ML20206B067
| ML20206B067 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/23/1999 |
| From: | Kelley J TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-MA4436, TAC-MA4437, TXX-99105, NUDOCS 9904290132 | |
| Download: ML20206B067 (23) | |
Text
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P7 Log # TXX-99105
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File # 10010 r
C Ref # 10CFR50.36 1UELECTRIC April 23,1999 c.14.= nery Senior Mce President
& PrincipalNuclear Of,cer U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON LICENSE AMENDMENT REQUEST 98-010 (TAC Nos. MA4436 and MA4437)
REF:
TU Electric letter, logged TXX-98265, from C. L. Terry to the NRC dated December 21,1998 Gentlemen:
Per Reference 1. TU Electric submitted a request to amend the CPSES Unit 1 Operating License (NPF-87) and CPSES Unit 2 Operating License (NPF-89) by incorporating changes into the CPSES Units 1 and 2 Technical Specifications and the CPSES Unit 2 Operating License to increase the licensed power for operation of CPSES Unit 2 to 3445 Mwt; an increase of approximately 1%. Per telephone conversation with NRR on April 13,1999, TU Electric received a request to provide the attached additionalinformation regarding License Amendment Request 98-010 by April 30,1999. Attachment 1 is the affidavit for this information supporting License Amendment Request 98-010, per telephone conversation, April,1999, with Dave Jaffe. Attachment 2 provides LJr response to the information l[/
requested.
If you have any questions regarding the attached information, please contact Mr. J. D. Seawright at (254) 897-0140.
SG..Ul 9904290132 990423
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PDR ADOCK 05000445 P
PDR COMANCHE PEAK SILAM ELECTRIC STATION P.O. Box 1002 Glen Rose. Texas 76043-1002 L
- TXX 99105 Pag 6 2 of 2 This communication ccatains no new commitments regarding CPSEE Units 1 and 2.
Sincerely, C. L. Terry By:
4 Jabes J. Kelley, Jr.
Vice President Nuclear Engineering & Support JDS/jds Attachment c-Mr. E. W. Merschoff, Rsgion IV Mr. J.1. Tapia, Region IV Mr. D. H. Jaffe, NRR Resident inspectors, CPSES Mr. Arthur C. Tate Bureau of Radiation Control Texas Department of Public Health 1100 West 49th Street Austin, Texas 78704 l
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. to TXX-99105 Page'1 of 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
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Texas Utilities Electric Company
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Docket Nos.
50-445
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50-446 (Comanche Peak Steam Electric
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License Nos. NPF-87 Station, UrJts 1 & 2)
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NPF-89 AFFIDAVIT James J. Kelley, Jr. being duly swom, hereby deposes and says that he is the Vice President -
Nuclear Engineerhg & Support of TU Electric, the licensee herein; that he is duly authorized to sign and file with me Nuclear Regulatory Commission this Request for Additional Information regarding License Amendment Request 98-010; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
James J. Kelley, Jr.
Vice President -
Nuclear Engineering & Support STATE OF TEXAS
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COUNTY OF Ent4M )
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Subscribed and sworn to before me, on this 83 day of bIMI)
.1999.
JEAN AMUNDSON d
W G MyCommsskmExpires E31-2002 l NOTARY PUBLIC Not#y Public STATEoFTEXAS l
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, Attachment 2 to TXX-99105 Page 1 of 20 Response to Questions from NRC j
Questions 1 - 13, from the Electrical, Materials and Human Factors branches, are ilsted first. Questions 14 through 18 discuss responses to the l&C branch regarding Caldon's Topical Report SER. Questions 19 - 25 are from Reactor Systems Branch.
Question 1:
Provide a discussion that addresses the impact of the proposed power uprate on the load, voltage, and short circuit values for all levels of the station auxiliary electrical distribution system. Include in this discussion any impact on the direct current power systems.
Response
As a result of this uprate, no auxiliary load ratings are expected to change, and the loads are not expected to experience demands above their ratings. Therefore, the plant auxiliary electrical load will not change. The main generator electrical parameters remain i
the same, and the uprate capacity remains within the generator rating. The voltage controls and grid source impedance at the CPSES 345 kv grid will not be affected by this uprate; therefore, the evaluated voltages and short circuit values at different levels of station auxiliary electrical distribution system will not change as a result of this uprate.
Question 2:
For the power uprated conditions, discuss environmental qualification for the safety related electrical equipment located in harsh environmental areas. For this safety-related electrical equipment, address the continued environmental qualification and the process for establishing qualification for any increased temperature, pressure, humidity, and radiation values.
Response
The normal environments for the plant buildings were qssessed. The 1% uprate has an insignificant effect on process fluid temperatures in the auxiliary, safeguards and electrical and control bui! dings. With the exception of the main feedwater, the increase in the heat loads is caused by the increase in the decay heat load as it in transferred to the Component Cooling Water and Station Service Water Systems. Thu increase in these system temperatures has been shown to be fractions of a degree. The main feedwater temperature is changing by approximately 1*F. This small change in fluid temperatures has an insignificant affect on the area temperatures. Similar conclusions were reached following the evaluations of the normal environmental conditions in the containment and fuel building.
The post-accident thermal environmental parameters were generated from computer models of the building structures that calculate the environment created by mass and energy releases during postulated pipe bresks. Evaluations concluded that through the
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, Attachpient 2 to TXX-99105
- Page 2 of 20
- use of the reduced 1% power calorimetric uncertainty to offset the'1% increase in reactor power, the existing mass and energy releases used in the environmental analyses for both inside and outside containment would remain valid. Because the mass and energy releases are not changed, the resulting environments are also unchanged.' Therefore, the 1% power uprate has no impact on the CPSES non-radiological equipment qualification program.
' Generally, postulated radiation doses impacting equipment qualification depend piimarily on post-accident contributions. However, normal-operating dose rate contributions are included in the design basis calculations. These normal-operating contributions are, in all cases, based on Westinghouse source terms which were originally generated for a power level of 104.5% RTP (i.e.,3565 MWt) and assumed 1%
' fuel defects. The assumption of 1% fuel defects is considered to be very conservative inasmuch as operation with that level of fuel leakage is not anticipated. Therefore, in regard to cases where normal operating equipment qualification dose rate contributions may be significant, it can safely be concluded that a power uprating of 1% would not cause dose rates or accumulated doses to exceed design basis values.
The effects of post-accident radiological consquences on equipment qualification were also evaluated. The source term used in the original analyses was generated for -
operation at a thermal power of 3565 MWt. Revised core fission product inventnry calculations were performed; it was concluded that the original source term remains bounding. Based on the revised core fission product inventory, the post-accident i
gamma source strengths for some energies were found to slightly increase as a result of 1
the power uprate; however, when applied in specific dose rate computations, it was 1
' shown that the accumulated doses at all times remain lower than current design-basis
' values. Therefore, it was concluded that all doses used for equipment qualification remain within existing design. basis values,
. In summary, the 1% thermal power uprate has a negligible effect on normal environmental conditions and no effect on the environmental conditions currently used for equipment qualification.
Question 3:
Discuss and verify the assumptions for the station blackout analysis are valid for
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the power uprate conditions, particularly as they relate to issues such as the heat 1
-up analysis, equipment operability, and battery capacity.
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Response
The existing calculations used to demonstrate the capability to withstand a Station Blackout event of four hours duration without uncovering the corb were reviewed for the 1% uprate conditions. The later stages of the existing analysis credit operator action to maintain the RCS temperature and pressure below specified limits; the steam generator atmospheric relief valves (ARVs) are used to accomplish this action. The capacity of the ARVs was evaluated and determined to be sufficient to accommodate the 1%
. uprated condition; therefore, the conclusions of the calculation remain valid, i.e., the time to uncover the core following a Station Blackout event is greater than four hours.
. to TXX-99105 Page 3 of 20 The existing loss of ventilation analyses for the CPSES Station Blackout (SBO)is a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> transient. The evaluation of the SBO transient is based on emergency operating procedures. Using these orocedures, a basic list of the equipment necessary to achieve safe shutdown and iestore AC power was developed. The SBO room temperatures identified in the equipment lists were calculated using transient heat-up computer models. The temperatures identified were the peak temperatures calculated for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping period. Equipment operability was assessed at those peak temperatures, except as noted for individual pieces of equipment.
The areas where the equipment environment was evaluated can be summarized as follows:
The UPS and battery rooms The control room The electrical and switchgear rooms The cable spread rooms The diesel generator rooms Tiie pipe tunnel The containment The pipe penetration area rooms The main steam and feedwater penetration areas The turbine driven feedwater pump room The instrument air compressor room The outdoors (turbine building, safeguards roof)
The expected increase in the RCS, mair steam, feedwater, and steam generator blowdown operating temperatures associated with the power uprating does not affect the heat loads used to calculate the temperature transients for the first six items. This is because these areas are primarily electrical areas that are not exposed to these process i
fluids.
The containment environment during a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event is significantly less limiting (by j
greater than 120 F) than the thermal profiles considered for LOCA/MSLB events. A small change in decay heat and initial process temperatures cannot result in a change of such magnitude that the calculated LOCA/MSLB environment will be exceeded.
Therefore, it was concluded that a small change in RCS temperature, decay heat, main steam and feedwater temperatures would have no effect on the equipment as evaluated for the SBO event.
The concern for the pipe penetration area rooms is the potential increase in the room heat load resulting from an increase in the steam generator blowdown line temperature.
The room temperatures used in the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO evaluation were obtained from the 30 day loss of HVAC analysis. The piping heat load input used in the loss of ventilation analyses assumed that the unit was also in a LOCA (since the signals obtained from the LOCA tripped the non-safety HVAC). In the 30 day loss of HVAC analyses, the piping heat loads in these rooms included RHR, CVCS, and/or containment spray fluid flowing from the containment sump at post LOCA temperatures (in excess of 200 F), as well as component cooling water (CCW) post accident temperatures. Steam generator blowdown piping is also routed through these rooms. A temperature of 550 F was used in the piping heat load and was held constant for the duration of the transient. The
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, to TXX-99105
' Page 4 of 20 steam generator blowdown is isolated in the SBO scenario and only that portion of the piping up to the blowdown isolation valve would remain hot. The heat source from the remaining piping would decay throughout the transient.
Since the operation of the containment spray, RHR, CCW systems are not postulated in the SBO teenario, it can be concluded that the effects of small changes in steam generator blowdown temperatures are bounded by the significantly larger post-accident piping heat loads. Therefore, small changes in steam generator blowdown temperatures do not impact the environment and the equipment already evaluated fot the SBO event.
The primary heat loads in the main steam and feedwater piping penetration areas are obviously from the main steam and feedwater piping. The power uprate results in a lower operating steam temperature and no change to the no-load steam temperature.
Therefore, the heat load resulting from the main steam lines will actually decrease during power operation and remain constant at zero power.
The feedwater temperature used in the SBO loss of HVAC analyses was 440*F. The estimated feedwater ternperature for the 1% uprate is 441 *F. The increase of 1 *F is expected to have an insignificant affect on the results of these analyses, especially when it is considered that the feedwater is isolated in the transient and the piping is insulated.
Based on the preceding discussions, it is concluded that the small changes in steam temperature and feedwater temperatures do not adversely impact the environment and the equipment already evaluated for the SBO event.
The primary heat load in the turbine-driven auxiliary feedwater pump room is from the main steam piping feeding the turbine. The power uprate results in a lower operating steam temperature and no change to the no-load steam temperature. Therefore, the heat load resulting from the main steam h.es will decrease during power operation and remain constant at zero power.
The primary concern in the instrument air compressor room is the potential increase in the room heat load resulting from an increase in the steam generator blowdown line temperature. This room contains the steam generator blowdown heat exchangers and associated piping for both units. The existing calculations modeled the fluid temperature
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as a constant heat source of 550 F for the duration of the transient. In addition, the instrument air compressors heat loads were included in the calculations. The affected calculations were reviewed, and it was concluded that the conservative modeling of the heat load sources bound small increases in the blowdown temperature.
1 The power uprate modification does not change the environment outdoors, therefore, there is no impact to the eppment evaluated for SBO.
l To provide for an oraerly and safe cooldown of the unit during a station blackout event, the following conditions must be met.
j the turbine driven auxiliary feedwater pump must operate to provide feedwater to the SGs, l
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- , Attachment 2 to TXX-99105
- Page 5 of 20 the SG atmospheric relief valves (ARVs) must cycle open to relieve steam for unit cooldown, and
. an adequate supply of water from the' condensate storage tank must be available to maintain adequate water level in the steam generators.
To accomplish these tasks, specific air operated valves in the main steam system and
- the auxiliary feedwater system must be able to be operated from air accumulators that have sufficient capacity to cycle the valves as needed during the controlled unit cooldown. In each case, the required number of valve cycles was established independent of and was determined to be reasonably insensitive to the actual power level. Accordingly, there is no change in the required operation of the AOVs for unit cooldown during a station blackout event as a result of either the 1% or the 4.5% power upratings and the AOV accumulator sizes are therefore sufficient to provide a safe cooldown during a SBO event.
An evaluation was also performed in which it was concluded that the current minimum available safety grade condensate inventory in the condensate storage tank is sufficient for the 1% and 4.5% uprate conditions.
Question 4:
Provide a discussion addressing the impact of the CPSES Unit 2 power uprate on the turbinelgenerator, isophase bus, main transformers, and switchyards.
- Address in detail any non hardware changes for these items as a result of the CPSES Unit 2 power uprate.
Response
The following major turbine system components have been evaluated and were determined to be acceptable for continuous operation under the new operating conditions associated with a core rated thermal power of 3445 MWt. The potential missile energy from the high-pressure turbine is less than that from the low pressure turbine because of its much smaller potential missile mass and thicker turbine casing.
Thus, the high-pressure turbine potential missiles are bounded by the low pressure turbine potential missiles.-
The turbine drain system was evaluated. The required drain system capacity is based on start up and low load conditions, which are not affected by the power uprate.
Therefore, it was concluded that the system capacity is adequate and no changes are required.
The moisture separator-reheaters (MSRs) and their sub-systems and components were reviewed and found to be adequate for service under the 1% uprate operating conditions. The MSR and hot reheat piping design pressures and temperatures are not changed by the uprate modification. Therefore, the lift pressure of the three shell side MSR safety valves is within the design requirements and no changes are required.
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. Attachment 2 to TXX-99105 l
Page 6 of 20 The turbine and its auxiliaries have been reviewed for operational impacts of the 1%
uprate on the relevant normal and abnormal modes of operation, and it was determined that there is no adverse impact.-
The following operational parameters and systems are monitored during start-up and operational changes. Each has been reviewed and the determination was made that the 1% uprate does not require any operational changes:
Main Steam Pressure Stage Pressures Stage and Exhaust Temperatures
_ Casing Temperature e'
Component Stresses Expansions The Bentley Nevada Turbine Supervisory Instrumentation System The Siemens Berhhrungsloses Schaufel-Schwingungs information System (BeSSI) a LP Blade Vibration Monitoring System The turbine system instrumentation and controls equipment does not constitute a hardware-imposed operating limit. However, specific setpoints will be revised for impacts on turbine operating parameters that are affected by the uprate. The original
. turbine design encompassed a " Valves Wide Open" operating condition representing a 5% increase in steam supply to the turbine relative to the 100% nominal operating point.
This bounding design condition provides assurance that instrument ranges and capabilities retain adequate margb to accommodate the 1% uprate.
The T,,, program usas turbine first stage (impulse chamber) pressure as an input to 1
maintain the Reactor Coolant System at the appropriate temperature (T,l) load turbine
. The installation of the new high-pressure turbine is expected to change the fu first stage (impulse chamber) pressure, resulting in the need to adjust the scaling of the turbine first stage (impulse chamber) pressure instrument channels, in order to more easily make these adjustments and to accommodate making them at power, if necessary, these instrument channels will be re-scated in units of percent turbine load.
The electrical systems associated with the turbine auxiliary systems are not affected by the uprate.
- The Unit 2 steam turbine-driven polyphase generator is a four polo machine rated at 1350 MVA, with an operating point of 1215 MWe at a 0.9 power factor. This rating is based upon 60 psig hydrogen pressure, which is supplemented with water cooling for
. the stator and rotor.
- Historically, Unit 2 has operated at a peak of 1167 MW. As the anticipated net increase of 20 MW (12 from the 1% uprate, and 8 from the HP turbine uprate) lies within the i
. Attachment 2 to TXX-99105 Page 7 of 20 nameplate rating of the generator, there will be no equipment limitations to prevent operation at a core power of 3445 MWt.
A review of applicable calculations identified no need for any changes to equipment i
protection relay settings for the generator; although some process alarm setpoints for the generator and the exciter may require adjustment.
To deliver electrical power provided by the generator to the transmission system, each unit is equipped with an isolated phase bus, two main transformers, cabling, and two 1
switchyard breakers. With the exception of the Unit 2 main transformers, which are rated for 650 MVA each, the remaining components are rated to deliver electrical power at or in excess of the main generator nameplate rating of 1350 MVA.
i The isophase bus main section is rated at 37,000 amps, with each main transformer branch rated at 18,000 amps. The bus conductor will permit a temperature rise of 55*C, with the enclosure rated at 30*C rise. This will permit a total load (assuming a nominal voltage rating of 22 kV and 36,000 amps) of 1372 MW. These figures are well in excess of the anticipated generatcr output of 1187 MW. The Isophase Bus will support the power increase with no modifications.
The Unit 2 transformers have a total capacity of 1300 MVA, just slightly less than the output rating of the generator (1350 MVA.) Since the reactive power of the generator must remain below approximately 400 MVARs due to the voltage rating on the primary windings, most of the MVA capacity can be utilized for real power. With an anticipated increased output of 1187 MW from the generator, and assuming a maximum reactive output of 400 MVARs, this will result in an apparent output of 1253 MVA. Therefore, the transformers will operate within all applicable limits at the 1% power uprating conditions.
Standard design practice at TU Electric requires that switchyard equipment at least meets, but often exceeds, the nameplate rating of the main generator. The switchyard will accept the additional load without the need for any hardware modifications, in summary, the turbine / generator and major electrical components extending from the isophase bus to the switchyard have adequate design margin to accept the additional power anticipated by the 1% uprate.
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. to TXX-99105 Page 8 of 20 Question 5:
Discuss the impact of the CPSES Unit 2 power uprate electrical conditions on the
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current grid stability and reliability analysis. Describe in this discussion, how the i
station continues to be in conformance with General Design Criterion 17 with CPSES Unit 2 at the power uprated electrical conditions.
Response
The capacity of a single CPSES unit is less than 3% of the Electric Reliability Council of l
Texas (ERCOT)-estimated peak load. The uprate of Unit 2 capacity is negligi' le and u
Unit 2 capacity wi!! still remain less than 3% of the ERCOT-estimated peak load. Actual disturbances on the ERCOT system have occurred where large amounts of capacity were lost, as high as 10%, with no integrity degradation of the transmission system observed. The power uprate of Unit 2 will not impact grid stability and reliability. Finally, availability and reliability of electric power from the transmission network to CPSES will
- not be affected, therefore the station will continue to be in conformance with GDC 17.
Question 6:
Provide a pressurized thermal shock evaluation for the CPSES Unit 2 reactor vessel before implementing the power uprate and after implementing the power uprate.
Response
The highest current RTns end of license value for the Unit 2 reactor vessel has been previously docketed as 94*F which is 170*F below the screening criteria of the PTS Rule (10CFR50.61). Two Unit 1 and one Unit 2 surveillance capsules have been analyzed confirming the similarity between the two vessels in irradiated and non-irradiated material properties. The results of these surveillance capsule evaluations have confirmed that the early projections for CPSES vessel materials were conservative. In addition, the majority of the irradiation-induced shift in vessel material properties occurs early in life. Therefore, with substantial margin to the RTns screening criteria and a nominal 1% increase in fluence, the change in the RTns value would not be significant and a revised Pressurized Thermal Shock report is unnecessary.
, to TXX-99105 Page 9 of 20 Question 7:
I What is the calculated end-of-life fluence in the current vessel design of CPSES Unit 2? What is the expected fluence for pressurized thermal shock with the revised design conditions / power uprate for CPSES Unit 27 1
Response
l The existing fast neutron f!uence data used in the reactor vessel design remains bounding for the uprated power conditions. This conclusion is based on a recent fluence i
evaluation performed in conjunction with the withdrawal of surveillance capsule Y (second capsule) from the CPSES Unit i reactor. In this evaluation, the inclusion of the impact of low leakage fuel management reduced the Unit 1 fluence projections by approximately 33% relative to the values used in the Unit i reactor vessel design. A similar reduction is anticipated for Unit 2 when the upcoming second surveillance capsule evaluation is performed for the Unit 2 reactor. This 33% margin more than offsets the 1% increase in fluence that could be caused by a 1% uprating. Thus, the fluence values used in the design bound the new best estimate fluence projections including consideration of a 1% uprating.
Question 8:
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Does the power uprate for CPSES Unit 2 change the cold leg temperature? If so, please provide details.
Response
The reference value for T,, will not be changed as part of the power uprate modification. However, because the core power will be increased by 1%, the AT will increase by 1%. The current AT is less than 60*F. Thus, with a constant value of T,
% s expected to increcse T,is expected to decrease by approximately 0.3*F, and T i
approximately 0.3*F. The expected cold leg temperatures remain within the range assumed in the development of the equations and tables which form the bases for evaluating the neutron irradiation effects on vessel integrity.
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Question 9:
Discuss whether the power uprate will change the type and scope of plant emergency and abnormal operating procedures. Will the power uprate change the type, scope, and nature of operator actions needed for accident mitigation and will new operator actions be required?
Response
The power uprate represents a modest 1% increase in the core thermal power level. A review of plant operations has concluded that an increase of this magnitude does not require any material modifications to plant procedures. Further, the responses of the reactor operators to any event will be unaffected by a change of this magnitude.
Question 10:
Provide examples of operator actions that are particularly sensitive to the proposed increase in power level and discuss how the power uprate will effect operator reliability or performance. Identify all operator actions that will have their response times changed because of the power uprats. Specify the expected response times before the power uprate and the new (reducedlincreased) response times. Discuss why any reduced operator response times are needed.
Diccuss whether any reduction in time available for operator actions, due to the power uprate, will significantly affect the operator's ability to complete the required manual actions in the times allowed. Discuss results of simulator observations regarding operator response times for operator actions that are potentially sensitive to power uprate.
Response
The power uprate represents a modest 1% increase in the core thermal power level. A review of plant operations has concluded that an increase of this magnitude does not require any material modifications to plant procedures. Further, the responses of the reactor operators to any event will be unaffected by a change of this magnitude.
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, Attachment 2 to TXX-99105 Page 11 of 20 Question 11:
Discuss all changes the power uprate will have on control room alarms, controls, and displays. For example, will zone markings on meters change (e.g., normal range, marginal range, and out-or tolerance range)? If changes will occur, discuss how they will be addressed.
Response
No changes to control room alarms, controls and displays are required as a direct result of the power uprate. When the power uprate is put in place, the Nuclear instrumentation System will simply be adjusted to indicate the new 100% RTP in accordance with Technical Specification requirements and plant administrative controls. Because this power uprate is predicated on the availability of the LEFM/, procedural guidance, supplemented by plant computer displays, will be developed to facilitate operation when the LEFM/ is unavailable. The reactor operators will be trained on the changes in a manner consistent with any other design modification.
Question 12:
Discuss all changes the power uprate will have on the Safety Parameter Display System (SPDS) and how they will be addressed.
Response
l The SPDS is unaffected by the proposed 1% increase in Rated Thermal Power.
' Question 13:
i Describe all changes the power uprate will have on the operator training program and the plant simulator. Provide a copy of the post-modification test report (or test abstracts) to document and support the effectiveness of simulator changes as required by American National Standards institute /American Nuclear Society (ANSI /ANS) 3.5-1985, Section 5.4.1.
Specifically, please propose a license condition and/or commitment that stipubias the following:
(a)
Provide classroom and simulator training on all changes that effect operator performance caused by the power uprate modification.
(b)
Complete simulator changes that are consistent with ANSI /ANS 3.5-1985. Simulator fidelity will be re-validated in accordance with
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. to TXX-99105 Page 12 of 20 ANSI /ANS 3.5-1985, Section 5.4.1, " Simulator Performance Testing."
Simulator revalidation will include comparison of individual simulated systems and components and simulated integrated plant steady state and transient performance with reference plant responses using similar startup test procedures.
(c)
Complete all control room and plant process computer system changes as a result of the power uprate.
(d)
Modify operator training and the plant simulator, as required, to address all related issues and discrepancies that are identified during the startup testing program.
Response
The CPSES simulator uses Unit 1 as the reference plant. Because the power uprate is only for Unit 2, no simulator modifications are required. The reactor operators will be trained on the unit differences, in the same manner as is currently used. Further, the modest 1% power uprate is not expected to have any significant effect on the manner in which the operators control the plant, either during normal operations or transient conditions.
Question 14:
The licensee should discuss the maintenance and calibration procedures that will be implemented with the incorporation of the LEFM. These procedures should include processes and contingencies for inoperable LEFM instrumentation and the effect on thermal power measurement and plant operation.
Response
4 The Caldon LEFM 8300 system is currently installed at CPSES. The existing maintenance requirements (scope and frequency) and calibration procedures for the LEFM 8300 system will be revised per the design control process to incorporate the vendor's requirements for the new LEFM/ system.
Current Operations procedures are used to perform a unit calorimetric measurement for the purpose of calibrating the Power Range NIS and N-16 channels. Contingencies and instructions are currently in the procedure in the event that the LEFM system is unavailabic. This procedure will be revised per the design control process to incorporate the requirements for the new LEFM/ system. In addition, more formal guidance, including routine surveillance requirements for the LEFM/ and appropriate l
contingency actions, will be provided in the Technical Requirements Manual. This l
guidance directs the operators to operate the plant consistent with the accident analyses and the uncertainty associated with the attemate methods of determining the plant t
Thermal Power (i.e.,LEFM/ or venturi-based indications of feedwater flow).
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, Attachment 2 to TXX-99105 Page 13 of 20 Question 15:
For plants that currently have LEFMs installed, the licensee should provide an evaluation of the operational and maintenance history of the installed installation
- and confirm that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Topical Report ER-80P.
Response
The LEFM 8300 system is currently installed at CPSES. Since 1995, CPSES has experienced no operational or maintenance problems with the LEFM except for occasional failure of a transducer. (The failure of a transducer results in loss of signal and does not cause an error in flow measurement. Transducers can be replaced on-line). The existing LEFM 8300 system, even though it is as accurate and reliable as the new LEFM/ system, does not contain all the intemal reliability checks and thus, is not bounded by the analysis and assumptions set forth in the Topical Report ER-80P.
Installation of the new LEFM/ system at CPSES will meet the requirements of the Topical Report. The new LEFM/ system at CPSES is the same LEFM/ system that formed the basis of the analysis in the Topical Report. Commissioning to be completed following installation of the new electronics and software will document that the new system is bounded by the Topical report. This documentation will be available for inspection.
i Question 16:
The licensee should confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current feedwater instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If an alternative methodology is used, the application should be Justified and applied to both venturi and ultrasonic flow measurement instrumentation installations for comparison.
Response
The LEFM/ methodology used to calculate the power calorimetric uncertainty based on the improved LEFM is exactly the same as the methodology presented in the Topical Report. This methodology is as recommended by ASME PTC 19.1 - 1985, Measurement Uncerfainty. In addition, the calorimetric power uncertainty analysis currently in place for use of the feedwater venturi-based instrumentation at Comanche Peak is also in accordance with ASME PTC 19.1. This method is the same as that precented in the Topical Report for the venturi-based lastrumentation, with the exception
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., to TXX-99105.
Page 14 of 20 that TU Electric treats the venturi coefficient error from loop to loop as a dependent, or systematic, error in flow. As a result, the feedwater flow uncertainty is act reduced by a factor of two to account for the four individual loop measurements, as is believed to be a more typical practice.
Question 17:
L'censees for plant installations where the ultrasonic meter (including LEFM) was not installed with flow elements calibrated to a site specific piping configuration (flow profiles and meter factors not representative of the plant specific installation), should provide additional Justification for use. This justification i
should show that the meter installation is either independent of the plant specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and plant configurations for the specific installation including the propagation of flow profile effects at higher Reynolds
. numbers. Additionally, for previously installed calibrated elements, the licensee i
should confirm that the piping configuration remains bounding for the original i
LEFM installation and calibration assumptions.
Response
The present system installed at CPSES uses an in-line flow meter spool piece installed in the feedwater header. A plant specific evaluation, including test results and calculations, were used to develop the installation-specific profile factor. There has been no change to the feedwater configuration which has impacted the assumptions, analysis, or the calculations which determined the site specific profile factor since the installation of the LEFM 8300 system. The design modification to install the LEFM/
system will not impact the spool piece hydraulic configuration. Only the electronics and software will be impacted. Therefore, the piping configuration remains bounded for the original installation and calibration assumptions. The justification for the profile factor uncertainty at Comanche Peak is described in some detail in the response to Question No.19 in Responses and Further Clarifications to NRC Questions from September 29,1998 Meeting, dated December 15,1998, and is incorporated here by reference.
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~ Page'15 of 20 Question 18:
j Based on the above, the staff finds that feedwater flow measurement using the LEFM can provide a thermal power measurement that will remain bounding within an uncertainty of 1% of rated thermal power. This is premised on the assumption that no additional uncertainties beyond those included in Topical Report ER-80P are assumed to be included in the 10 CFR Part 50, Appendix K 102% thermal power margin requirement.
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Response
TU Electric concurs that no additional uncertainties, beyond those included in the Caldon Topical Report ER-80-P, are assumed to be included in the 10CFR Part 50, Appendix K 102% thermal power margin requirement.
l Question 19:
The amendment request proposes to reduce the margin for assumed power level for non-LOCA accident and transient analysis on the same basis as the proposed exemption to the Appendix K ECCS evaluation requirement. Staff consideration of the related Appendix K exemption request was in part based on the premise that the power level requirement is one of several conservative features that, taken together, provide substantial conservatism in ECCS analyses.
Justify the proposed margin reduction for non-LOCA analyses that currently i
assume 102% power. The justification should include a quantitative or qualitative discussion of conservative analysis assumptions for the non-LOCA accidents and transients and the safety margin they provide relative to the power level margin assumption.
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Response
TXX-98265 requested approval for an exemption to the 2% uncedainty value that is I
both imposed by 10 CFR 50, Appendix K and also used in the topical reports identified in CPSES Technical Specification 6.9.1.6b. These topical reports describe the NRC-approved methodologies that support the CPSES safety analysis, including the small break and large break loss of coolant accident analyses, in many of these topical i
reports, reference is made to the use of a 2% uncertainty applied to reactor power, consistent with Appendix K. TU Electric proposed that these topical reports be approved for use with a 1% uncertainty consistent with the exemption request. The NRC-approved Caldon Report ER-60P was provided as the basis for this change. In this report, it is argued that with the use of the improved instrumentation, the probability that the actual reactor power would exceed 102% of the current RTP, when operating at a steady-state power level of 101% RTP, was less than the probability of exceeding t
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. Attachment 2 to TXX-99105 Page 16 of 20 L
' 102% of the current RTP, when operating with the traditional instrumentation at 100%
RTP. Based on this argument, there is no margin reduction associated with operation at l
"101% RTP" with a 1% calorimetric uncertainty when using the improved LEFM i
l instrumentation.
J J Other conservative assumptions in the accident analyses are unaffected by the change In the uncertainty allowance applied to the initial core power level. For the non-LOCA events presented in FSAR Chapter 15, conservative initial conditions are used as described in FSAR Section 15.0.3.2. Other conservative assumptions considered in the non-LOCA transient analyses are described in FSAR Section 15.0.3. These assumptions include conservative core power distributions and peaking factors,-
conservative moderator and Doppler fuel temperature reactivity feedbacks, a small value of the control rod trip reactivity worth and conservatively-skewed trip reactivity insertion characteristics. With respect to available equipment and instrumentation, the beneficial effects of control systems are not credited in the analyses, and, in addition, a
. single failure of equipment or instrumentation required to mitigate the transient is assumed.
The generic models and methods used to analyze the transients are described in the methodology topical reports listed in Improved Technical Specification 5.6.5 and incorporated by reference into the FSAR. Factors that make up the inherent conservatism of the models and methods include the use of the point-kinetics approximation in lieu of a multi-dimensional representation of the reactor core and simplified steam generator models developed to conservatively predict the primary-to-secondary heat transfer rate.
In addition to these somewhat generic assumptions, additional conservative assumptions or models may be applied to specific transients.' The more significant of these assumptions are delineated in the " Method of Analysis" discussion provided for each transient analysis presented in FSAR Chapter 15. Examples include maximized main and auxiliary feedwater flows for the main steam line break analysis and minimized auxiliary feedwater flows for the loss of feedwater analysis.
As described in FSAR Section 6.2.1.4, many of these same types of conservative assumptions are applied in the development of the steam line break mass and energy releases used to evaluate the containment response. The steam line break mass and energy release calculation developed for use in the environmental analysis outside of containment is described in FSAR Section 3.6B.2.5.2 Subsection 1. D. This analysis contains the typical, generic, conservative assumptions, as well as additional assumptions designed to increase the severity of the event with respect to the acceptance criteria for this specific application.
in summary, the allowance provided for the power calorimetric uncertainty is but one of sevural conservative assumptions that are applied to each of the safety analyses.
However, through the use of the improved LEFM/ instrumentation, the use of a smaller value of the power calorimetric uncertainty does not result in a reduction of analytical i
margin in the safety analyses.
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'Page i7 of 20 Question 20:
Increasing licensed power level would result in an increased heat source that could affect the progression of certain accidents. Discuss the potential impact of plant operation at the higher proposed power level on ATWS progression, containment integrity analyses, and on overall IPE results.
Response
The proposed increase of 1% in the Rated Thermal Power is not sufficient to materially affect the proqression of any event. Discussions of the effects of the proposed uprate on specific events are provided below; a)
ATWS progression: The current CPSES analytical basis for this "beyond design basis" event is provided in FSAR Section 15.8. Incorporated by reference is a letter from T. M. Anderson of Westinghouse to S. H. Hanaver of the NRC, "ATWS Submittal," NS-TMA-2182, dated December,1979. The applicable analyses described in that document are based on an NSSS power of 3427 MWth; however, an additional sensitivity study is provided to address a power level 2% higher. Acceptable results were obtained for this case. The proposed 1% power increase falls within the parameter range analyzed in this document. Therefore, there is no significant effect on the ATWS progression.
b)
Containment integrity analyses: The mass and energy release calculations used to evaluate the containment integrity were performed at power levels of up to 102% RTP. For the LOCA mass and energy release calculations, a higher power level of 104.5% RTP was used. A spectrum of lower initial power levels was also considered. Analyses initiated from lower power levels were found to be limiting for most containment analyses; thus, these analyses remain unaffected. Those analyses initiated at the current RTP, or higher power levels, included a 2% power uncertainty. As previously described, through the use of the improved LEFM instrumentation, the 1% power uncertainty is used to offset the increase in the operating power level. In all cases, it was determined that the mass and energy release calculations remained valid; therefore, the containment integrity analyses are unaffected by the proposed 1% uprate.
c)
The success criteria used for the CPSES IPE were reviewed and found to not be materially affected by the modest 1% power uprate. Therefore, it is concluded that the overall IPE results are similarly unaffected, and the proposed power uprate is not risk-significant.
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Discuss the impact on LOCA and non-LQCA analysis results (e.g., main steam line break) of the revised values for RCP heat addition and RCS flow rate included in the amendment request.
Response
The RCS flow rate used in the LOCA and non-LOCA accident analysis is not being revised through this amendment request. In order to provide flexibility for future RCS flow reductions (e.g., due to significant SG tube plugging), the RCS stress evaluations were performed for a range of parameters that bound both current operation and potential future operation.
In the computer codes used to analyze the non-LOCA and LOCA transients for the last several cycles of operation, the RCP performance during steady-state and transient conditions has been modeled through the use of homologous curves. The net RCP heat addition is the difference between the RCP heat addition and the various RCS heat losses (e.g., charging and letdown). The RCP heat addition is calculated, based on these homelogous curves, to be greater than 19 MWth. The heat losses are not modeled. Therefore, the effect of the increase in the net RCP heat addition has been conservatively considered in the current accident analyses.
Question 22:
Provide the detailed calculational basis to substantiate the statement made in the amendment request that a 10-percent SG tube plugging level supports a peak plugging level of 15% in any one SG, provided that the average level of plugging of all four SGs is no greater than 10 percent. Explain the difference between the plugging level used in the analysis discussed in the amendment request and the plugging level assumed in the current LOCA analysis?
Response
The SG tube plugging level assumed in the current LOCA analyses of record is 5%
The LOCA analyses are not being revised as a result of the proposed 1% power increase. However, the RCS stress evaluations were performed to support a wide range of parameters that bounds both current operation and potential future operation.
The allowance for the asymmetric tube plugging is based on the requirement that the
- effects of the asymmetry on the NSSS parameters are within the ranges considered in the RCS stress ovaluations. Although the effects on the NSSS components have been evaluated, there are clearly other analyses that must be considered before the plant will be operated with significantly asymmetric tube plugging levels.
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l Question 23:
Plant response to SGTR and other events depends on SG atmospheric relief valve operation. Reactor operation at higher power levels may cause these valves to operate more often in the event of certain events, thereby affecting their reliability. Discuss the effects of operation at the proposed new power level on 3
l the possible increased challenge to these valves and their expected failure i
frequency during a SGTR event (and other events requiring their operation).
Response
While it is true that transients initiated from a higher power level may present more challenges to the ARVs, the frequency of such challenges is not considered to be significant. The proposed increase in reactor power is very modest. The capacity and i
reliability of the Steam Dump System are such that the ARVs are not anticipated to be operated any more frequently than they are currently cycled.
l Question 24:
When considered in terms of core power, the proposed changes in power range neutron flux, and overpower N-16 nominal and allowable reactor power trip levels appear slightly non-conservative. Explain the basis for the proposed revision to the N-16 overpower and power range neutron flux trip set points given in the amendment request. Provide justification for the apparently non-conservative set point changes.
Response
The values for the nominal trip setpoint and allowable value were calculated using the essential elements of the methodology described in the Westinghouse report WCAP-12123, Revision 2, which has been previously placed on the CPSES dockets. This methodology is based on a statistical combination of the uncertainties associated with a particular trip function. Sufficient margin is provided between the safety analysis limits (the value of the trip setpoint assumed in the accident analyses) and the nominal trip setpoint to accommodate the channel's uncertainties. The M >wable Value is then calculated as described in the aforementioned report.
The safety analysis limits for both the overpower and high neutron flux reactor trip i
functions were set at 118% of 3411 MWth. When operating at the uprated condition, i
the "new" safety analysis limit for each of these functions is 116.9% of 3445 MWth.
l Sufficient margin was available in the uncertainty analyses supporting the Technical Specification values of the high neutron flux reactor trip function to absorb the effects of the " reduction"in the safety analysis limit. New values of the Nominal Trip Setpoint and
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Page 20 of 20 Allowable Value were necessary to provide the required margin for the overpower N-16 l
' reactor trip function.
Question 25:
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- The N-16 overtemperature trip setpoint was not changed in the amendment request, based on the statement that it was previously analyzed at the power level l
requested in the proposed amendment. Confirm that the other proposed changes l
to plant parameters such as RCS flow and coolant temperatures do not result In a change to the N-16 overtemperature trip setpoint. Explain how the proposed changes in core flow rate and coolant temperatures affect the calculation of the N-16 overtemperature trip setpoint.
Response
The overtemperature N-16 reactor trip setpoint provides DNB protection from events that affect power, pressure, temperature or axial power distribution, do not affect the RCS flow rate, and are slow relative to loop transit times. The overtemperature N-16 reactor trip setpoint, calculated to su ) port Unit 2 Cycle 5 operation and approved via Amendments 63/49 to the C 'SES Technical S secifications, considered operation at the uprated power level of 3445 Mw;h. The base temperature used in the overtemperature setpoint, T/, was selected to be conservative relative to operation of 3411 Mwth and consistent with operation at 3445 MWth. The RCS flow rate upon which the ovedemperature setpoint is based is consistent with the current Technical Specifications. As previously described, the minimum required RCS flow rate specified in the Technical Specifications is not being revised at this time. Thus, the current ovedemperature N-16 seboint remains valid for operation during Unit 2 Cycle 5 at the uprated power of 3L45 MWth, l
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