ML20062A178
| ML20062A178 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 07/16/1982 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Bauer E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| References | |
| NUDOCS 8208030094 | |
| Download: ML20062A178 (12) | |
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Document Control 50-352/353 JUL 16 1982 NRC PDR L PDR NSIC PRC ggggqg;ggp LB#2 Reading EHylton Habel son OELD) g u/3 i&E Mr. Edward G. Bauer, Jr.
Vice President & General Counsel ACRS (16)
Philadelphia Electric Company Region I 2301 Market Street Philadelphia, Pennsylvania 19101
Dear Mr. Bauer:
Subject:
Limerick - Draft SER - Section 4.4 The. Thermal-Hydraulics Section of the Core Performance Branch has prepared the draft SER input for Section 4.4 of the FSAR. This input, provided in describes the position of the Branch, and the current status of the review relevant to this portion of the FSAR.
Any questions concerning the enclosed material should be directed to Dr.
Harvey Abelson, (301) 492-9774, the Licensing Project Manager.
Sincerely,
[
J A. Schwencer, Chief Licensing Branch No. 3 Division of Licensing
Enclosure:
As stated l
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esomim_ m m PeRC FORM 318 (10-80) NRCM O24a
Lime rick Mr. Edward G. Bauer, Jr.
Vice President & General Counsel Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101
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cc:
Troy B. Conner, Jr., Esquire Mr. Marvin I. Lewis Conner and Wetterhahn 6504 Bradford Terrace 1747 Pennsylvania Avenue, N. W.
Philadelphia, PA 19149 Washington, D. C. 20006 Frank R. Romano, ' Chairman Mr. Robert W. Adler Air & Water Pollution Patrol Assistant Counsel 61 Forest Avenue Commonwealth of Pennsylvania, DER Ambler, PA 19002 505 Executive House P. O. Box 2357 Charles W. Elliott, Esquire Harrisburg, Pennsylvania 17120 Thones & Hair 123 North Fifth Street Honorable Lawrence Coughlin Allentown, PA 18102 House of Representatives Congress of the United States Judith A. Dorsey, Esquire Washington, D. C. 20515 Limerick Ecology Action 1315 Walnut Street, Suite 1632 Roger B. Reynol ds, Jr., Esquire Philadelphia, P.A 19107 324 Swede Street Norristown, Pennsylvania 19401 Mr. Karl Abraham Public Affairs Officer Lawrence Sager, Esquire Region I Sager & Sager Associates U.S. Nuclear Regulatory Commission 45 High Street 631 Park Avenue Pottstown, Pennsylvania 19464 King of Prussia, PA 19806 Joseph A. Smyth Mr. Jacque'Durr Assistant County Solicitor Resident Inspector County of Montgomery -
U.S. Nuclear Regulatory Commission Courthouse P. O. Box 47 Norristown, Pennsyl vania 19404 Sanatoga, PA 19464 Eugene J. Bradley James M. Neill, Esquire Philadelphia Electric Company Associate Counsel for Del-Aware Associate General Counsel Box 511 2301 Market Street Dublin, PA 18917 Philadelphia, Pennsylvania 19101 Joseph H. White III Mr. Vincent Boyer 11 South Merion Avenue Senior Vice President Byrn Mawr, PA 16801 Nuclea r ~0perations Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101
' Thomas Gerusky, Director Sugarman & Denworth Bureau of Radiation Protection Suite 510 Dept. of Environmental Resources.
North American Building 5th Floor, Fulton Bank Bldg.
121 South Broad Street Third & Locust Streets Philadelphia, PA 19.107' Harrisburg, PA 17120 Donald S. Bronstein, Esq.
Director, Pennsylvania Emergency The National Lawyers Guild Management Agency Third' Floor..
Basement, Transportation &
1425 Walnut Street Safety Building Philadelphia, PA 19102 Harrisburg, PA 17120 Lawrence Brenner, Esq., Chairman
- John Shniper Administrative Judge Meeting House Law Bldg. & Gallery U.S. Nuclear Regulatory Commission Mennonite Church Road Washington, D.C.
20555 Schuykill Road (Rt. 724)
Spring City, PA 19475 Dr. Richard F. Cole
- Administrative Judge Robert L. Anthony U.S. Nuclear Regulatory Commissio'n Friends of the Earth of the
' Washington, D.C.
20555 Delaware Valley 103 Vernon. Lane, Box 186 Dr.- Peter A. Morris
- Moylan, PA 19065 Administrative Judge U.S. Nuclear Regulatory Commission W. Wilson Goode Washington, D.C.
20555 Managing Director City of Philadelphia Philadelphia, PA 19107 William A. Lochstet 119 E. Aaron Drive State College, PA 16801
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Walter W. Cohen -
Consuner Advocate Office of Attorney General 1425 Strawberry Square Harrisburg, PA 17120 Steven P. Hershey, Esquire C'onsumers' Educe. tion & Drotective As sociation Sylvania House Juniper & Locust Streets Philadelphia, PA 19107 Alan J. Nogee The Keystone Alliance 3700 Chestnut Street Philadelphia, PA 19104 DRAFT SER FOR LIMERICK Thermal-Hydraulics Section, Core Performance Branch 4.4 Thermal and Hydraulic Design The scope of the review included the design criteria, implementation of design criteria as presented by the final core design, and the analyses of core thermal-hydraulic performance.
The acceptance criteria used as the basis for the staff's evaluation are set forth in Chapter 4.4 of the Standard Review Plan (NUREG-0800).
4.4.1 Thennal-Hydraulic Safety Design Bases The thermal-hydraulic safety design bases for Limerick can be summarized as follows:
1.
Fuel failure due to cladding overheating should not occur during normal operation or anticipated operational occurrence.
Specifically, the minimum critical power ratio (11CPR) operating limit is specified such that at least 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during normal operation or anticipated operational occurrence.
2.
Core and fuel thermal-hydraulic design limits for steady-state opera-tion, i.e., MCPR and LHGR limits, should be defined to provide margin between' steady-state operating conditions and any operating condition that could result in fuel failure.
Specifically, they should assure that no fuel failure results during normal operation or anticipated operational occurrence even with uncertainties taken into account.
3.
No undamped oscillations or other hydraulic instabilities should occur either for normal operation or for anticipated operational occurrence.
The applicant's thermal-hydraulic design bases are consistent with the Standard Review Plan and are, therefore, acceptable.
4.4.2 Critical Power Ratio (CPR) Correlation The GEXL correlation is used to predict the (critical) quality at which boiling transition occurs as a function of the distance from the equilibrium boiling boundary. The GEXL correlation and design application ~are described in the General Electric Company Topical Report, NED0-10958A. The figure-of merit chosen for reactor design and operation is the critical power ratio, the ratio of the critical bundle power to the operating bundle power.
This correlation and supporting topical report have been previously reviewed and approved by the staff (Butler, October 21,1974).
Critical power tests have been run on prototypical 8x8 bundles with two water rods. Test data for cosine axial heat flux shapes indicate that the water rods do not affect the GEXL capability of predicting the bundle critical power performance for bundle radial peaking patterns typical of 8x8 retrofit fuel. The staff has previously found that the GEXL data base, which includes top and bottom peak axial heat flux distributions canbined with the two-water rod data, demonstrates the adequacy of the GEXL corre-lation.to predict critical power in both 8x8 and 8x8 retrofit bundles.
The staff has previously concluded that the GEXL correlation is acceptable for both 8x8 and 8x8 retrofit fuel applications (Tedesco, April 16,1981).
4.4.3 Thermal-Hydraulic Analysis Methods The MCPR limit originally proposed was based upon calculations using the
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REDY model described in NED0-10802.
The. results from the tests performed at Peach Bottom-2 revealed that in certain cases the results predicted by
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REDY are non-conservative.
Therefore, the staff required the applicant to use ODYN methods to analyze the following transients:
(1) feedwater 2
controller failure at maximum demand with and without bypass, (2) generator load rejection, and (3) turbine trip with and without bypass.
The applicant is required to submit the operating limit MCPR as calculated by including the ODYN methods for review and approval prior to issuance of an operating license.
4.4.4 Themal-Hydraulic Stability Recent BWR fuel design changes which affect stability include decreasing the rod size and increasing'the gap conductance because of pre-pressuriza-tien. As a consequence, the maximum decay ratio for most BWRs increases and becomes larger than 0.5, which is the original GE design criterion for BWR stability.
Therefore, GE now proposes a decay ratio of 1.0 for their
- criterion. The staff has not agreed that the proposed criterion of a 1.0 decay ratio calculated by using the FABLE code is acceptable.
To further evaluate this criterien and other stability criteria, the staff is perfoming a generic study of the hydrodynamic stability characteristics of light water reactors under nomal~ operation, anticipated transients, and accident conditions.
The results of this study will be applied to the staff's review and acceptance of stability analyses and analytical methods now in use by the reactor vendors.
The stability analysis resulted in a maxium decay ratio of 0.7 for the end-of-life cycle, which is the limiting cycle with respect to stability.
The staff has approved for operation the Susquehanna core design (NUREG-0766),
which has calculated a maximum decay ratio value of 0.7 for the initial cycle. Since Susquehanna and Limerick have similar core designs, the staff concludes that the Limerick core design stability is acceptable for Cycle 1.
However, in order to provide additional margin for stability, natural c'ircu-lation operation will be prohibited until the staff's review of these con-ditions is completed. Any action resulting from the staff's study will be applied to Limerick. Theapplicant$operatinglicensewillbeconditionedto 3
require that a new stability analysis be submitted and approved prior to second-cycle operation. Also, since no analysis has been presented for minimum critical power ratio limits or stability characteristics for single loop operation, the staff will require by Technical Specifications that single loop operation not be pemitted until supporting analyses are provided and approved.
4.4.5 Crud Deposition Crud deposition causes gradual flow reductions in some light water reactor cores.
However, measurement of core flow by jet pump pressure drop and core plate pressure drop will provide adequate indication of such flow reductions, if they should occur.
Technical Specifications will require that the core flow be checked at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to detect flow reduction.
For pressure drop considerations in design analysis, it is assumed that a conservative amount of crud is deposited on the fuel' rods and the fuel rod spacers.
This is reflected in decreased flow area, increased friction factors, and increased spacer loss coefficients. The effect of this crud is to increase the core pressure drop by approximately 1.7 psi.
The staff l
concludes that the assumptions regarding crud deposition used in the design l
analysis in conjunction with the required flow monitoring are acceptable.
4.4.6 Loose Parts Monitoring System The applicant has provided a description of the loose parts monitoring system (LPMS) which will be used by Limerick.
This infomation is still under review.
The results of our review will be included in the SER.
- The, system should be installed to meet the operability requirements of Reau-latory Guide 1.133, Revision 1 (May 1981).
The LPMS must be operational i
and capable of recording vibration signals for signature analysis at the time of initial startup testing. The staff also requires that the applicant 4
commit to evaluate the system to address conformance with Regulatory Guide 1.133, Revision 1.
The conformance evaluation should emphasize the following areas:
1.
A description and evaluation of diagnostic procedures used to confirm the presence of a loose part.
2.
A description of how the operators will be trained in the purpose and implementation of the system.
3.
A report on system calibration, including signature analysis, evalu-ation of background noise, and alarm settings.
4.
A description and evaluation of alert level establishment procedure with consideration of internal and external background noises.
4.4.7 THI-2 Action Plan Item II.F.2 A clarification of requirements for inadequate core cooling instrumentation, which is required to be installed and operational prior to fuel loading, was provided in a letter to all operating nuclear power plants (Denton, October 30,1979) and in Section II.F.2 of NUREG-0737, " Clarification of TMI Action Plan Requirements." The applicant is required to provide a response to Section II.F.2 of the TMI Action Plan for review and approval prior to issuance of an operating license.
4.4.8 Thermal-Hydraulic Comparison A summary of the thermal-hydraulic parameters for Limerick is given in Table 4.4-1.
A canparison with the parameters for the Hatch-2 core design is given for reference.
This core design was previously approved in the Safety Evaluation Report issued in June 1978 for Hatch-2, which is now an operating reactor. The primary difference in core design between Limerick and Hatch-2 is size.
Both are BWR/4 cores and use the improved 8x8R fuel assemblies.
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Table 4.4-1 Themal-Hydraulic Design Comparison Design Parameter Hatch-2 Limerick Design themal output (MWt) 2,436 3,293 Final Feedwater temperature (FFWT)( F) 420 420 6
Steam flow rate at FFWT (10 lbs/hr) 10.47 14.16 6
Core coolant flow rate (10 lbs/hr)
~ 77.0 100 6
Feedwater flow rate '(10 lb/hr) 10.44 14.13 Steam pressure, nominal in steam dome (psia) 1,020 1,020 Steam pressure, nominal core design (psia) 1,035 1,035 Average power density (kW/ liter) 49.15 48.7 Maximum linear themal output (kW/ft) 13.4 13.4 Average linear themal output (kW/ft) 5.38 5.3 2
Core total heat transfer area (ft )
54,879 74,871 Fuel type P8x8R P8x8R Water rods per bundle 2
2 Core inlet enthalpy at FFWT (Btu /lb) 526.9 526.1 Core maximum exit void within assemblies (%)
76.3 77.1 Core average void, active coolant (%)
42.2 41.8 2
Active coolant flow area per assembly (in )
15.82 15.82
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Core average inlet volocity (ft/sec) 6.6 6.44 Total core pressure drop (psia) 23.9 21.95 Core support plate pressure drop (psia) 19.46 17.52 Average orifice pressure drop (psia):
Central region 8.0 5.10 Peripheral region 16.52 14.32 Number of fuel rods per bundle 62 62 Rods outside diameter (in.)
Fuel rod 0.483 0.483 Water rod 0.591 0.591 Active fuel length (in) 150 150 Rod pitch (in) 0.640 0.640 6
The canparability of Limerick design with Hatch-2 supports the conclusion that the Limerick thermal-hydraulic design is acceptable.
4.4.9 Evaluation Findings The Limerick thermal-hydraulic design has been reviewed according to Section 4.4. of the Standard Review Plan (NUREG-0800). The scope of review included the design criteria, implementation of the design criteria as pre-sented by the final core design, and the steady-state analysis of the core thermal-hydraulic performance. The review concentrated on the differences between the proposed core design (and criteria) and those designs and criteria that have been previously reviewed and found acceptable by the staff.
It was found that all such differences were satisfactorily justified by the applicant. The applicant's thermal-hydraulic analyses were performed using analytical methods and correlations that have been previously reviewed by the staff and found acceptable.
However, the operating license should be restricted with the following conditions:
1.
Single loop operation is not permitted unless supporting analyses are provided and approved.
2.
Operation beyond Cycle 1 is not pennitted until a stability analysis is provided and approved for the additional cycles of operation.
3.
The natural circulation operation mode is not permitted.
4.
The core flow should be checked at least once per day.
The above restrictions should be incorporated into the proposed Technical Specifications, except for Item 2 which should be incorporated as a license condition.
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- The following documents should be submitted for review and approval prior to issuance of an operating license:
1.
The applicant should commit to evaluate the LPMS to address confor-mance with Regulatory Guides 1.133, Revision 1.
2.
A response to Section II.F.2 of the Till Action Plan for review and approval.
3.
The operating limit MCPR as calculated by including ODYN methods must be provided for review and approval.
With the exception noted above, the staff concludes that the thennal-hydraulic design of the core conforms to the requirements of General Design Criterion 10 of 10 CFR Part 50, Regulatory Guides 1.68 and 1.133, and Section 4.4 of the Standard Review Plan and is, therefore, acceptable.
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References NED0-10958A, " General Electric Thermal Analysis Basis (GETAB) Data, Correlation, and Design Application," January 1977.
W. R. Butler (NRC) letter to Stewart (GE), " Acceptance for Referencing Licensing Topical Report NEDE-10958," October 1,1974.
R. L. Tedesco (NRC) letter to G. Sherwood (GE), " Conditions Removal from Acceptance for Referencing of Licensing Topical Report NEDE-24011P.
NEDE-10802, " Analytical Methods Of Plant Transient Evaluations for General
' Electric Boiling Water Reactor," February 1973.
H. Denton (NRC) letter to All Operating Nuclear Power Plants, " Discussion of Lessons Learned Short-Term ReqLirements," October 30, 1979.
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