ML20057A473
| ML20057A473 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/31/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9309140267 | |
| Download: ML20057A473 (5) | |
Text
. _ _.
GE NucIcar Energy c,.. n 5c..,cw s., e ; os. s August 31,1993 Docket No. STN 52-001 f
r l
Chet Posiusny, Senior Project Manager Standardization Project Directorate l
Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Schedule - Fuel and Control Rods
Dear Chet:
Enclosed are SSAR markups of two of the three items discussed at the August 20,1993 phone call between Caroline Smith and floward Richings pertaining to fuel and control rods. These markups will be included in Amendment 32.
Please provide a copy of this transmittal to floward Richings.
Sincerely, 4
Jack Fcx Advanced Reactor Programs Alan Beard (GE) cc:
Norman Fletcher (DOE)
Careline Smith (GE)
I 080040 Ql nm m p-t 9309140267 930831 PDR ADOCK 05200001 A
23A6100 Rev.1 ABWR Standard Safety Analysis Repor?
The boron carbide powder in the absorber tubes is compacted to about 70% ofits theoretical density and contains a mimmum of 76.5% by weight of natural boron. The boron-10 minimum content of the boron is 18% by weight. The absorher tubes are scaled by a plug welded into each end. The boton carbide is longnudinally separated into individual compartments by stainless steel balls at approximately 10-inch in tervals.
The steel balls Are held in place by a slight crimp of the tube.
4.2.3 Design Evaluation 4.2.3.1 Fuel Assembly 4.2.3.1.1 Evaluation Methods The thermal-mechanical evaluation's described in Section 41L3 ofAppendix 4B.with the exception of the stress / strain analyses in 4B.3(2)(a), are all performed using the NRC-approved GESTR-MECHANICAL fuel rod thermal-mechanical performance model.
The stress / strain methodology is described later in this subsection. A3 clo v cgc.3 ic I4ew 3 O'S W O*dobh ~^ 5 A L *S-Poor NR.c avsew awJ pov a l.
GESTR-MECHAMCAL The GESTR-MECllANICAL fuel rod performance model performs best estimate coupled thermal and mechanical analyses of a fuel rod experiencing a variable operating history. The model explicitly addresses the effects of:
Fuel and cladding thermal expansion a
Fuel and cladding creep and plasticity e
Cladding irradiation growth a
Claddi.y irradiation hardening and thermal annealing of that irradiation a
hardening Fuel irradiation swelling a
a Fuel irradiation-induced densincation Fuel cracking and relocation a
Fuel hot pressing a
Fission gas generation and exposure <nhanced fission gas release including fission a
product helium release Differential axial expansion of the fuel and cladding reDecting axial slip or lockup a
of the fuel pellets wah the cladding Fuel phase change solumenic expansion upon melting a
fuel System Desgn - Amendment 31 4 2-5
23A6100 Rev r ABWR Standard Safery Analysis Report The GESTR41ECHANICAL material properties and component models represent the latest exper: mental information available.
NRC approval of the GESTR-MECH ANICAL model and its application methodology is provided in Reference 4.2-2.
Stress / Strain Analyses The fuel rod cladding stress analyses are performed using a Monte Carlo statistical method in conjunction with distortion energy theory. Fuel cladding plasticity analyses are also performed when required by the loading conditions.
NRC approval of the stress / strain analyses methodology is provided in Reference 4.2-2.
4.2.3.1.2 Evaluation Results The fuel rod thermal-mechanical evaluations described in Section 4B.3 0f Appendix 4B have been completed for the reference fuel design (GE P8x8R design) using the methodologies described in Subsection 4.2.3.1.1. The evaluations demonstrate that the criteria of Appendix 4B are satisfied for the PSxSR fuel design. The NRC previously reviewed and approved this in Reference 4.2-2.
The approval in Reference 4.2 2 contains the following conditions:
(1) The hcense/ applicant must provide a plant-specific analysis of combined seismic and LOCA loading using NRC-approved methodology or another acceptable method to demonstrate conformance to the stnictural acceptance requirements described in Appendix A of Standard Review Plan Section 4.2.
(2) The license / applicant must provide an acceptable post-irradiation suneillance program or endorse the approved GE fuel surveillance program.
For the P8x8R fuel design, the second condition is satisfied by the fuel suneillance program described in Section 4B 2(3) of Appendix 4B (see also Subsection 4.2.4).
4.2.3.2 Control Rod9
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f 4.2.4 Testing, inspection, and Surveillance Plans GE has an active program of suneillance of fuel, both production and desclopmental.
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The NRC has reviewed the GE program and approved it in Reference 4.2-3.
4.2.5 References 4 2-1 GE Fud Bund!r Dnigns, NEDE-31152P.
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O b a h vat Av4 w (.A. J'.c W b ok e e d. T b g qvcde h o w o <kom o w t h ch. k b e o v-4.2-6 Cvs A&
- E ^ f Q w chw 4C crv-a. 5 sk\\s $ d f,f ir k H f C M 4 6q C Cow A vo\\ ch. FuelSynem Design ~ A,nendment 31
23A6100 Rev 1 ABWR Standard Safety Analysis Report 4B Fuel Licensing Acceptance Criteria 48.1 Introduction A set of fuel licensing acceptance cnteria has been established for evaluating fuel designs and for deterr ining the applicability of generic analvses to these designs. Fuel a
design compliance with the fuel lMensing acceptanc:. cnteria constitutes USNRC acceptance and approval of the fuel design for initia' core and reload applications without specific USNRC resiew. The fuellicensing acceptance criteria are presented in i
-,9 c
the following subsections. An Chin.4 io Wie GMen w Dws+
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l 48.2 General Criteria (1) NRC-approved analytical models and analysis procedures are applied.
Consistent with current practice, NRC-approved procedures and methods are f
used to evaluate new fuel designs.
(2) New design features are iricluded in lead test assemblies.
GE's " test before use" fuel design philwophy includes irradiation expenence with new fuel design features in full-scale fuel assemblies (Lead Test Assemblies) in operadng reactors prior to standard reload applicadon.
(3) The generic post-irradiation fuel examination program approved by the NRC is maintained.
Section 4.2.ILD.3 of the Standard Review Plan (SRP) requires each plant to f
implement a post-irradiation fuel surveillance program to detect anomalies or to confirm expected fuel performance. The USNRC has found that the fuel surveillance program described in Reference 4B-1 is an acceptable means for licensees to satisfy the post-irradiation surveillance requirement of the SRP.
l This program includes examination of LTAs and selected discharged bundles with the results reported to the USNRC in a yearly operating experience
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report.
4B.3 Therrnal-Mechanical (1) The fuel design thermal-mechanical anahses are performed for the following l
conditions:
(a)
Either worst tolerance assumptions are applied or probabilistic anahses are performed to determine statistically bounding results (i.e., upper 95c7c confidence)
(b) Operating conditions are taken to bound the condinons anucipated dunng nonnal steady-state operation and anticipated operanonal i
l occurrences Fuel Licensing Acceptarnce Crnens - Amendment 31 48-1
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23A6100 Rev.1 ABWR
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Standard Safety Analysis Report 4C Control Rod Licensing Acceptance Criteria 4C.1 Introduction A set of acceptance enteria has been established for evaluating new control rod designs.
l Control rod compliance with these criteria constitutes the basis for USNRC acceptance I
and approval of the design. The control rod licensing acceptance criteria and their bases are provided below. Hn spu s
,'o ihcde, C,ra hrk r<ud iW'9
we prior NRd. Nv;ew ard cwproval y
C.2 General Criteria Control rod designs must meet the following acceptance criteria-(1) The control rod stresses, strains.and cumulative fatigue shall be evaluated to i
not exceed the ultimate stress or strain of the material.
(2) The control rod shall be evaluated to be capable ofinsertion into the core during all modes of plant operation within the limits assumed in the plant 5
analyses.
l (3) The material of the control rod shall be shown to be compatible with the
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reactor emironment.
i (4) The reactivity worth of the control rod shall be included in the plant core analyses.
(5) A surveillance program shall be implemented if a change in design features such as new absorber material or structural material not previously used in reactor cores could impact the function of the control rod.
4C.3 Basis for Acceptance Criteria i
J The following comprise the basis for the licensing acceptance criteria given in Section 4C.2.
4C.3.1 Stress, Strain and Fatigue i
The control rod is evaluated to assure that it does not fail because ofloads due to l
shipping, handling, and normal, abnormal, emergency, and faulted operating modes.
To assure that the control rod does not fail, these loads must not exceed the ultimate I
stress and strain limit of the material. Fatigue must not exceed a fatigue usage factor of 1.0.
The loads evaluated include those due to normal operational transients (scram and jogging), pressure differentials thermal gradients, flow-and system-induced vibration, j
and irradiation growth in addition to the lateral and verticalloads expected for each Control Rad Lacensing J.cceptance Cnteris - Amendment 31 4C4 5