room Pressurization System and the Emergency Exhaust System comply with this section of the ANSI standard. This is being revised to reflect compliance with the current Section 5.5 of ANSI N509-1989. The specific difference being implemented is the reset method for the charcoal adsorber heater cutouts. The previous commitment to the 1976 revision of ANSI N509 requires a manual reset for the cutout. The 1989 revision of ANSI N509 states that manual resets are not recommended for ESF air cleaning units located in areas not accessible following a Design Bases Accident.

The function of the heaters is to control humidity in the air passing through the charcoal adsorbers. Increased humidity results in decreased efficiency of the charcoal adsorbers, which could result in an increase in accident consequences to the public and Control Room personnel.

Summary of Evaluation:

This evaluation demonstrates that the proposed change does not require NRC approval prior to implementation. The evaluation further demonstrates that there is no increase in accident consequence, or frequency of occurrence, and no increase in the frequency or consequence in the failure of an SSC important to safety. Current plant design incorporates two cutouts in series, one with an automatic reset and one with a manual reset. Implementation of this FSAR change will allow the elimination of the manually reset cutout. This is not a reduction in the level of redundancy credited in the FSAR. The ANSI standard and FSAR descriptions of the affected systems only describe one cutout.

Page 11 of 15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 MP 07-0066 (Callaway 50.59 Evaluation Log No. 07-04)

Replacement of Buried ESW Piping with HDPE Material Activity

Description:

This 50.59 evaluation is applicable to a change being made as a result ofMP 07-0066 for the replacement of Essential Service Water (ESW) buried piping. Specifically, the change being addressed by this 50.59 evaluation involves the use of several computer programs related to piping stress analysis that are not currently listed in the FSAR.

Summary of Evaluation:

The following computer codes will be used for the design change that are not currently identified in the FSAR: PIPSYSW is used for piping stress analysis and determination of support loads; W ATPRO is used for piping stress analysis involving welded attachments; ANCHOR is used for piping stress analysis involving anchor attachments; and SAP2000 is used for finite element analysis of structural systems.

The use of these computer programs revises FSAR-described evaluation methodologies used to establish the design bases, but does not introduce new design methodologies. The program algorithms are based on established and accepted calculation methods, including ASME B&PV Code,Section III, and related Code Cases. The use of the ASME Code and acceptable Code Cases is established in the Callaway FSAR, as is the use of finite element software for structural analysis. There are no unique plant operating conditions or site characteristics that would limit use of these computer programs for Callaway. In addition, the programs have been verified and validated in accordance with the Sargent & Lundy, LLC software QA process. Therefore, the 50.59 evaluation determined that the use of these computer programs, not currently identified in the FSAR, does not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses.

FSARCN 08-028 (Callaway 50.59 Evaluation Log No. 08-01)

Subcritical Time Requirement Prior to Fuel Off-Load Activity

Description:

FSARCN 08-028 will reduce the amount of subcritical decay time required prior to initiating core offload during refueling outages. The fuel handling decay time requirement is contained in FSAR Specification 16.9.5. The current value is 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />; FSARCN 08-028 will change the value to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. There are no physical changes to the facility associated with this FSARCN.

Summary of Evaluation:

The change involves an increase in Fuel Handling Accident (FHA) consequences. However, the increase was found to be not more than minimal.

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10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 3O-S eptember-2 0 10 Enclosure to ULNRC-05734 Radiological consequences for the Reactor Building FHA (RBFHA) and Fuel Building FHA (FBFHA) were re-analyzed for the proposed change in subcritical decay time. Whole-Body consequences for the RBFHA described in the FSAR will be increased, but the increase is not more than minimal.

MP 06-0144 (Callaway 50.59 Evaluation Log No. 08-02)

Implement Westinghouse OPDT/OTDT Margin Recovery Program (MRP)

Activity

Description:

MP 06-0144 involves modifying the parameters that are used to calculate the Overtemperature Delta-T (OT~T) setpoint in order to increase the margin between the normal operating point and the setpoints for the OT~T Rod Stop/Turbine Runback alarms. This will eliminate the spurious alarms that are occurring due to periodic upward spikes in Hot Leg temperature measurement that are being caused by Upper Plenum Flow Anomaly.

The specific parameters that are changing that will affect the OT~T setpoint are K}, K 2 , K 3, 1'},

1'2, 1'3, 1'4, & 1'6. The method by which these parameters are combined to determine the OT ~ T setpoint is described in an equation given in Note 1 of Table 3.3.1-1 of the Technical

. Specifications. The OT ~ T trip occurs if OT ~T is indicated in two reactor coolant loops. The OT ~ T also provides a signal to generate a turbine runback prior to reaching the trip setpoint. A turbine runback will reduce turbine power and reactor power, either through automatic rod insertion or through operator action. A reduction in power will normally alleviate the OT ~T condition and may prevent an unnecessary reactor trip.

The evaluation of the new setpoint parameters also includes changes to the Overpower Delta-T (OP ~ T) setpoint due to common parameters in the two setpoints, 1}, 12, 1'3, & 1'6. These changes do not affect the Safety Analysis Limit for the OP~T.

This evaluation covers MP 06-0144 as well as FSARCN 08-036 for reflecting these changes in the FSAR.

Summary of Evaluation:

This change will not result in the capability of the OT ~T setpoint spuriously initiating any transient. The changes to the OT ~ T setpoint parameters will have no impact on any plant conditions that could cause an accident. So, this change will not increase the frequency of any accident that has previously been evaluated, nor will it create the possibility of an accident of a different type than any that has previously been evaluated. It will not introduce any new failure modes nor create any new plant conditions that could create the possibility of causing an accident of a different type than any previously evaluated in the FSAR.

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10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 The changes to the aT ~T setpoint parameters will result in a change in the timing of a reactor trip, which will have a small impact on the RCS temperature and power levels at which the reactor trip occurs. However, this will not affect any system, structure, or component (SSC) important to safety in any way that would increase the probability or the consequences of a malfunction of the SSC. Also, this change will not create the possibility of a malfunction of any SSC important to safety with results that are different than any that have been previously evaluated.

Westinghouse analyzed the various accidents that rely on either the OP ~T or the aT ~T setpoints. These analyses show that with the changes to the aT~T and the OP ~T setpoint parameters, the design basis limits for fission product barriers are not being exceeded or altered.

Prompt on associated with CAR 201001813 (Rev. 3) (Callaway 50.59 Evaluation Log No.

10-01)

EOP E-1, "Loss of Reactor or Secondary Coolant," (Rev. 13); EOP Addendum 17, "Securing ESW Train Due to UHS Cooling Tower Trouble," (Rev. 0); OTA-RK-00016, "Annunciator Response Procedure - MCB panel RKOI6" Activity

Description:

The Prompt Operability Determination (POD) for CAR 201001813 is being revised such that operator action will be credited (as a compensatory action) to diagnose and mitigate the postulated single failure of components within the UHS Cooling Tower. This manual operator action to manage UHS heat load will be credited to be taken within two hours of initiation of the LOCA sequence. These additional operator manual actions will be implemented via plant emergency operating procedure (EOP) and annunciator response procedure revisions. The compensatory actions for Revision 3 of the POD include administrative limits for UHS pond level and temperature that are more restrictive than the corresponding Technical Specification limits. Those more restrictive UHS pond level and temperature limits have been reviewed and were screened pursuant to 10 CFR 50.59. Those changes are not included in the scope of this evaluation.

Summary of Evaluation:

Revision 3 of the POD for CAR 201001813 requires a compensatory action in which credit for a manual operator action to diagnose and mitigate the postulated single failure of components within the UHS cooling tower is taken. Currently, Callaway's FSAR credits manual operator action to manage UHS heat loads within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the initiation of the Large Break LOCA accident sequence. Revision 3 of the POD requires that manual action to manage UHS heat load be taken within two hours of the initiation of the LOCA sequence. From 10 CFR 50.59 evaluation of this modification to credited operator action, including consideration of the impact of the change on other aspects of the facility, it was determined that the proposed change is consistent with operator action for managing UHS heat load (as already contained in the accident analysis) and does not impact other operator actions credited in the accident analyses. All Page 14 of 15

10 CFR 50.59

SUMMARY

REPORT FOR CALLAWAY PLANT 30-September-2010 Enclosure to ULNRC-05734 evaluation questions are answered "No" such that Callaway may implement this proposed change without prior NRC approval.

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