ML19350A187
| ML19350A187 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/31/1980 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19350A186 | List: |
| References | |
| 80NED278, NEDO-24243-S1, NUDOCS 8102260555 | |
| Download: ML19350A187 (14) | |
Text
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- CLASS 1 SUPPL ET
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GENERAL ELECTRIC BOILING WATER REACTOR
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LOAD LINE LIMIT ANALYSIS FOR
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JAMES A. FITZPATRICK
, NUCLEAR POWER PLANT CYCLE 4
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GENERAL ELECTRIC BOILING WATER REACTOR LOAD LINE LIMIT ANALYSIS FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT CYCLE 4 l
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NEDO-24243 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREULLY This report was prepared by General Electric solely for The Power Authority of the State of New York (The Authority) for The Authority's use with the U.S.
Nuclear Regulatory Commission (USNRC) for amending The Authority's operating license of the James A. FitzPatrick Nuclear Power Plant. The information con-tained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provideo to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the contract between The Authority and General Electric Company for nuclear fuel and related services for the nuclear system for The James A. FitzPatrick Nuclear Power Plant, dated June 12,1970, and nothing contained in this document shall be construed as changing said contract.
The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe' privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result fn>m such use of such infonnation.
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NEDO-24243 CONTENTS Page
- 1. INTRODUCTION 1-1
- 2. SUKHARY 2-1
- 3. DISCUSSION 3-1 3.1 Background 3-1 3.2 Analytical Basis 3-1 3.3 Analysis Results 3-1 3.3.1 Stability 3-2 3.3.2 Transients 3-3 3.3.3 Lost-of-Coolant Accident 3-3
- 4. REFERENCES 4_1 iii/iv
e NEDO-24243 FIGURES Figure Title Page 4
2-1 JAFNPP Power / Flow Map 2-2 3-1 Operating Map 3-4 3-2 Reactor Core Stability Decay Ratio 3-5 Y
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- 1. INTRODUCTION
' The flexibility of a typical Bh'R in proceeding from low-power / low-core flow to high-power /high-core flow is affected by many factors; two of these are (1) the power / flow ratios defined by the power flow curve and (2) Pre-conditioning Interim Operating Management Recommendations (PCIOMRs) .
The power / flow curve is the locus of power'from a fixed rod pattern as a function of flow from which the occurrence of certain nonstandard events (abnormal operating transients) will yield results within defined plant safety limits; PCIOMRs reduce the rate of change of power in the fuel to improve the fuel performance. The combination of these two effects can result in the inability to attain full reactor power directly.
Recent analyses (References 1, 2 and 8) justify the modification of the operating envelope defined by the power / flow curve while remaining within previously established operating limits and the PCIOMRs.
Reference 8 provided the analytical basis for JAFNPP, Cycle 3, operation under a modified power / flow line designed to permit the direct ascension to full power within the design bases previously applied. This report provides the information necessary to extend that basis to include Cycle 4.
t 1-1/1-2
- 2.
SUMMARY
)- A method has been derived whereby the power / flow curve can be modified to provide relief from the difficulties associated with the effects of the existing power / flow curve and the PC10MRs.
f Analyses show that reactor ascension to full power may proceed along a modified power / flow line bounded by the 108% rod block line*, up to the point labeled,85/61 (rod block intercept point), from which continued power increases may proceed along the rod block intercept line to the 100% power, 94% flow point (100% intercept point), as shown in Figure 2-1.
The discussion and analyses in the following sections of this report show that all safety bases normally applied to JAFNPP are satisfied for operation within this envelope throughout Cycle 4.
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- If the rod block is reduced for power distribution, the allowable region is bounded by the reduced rod block line.
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l 100% INTERCEPT l POINT (100/94) l APRM ROD BLOCK UNE (108/1001 iPROPOSED) 10.58W + 50%l g,gf 3g
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100 - 1100/1006 ROD INTERCEPT LINE ROD 8 LOCK INTERCEPT POINT (85/61)
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OPERATE IN THIS REGION TYPICAL t00% POWER /
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Figure"2-1. JAFNPP Power / Flow Map 2-2
- 3. DISCUSSION j 3.1 BACKROUND
/ Operation of the JAFNPP utilizing the power / flow map is described in Section 3.7 of the FSAR (Reference 3). This section of the FSAR describes the basic oper-ating envelope (Figure 3.7-1) within which normal reactor operations are conduc-ted, and provides the basic philosophy behind the power / flow curve. FSAR Figure 3.7-1 is reproduced as Figure 3-1, 3.2 ANALYTICAL BASIS To provide relief from the operating restrictions inherently imposed during ascen-sion to power by the original power / flow curve and PCIOMRs, a modified power / flow curve was derived and presented in Reference 8. This curve is shown herein as
, Figure 2-1. The analyses and results presented in Reference 8 are, with three exceptions, bounding for JAFNPP and need not be performed again.
There are, however, three areas where bounding conclusions could not be made and cycle-specific analyses are required. The required supplementary analyses for JAFNPP, Cycle 4, are presented in Section 3.3 of this document.
3.3 ANALYSIS RESULTS i
The three items specified for plant / cycle-specific investigation are:
! (1) Stability -- A stability analysis was performed at the extended APRM rod block line power at natural circulation flow for JAFNPP, l
Cycle 4.
(2) Loss-of-Coolant Accident (LOCA) -, The applicability to the extended r
operating region of LOCA analyses previously approved by the NRC has been verified for JAFNPP, Cycle 4.
3-1
NEDO-24243 (3) Transients - If either cold water injection Loss of Feedwater Heating (LOFWH) or High Pressure Coolant Injection (HPCI) or Feedwater Con-troller Yoilure (FL'CF) is limiting, special analyses must be performed to determine if operating limit adjustments are necessary for operation in the extended operating region.
3.3.1 stability 1
3.3.1.1 Channel Hydrodynamic Conformance to the Ultimate Performance Criterion The channel performance calculatica for JAFNPP, Cycle 4 yields decay ratios are presented below:
Extrapolated Rod Block Line -
Channel Hydrodynamic Performance Natural Circulation Power i
Decay Ratio, Xy /X, 8x8R Channel 0.30 8x8 Channel 0.38 8x8 Channel 0.22 L -
l At this most responsive condition, the most responsive channels are clearly L within the bounds of the ultimate performance criteria of <1.0 decay ratio at all attainable operating conditions.
3.3.1.2 Reactor Conformance to Ultimate Performance Criterion The decay ratios determined from the limiting reactor core stability con-ditions are presented in Figure 3-2. The most responsive case for this analysis is the extrapolated rod block line -. natural circulation condition.
l Extrapolated Rod Block Line -
l Reactor Core Stability Natural Circulation Power ,
l Decay Ratio, X2 /X, 0.85 i.
These calculations show the reactor to be in compliance with the ultimate performance criteria, including the most responsive condition.
3-2
NEDO-24243 3.3.2 Transients Neither cold water injection nor feedwater controller failure are limiting for JAFNPP at any time during Cycle 4 (Reference 7); no apecial transient analysis is required.
3.3.3 Loss-of-Coolant Accident A discussion of low-flow effects on LOCA analyses for all operating plants (Reference 4) has been presented to and was approved by the NRC (Reference 5).
The LOCA analysis for JAFNPP (Reference 6) is applicable in the power flow domain discussed in this submittal.
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ULTIMATE PERFOAMANCE LIMIT sn -----------__---__--.
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EXTRAPOLATED ROD BLOCK LINE
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! 4. REFERENCES t
- 1. Millstone Unit 1. Load Line Limit Analysis, Revision 1, November 1977 (NEDO-21285-1).
- 2. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting j Amendment No. 52 to Provisional Operating License No. DPR-21, Northeast
- Nuclear Energy Company, Millstone Nuclear Power Station Unit 1, Docket
! No. 50-245, July 1978.
- 3. Final Safety Analysis Report, James A. FitzPatrick Nuclear Power Plant, Docket No. 50-333.
- 4. Letter, R. L. Cridley to D. G. Eisenhut (NRC), " Review of Low-Core Flow
'l Effects on LOCA Analysis for Operating BWRs", May 8, 1978.
j 5. Letter D. G. Eisenhut (NRC) to R. L. Gridley, enclosing " Safety Evaluation Report Revision of Previously Imposed MAPLHGR (ECCS-LOCA) Restrictions for BWRs at Less Than Rated Flow", May 19, 1978.
- 6. " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)", NEDO-21662, July 1977.
- 7. " Supplemental Reload Licensing Submittal for James A. FitzPatrick Nuclear Power Plant, Reload 3, NEDO-24242, February 1980.
s 8. " General Electric Boiling Water Reactor Load Line Limit Analysis for James A. FitzPatrick Nuclear Power Plant", NEDO-24243, February 1980.
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