ML20215H689
| ML20215H689 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/30/1986 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML19292G101 | List: |
| References | |
| L-CE-R-042, L-CE-R-42, NUDOCS 8610240089 | |
| Download: ML20215H689 (11) | |
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ENCLOSURE 1-NP TO L-CE-R-042 Y
' RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION ON WATERFORD-3 CYCLE 2 RELOAD i
l SEPTEMBER 1986 I
i-Combustion Engineering, Inc.
Nuclear Power Systems Windsor, CT 8610240089 861013 PDR ADOCK 05000382 P PDR
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LEGAL NOTICE ,
This report was prepared as an account of work .
sponsored by Combustion Engineering, Inc. Neither Combustion Engineering nor any person acting on its behalf:
- a. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process dis-closed in this report may not infringe privately owned rights;.or
- b. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method or process disclosed in this report.
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t i-r ABSTRACT t .
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This report contains responses to NRC questions on the l
Waterford-3 Cycle 2 reload submittal.
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Responses to Request for Additional Information Waterford 3 Cycle 2 Reload QUESTION 1:
Please explain in more detail the bases for not performing clad collapse analyses for Cycle 2 and reference any applicable NRC approval.
RESPONSE
Justification for not performing clad collapse analyses is based on the work-reported by EPRI (Reference 1-1). An evaluation specifically for C E modern fuel and based on the EPRI work was submitted to the NRC on ~another Docket
' (Reference 1-2). It concluded that collapse analyses are not necessary for C-E fuel because of the absence of gaps between the pellets. The NRC con-curred in this in Reference 1-3, with certain provisions regarding the con-tinued validity of data and the consistency of the fuel manufacturing process.
An evaluation for the Waterford 3 fuel has shown that these provisions are met and therefore no clad collapse analyses were performed for Cycle 2.
Reference 1-1 EPRI NP-3966-CCM, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding, Volume 5: Evaluation of Interpellet Gap Forma-tion and Clad Collapse in Modern PWR Fuel Rods", April,1985.
Reference 1-2 A. E. Lundvall, Jr. (BG&E), to J. R. Miller (NRC), "Calvert Cliffs Nuclear Power Plant Nos. I and 2, Docket Mos. 50-317 and 50-318, Request for Amendment", December 31, 1984.
Reference 1-3 " Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.104 to Facility Operating License No.
DPR-53, Baltimore Gas a:d Electric Company Calvert Cliffs Nuclear Power Plant Unit No. 1 Docket No. 50-317", May, 1985.
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Responses to Request for Additional Information Waterford 3 Cycle 2 Reload QUESTION 2:
The results of the burnup dependent fuel thermal performance calculations are not inentioned. Please ver:fy that the internal pressure in the most limiting hot rod satisfies the internal ~ gas pressure requirement of Standard Review Plan 4.2, Section 11.A.1(f).
RESPONSE
The hot rod analyzed for maximum rod internal pressure bounded all the fuel in i
the Waterford 3 Cycle 2 core. The rod internal pressure is predicted to remain below RCS pressure throughout Cycle 2.
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Responses to Request for Additional Infonnation Waterford 3 Cycle 2 Reload QUESTION 3:
Does the most restrictive value of required shutdown margin still occur at EOC hot zero power conditions? If so, verify that sufficient shutdown margin exists to control the steam line break event initiated at this condition during Cycle 2. Also, verify sufficient CEA worth is available to meet the i required shutdown margin below 200'F.
RESPONSE
In response to the specific question, the most restrictive (i.e., largest) value of required shutdown margin occurs at HZP for Cycle 2. The maximum value of the minimum required shutdown margin is dictated by the Steam Line -
Break (SLB) accident at End-of-Cycle (E0C). This is true for both Cycle 1 and Cycle 2. In Cycle 1, the most limiting SLB was initiated from Hot Zero Power (HZP) and required 5.15% shutdown margin, which is the minimum required value in the Technical Specifications for Modes 1 and 2. This value was also used in the Cycle 2 analyses of the HZP SLB, although the minimum calculated scram worth actually exceeds 5.15% for Cycle 2 at HZP E0C conditions after allowing for all biases and uncertainties and for the stuck rod.
However, it is the EOC HFP initial condit. ion which yields results which are closest to the acceptance criteria for the SLB accident. Results for the Cycle 2 E0C HFP SLB with concurrent loss of AC, initiated with the minimum scram worth calculated at the PDIL limit, have been shown to bound the results for the HZP case. The analyses for this HFP case are presented in Section 7.1.5b of the Cycle 2 Reload Analysis Report.
The second portion of the question refers to requirements on CEA worth below i 200*F. Requirements for shutdown margin in subcritical modes, including those below 200*F are satisfied by a combination of boration and of available CEA l worth. It is this combination which assures the required shutdown margin is t
available.
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Responses to Request for Additional Information Waterford 3 Cycle 2 Reload QUESTION 4:
The staff has previously approved the CETOP-D Computer Code with appropriate hot assembly inlet flow starvation factors to assure conservatism relative to the TORC Code. Since these flow starvation factors are plant specific, describe the analysis performed for Waterford Cycle 2 and verify the conserva-tism of CETOP-D relative to TORC.
RESPONSE
The C-E method for core thermal margin analysis utilizes two models. The first model, called the TORC model, is a detailed representation of reactor core for accurate prediction of core margin to DNB. The second model is the simplified CETOP-D for design analyses.
In the thermal margin analysis for Waterford Cycle 2, both models are used.
Values of DNBR are compared at each of a set of oper.ating conditions encom-passing the operating space. The "most limiting" operating condition is that for which CETOP-D contains the least conservatism relative to TORC. The hot assembly inlet flow starvation factor in the CETOP-D model is then adjusted in order to match the DNBR from CETOP-D to that from TORC at the most limiting condition. The CETOP-D model containing this flow starvation factor is therefore conservative relative to the TORC model over.the operating space and I applies specifically to the Waterford 3 Cycle 2 core.
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Responses to Request for Additional Information Waterford 3 Cycle 2 Reload QUESTION 5:
Since some of the variables whose uncertainties are statistically combined to generate a new DNBR limit are plant dependent, describe the specific SCU analysis performed for Waterford including the most adverse state parameters assumed.
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RESPONSE
~ The system parameter SCU analysis makes use of response surface methods and stochastic simulation (Monte Carlo) techniques to generate the 95/95 system parameter SCU DNBR limit. The response surface is essentially a curve fit to DNBR data generated from detailed TORC analyses. The Monte Carlo methods are numerical techniques to sample from the probability density functions (P.D.F.'s) of several variables (e.g., system parameters) in order to generate the p.d.f. of a parameter which -is a function of the variables that were ,
sampled (e.g., DNBR is a function of system parameters).
The system parameter uncertainties included in the Waterford 3 Cycle 2 SCU analysis are:
- 1. Inlet flow distribution uncertainties
- 2. Enthalpy rise factor
- 3. Systematic pitch uncertainties 4 Systematic clad 0.D. uncertainties-
- 5. Heat flux factor
- 6. CE-1 CHF correlation uncertainties
- 7. TORC Code uncertainty
- 8. Fuel rod bow penalty on DNBR*
- 9. HID-1 gr_id penalty *
- applied detenninistically The most adverse state parameters have been determined as:
, The SCU analysis performed for Waterford 3 Cycle 2 demonstrates that there J. will be at least 95% probability with at least 95% confidence that the limit-( ing fuel pin will avoid departure from nucleate boiling (DNB) so long as the
! minimum DNB ratio found with the best estimate design CETOP-D model ~ remains at or above 1.26.
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Responses to Request for Additional Information Waterford 3 Cycle 2 Reload QUESTION 6:
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Are sta'tistical combination of uncertainties relevant to the LHR and DNBR LSSS (Part 2) and the LHR and DNBR LC0 (Part 3) to be implemented for Cycle 2? If
. so, discuss the plant specific aspects of these analyses and provide the derived uncertainties.
RESPONSE
Statistical Combination of Uncertainties (SCU) will be applied to the deter-.
mination of Limiting Conditions for Operation. (LCO) and Limiting Safety System Setpoints (LSSS) on Linear Heat Rate (LHR) and Departure from Nucleate Boiling Ratio (DNBR) for Waterford Unit 3 beginning with Cycle 2. The methodology used for Waterford 3 is similar to that employed for. SONGS Units 2 and 3.
Minor changes were made to the SCU methodology as part of'the CPC improvement Program (CIP). These changes, briefly discussed with NRC in Nov. ember 1985, involve no substantial difference in basic methodology or results. The changes made to the SCU methodology as part.of the CIP involve the modifica-tion of the COLSS and CPCS simulators in order to simplify the SCU analysis process. Additional' details on the application of the SCU methodology to Waterford 3 may be found in Reference 6-1.
Input and calculated data appropriate for Waterford 3 will be used to deter-mine final COLSS and CPCS uncertainty factors. Final values will be available no sooner than one month before startup of Cycle 2.
Reference 6-1 CEN-343(C)-P, " Statistical Combination of Uncertainties for Waterford-3," October,1986.
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% s Responses to Request for Additional Information Waterford 3 Cycle 2 Reload QUESTION 7:
In the issuance of Amendment 5 to the Waterford 3 Facility Operating License (May 30, 1986) certain credits were allowed when using CPC to monitor LCOs and COLSS is out-of-service. Since the CPC changes associated with the CPC Improvement Program described CEN-308-P and CEN-310-P, which involved some penalty reduction as well as possible use of SCU, will now be implemented at Waterford 3, the previous credits for COLSS out-of-service operation should be J reevaluated. For example, the addition of the CPC variable overpower trip allowed certain power penalties used to compensate for potential nonconserva-tisms during rapid transients to be removed. Since these penalties no longer exist, they cannot be reduced to obtain additional credit for CPC monitoring.
Also, the addition of ASI dependent penalty factors, as described in CEN-308-P, removed the need to penalize operation at all ASI values for phenomena that occur over only'part of the allowed ASI range. Therefore, additional credit for LHR and DNBR due to the difference between the uncer-tainty(over range themay
.3 ASI) LSSS range be no longer ( .6appropriate.
ASI) and the Please uncertainty coment onover the reduced LCO this and justify the continued use of the COLSS out-of-service credits previously approved for Waterford 3.
RESPONSE
Many of the credits applied in Cycle 1 to the COLSS out-of-service Technical Specification limits on DNBR.and LHR are not valid for Cycle 2 because of changes to CPC software and methodology associated with the CPC Improvement Program (CIP). Only the following credits are valid for Cycle 2:
a) Credit for conservatism of neutron flux power relative to thermal power.
l b) Credit for dynamic pressure transient offset applied to thermal power.
These credits, worth'approximately 3.0% power margin for Cycle 2, are applica-ble to the COLSS out-of-service DNBR and LHR limits.
New COLSS out of service curves which incorporate only the credits which are valid for Cyc-le 2 operation have been prepared and are included with proposed changes to Technical Specification 3/4.2.4, DNBR margin, and 3/4.2.1, linear heat rate.
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Response to Request for Additional Information Waterford2 f Cycle 2 Reload QUESTIONS FROM LARRY K0PP - REACTOR SYSTEMS BRANCH:
QUESTION 1: ,
In the evaluation of the adequacy of the Waterford fuel assembly shoulder gaps J for Cycle 2 operation, the limiting fuel rod growth rate observed for Batch C fuel in ANO-2 was assumed to be bounding for the Waterford fuel rods. Please justify this assumption.
RESPONSE
Combustion Engineering has inspected the shoulder gaps of seven batches of 16X16 fuel (ANO-2 Batches A,B,C D and E, and SONGS-2 Batches B and C). The inspections have included the measuring of thousands of individual shoulder gaps from a' total of over 100 fuel bundle assemblies. The fuel-with the most limiting growth behavior was the ANO-2 Batch C fuel. Although the fuel rod designs of all the 16X16 fuel are essentially identical, the limiting ANO-2 Batch C fuel rods exhibited essentially a linear growth rate while all the other fuel types (including 14X14 fuel) have exhibited a decrease in fuel rod growth rate at increasing exposures. Although the decreasing growth behavior is expected for the Waterford-3 fuel, the more conservative linear growth rate of the limiting ANO-2 Batch C fuel rods was used in the shoulder gap justification for Waterford-3 Cycle 2.
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