ML20040A315
| ML20040A315 | |
| Person / Time | |
|---|---|
| Site: | Green County  |
| Issue date: | 02/07/1977 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200752 | |
| Download: ML20040A315 (34) | |
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p UNITED STATES y*
'4 NUCt. EAR REGULATORY COMMISSION jI Mij WASHINGTON, D. C. 20555
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i G7 g3 MEMORANDUM FOR:
D. B. Vassallo, Assistant Director for Light Water Reactors FROM:
D. F. Ross, Assistant Director for Reactor Safety, DSS
SUBJECT:
.95R INPUT FOR GREENE COUNTY Plant Name:
Greene County Nuclear Power Plant Docket Number:
50-549 Milestone Number:
24-24 Licensing Stage:
CP Responsible Branch LWR-2 and Project Mana5er:
S. Burwell Systems Safety Branch Involved:. Core Performance Branch Description of Review:
SER Input Requested Completion Date:
January 14, 1977 Review Status:
Complete The Reactor Fuels Section and Reactor Physics Section of the Core Performance Bra.ach have prepared the enclosed SER Input for the Greene County Nuclear Power Plant
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DE'nwood F."Ross, Assistant Director for Reactor Safety Division of Systems Safety
Enclosures:
As stated cc:
S. Hanauer R. Heineman R. Boyd K. Kniel e
S. Burwell P. Check D. Fieno R. Meyer W. Mcdonald W. Brooks M. Tokar T. Novak Z. Rosztoczy
Contact:
W. Brooks, NRR, 27577 8201200752 810403 PDR FOIA MADDEN 80-515 PDR
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SER Inout from Fuels Section for Greene County 4.2.1 Fuel l
4.2.1.1 Descriotion The Greene County Nuclear Power Plant will contain the Mark C fuel assembly, which is a new design incorporating a 17x17 square array of fuel rods.
The basic Mark C fuel assembly consists of 264 fuel rods, 24 control rod guide tubes, 1 instrumentation tube assembly, 8 spacer grids and 2 end fittings.
The guide tubes, spacer grids, and end fittings form a structural cage for the rods and tubes.
The guide tubes are rigidly atta.hed to the upper and lower end fittings.
The upper end fitting positions the upper end of the fuel assembly in the upper grid plate structure and provides means for coupling the handling equipment.
Integral with each upper end fitting are holddown springs and a spider to provide a positive holddown margin to oppose hydrculic forces.
The lower end fitting positions the fuel assembly in the lower core grid plate.
The lower end fitting grillage provides a support surface for the bottom end of the fuel rod.
The spacer grids are constructed from strips which are slotted and fitted together in an " egg-crate" fashion. Each j
grid has 36 strips -- 18 perpendicular to 18 -- which form the 17 by 17 lattice.
The square cells formed by the interlaced strips provide support for the fuel rods in two perpendicular directions through contact points on each wall of each cell.
The contacts are in the form of protruding dimples, which are integrally punched from the strips on the walls of each square opening. On each of the two end spacer grids, the peripheral-strip is extended and mechanically attached to the respective end fitting.
The Zircaloy guide tubes provide a guidance envelope for the control rods, which are moved in and out of the fuel assembly during operation.
They also provide the structural continuity for the fuel assembly. Welded to each end of a guide tube are threaded sleeves, which secure the guide tubes to each end fitting by lock-welded nuts. Transverse posi-tioning of the guide tubes i: provided by the spacer grids.
The Zircaloy instrumentation tube serves as a channel to guide, position, and contain the incore instrumentation within the fuel assembly.
The instrumentation tube is located on the centerline of the fuel assembly and is axially retained at the lower end fitting.
The spacer tubes fit around the instrumentation tube batween spacer grids and restrict axial movement of the spacer grids.
Each fuel rod is comprised of cladding, fuel pellets, end caps, and internal support components.
The fuel is in the e
form of sintered and 6round pellets of low enriched uranium dioxide.
Pellet ends are dished and chamfered.
The pellets are inserted in Zircaloy-4 tubing and sealed by Zircaloy-4 J
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. end caps, welded at each end of the tubes.
All fuel rods are internally pressurized with helium, which reduces the differ-ential pressure across the cladding and improves heat transfer within the rods. The level of prepressurization is designed both to preclude cladding tensile stresses resulting from a net internal pressure and to reduce cladding flattening (" creep-down") during normal operation.
Above the fuel column in each rod is a spring that separates the fuel from the fuel rod upper end cap.
This a
spring maintains the fuel column in place during shipping and handling.
In operation, the, spring permits axial differential growth and thermal expansion between the fuel rod and the cladding. Below each fuel column there is also a spring which axially locates the bottom of the fuel column and separates the fuel free the lower fuel rod end cap.
This spring is designed 9
to deflect under high column loads to reduce axial strain in the cladding.
This 2-spring, 2-plenum design is unique to B&W fuel as the rest of the U.S. fuel manufacturers utilize only an upper plenum and spring.
The " Mark C" fuel assembly design (17x17 array) is N
mechanically similar to the Babcock and Wilcox " Mark B" fuel i
assembly (15x15 array).
A comparison of critical dimensions l
l is indicated in Table 4.2-1 of this report.
The differences i
l l
are essentially geometric and will result in a lower linear power density, with concomitantly lower average and maximum centerline fuel temperatures for the Mark C design compared with the Mark B.
The evaluation of the Babcock and Wilcox Mark C fuel
=echanical design is based upon the assessment of mechanical tests, in-reactor operating experience with lead assemblies, and engineering tests.
Additionally, the in-reactor per formance of the Mark C fuel design will be subject surveillance programs carried out by Babcock A individual utilities.
4.2.1.2 Thermal Performance In our evaluation of the thermal performance of ths reactor fuel, we assume that densification of the uranium dioxida fuel pellets may occur during irradiation in light water reactors. The initial density of the fuel pellets and the size, shape, and distribution of pores within the fuel pellets influence the densification pheno =enon.
Briefly stated,' in-reactor densification (shrinkage) of oxide fuel pellets (a) may reduce gap conductance, and hence increase-fuel temperatures, because of a decrease in pellet diameter; y-(b) increases the linear heat generation rate because of the decrease in pellet length; and (c) may result in gaps in the-y l
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fuel column as a result of pellet length decreases -- these gaps produce local power spikes and the potential for cladding creep collapse.
The engineering methods to be used by Babcock and Wilcox to analyze the densification effects on fuel thermal performance have been previously submitted (Ref. 1) to the NRC staff and approved for use in licensing. The results of our review are reported in two memoranda (Ref. 2,3), and additional information on densification methods can be found in "The Analysis of Fuel Densification," NUREG-0085, (Ref. 4).
The approved B&W model for fuel densification has been incorporated, along with companion models for fuel swelling, gas release, gap closure, and cladding creep, in a B&W computer code, called " TACO."
This code was written to calculate gap conduct-ance, fission gas pressure and stored energy over the lifetime of the fuel rod. A description (Ref. 5) of the TACO code has been submitted for review. Our review has prompted several modifica-tions of this code, including a revision of the fission gas release model to account for enhanced release at high exposures.
Issuance of an approved version of TACO is imminent.- Thermal performance calculations will be performed by B&W with the NRC approved version of the TACO code.
c A topical report (Ref. 6) describing the ane.lytical procedure and supporcing data developed by Babcock and Wilcox to determine the minimum time for B&W fuel cladding to collapse
under operating conditions was reviewed, and, subject to provi-sions specified in our evaluation report (Ref. 7), was accepted for use in safety analyses related to licensing. The computer
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code used to perform these calculations is referred to as "CROV."
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In our evaluation of the CROV topical report we stated that A
it was acceptable for use in safety analyses related to
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licensing provided that (1) the creep related material proper-ties were similar to those characteristic of current B&W cladding, (2) the initial ovality input to CROV both bounds s
the as-fabricated cladding and is not less than 0.0005 inches min), and (3) the results of long-term, in-(Ovality =0D
-0D max
. reactor confirmatory tests continue to be favorable.
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Recently, a revision (Ref. 8) to the CROV report was reviewed. The revised report was unchanged from the original except for the creep correlation and its effects on predicted times to cladding collapse.
In our evaluation (Ref. 9) of the revised report we have noted that the cladding creep s
' correlation provides a more phenomenologically sat'sfying i
s expression, reduces some of the former uncertainties, eliminates the need for some assumptions and still conservatively represents appropriate creep data.
The revised report has, therefore, been found acceptable for use in safety analyses related to e
licensing, subject to the same conditions noted above for the original report.
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4.2.1 3 Mechanical Performance Although limited operating experience exists on the Mark C 17x17 fuel assemblies, substantially all of the in-reactor operating experience with Babcock & Wilcox fuel rods and assemblies is applicable to the Greene County Hark C fuel design, since the 17x17 fuel assembly is only a slight mechanical extrapolation from the 15x15 fuel assembly.
The current use of similar fuel rods and assemblies has yielded operating experience that provides confidence in the acceptable perform-ance of the fuel, assembly design.
In addition, B&W is irradiating two Mark C demonstration asse=blies in the Oconee 2 reactor. Non-destructive post-irradiation examinations are being performed at each refueling outage, and a destructive examination will be conducted after three cycle's of operation.
The irradiation of these two 17x17 demonstration assemblies is described in a B&W report (Ref. 10).
Many of the tests and analyses previously performed for the Mark B fuel assembly are applicable for the Mark C assembly.
The development program for the Mark C design is drawing upon the experience gained in the Mark B development program. B&W has stated that the Mark C program results will be used to demonstrate that Mark B tests and analyses are applicable for the safety analysis of the Mark C fuel and that, where neces-sary, supplemental tests have been planned to demonstrate the performance of the Mark C design.
In the event that any
experimental results fall outside the design values used in the a_otysis of Mark C assembly performance, changes in the Mark C design may be required.
The Mark C fuel assembly program objectives are to obtain data for analytical models, to confirm analytical predictions, and to verify the adequacy of the design.
Included in this effort are mechanical and flow tests (scheduled for completion in September,1977) and critical heat flux and reficod heat transfer testing (to be completed in September, 1978).
These programs are described briefly in section 1.5 of the PSAR and in more detail.in a topical report (Ref. 11).
This topical report was reviewed and found acceptable for reference in licensing applications, provided that B&W submits semi-annual status reports, beginning January, 1976, on the Mark C research and development programs.
In our evaluation (Ref.12) of the report on the Mark C R&D program, however, we also noted that a thorough safety review of the Mark C design and issuance of an operating license for a Mark C plant would require addi-tional information in some areas that were not addressed in the R&D topical report. One such area is rod bowing.
Rod-to-rod gap spacing measurements have been taken through two cycles of operation on the lead Mark B plaat c:
on fuel assemblies up to 20,000 mwd / tU burnup.
In addition, visual examinations of peripheral rods for bowing have been performed at five different plants, and bow profile measurements
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-9 have been taken on rods on the lead Mark B plant.
The schedule and scope of the 15x15 examination program includes three-cycle examination of 15x15 fuel assemblies in the lead Mark B plant.
End-of-core-1 and end-of-core-2 examinations have been completed.
and the data and evaluations have been reported (Ref. 13) to the the NRC staff.
B&W has stated that a topical report covering the results of three cycles of operation will be submitted in the first quarter of 1978. To project 17x17 Mark C bowing from Mark B data, B&W proposes to follow the technique discussed in an NRC memorandum (Ref. 14). This technique will be used for interim evaluation until actual Mark C bow data are available.
Methods used by B&W to analyze the effects of rod bowing in Mark B (15x15) fuel have been reviewed and approved (Ref. 15).
Revised methods for Mark C (17x17) fuel, which may be presented in a supplement to the Mark C topical report or as a separate report, will be incorporated in the FSAR. Guidelines for treating rod bowing in applications for construction permits and operating licenses were recently established (Ref. 16).
In general, no interim rod bowing DNB penalties are being applied to CP applications.
In addition to rod bowing, several other fuel performance c
issues are being studied as generic issues.
These include the mechanical response to seismic and LOCA forces, the potential for waterlogging rupture, fretting and wear, fatigue, and pellet /
cladding interaction (PCI). Although each of these issues has been identified as an area requiring furthe'r study, final resolution is not considered necessary for the preliminary design approval requested by B&W. Each item is addressed below.
With respect to the seismic and loss-of-coolant analyses, we have requested (Ref. 17) an in-depth safety analysis of the seismic and LOCA response of the Mark C (17x17) fuel assembly.
Although a commitment to submit a topical report in early 1976 (at least one year prior to the filing of the first FSAR incorporating the Mark C fuel assembly) was made by B&W (Ref. 18),
this schedule has slipped to March, 1977 At that time B&W is expected to submit a topical report entitled " Mark C Fuel Assembly LOCA Seismic Analyses." The analysis will include asymmetric loads that may result from a pipebreak within a biological shield.
During the course of the review of the PSAR,the potential for waterlogging rupture emerged as an item requiring further consideration.
The original PSAR (p.4.2-9) contained a 4
statement that " increased reactor coolant activity would identify a fuel rod rupture and allow adequate time for corrective action to be taken." A question was raised (Q231.14) on the " corrective action" suggested in the PSAR.
The response was that the recommended startup power rate in the 0-205 power range would be reduced.
We, thereupon, asked another question (Q 231 37) on the reco= mended reduction in startup power rate.
The response to this question indicated
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_11 that the PSAR section " Potential for Water-logging Rupture" (4.2.1.3.2) never addressed the potential but rather the consecuences of rupture. B&W thus contended that rupture due io waterlogging cannot occur.
B&W has also cited test results (PSAR section 4.4.3 9) that indicate that the effects of waterlogging rupture are not severe and should not result in failure propagation.
We have reviewed the safety aspects of waterlogging failure, not only as a result of our PSAR review activity, but also as a consequence of our inquiry into the broader issue of pellet / cladding interaction (PCI) as a potential failure mechanism. A survey of the available information, which includes (1) test results from SPERT and Japanese test reactor NSRR, and (2) observations of waterlogging failures in commercial reactors, indicates that the assessment of the consequences of.
waterlogging failures made by B&W is correct. We believe, however, that the probability of waterlogging rupture, while small, is still finite (i.e., not zero).
Since waterlogging ruptures can lead to the dispersal of fuel into the primary coolant, consideration should be given to methods of reducing both the potential and the consequences of waterlogging failure.
One such method could involve a reduction in the rate of power c
increase allowed during startup. We will continue to monitor the waterlogging test programs and study this phenomenon generically.
Limitations on power rate changes could also affect pellet /
cladding interaction,'which is being reviewed as a generic item.
The B&W fuel rod design incorporates features directed at reducing cladding strain due to PCI.
These include pellet chamfering, rod prepressurization, plenum regions at both the top and bottom of the fuel rod, and greater cladding thickness-to-diameter ratio (in Mark C fuel as compared with Mark B).
Based on the available experimental and commercial reactor data, the design features adopted by B&W should result in a reduction or delay of PCI failures to late in the fuel design life.
While the failure thresholds are probably lower at high burr.up than at low burnup, the fuel duty is also less severe. Our review of the consequences of PCI failures has so far not resulted in the identification of safety problems.
Therefore, no operating restrictions are currently warranted.
If any safety issues are identified in the future, however, appropriate restrictions will be implemented.
In the treatment of fuel assembly fretting and wear, B&W contends that for Mark B fuel the potential for fretting is low so long as interference between fuel rod and spacer grid is maintained throughout life.
This condition is proposed as a design limit for Mark C as well as Mark B c
assemblies because of their similarity in design.
A test program, outlined in section 1.5 2.1.2 of the PSAR, is being performed to verify the adequacy of this limit for the Mark C
, fuel assembly. Fuel assembly resistance to fretting and wear is to be evaluated on the basis of' test results from exposure at reactor operating conditions in the control rod drive line (CRDL), which is described in reference 11.
In addition, evidence of fretting and wear will be checked for a range of flow rates, temperatures and pressures in both the CRDL and the cold water facility (CWF).
Results and interpretation of test data will be presented to the NRC in a topical report at least one year prior to submittal of an FSAR on a reactor incorporating Mark C fuel assemblies.
In the topical report B&W will also specify the criteria used in judgins the applicability and adequacy of the tests.
In the PSAR discussion of cycling and fatique (p 4.2-11),
B&W states that " fatigue analyses, based on conservative assumptions, will be performed to show that design limits.
are met."
In view of this statement, we have requested a topical report on the cycling and fatigue analyses.
This topical report will be submitted for review at least one year prior to the date of submittal of an FSAR on the Greene County Nuclear plant.
Descriptions of the experimental facilities used in the Mark C fuel assembly mechanical and hydraulic testing program C}
are provided in reference 11 along with the test descriptions.
These include assembly mechanical tests, a lateral stiffnese test, frequency and damping test, tilt test, shim test,
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tensional stiffness test, axial stiffness test, assembly hydraulic test, and tests on control rod wear, assembly wear and life, and several tests on the grid spacer and springs.
In addition critical heat flux and rod swelling and burst test are being performed. Since the Mark C assembly R&D reports are being up-dated on a semi-annual basis, we expect that the reports, coupled with meetings with B&W to discuss program plans and developments, will keep us adequately advised of the progress of the Mark C assembly development program.
4.2.1.4 Fuel Surveillance Performance of the fuel during operation will be indirectly
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e monitored by measurement of the activities of both the primary and secondary coolants for compliance with technical specifi-cation limits. For new fuel designs for which there is no operating experience, we require that a supplemental fuel surveillance program be conducted.
The supplemental fuel surveillance program is directed at monitoring the behavior of the actual fuel systers as they perform in reactor, thus demonstrating the adequacy of the conclusions reached in the design evaluation.
We are, therefore, requiring a fuel surveillance program for the first two plants with Mark C fuel.
The details of cq the surveillance program requirements are provided in reference 19 Briefly, the program will consist of a visual inspection
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l 1
L
. of all the peripheral rods in the initial-core fuel assemblies, as they are discharged into the spent fuel pool. The visual inspection will include observations for cladding defects, fretting, rod bowing, corrosion and deposition and geometric distortion.
If any anomalies are detected during the visual examination, further investigation will be performed, including, under unusual circumstances, destructive examination of a fuel I
assembly or individual fuel rods as required.
4.2.1.5 Fuel Mechanical Desian Evaluation Conclusion Subject to the following conditions:
l (1) favorable results of the Mark C assembly development
- program, (2) the successful operation and subsequent PIE of the two Mark C demonstration assemblies in the Oconee 2 reactor, we conclude that there is reasonable assurance that the cladding integrity of Greene County Mark C (17x17) fuel will be maintained, that significant amounts of radioactivity w'ill not be released, and that neither accidents nor earthquake-induced loads will result in either an inability to cool the fuel or interference with control rod insertion. Our conclusion is based on (1) analytical results, (2) operating experience with similar c
Mark B (ISx15) fuel, (3) increased thermal margins of 17x17 fuels, (4) technical specifications that will be in effect l
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. j to limit offgas and effluent activity, (5) the on-going development and demonstration test program, and (6) the j
commitment to perform the Mark C fuel rod surveillance program and post-irradiation examinations.
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Table 4.2-1 Dimensional Comoarison i
Mark B Mark C (15x15)
(17x17)
I No. of fuel rods per fuel assembly 208 264 No. of guide tubes per assembly 16 24 No. of instrument tubes per assembly 1
1 Fuel rod OD, in.
0.430 0 379 Cladding thickness in.
0.0265 0.0235 Fuel rod pitch, in.
0 568 0 501 Fuel assembly pitch spacing, in.
8.587 8.587 Guide tube OD, in.
0.530 0.465 Instrument tube spacer sleeve OD, in.
0.554 0.480 Fuel pellet OD, in.
0 370 0 324 Fuel pellet length, in.
0 70 0 375 Fuel stack length, in.
144 143 d
References (Section 4.2.1) 1.
B.J. Buescher and J.W. Pegram, " Babcock & Wilcox Model for Predicting In-Reactor Densification," BAW-10083 (Proprietary Version BAW-10083P),
Rev. 1, November 1976.
2.
P.S. Check (NRC), Memorandum to S.A. Varga, " Review of Revision 1 to BAW-10083P," December 27, 1976.
3 V. Stello (NRC), Memorandum to V.A. Moore, " Review of B&W Densification Model," September 8, 1975.
4.
R.O. Meyer, "The Analysis of Fuel Densification," NUREG-0085, July 1976.
5 R.H. Stoudt, 31,al., " TACO-Fuel Pin Performance Analysis," BAW-10087 June 1976.
6.
A.N.J. Eckert, H.W. Wilson, and K.E. Yoon, " Program to Determine In-Reactor Performance of B&W Fuels-Cladding Creep Collapse, "BAW-10084, May 1974.
7.
Victor Stello, Jr. (NRC) Memorandum to V.A. Moore, "A Generic Review of the B&W Cladding Creep Collapse Analysis Topical Report BAW-10084,"
August 9, 1974.
8.
A.N.J. Eckert, H. W. Wilson, and K.E. Yoon, " Program to Determine In-Reactor Performance of B&W Fuels-Cladding Creep Collapse," BAW-10084 (Rev 1), October 1976.
9 V. Stello (NRC), Memorandum to D. Vassallo, " Program to Determine In-Reactor Performance of B&W Fuels-Cladding Creep Collapse," December 1976.
10.
" Irradiation of Two 17x17 Demonstration Assemblies in Oconee 2, Cycle 2 - Reload Report," BAW-1424, January 1976.
11.
R.H. Stoudt, P.H. Klink, and L.A. Walton, " Mark C (17x17) Fuel Assembly - Research and Development - Revision 2" BAW-10097 (Proprietary Version 10097P), July 1975.
12.
A. Schwencer (NRC), Letter to K.E. Suhrke (B&W), November 25, 1976.
13 K.E. Suhrke (B&W), Letter to D.F. Ross (NRC), " Fuel Rod Bow Projection, 95x95 Upper Tolerance Limit, 15x15 Fuel Assembly (Mark B)," September 10, 1976.
14.
D. F. Ross (NRC), Memorandum to R.C. DeYoung and D.G. Eisenhut,
" Westinghouse Fuel Rod Bowing," April 14, 1976.
15 a)
S. Kim (NRC), Memorandum to D.F. Ross, " Babcock and Wilcox Rod Bow Model," November 26, 1975 b)
S. Kim (NRC), Memorandum to Paul S. Check, " Babcock and Wilcox Rod Bowing," April 5, 1976.
16.
D.F. Ross and D.G. Eisenhut (NRC), Memorandum to R.S. Boyd, R.E. Heineman, and V. Stello, " Guidelines for Treating Upper Head Temperature and Rod Bow in Applications for Construction Permits Operating Licenses," December 15, 1976.
17 D.F. Ross (NRC), Memorandum to A. Schwencer, July 25, 1974.
C 18.
J. Malloy (B&W), Letter to A. Schwencer (NRC), September 7, 1974.
19 D.F. Ross (NRC), Letter to Kenneth E. Suhrke (B&W) September 20, 1976.
t
r SER Inaut fredh,ypig_s Section for Gr,e_eae County 43 Nuclear Design Our review of the nuclear design of the Greene County Nuclear Power Plant (GCNP) was bared on information supplied by Power Authority of the State of New York (PASNY) in the Preliminary Safety Analysis Report and acendmenta thereto.
Our review was conducted in accordance with Section 4.3 of the Standard Review Plan (NUREG-75/087).
The nuclear design of GCNP is similar to that of other B&W plants having 205 fuel assemblies (e.g, Pebble Springs Nuclear Plant, Docket 50-514/515) which have been reviewed and approved.
431 Design Bases Design bases are presented which comply with the appli-cable General Design Criteria. Acceptable fuel design limits are specified (GDC 10), a negative prompt feedback coefficient is specified (GDC 11) and power oscillations are required either to be not possible or to be detected and suppressed by the control system (GDC 12). Design bases are presented which require control and monitoring system (GDC 13) which automatically a
initiates a rapid reactivity insertion to prevent exceeding fuel design limits in normal operation or anticipated transients C
(GDC 20).
The control systen is required to be designed so that a single malfunction or single operstor error will cause no
violation of fuel design limits (GDC 25).
A chemical shis system is provided which is capable of bringing the reactor to cold shutdown conditions (GDC 26) and the control system is required to control reactivity changes during accident conditions when combined with the engineered safety features (GDC 27). Reactivity accident conditions are required to be limited so that no damage to the RCS occurs (GDC 28)
We find the design bases presented in the PSAR to be acceptable.
4.3 2 Design Description Descriptions of the first cycle enrichment distribution, burnable poison loading, soluble boron concentrations, pluto-nium buildup, delayed neutron fraction, neutron lifettte, and core burnup have been provided. The values presented for these parameters are consistent with the design bases and are acceptable.
Power Distribution We have reviewed the methods used by Babcock and Wilcox to calculate power distributions for both steady state and transient conditions (see Section 4 3 3, below).
These methods have been c
compared to experiment to determine the uncertainty in the core peaking factor for plants with 177 fuel assemblies.
In addition the, power distributions at various core powers have been compared
- to calculated values during startup testing of several B&W plants with 177 fuel assemblies.
These comparisons have shown that assumption of a 7.5 percent uncertainty on the calculated core peaking factor is conservative. Fcr one plant (Rancho Seco Unit 1) our consultant, Brookhaven National Laboratory, has performed an independent audit of heat generation rates at beginning of life.
The audit calculation agreed to within approximately 3 5% with those calculated by B&W. While these comparisons have been for plants with 177 fuel assemblies, it is expected that compar-isons will not be appreciably different with plants having 205 fuel assemblies. In addition, the peaking factors will be measured at several power levels during the escalation to full power in the startup tests.
Monitoring of power distributions for GCNP will be performed by excore detectors (axially split ionization chambers) or by incore self-powered neutron detectors.
The former are used for the reactor protection system and may be used for monitoring operating conditions.
The incore detectors are used for calibrating csrtain functions (axial imbalance and quadrant tilt) of the excore system and may be used for C
monitoring operating conditions.
Both sets of instrumentation are normalized to the calorimetric value of total reactor power. Functionally identical systems have been successfully employed on other B&W plants, and we find their use acceptable.
-4 Peaking factor limits are determined by the requirement that fuel design limits not be exceeded during normal operation or anticipated transients and that a fuel clad temperature of 2200* F be not exceeded during a LOCA.
In operation, peaking factors are controlled by the application of limits on control rod insertion and on axial L7 balance.
Operating experience with B&W reactors has shown that operation within these limits (which are included in the plant Technical Specifi' cations) is sufficient to assure that peaking factor limits are not violated.
On the basis of good agreement between calculation and measurement in operating B&W reactors, we conclude that.
acceptable predictions of core power distributions have been made by PASNY for GCNP.
On the basis of satisfactory operating experience in B&W reactors, we conclude that adequate instru-mentation exists to monitor power distributions during plant operation.
Reactivity Coefficients Babcock and Wilcox has presented values of various reactivity coefficients in the PSAR which are used in the analysis of normal, transient, and accident conditions.
e Included are values for moderator temperature, Doppler, and power coefficients of reactivity as well as soluble poison worths.
The calculation methods used to obtain these quantities
' have been reviewed (see Sect 4 33 below). These quantities are routinely measured as part of the startup program at various reactor power levels.
In particular, measurements of the power coefficient and moderator temperature coefficients are usually within %10% of the calculated value. Measured values are always conservative with respect to values which are used in safety analyses.
With respect to operation with positive moderator i
temperature coefficient, a Technical SpecificatLon requirement that this coefficient be negative above 95% of rated power is typically imposed.
Analyses show that transients or accidents initiated below 95% power do not exceed safety limits even for positive temperature coefficients within Technical Specifica-tion limits.
On the basis of good agreement between calculation and measurement in operating B&W reactors, and on the fact that startup tests in new plants will confirm that conservative reactivity coefficients were used in safety analyses, we conclude that suitably conservative values of reactLvity coefficients have been provided.
i l
l Control Requirements ch To allow for. changes in rsactivity due to reactor heatup, load following, and fuel burnup with consequent TLssion product buildup, a significant amount of excess reactivity is built
4 into the core.
This excess reactivity is controlled by a combination of soluble boron and control rods.
Soluble boron is used to control reactivity changes due to:
. Moderator deficit from ambient to operating temperatures
. Equilibrium xenon and samarium buildup
. Fuel depletion and fission product buildup through-out cycle life that part not controlled by burnable poison
. Transient xenon resulting from load following Regulating rods are used to control reactivity changes due to:
. Moderator deficit from hot zero poder to full power
. Power level changes (Doppler)
Lumped burnable poison rods are used for radial flux shaping and to control part of the reactivity change due to fuel burnup. Part length control rods are used to maintain an axially balanced power distributton.
PASNY has provided data to show that adequate control exists to satisfy the above requirements with enough additional control to provide a hot shutdown effective multi-plication factor (k,gg) ;; 0.99 during the initial and equilibrium cycles with the most reactive control rod stuck e
out of the core.
Comparisons between calculated and measured control rod worths have been made for As-In-Cd control rods in 177 fuel assembly plants.
The agreement obtained was well within the 10 percent error assigned to total cod worth in the
analysis.
It is not anticipated that the expansion of the core size to 205 fuel assemblies will affect the agreement between calculation and exportsent by a significant amount.
In addition, control rod worth measurements, with particular emphasis on shutdown margins, is a part of the startup progran for each plant.
On the basis of our review, which has included the considerations described above, we conclude that the applicant has made suitably conservative assessments of reactivity control requirements for GCNP and that adequ' ate reactivity has been provided to assure shutdoen capability.
The soluble boron (chemical shim) system has sufficient capability to shut down the reactor and to naintaLn it in the cold shutdown state at any point in core life.
This satisfies the requirement of General Design Criterion 26.
Control Rod Patterns and Reactivity '4 orth The full length control rods are divided into two classes - shutdown (or safety) rods and regulating rods.
The safety rods are always completely out of the core when the reactor is at operating conditions. Load (core power) changes are made with regulating rods and/or soluble boron.
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Regulating rod insertion will be controlled by power dependent insertion limits given in the Technical Specifi-cations. These limits will ensure that:
. There is sufficient negative reactivity available to permit the rapid shutdown of the reactor with adequate margin.
. The worth of a control rod that might be ejected in the unlikely event of failure of a pressure barrier in a control rod drive mechanism is not greater than that which has,been shown to have acceptable conse-quences in the safety analysis.
. The overall peaking factor does not exceed that used in the safety analysis as the initiating value for transients or accidents.
We have reviewed the calculated rod dorths and the methods used by B&W to obtain the worths. Our consultant, Brookhaven National Laboratory, has performed independent calculations of regulating bank worths and their calculations agree to within N2 percent with those calculated by B&W.
The effects of fuel densification on peaking factors are reflected in the Technical Specifications on rod insertion limits.
These effects are considered in the OL licensing stage since "as-built" fuel characteristics inust be used.
Babcock and Wilcox has an acceptable codel for fuel densifi-cation which will be applied.
It is expected that modification
_g.
may be made in this model as more data are obtained.
These modifications are reviewed generically and the latest accept-able model will be used to account for denstfloation effects at the appropriate stage in the licensing process.
On the basis of our review, we have concluded that the rod groupings proposed for GCNP satisfy the requirements for safe shutdown, ejected rod worth and poder distribution control and are acceptable.
Stability The stability of the 205 fuel assembly core with respect to xenon oscillations has been analyzed.
Azimuthal xenon oscillations are predicted to be daoped but sustained axial xenon oscillations may occur under certain conditions if no remedial action is taken.
The stability of 177 fuel arsembly reactors was investigated during the startup tests for the Oconee Unit 1 reactor.
A diagonal (combination of azimuthal and axial) oscillation was induced at 75 percent power and the reactor response was monitored for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The azimuthal component was strongly damped but the axial component was divergent.
At 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> into the transient the part length rods were used to return the axial imbalance to near zero, c
where it was successfully kept.
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The increased lateral dimensions of the 205 fuel assembly plants is expected to increase the tendency toward azimuthal oscillations but these cores a're still expected to be stable azimuthally. The heicht of these plants is the same as that of the 177 fuel assembly plants and the part length rods are expected to be as effective in suppressing axial oscillations as in the 177 fuel assembly plants.
On the basis of our review we conclude that these plants will not experience uncontrolled oscillations.
This conclusion will be verified on an early 205 fuel assembly plant.
4.3 3 Analytical Methods Babcock and Wilcox has submitted a series of topical reports (3AW-10111 to 10115, BAW-10117 and BAW-10124) which describe the analytical methods used by B&W in the nuclear design of reactor cores. These topical reports describe the data base used, the manner in which the data are treated to obtain calculational parameters for use in the design codes, PDQ07 and FLAME, to perform the design calculations.
Sufficient detail is given concerning data sources, physics of the calculations and calculation procedures to allow a knowledge-able person to conclude that the methods employed are " state-C of-the-art."
We have reviewed these reports and have concluded that they are acceptable for reference in licensing submittals.
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. 15 0 Accident Analysis 15.1.1 Startup Accident, and 15 1.2 Rod Withdrawal at Power We have reviewed the reactivity coefficients and reactivity insertion rates used in the analysis of these transients.
The nominal values quoted for the analysis are conservative with respect to those given in Section 4 3 Further conservatism in the analyses is provided by sensitivity analyses which yield the consequences of the transient as a function of:
. Reactivity insertion rate
. Doppler coefficient
. Moderator coefficient
. Trip delay time 15.1 3 Control Rod Misoperation The case of control rod misoperation - rod stuck or dropped - has been examined.
The limitin5 case - that of a rod dropped into the core - has been analyzed.
The Integra-1 l
ted Control System (ICS) acts to prohibit rod withdrawal and l
to initiate a power runback to 60% power whenever a rod is more than 9 inches removed from the average position of its d
group. No credit is taken for this action in the analysis.
l
Analyses have been performed for both BOL and EOL conditions.
At BOL the core power drops to S60% of full power and remains there.
The core pressure and temperature fall since heat is being removed at a rate greater than it is being produced (ICS action to reduce the load is assumed not to occur). The reactor TLnally trips on low pressure or low temperature. The change in DNBR during this transtent has been analyzed and it has been shown that DNB does not occur.
I At EOL the moderator and Doppler coefficients are more negative and the core power at first drops rapidly but then returns to approximately full power where it remains.
The i
average moderator teraperature decreases by N15' F and the system pressure by %150 pai.
The reactor stablises in this l'
new condition and a trip does not occur. In the new reactor i
state peaking factors may be higher than normal.
- However, tests in the Rancho Seco Unit 1 reactor, a 177 fuel assembly plant, have shown that no thermal limits are exceeded at full power when the rod causing the largest peaking factor change is dropped into the core. It is expected that this will also be true for 205 fuel assembly plants, and this fact will be verified during startup tests.
On the basis of our review, we conclude that the analysis c
of the rod drop accident is acceptable.
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, 15.1.4 Rod Ejection Accident We have reviewed the analysis of the rod ejection accident presented in the Preliminary Safety Analysis Report.
The requirements of Regulatory Guide 177 have been met.
Analyses were performed at BOL and EOL and for full power and near zero power cases. The maximum anticipated potential ejected rod worth is 0.5% Ak/k for both full power and zero power cases.
However, analyras have been carried out at worths up to 1.0% Ak/k at zero power and 0.65% Ak/k at full power.
The rod ejection at full power and beginning of life results in the largest enthalpy increase in the fuel. The peak enthalpy
. values are N150 cal /gm for a rod worth of 0.5% ak/k and N 170 cal /gm for a worth of 0.65% Ak/k. Both these values are far below our acceptance criteria of 280 cal /sm.
The rod ejection accident analysis is performed with a point kinetics code. Doppler feedback in the analysis is based on core average power - no credit is taken for the h!gh feedback in the hottest rod.
This procedure has been shown to be conservative by comparing point kinetics calculations with two-dimensional WIGL2 calculations. Further, we have compared the B&W point kinetics calculation results with those from three-dimensional calculations and have confirmed the conserv-c ative nature of the point kinetics results.
l l
On the basis of our review we conclude that the analysis l
of the rod ejection accident is acceptable.
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_14 15.1.5 Fuel Misloading s
Three types of fuel misloadings are considered:
1.
Misloading a fuel pellet with an incorrect enrichm nts
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in a fuel rod, 2.
Misloading a fuel rod with an incorrect enrichment x_
/
into a fuel assembly, and 3
Misloading a fuel assembly with an incorrect enrichinent'
,into the core.
Babcock and Wilcox has performed analyses of the effects.of these types of misloadings (" Effects of Asymmetries in Fuel -
Loading, BAW-10028, July 1971). The first two types of loading errors do not have serious consequences on DNB (DNBR does not go belo'w 1 32 for design overpower conditions).
The third type of error may, however, produce DNB conditions on design overpower when the most reactive fuel assembly is loaded into certain positions intended for the least reactive assembly..
The probability of a fuel misloadin6 error is reduced to a very low value by careful procedures during the manufacture of the fuel assemblies and the loading of the reactor.
In addition, a careful comparison of core power distributions (obtained from the incore detectors) to calculated values is C
made.
In particular, the power distributions are carefully examined to determine whether a tendency to an'acimuthal power tilt is developing.
In this manner there is a high probability of detecting a misloaded fuel assembly.
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t Finally the undesirable consequences of a misloaded fuel assembly are limited to the misloaded assembly which constitutes less than 0.5% of the fuel in the core.
In view of the fact that operaticn with 1.0% failed fuel is permitted and of the low probability of occurrence of a misloaded
~
assembly, we find the analysis of the misplaced fuel accident to be acceptable.
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Appendix C Information for Greene County The following is a list of comments concerning items from the list of ACRS Generic Items.
Item IB-1 Section 4 3 3 Reactivity Coefficients Item IB-2 Fixed incore detector are not required in PWR's since reviews have not revealed a clear need for continuous monitoring (letter, Moeller to Rowden, April 16, 1976)
Item IB-7 Section 15.1.4 ktemIC-1 Section 4.3 3 - control Rod Patterns and reactivity Worth Item IC-3 Not applicable Item IIA-2 Not applicable 4
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