ML20097D257
| ML20097D257 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/02/1996 |
| From: | Woodlan D TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M94167, TAC-M94204, TXX-96040, NUDOCS 9602120276 | |
| Download: ML20097D257 (13) | |
Text
-. . . _ _ _ _ _ = _
i o
Log # TXX 96040 F9 File # 916 (2)
_ = 10010
= = Ref. # 10CFR50.90 10CFR50.36 i M/ELECTR/C February 2, 1996
- c. Lance Terry Group Mce President U. S. Nuclear Regulatory Commission Document Control Desk Washington. DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50 445 AND 50-446 SUPPORTING INFORMATION FOR LICENSE AMENDMENT REQUEST 95 008 UNIT 2 RELOAD ANALYSES AND UNIT 1 REACTOR C0OLANT FLOW (TAC NOS. M94167 AND M94204)
REF: 1) TV Electric letter logged TXX 95288 from C. L. Terry to the NRC dated November 21, 1995 I
- 2) NRC letter requesting addition information concerning CPSES License Amendment Request 95 008 from T. J. Polich to C. L. Terry dated January 30, 1996. l Gentlemen:
By Reference 1) above. TU Electric requested an amendment to the CPSES Unit 1 Operating License (NPF 87) and CPSES Unit 2 Operating License (NPF-89). The License Amendment would change the CPSES Units 1 and 2 Technical Specifications by revising the core safety limit curves and the N 16 Overtemperature reactor trip setpoints. In addition, the minimum required Reactor Coolant System (RCS) flow is increased and an administrative '
enhancement is included in the footnotes of the RCS flow low reactor trip function setpoint.
By this letter TV Electric provides information to facilitate review of the License Amendment Request in response to Reference 2) above.
The attached information is typically included in each Reload Safety Evaluation performed in accordance with 10CFR50.59 prior to the start of a specific operating cycle and, for Unit 2 Cycle 3, is predicated on the approval of the License Amendment Request. The 50.59 evaluation and supporting calculations will be available for audit upon completion of Cycle 2 and finalizing of Cycle 3 design for Unit 2: however, the requested information is summarized and reproduced here to support the request for a timely review.
9602120276 960202 PDR P
ADOCK 05000445 PDR 120061 Energy Plaza 1601 Bryan Street Dallas, Texas 75201 3411
- - _ . .- . . . . _ . - _ ~ - . - . . . - . . - .-. - . .. .._ - - . ~
! Page 2 of 2 Please contact My. J. D. Seawright at 214/812-4375 if further information is needed to complete the review.
Sincerely.
C. L. Terry By:
D. R. Woodlan Docket Licesing Manager JDR/jr Attachment cc: Mr. L. J. Callan. Region IV Mr. W. D. Johnson. Region IV Hr. T. J. Polich. NRR w\encls (2)
Resident Inspectors clo Mr. Arthur C. Tate Bureau of Radiation Control Texas Department of Public Health 1100 West 49th Street Austin. Texas 78704 I - _ _ ..
J
. 1 Attachment to TXX-96040 Page 1 of 11 NRC Ouestion 1:
You have discussed the use of different co resident fuel assembly designs in reference 1 (page 1 of 13, Attachment 2). Please provide the reference for the method that has been used for the core reload with mixed fuel for CPSES Unit 2. Cycle 3. Have all the provisions from the reference been satisfied such as that required for the analysis for the effect of stress from seismic forces between the different fuel types (Siemens and Westinghouse) and the DNBR penalty factors required for transition cores?
TU Electric Resoonse: )
1 A. The methods used for the calculation of the mixed core DNB penalty are the same as those in the TV Electric report "VIPRE-01 Core Thermal Hydraulic Analysis Methods for CPSES Licensing Applications," which is identified in TS 6.9.1.6b, Item 12. A Unit 2, Cycle 3 full core model was developed and used to assess the effects of the mixed core on the DNBR.
B. The effects of the mixed core on the large break LOCA analysis were evaluated in accordance with TV Electric report, "Large Break Loss of Coolant Accident Analysis Methodology " TS 6.9.1.6b, Item 15.
C. Both mechanical and thermal-hydraulic compatibility between the co resident Westinghouse (W) and Siemens Power Corporation (SPC) supplied fuel assemblies are evaluated in the Reload Safety Evaluation. Both SPC and W have performed evaluations which demonstrate that their respective fuel assembly designs meet all applicable design criteria. In addition SPC has evaluated the interaction between the co-resident W supplied and SPC supplied fuel assemblies and has confirmed that all applicable design criteria are satisfied.
NRC Ouestion 2:
You have discussed meeting the minimum measured flow requirement in Technical Specification (TS) 3.2.5c in Reference 1. Will this reload incorporate low leakage core loading? If so, this type of loading has resulted in increased hot [ leg] streaming in many plants that has resulted in reduced indicated Reactor Coolant System (RCS) flow rates. Will this reduced indicated RCS flow be a problem for CPSES Unit 2, Cycle 37 Please provide the total flow rates in ppm measured from the calorimetric heat balance for the current cycles for Units 1 and 2. Also please provide the references that approved the 1.8% uncertainty for the flow measurement and the 0.5% for the effects of the lower plenum flow anomaly.
Attachment to TXX 96040 Page 2 of 11 TU Electric Resoqnigi A. The Unit 2 Cycle 3 core configuration is a " low leakage" core design, as were the Unit 2 Cycle 2 and Unit 1 Cycles 2 through 5 core configurations. The reduction in indicated RCS flow seen with i low leakage core designs is caused by hot leg temperature streaming. i At CPSES the N 16 based Transit Time Flow Meter (TTFM) is used to i perform the precision flow calorimetric measurement. This l measurement technique, and the associated accuracy of _ the flow measurement, is unaffected by the hot leg temperature streaming '
phenomenon. The evaluation of the existing flow margin is based on Unit 2 Cycle 2 operation, in which a low leakage core configuration is used. No significant degradation in flow in anticipated for .
Cycle 3.
]
B. For CPSES-1, Cycle 5, the "as measured" RCS flow rate was 410,948 gpm. For CPSES 2, Cycle 2, the "as measured" RCS flow rate was 421,610 gpm.
C. The 1.8% uncertainty for the RCS flow measurement is incorporated into Technical Specification 3.2.5 and is based on uncertainty calculations originally performed for CPSES 1 by Westinghouse. The calculations have since been updated by TU Electric for both CPSES units using the Westinghouse methodology. The 1.8% allowance remains valid for CPSES 2. The 1.8% uncertainty is approved in the CPSES Technical Specifications (through Amendments 44/30) and in NUREG 0797, SSER 12, Pages 4 1 and 4-2.
D. The allowance for the lower plenum flow anomaly was obtained from WCAP-11528. "RCS Flow Anomaly Investigation Report." April 1988 and confirmed through plant specific measurements.
NRC Ouestion 3:
Please provide the reference for the approved method used for obtaining the Overtemperature N 16 reactor trip setpoint [and] for obtaining the total uncertainty as discussed in reference 1 (pages 1, 2, and 3 of 13. Attachment 2 and TS Table 2.2 1) and the Overpower N 16 trip setpoint (page 5 of 13 Attachment 2. and TS Table 2.2-1).
TU Electric Resoonse:
The Unit 2 Cycle 3 overtemperature N 16 setpoint was developed in accordance with TS 6.9.1.6b, Item 9. " Power Distribution Control Analysis and Overtemperature N 16 and Overpower N 16 Trip Setpoint Methodology."
The Westinghouse setpoint application to CPSES 1 is summarized in WCAP-12123 and was reviewed by the NRC prior to issuing the Unit 1
Attachment to TXX 96040 1 Page 3 of 11 operating license. This review is documented in NUREG 0797 SSER 22 (Page 7 7). This setpoint methodology was used by Westinghouse in the calculation of the Reactor Trip System (RTS) and the Engineered Safety Features Actuation System (ESFAS) setpoints for the CPSES-1 Technical Specifications. TU Electric applied this methodology in the calculation of the RTS and ESFAS setpoints which were approved for incorporation into the original CPSES-2 Technical Specifications and also into past revisions to the CPSES-1 Technical Specifications. TV Electric used the same methodology for the revised overtemperature N-16 uncertainty calculations that was used '
for the original CPSES-2 Technical Specifications which have been approved by the NRC.
NRC Ouestion 4:
Please explain the difference between how the power is calculated l using the N 16 power indication and that from the calorimetric power '
indication as discussed in Reference 1 (page 11 of 21, Attachment 2).
TU Electric Resoonse:
The calorimetric indication of core power is developed from a secondary plant heat balance. The secondary plant thermal power is determined from measurements of the feedwater flow and enthalpy and steam enthalpy. The reactor coolant pump heat addition and '
charging / letdown losses are then used to determine the core thermal power. The N 16 power indication is then normalized to indicate the calorimetric power. This process is identical to that used to normalize the Nuclear Instrumentation System excore power indication.
NRC Ouestion 5:
Please provide a list of the NRC approved codes, with the titles of the approved reports, used for the Unit 2, Cycle 3 reload analysis.
TU Electric Resoonse:
The NRC approved methods are listed in Technical Specification 6.9.1.6b. and is also detailed in the cycle-specific Core Operating Limits Report. The relevant information (excerpts from Technical Specification 6.9.1.6b) is reproduced below.
Attachment to TXX 96040 Page 4 of 11
- 6) . WCAP 10079 P A, "NOTRUMP, A N0DAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," August 1985, (W Proprietary).
7). WCAP 10054 P A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE," August 1985, (W Proprietary).
8). WCAP-11145 P A, " WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUDY WITH THE NOTRUMP CODE," October 1986 (W Proprietary).
f 9). RXE 90 006 P, " Power Distribution Control Analysis and Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology," February 1991. i 10). RXE 88-102 P, "TVE 1 Departure from Nucleate Boiling Correlation," January 1989.
l 11). RXE 88 102 P, Sup. 1 "TUE-1 DNB Correlation - Supplement 1,"
December 1990.
12). RXE-89 002, "VIPRE-01 Core Thermal-Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing l Applicatic.'s," June 1989. '
13). RXE 91 001, ' Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications," February 1991, 14). RXE 91 002, " Reactivity Anomaly Events Methodology," May 1991.
15). RXE 90-007, "'.arge Break Loss of Coolant Accident Analysis Methodology," December 1990.
16). TXX-88306, " Steam Generator Tube Rupture Analysis," March 15, .
1988. 1 17). RXE 91 005, " Methodology for Reactor Core Response to Steamline Break Events," May, 1991.
- 19) RXE 94 001 A, " Safety Analysis of Postulated Inadvertent Boron Dilution Event in Modes 3, 4,and 5." February 1994.
NRC Ouestion 6:
You mention on page 11 of 13 of Attachment 2 that the most relevant design basis analysis in Chapter 15 of the CPSES Final Safety Analysis Report (FSAR) which is affected by the change in the safety analysis value for the CPSES Unit 2 Overtemperature N 16 reactor trip setpoint is the Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power (FSAR Section 15.4.2) and that all acceptance criteria were satisfied. Please provide information on all the
l .
. . i
[ Attachment to TXX-96040 l
Page 5 of 11 i
Chapter 15 accident analyses that were performed for CPSES Unit 2 Cycle 3 and indicate what approved codes were used for each accident l or traneMt and why the results were acceptable (i.e.. met the DNBR l requir< snt, met the pressure requirement, etc.).
1 l
TU Electric Resoonse:
A. When preparing the License Amendment Request, all events were
! reviewed. Those events for which the Overtemperature N 16 trip function provides a primary protective or mitigative function were identified. With the exception of the Uncontrolled Rod Withdrawal from Power (RWAP) event, none of the events are " limiting" with respect to the DNBR event acceptance criterion. Therefore, the ;
discussion in the License Amendment Request is based only the RWAP 1 event, which is the most limiting of those events for which DNBR -
protection is provided by the overtemperature reactor trip function.
The analyses of this event demonstrated that the 'evant event l acceptance criteria (DNB and overpressure) are sa- .fied. Note that the RWAP event was analyzed using deterministic Dhd methods.
B. A table of the relevant event acceptance criterion for each non LOCA l FSAR Chapter 15 event considered during the core reload evaluation process is erovided in the NRC's Safety Evaluation Report " Comanche Peak Steam Electric Station Units 1 and 2. Topical Report RXE 91-001, ' Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications' (TAC No. M79866)," T. A. Bergman (NRC) to W. J. Cahill (TUEC), July 16, 1993. This table is attached fer your convenience. The LOCA evaluations are performed in
! accordance with Technical Specification 6.9.1.6b Items 6, 7, 8. and l 15. The most relevant event acceptance criterion, the peak clad i temperature, is reported to the NRC ir, accordance with 10CFR50.46.
During the Reload Safety Evaluation performed in accordance with l 10CFR50.59 prior to the start of a specific operating cycle, it is confirmed that the methods listed in Technical Specification 6.9.1.6b are used to ensure that each of the relevant event acceptance criteria are satisfied.
4 i
l
Attachment to TXX-96040 - -
Page 6 of 11 f*t' IV-8 Page 1 of 3 FSNL SECil0N 15.1.3,2 15.1.3 15.t.4 35.3.5 35.2,3 35,2,3 15.2.6 -
Event krorve FWM ELI MSSV MSLB TT if LOAC Aceptante Cr1terton DMist DNUR DNOR DMWt DNBR RCS Press Dret' Power' Higylero High lero Zero High High High Pr r Pressure
- tow low Ilmelnal teominat low High High Prir tevet High High Hominal teaminal High High High RCS T-avg' High High leuminal loominal High High High RCS toop Flow' tow low low tow tow ,
tow to.
SG tevet low High High High High High High Fuel Iemo low tow - - High High High Prrr Prs CntI~ b Off off Off Om Off off SC vtr tvl Cntt On On - - - - -
Rod Cnt t OryOff OrV0ff Off Off Off off off furblne Cntt toad tand - - - -
toad Rs Trip Signat OTM16 St 8eone St HPP.to SWt Hi Pres to Srkt ECCS Act Signat score enom to Step to StaP Denne teore kwe Iurbine Trip Hi SOWL Eme SI SI Rs irlp -
Rs Trip St Iso. Signet None None to SV* to Step Ianne tenne kwe tfW iso. Signal HI SGWL Dbne MS 1941 MS Isot SQ4C SWlC -
tbderator Temp Coef Range Range Most leegative Most leegative Range Range least fergatIve Doppter Fuel Teme Coeff tesSt Bergat Eve least elegatIve tesst 8ergatIve teast toegatIve Most fergetIve fbst IergatIve fbst sergatIve Ef f Detmyed Iseutron Frac Minlease Range Minlansa Minlem Range Range Masianan Pressur tier Sprsys and PORVS onty.
8 Demelnal Values are used if KU Des methodology is used.
3 Decay Heat Removal with Aunit lary Feedwater System
- 153
Attachment to TXX-96040 . .
Page 7 of 11 .
Table IV-l Page 2 of 3 FSAR SECTION 15.2.F 15.2.8 15.3.1 15.3.2 Event Acronya 15.3.3 15.4.8 15.4.2 10FW _ ffte r. 0F CLOF tR/55 RUS Rw Acceptance Criter ?on C:R' tHt' Dulut tueR bM RCS Pres M Deer Doet High High High High High fero Prrr Pressure
- Range High High tow tav High/tav tow tow Prrr level High High High High High High High RCS 1-avg
- Migh High High High High High High RCS toop Flow' Low tow Low tow tow tow tow SG tevet High High/ low -
Fuel Temp High High High High High -
Prrr Frs Cntl* Off Off On High b Of FAh Off SG vtr tvt Cntt -
th On On (h Rod Cntt th on Off Off Off Off Off off furbtre CntI toad ioed -
off Ra irlp Sl1Put toad inad to SWL to SGWL to Flow =-
W/tf to Flow HI Ftua to Hi Flua Stpt Of/1F M16 ECCS Act Slanet tenne to Stue peone none Turbine Trip Rn Trio tenne tenne aw Rs Trio Rs frio Rm Trip Rs irlo scorie Ru Trip Ste Iso. Signet hone
. _ _ to Star esone stone saane tenne Hon, ifW !so. Si mit -
to Star soone saare esone stone econ, Nderansviemy_ Coef least streatlwe flest serestive least toegattwe teest sereatlwe teast sere.?tive least streat tw Range Doppler fuel feue Coef Pbst Meestlwe fbst sneestive fbst seroative Most eneestive feast farestive teast steeative Ef f Delayed earutron Frac Ranae i Haa laasm ftse leman Mas lanas flos lmass feastmass stea laman I Has laman Pressurlier Sprays and PORYS only.
3 esaminal Values are used if SCU Die methodology is used.
a Decay Heat Remawal with Aunit lary Feedwater System I I
i 154 '
Attachment to TXX-96040 Page 8 o f 11 ^ '
Table IV-t Page 3 of 3 FSMt SECit(DI 15.4.3 15.4.3a 15.4.8 15.5.1 15.6.1 Event Acronym Dropfted 9ttdP OtE ECCS RCS DP Acceptonce CrIterlon INEut DNiut Peitet !!aENt DIENt enthalpr/ PCT Power' High High High/Zero High High Prrr Pressure
- tow tow High tow Low Prrr tevet low High High High High RCS T-avg' High High High High High RCS tcop Ftow' tow iow tow tow tow SG teveI - - - - -
Fuel temp High High High High High Prir Prs Cntt* th On -
Off Off SG War tvt Cntt b b On N %
Rod Cntt On On -
Off Off/Un furbIne CntI toad toad -
toad ioad Re Trip Slgnet Hi Flus HI Flue: OIN16 NI Flus to Pr:P to PrrP OfM16 (CCS Act Slanat None Ame unne None I*we Iq 5Bne Trip Rs Trip Rn Trip Rs Irlo su Trip Rn Trip 57 f w . Signal peone Ilone More Itore Nnne ffW l*o. Signat leone None leone stone sanne 79aderator Teme Coef Range Range teast Ierestive teast NeestSve Range Doppler Fuel Tese Coef least Negative Range teast lernative least Negative Range Ef f Delayed terutron Frac Man lanan Maa laman Minlemme haslaase Man 6 mum Pressurleer Sprsys and PatVS onty.
5 Hominst Values are used if 500 DNB methodology Is *oed.
155
Attachment to TXX 96040 Page 9 of 11 NRC Ouestion 7:
Provide input parameters for (power, pressure, temperature, flow, and power density) used to calculate DNBR and other Chapter 15 analyses for Unit 2 Cycle 3 and the resultant DNBR value.
TU Electric Resnonse:
The initial conditions used in the transient analyses for Unit 2 Cycle 3 are separated into 2 columns. When deterministic DNB methods are to be used, the left-most column is applicable. When statistical DNB methods are used, the right-most column is applicable for Unit 2 Cycle 3.
Parameter Deterministic Statistical Maximum Rated Thermal 102% 100%
Power (% of tall MW)
Pressurizer Pressure 2220 (DNB limited) 2250 (System analysis) !
(psia) 2280 (overpressure limited) 2280 (DNB analysis) i l
T average at 100% RTP 595.7 589.2
( F)
RCS Flow (gpm) s 400.800 408,000 Average Power Density 5.55 5.445 l (kw/ft)
DNBR Limit Value 1.16 1.429 Experience has shown that, with the use of TV Electric methods and i CPSES tore designs, the event for which the calculated minimum DNBR j most closely approaches the DNBR limit value is typically the dropped rod event; although any event can be made to appear
" limiting", depending on how much conservatism is included in the evaluation. For the preliminary, Unit 2 Cycle 3 analysis of the dropped rod event, in which the Statistical Combination of Uncertainties (SCU) DNB methodology is used, a minimum DNBR of apprLximately 1.50 was calculated. This value is greater than the DNBR limit value and is, therefore, acceptable.
Of more relevance to this License Amendment Request is the analysis of the RWAP event. In order to provide allowances for future uses, these evaluations were performed with peaking factors greater than expected to be required for Unit 2 Cycle 3 operation. For the preliminary, Unit 2 Cycle 3 analysis of the RWAP event. in which the deterministic DNB methodology is used, a minimum DNBR of approximately 1.34 was calculated for the case initiated from 102%
RTP. For the case initiated from 12% RTP, the minimum DNBR was calculated, on a preliminary basis, to be approximately 1.17. Both
I I
Attachment to TXX-96040 Page 10 of 11 l
l of these values are greater than the deterministic DNBR limit of i 1.16 and are, therefore, acceptable.
The " resultant" DNBR for each transient is confirmed to be greater than the appropriate DNBR limit value (1.16 for deterministic methods,1.429 for statistical methods for Unit 2 Cycle 3). Through the Reload Safety Evaluation evaluation process, it is confirmed, prior to the start of a specific operating cycle, that all analyses are performed in accordance with the methodology approved by the NRC i and listed in Technical Specification 6.9.1.6b. !
NRC ouestion 8: i Provide the uncertainty values and bases used in the statistical combination of uncertainties as required by the safety evaluation 1 report that approved RXE-91002, " Reactivity Anomaly Event Methodology," dated January 19.1993.
TU Electric Resoonse:
TV Electric's topical report RXE 91-002 contained a demonstration application of the Statistical Combination of Uncertainties (SCU) methodology. The values used in the demonstration application were applicable to Unit 1 Cycle 1. As stated in the NRC's Safety Evaluation Report. Technical Evaluation Report, and the Responses to the Request for Additional Information related to the TU Electric's report RXE 91002, TV Electric sill use unit and cycle specific values when applying the SCU methodology. This information is provided, as required by the forgoing documents, in the applicable ,
TU Electric calculations.
When the Unit 2 Cycle 3 Reload Safety Evaluation is completed, in accordance with 10CFR50.59 and predicated on the approval of License Amendment Request 95 008, no unreviewed safety questions will exist and it is anticipated that no additional changes to the plant Technical Specifications will be required. All analyses will be performed in accordance with Technical Specification 6.9.1.6. If, as expected, no unreviewed safety questions are identified, no additional licensing submittals will be required.
For Unit 2 Cycle 3 the uncertainty values expected to be used in the SCU applications are reproduced below. The DNBR uncertainty factor is 0.8278. Temperature and flow biases, totaling - 0.027 DNB, are treated appropriately. The bases are described in the approved topical report, which includes the additional questions and responses and the NRC's safety evaluation report.
l '
Attachment to TXX 96040 1 Page 11 of 11 Parameter Sensitivity Coefficient (o DNB / of Variance a change in parameter) (o/ )
Pressure 1.577 0.00811 Temperature 9.793 0.00508 Power 2.605 0.01351 Flow -1.600 0.01160 FdelH 3.374 0.02432
)
4 1
1 1
1