ML20041E114
| ML20041E114 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/18/1982 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TAC-43380, TAC-47355, TAC-47389, TAC-47473, NUDOCS 8203100160 | |
| Download: ML20041E114 (34) | |
Text
1 FEB 1 b 1382 Docket No. 50-336 DISTRIBUTION Mr. W. G. Counsil, Vice President NSIC Docket File RAClark NRC PDR Nuclear Engineering & Operations Northeast Nuclear Energy Company ACRS.10 Local PDR P. O. Box 270 Gray File DEisenhut m
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Hartford, Connecticut 06101 K
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Dear Mr. Counsil:
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Tq This is the third and final letter addressing our review of the Bfil e k
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Safety Report (BSR) submitted by your letter dated March 6,1980.
The h8 C
BSR is intended to serve as a reference fuel assembly and safety 11ys16F%
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report for use of Westinghouse fuel assemblies in NSSS designed by 9m.-
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Mg bustion Engineering, specifically in Millstone Nuclear Power Station Unit No. 2. Our letters, dated June 22, 1981 and January 12, 1982, our acceptance of the physics portion of the BSR and of the transient analyses covered in Sections 5.3.2 through 5.3.9, 5.3.13 and 5.3.15 through 5.3.17 of the DSR. This letter transmits our Safety evaluation covering the reactor fuels and thermal-hydraulic sections of this same document. This completes our review of the BSR.
Based on our review, we conclude that the reactor fuel design and the thermo-hydraulic characteristics, including the remaining transient and accident analyses addressed in the BSR, are acceptable for reference in licensing actions for Millstone, Unit No. 2.
Sincerely, Goynal signed by Robert A. Clark Robert A. Clark, Chief Operating Reactors Branch #3 I
Division of Licensing cc: See next page l
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NRCFORM 318 (10-80) NHCM 024a OFFIClAL RECORD COPY usam. mi_.m yo
s Northeast Nuclear Energy Company CC:
William H. Cuddy, Esquire
'Mr. John Shedlosky Day, Berry & Howard Resident Inspector / Millstone Counselors at Law c/o U.S.N.R.C.
One Constitution Plaza P. O. Drawer KK Hartford, Connecticut 06103 Niantic, CT 06357 Mr. Charles Brinkman Regional Administrator Manager - Washington. Nuclear Nuclear Regulatory Commission, Region 1 Operations Office of Inspection and Enforcement C-E Power Systems 631 Park Avenue Combustion Engineering, Inc.
King of Prussia, Pennsylvania 19406 4853 Cordell Aven., Suite A-1 Bethesda, MD 20014 Mr. Lawrence Bettencourt, First Selectman Town of Waterford Hall of Records - 200 Boston Post Road Waterford, Connecticut 06385 Northeast Nuclear Energy Company ATTN:
Superintendent Millstone Plant Office of Policy & Managerent Post Office Box 128 ATTN: Under Secretary Energy Waterford, Connecticut 06385 Division 80 Washington Street Waterford Public Library Hartford, Connecticut 06115 Rope Ferry Road, Route 156 Waterford, Connecticut 06385 U. S. Environmental Protection Agnecy Region 1 Office ATTN:
Regional Radiation Representative John F. Kennedy Federal Building Boston, Massachusetts 02203 Northeast Utilities Service Company ATTN: Mr. Richard T. Laudenat, Manager Generation Facilities Licensing P. O. Box 270 Hartford, Connecticut 06101 e
E'nclo s"ure FINAL SAFETY EVALUATION WESTINGHOUSE BASIC S'AFETY REPORT 1
INTRODUCTION By letter (Ref.1) dated March 6,1980, Northeast Nuclear Energy Conpany (NNECo) submitted the Basic Safety Report (Ref. 2) for NRC review. The BSR was prepared by Westinghouse Electric Corporation and is applicable to Millstone, Unit 2 operation with Westinghouse reload fuel. Hence, the BSR is intended to be a reference FSAR that supplants sections of the original Millstone, Unit 2 FSAR, which was prepared by Combustion Engineering Company.
In many analyses, the BSR uses input specific to Millstone, Unit 2 Cycle 3, which is designated as the " reference cycle."
Previously the BSR constituted a substantial portion of the reload safety analysis for Cycle 4 operation. However, because of NRC manpower limitations the staff was unable to review the~BSR on a generic basis. Conse-quently, our approval of the BSR was limited to Cycle 4 operation only. We are now able to provide (enclosed) our generic input to the safety evaluation report that is currently being compiled by DL. Our 5,ER input allows NNEco to l
utilize the BSR as a referential report for future cycles of operation provided that the accrued fuel burnups and other important analytical inputs do not exceed those expliciti employed in the respective BSR analyses.
2 FUEL SYSTEM DESIGN The objectives of this fuel system safety review are to provide assurance that (a) the fuel system is not damaged as a result of normal operation and anticipated operacional occurrences, (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) cool-ability is always maintained.
"Not damaged" is defined as meaning that fuel rods do not fail, that fuel system dimensions remain within operational
~
tolerances, and that functional capabilities are not reduced below those assumed in the safety analysis.
This objective implements General Design Criterion 10 (Ref. 3) and the design limits that accomplish' this are called Specified Acceptable Fuel Design '.imits (SAFDLs).
" Fuel rod failure" means that the fuel rod leaks and that the first fission product barrier (the cladding) has, therefore, been breached.
Fuel rod failures must be accounted for in the dose analysis required by 10 CFR Part 100 (Ref. 4) for postulated accidents.
"Coolability," which is sometimes temed "coolable geometry",
means, in general, that the fuel assembly retains its rod-bundle geometrical configuration with adequate coolant channel spacing to permit removal of residual heat even after a severe accident. The general requirements to maintain control rod insertability and core coolability appear repeatedly in the General Design Criteria (e.g., GDC.!7 (Ref. 5) and 35 (Ref. 6)). Specific coolability requirements for the loss-of-coolant accidents are given in 10 CFR Part 50.46 (Re'. 7).
To meet the above-stated objectives of the fuel system review, the following specific areas are critically examined:
(a) design bases (and limits),
(b) description (and design drawings), (c) desian evaluation, and (d) testing, inspection, and surveillance plans.
In asser o ~ the adequacy of the design, several items involving operating experi m & totype testing, and analyti-cal predictions are weighed in terms of tr W.
ecceptance criteria for fuel system damage, fuel rod failure, and fuel coolability..
2.1 Design Bases '
With the exception of,the fuel rod internal gas pressure design basis, the specific Millstone, Unit 2 design bases are given in RESAR-414 (Ref. 8), which was reviewed (Ref. 9) and granted a Preliminary Design Approval in 1978
( Re f. 10).
These design bases are comparable to those previously established for Millstone, Ur:it 2.
With regard to fuel rod internal gas pressure, the Westinghouse reload fuel is designed such that the internal pressur'e will not exceed the nominal primary coolant system pressure during the design lifetime of the fuel.
This is an acceptable criterion; it is more conservative than the criterion used in RESAR-414, and it corresponds to the criterion in the current Standard Review Plan (Ref.11).
To ensure that the design bases of the Millstone, Unit 2 reload fuel are met, Westinghouse has used their standard evaluation techniques and methods.
Many of these are described in the BSR. To establish the Reactor Protection System (RPS) setpoints, which detemine the Limiting Safety System Settings (LSSS) and the Limiting Conditions for Operation (LCO), the Westinghouse fuel is designed to conform to the following SAFDLs.
1.
The peak linear heat rate must be below that which will cause incipient U
fuel centerline melting (i.e., 4700 F).
2.
The departure from r.ucleate boiling thermal limits will not be exceeded
(
(i.e., W-3 DNBR > 1.30 using THINC-I).
l These two SAFDLs have previously b'een evaluated versus applicable experimental data i
and were found acceptable for application to Millstone prior to Cycle 4 operation.
2.2 Design Description The 2700 Mwt Millstone, Unit 2 core is composed of 217 fuel assemblies and employs 73 control element assemblies (CEAs), of which 12 CEAs are of the dual i.
9
type (i.e., contain 10 neutron absorber elements and interface with more th The active fuel zone is 136.7 inches tall and each fuel one fuel assembly).
a:sembly uses 9 fuel rod spacer grids supported by 5 guide tubes.
A brief description of Westinghouse reload fuel assembly components, including fuel rods, upper and lower end fittings, guide tubes, and spacer Numerical values and drawings are provided for grids is contained in the BSR.
Since the Westinghouse reload fuel is intended to be some components.
mechanically and hydraulically compatible with the original Combustion The most notable Engineering NSSS fuel, many design aspects are identical.
difference is in Westinghouse's use of Inconcel-718 space.r grids as compared to Combustion Engineering's use of 8 Zircalay-4 grids and 1 Inconel-625 bottom Yet there are many subtle but not inconsequential design spacer grid.
differences such as (a) Westinghouse's use of spacer grid bow springs and dimples rather than leaf springs and arches, (b) Westinghouse's use of swa mechanical joints between spacer grids and guide tubes rather than welding, and (c) Westinghouse's raised pad feature on the center of top nozzle orifice This latter feature is intended to limit the amount of fuel assembly plates.
liftoff during an accident (e.g., blowdown) and thus preclude the coil springs from being compressed solid, and prevent the nozzle pin extensions from Table 1 provides a topping out in the upper core plate blind holes.
comparison of some pertinent fuel assembly design variables.
2.3 Design Evaluation _
Cladding Collapse Westinghouse reload fuel rods are internally pressurized with helium prior to The end-cap welding to reduce cladding compressive stresses during service.
combination of the level of prepressurization, theoretical density of fuel pellets, cladding wall load-carrying capacity, and restrictions on the degree of as-fabricated cladding ovality are designed to preclude cladding collapse during the fuel lifetime.
TABLE 1 FUEL MECHANICAL AND HYDRAULIC DESIGN COMPARISONS Combustion Engineering Westinghouse Design Parameter Reference Reload Fuel assembly Fuel rod array 14x14 14x14 Number of fuel rods 176 176 Number of spacer grids 9
9 Number of control rod guide tubes 4
4 Number of instrument tubes 1
1 Assembly pitch (inches) 8.180 8.180 Assembly. envelop (inches)(%)
8.19 8.19 Lower nozzle blocked area 64 64 Upper nozzle blocked area (%)
57 56 Grid blocked area (%)
22 20 Pressure drop flow factor (fl/De) 3.90 3.90 Fuel rod pitch (inch) 0.580 0.580 Fuel-rod-to-bottom-plate spacing (inch) 0 0.17-0.20 Number of coil springs 4
4 Fuel pellets Length (inch) 0.650 0.600 Colunn height, cold (inches) 136.7 136.7 Theoretical density (percent) 94.75 - 95.0 95.0 Diameter (inch) 0.3795 0.3805 Fuel cladding Outerdiameter(inch) 0.440 0.440 l
Thickness (inch) 0.026 0.026 Control rod guide tube Innerdiameter(inches) 1.035 1.035 l
Thickness (inch)-
0.040 0.038 Instrument tube Inner diameter (inches) 1.035 1.035 Thickness (inch) 0.040 0.038 i
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The Westinghouse analysis for cladding collapse is based on a model (COLLAP) described in WCAP-8377 (Ref. 12).
Our evaluation (Ref. 13) of the COLLAP code concluded that the model is acceptable if used without alteration l
to the specified curves which are used as input to the model.
Consequently, we conclude that NNECo should confirm for each cycle of application that (a) the specified input curves are used in accordance with the conditions of the safety evaluation report (Ref. 13) and (b) the predicted cladding collapse time exceeds the expected lifetime of the fuel.
Fuel Rod Bowing Fuel rod bowing is a phenomenom that alters design pitch dimensions between adjacent fuel rods.
Bowing affects local nuclear power peaking and the local heat transfer to the coolant.
Rather than placing design limits on the degree of bowing that is permitted, the effects of bowing are explicitly included in plant safety analyses.
The most recent Westinghouse methods for analyzing the effects of rod bowing are presented in WCAP-8691, Revision 1 (Ref. 14), which is under review by an NRC contractor.
Some of the methods proposed in this WCAP report are also employed in the BSR analysis.
We have thus reviewed those methods and our evaluation is discussed below.
The Millstone, Unit 2 rod bow analysis can be divided into three specific segments.
First, a baseline prediction of bowing-induced gap closure, which is that for the Westinghouse 15 x 15 fuel design; second, a formulation for extrapolating that burnup-dependent closure to a closure which would correspond to the Westinghouse reload fuel for Millstone, Unit 2 at the same exposure; and third, reference to a gap closure (viz., 50%) below which there is no observed DNB effect.
With regard to the first and third segments, we have previously approved (Refs.15 and 16) these for Westinghouse applications and their application here is acceptable.
The second segment assesses the differences in spacer grid span lengths and fuel rod cross-sectional moments of inertia between the two fuel designs. Though the particular formulation used is in contrast to (and less conservative than) that previously approved (Refs. 17, 18, and 15) by the NRC staff, our contractor's review of th'e revised WCAP-8691 _ _ _ _ _ _ _ _ _ _
report has progress:d to the' point that we are able to project a fav'orabfo safety evaluation report on this specific formulation that Westinghouse has proposed.
Consequently, we approve of the fue.1 rod bowing analysis and agree with the BSR's conclusion that no bowing penalty is required for Millstone, Unit 2.
Fuel Rod Fretting Wear The Westinghouse reload fuel for Millstone, Unit 2 employs a spacer grid / fuel rod support design (i.e., springs and dimples) similar to that in standard Westinghouse fuel assemblies.
Therefore, NNECo and Westinghouse have not seen a need to conduct long-duration flow tests to investigate the grid / cladding -
fretting wear potential of the new Westinghouse-supplied fuel for Millstone, Unit 2 reloads.
For their standard fuel, Westinghouse has found acceptable experience in (a) 1000-hour duration flow tests for several spacer grid / fuel rod configurations and (b) post-irradiation examinations of spent fuel assemblies, which have not shown evidence of appreciable wear (that level of wear which would threaten fuel rod hermiticity). We thus agree that the Westinghouse design and experience (which at this time includes the apparently favorable experience of about o'ne year of operation of 72 Westinghouse reload assemblies used in Millstone's Cycle 4) are sufficient to conclude that the Westinghouse reload fuel will have, an acceptable resistance to fretting wear.
Fuel Assembly Guide Tube Fretting Wear Fretting wear has been observed in irradiated fuel assemblies taken from Millstone, Unit 2 (Ref.19) and other Combustion Engineering NSSS plants (for example see Ref,erences 20, 21, and 22). These observations detected an unexpected wearing of guide tubes that are under CEAs.
Coolant turbulence was responsible for inducing vibratory motions
- in the normally fully withdrawn CEAs and, when these vibrating control rods were in contact with the inner surface of the guide tubes, a wearing of the guide tube wall
'has ensued.
The most substantial wear has been found to be limited to the relatively soft Zircaloy-4 guide tubes because the Inconel-625 cladding on the control rods provides a relatively hard wear surface. The extent of the observed wear has appeared to be plant dependent and has in some cases extended completely through the guide tube walls..
9
As an interim solution to this problem, NNECo and many other Combustion Engineering licensees have attached sleeve inserts to the interior of the uppermost portions of rodded fuel assembly guide tubes. The function of the sleeve inserts is to provide relatively fretting resistant barriers rather than to eliminate CEA vibrating motion.
Sleeves are also used in the Westinghouse supplied fuel assemblies to alleviate guide tube wear.
The Westinghouse design is similar to that of the Combustion Engineering design inasmuch as both designs are similarly dimensioned stainless-steel sleeves that are partially chrome plated and have series of slots and holes.
(The chrome plate provides a bearing surface for control rod vibration and the slots and holes preclude coolant entrapment between the guide tube and the sleeve.) Major differences, however, do exist in the design of the upper end of the sleeves and the method of sleeve attachments.
On the Combustion Engineering sleeve design, the upper ends of the sleeves are conically shaped to fit the contour of the upper end fitting posts. Because the conical section is not connected to the post, free movement under heatup, cooldown, and differential irradiation g.rowth exists between the guide tube and sleeve.
The sleeves extend from the top of the upper end fitting posts to several inches below the area where the ends of the control rods reside when in the fully withdrawn position. The sleeves are securely fastened in place by mechanically " bulging" both the sleeve and the guide tube near the lower end of the sleeve.
The Westinghouse design is completely cylindrical with no conica11y shaped end.
And the mechanical attachment of the sleeve is accomplished by outwardly deforming the sleeve into two swage grooves, which are located in the top nozzle extension.
(The Combustion Engineering terminology.for this component is upper end fitting post.) For this Westinghouse method of attachment, free movement of the sleeves is accommodated inversely to that of the Combustion Engineering method.
We have previously concluded that the use of sleeve inserts is an acceptable means of eliminating guide tube wear and does not produce undesirable changes in fuel assembly structural properties.
In addition, confirmatory CEA scram testing has not revealed any significant occurrences where the use of sleeve inserts produced unacceptable scram times.
Our previous approvals for the use l.
of sleeve inserts in Combustion Engineering NSSS plants were for Millstone, Unit 2 (Ref. 23) Calvert Cliffs, Units 1 and 2,(Refs. 24 and 25), Arkansas Nuclear One, Unit 2 (Ref. 26), and others.
We therefore conclude that guide tube fretting wear has been adequately addressed in the BSR and that any required surveillance will be addressed during our review of individual reload safety analysis reports.
Though the use of sleeve inserts precludes guide tube wear, this remedy adversely affects CEAs by exposing CEA cladding to a harder wear surface.
Consequently at this time, outage surveillance (for Millstone, Unit 2 discussions see Refs. 27, 28, 29, and 30) continues to be necessary to assess the degree of CEA fretting wear and hence the ability.to maintain hermiticity and to achieve design lifetime of CEAs.
Until this issue is generically resolved for Combustion Engineering NSSS plants or specifically resolved for Millstone, Unit 2, it should be considered in the preparation of reload safety analyses for Millstone, Unit 2.
~
Fuel Assembly Liftoff The holddown flower is the interface between the upper core plate and the fuel assembly coil springs.
The coil springs are designed to provide holddown force and accommodate axial expansion and contraction of.the fuel assemblies.
When a fuel assembly is in the core, the holddown flower contacts the underside of the upper core plate and preloads the fuel assembly in order to resist hydraulic loads.
The Westinghouse analysis shows that the coil springs provide sufficient holddown capability to preclude fuel assembly liftoff under nonnal operating condition's (which includes anticipated operational transients) throughout the fuel assembly lifetime.
This is consistent with the original design basis for the Millstone, Unit 2 fuel that was supplied by Combustion Engineering. We, therefore, conclude that fuel assembly liftoff has been adequately addressed.
-g-n
, Thermal Performance Analytical Methods The Westinghouse fuel thermal performance code as described in WCAP-8720 (Ref. 31) was used for the-Millstone, Unit 2 safety analysis. This Westinghouse code was approved with four restrictions as described in our safety evaluation of February 9, 1979 (Ref. 32). Three of those restric-tions deal with numerical limits and have been complied with. The fourth restriction relates to the use of the PAD-3.3 code for the analysis of fission gas release from uranium dioxide (U0 ) for power increasing conditions during 2
normal operation.
This restriction applies to the safety analysis of Millstone, Unit 2.
However, Westinghouse has stated that this restriction does not adversely affect the results of the safety analyses performed for Millstone, Unit 2.
We believe that this statement is essentially correct for the planned operation of Millstone, Unit 2, and Westinghouse has prepared and submitted a detailed evaluation (Ref. 33) of this restriction.
At this time, we have not completed our review of the Westinghouse evaluation of this restriction.
However, our review has progressed to the point where the following conclusions can be made:
1.
The Westinghouse evaluation of our restriction on the use of the PAD-3.3 code supports their earlier statement that the restriction does not adversely affect the results of the safety analyses performed for Millstone, Unit 2.
. ~
2.
Based on additiorial infomation stbmitted IRef. 34) by Westinghouse to confim this conclusion, we continue to believe that this result is essentially correct.
3.
Because the restriction pertains to the release o'f fission gases from the fuel, any change in our conclusions would not have significant impact at low burnups, when the fission gas. inventory in the fuel is small.
At this time, therefore, we conclude,that for Millstone, Unit 2, the restriction on PAD-3.3 is not significant and the analyses as presently docketed are acceptable.
We anticipate completion of our review of the Westinghouse evaluation prior to operation of the Westinghduse reload fuel at high burn 0ps.
We will reopen our review of this issue with NNECo at that time if the need should arise.
Seismic and LOCA Mechanical Response One of the NRC's generic safety issues deals with asymmetric blowdown loads in a LOCA (Re s. 35, 36, and 37).
For fuel assemblies, the e
asymmetric blowdown loads and the ' loads from the safe-shutdown earthquake are used to determine if fuel assembly components meet a'cceptance criteria (see Appendix A in Reference 11).
The analysis of Millstone, Unit 2 for a homogeneous core of Westinghouse reload fuel assemblies has been submitted (Ref. 38).
The analysis predicts that grid defomation occurs in fuel assemblies adjacent,to the core shroud, but NNECo has determined (Ref. 39) that the effect of assembly deformation on fuel rod heat transfer during the reflood stage will not invalidate (i.e., produce higher peak cladding temperatures than) the current LOCA analysis (Ref. 40).
9
The mechanical analysis for a homogeneous core of Westinghouse assemblies was performed with methods (Ref. 41) that we have reviewed and approved (Ref.42). However, for input this fuel assembly analysis utilized the core-plate-motion-time history developed as part of the generic asymetric LOCA loads evaluation program, and the primary systems asymmetric loads analysis for this program has not been reviewed. Since the Task Action Plan (Ref. 36) for this generic issue provides a basis for continued operation while the issue is being resolved, no further action is required at this time.
Should the review of the asymetric loads analysis show that the core-plate-time history used for Millstone, Unit 2.was incorrect, NNECo will be notified.
During the first few cycles of operation with Westinghouse reload fuel, the Millstone, Unit 2 core will contain a mixture of Westinghouse and Combustion Engineering fuel. The fuel assembly analysis mentioned above does not consider this design variation, and this mixed-core issue must be addressed in each reload safety analysis that uses both fuel designs.
Material Properties The BSR provides or makes reference (e.g., RESAR-414) to various important material properties that are used in the Millstone, Unit 2 core analysis. We have reviewed many of these properties, which include such parameters as Young's Modulus of elasticity for Zircaloy-4 cladding, thermal expansion of UO2 pellets, and thennal conductivity of Inconel-718 spacer grids.
We have found that the proposed values for these properties (a) do not strongly influence the analyses in which they are employed; (b) have been explicitly approved in other Westinghouse documents such as PAD-3.3 (Ref. 31) and remain acceptable, (c) are similar or conservative to properties that were proposed by other vendors and subsequently approved, (d) are a reasonable or conservative interpretation of publically available data and correlations such as in MATPRO (Ref. 43), or (e) have been challenced elsewhere and we are pursuing resolution to the issue on a separate basis (see the subsequent section dealing with Cladding Swelling and Rupture During LOCA). We, therefore, conclude that material properties have been adequately addressed..
Pellet / Cladding Interaction Fuel warranty limitations on power rate changes will affect fuel pellet /
cladding interaction (PCI), which is being reviewed as a generic issue. The Westinghouse reload fuel rod design for Millstone, Unit 2 incorporates features (such as pellet dishing and rod prepressurization) that reduce cladding strain due to PCI.
Based on the available experimental and commercial reactor data, we believe that these design features should result in a reduction or delay of potential PCI failures to later in the fuel operating life, albeit by a presently unquantified degree. While the failure thresholds are probably lower at high burnup than at low burnup, the fuel duty (e.g.,
peak power and rate of power change) is also reduced..
Our review of PCI has to date not resulted in a revision of the current licensing criteria that are applicable to PCI. As discussed in the Standard Review Plan (Ref.11), two acceptance criteria in current use are (1) no centerline fuel melting and (2) less than 1% transient-induced cladding s train.
As discussed in Chapter 6 of the BSR, the Westinghouse reload fuel for Millstone, Unit 2 is designed to assure that U0 centerline melting will not 2
occur through selection of a calculated fuel centerline temperat'are of 4700 F as an overpower limit.
The l' cal power density trip provides such overpower o
protection during anticipated operational occurrences.
The trip setpoint detennination for the " reference cycle" was made with general analysis methodology and computer co, des as @ scribed in the approved RESAR-414 report (Ref. 8) and are thus acceptable for this application as well.
With regard to the second criterion, Westinghouse has used the approved PAD-3.3 code to calculate transient-induced cladding creep strain, and the values calculated are found (Ref. 44) to be less than 1% plastic strain. This is consistent with the design criterion given in the Millstone, Unit 2 FSAR (Ref. 45). Therefore, the two existing licensing criteria for PCI have'been satisfied..
E
0xidation and Crud Buildup Mechanical properties of the cladding are not significantly affected by oxide and crud buildup.
However, cladding oxidation and crud buildup are potential fuel system damage mehanisms, inasmuch as the phenomena affect the cladding-to-coolant heat transfer coefficient and the temperature drop across the
' cladding wall.
Because of the increased themal resistance of these layers, there is an increased potential for elevated temperature within the fuel as well as the cladding'.
Because the effect of oxidation and crud layers on fuel and cladding temperature is a function of several different parameters (e.g.,
heat flux and themal-hydraulic boundary conditions), a design limit on oxide or crud layer thi:Eness does not, per se, preclude fuel damage as a result of these layers.
Rather, it is 'necessary that these layers be appropriately considered in other temperature-related fuel system damage and failure analyses.
In the BSR, there is no explicit discussion of cladding oxidation and crud buildup.
The applicable models for cladding oxidation and crud buildup are, however, discussed in the supporting document (Ref. 46) for the Westinghouse fuel performance code PAD.
These models were previously approved by the NRC staff, and we find the models applicable to the Westinghouse reload fuel for Millstone, Unit 2.
On the basis of our previous review of the oxidation and crud buildup models, we conclude that these effects have been adequately accounted for in the BSR.
Hydriding I'nternal bydriding as a cladding failure mechanism is precluded by controlling the level of moisture and other hydrogenous impurities during fabrication.
Westinghouse has us:d moisture and hydrogen control limits in the manufacture
.of earlier fuel types and has found that typical end-of-life cladding hydrogen levels are less than 100 ppm, a level below which hydride blister fonnation is not anticipated in fuel cladding. A specific design limit for external -
hydriding has been found unnecessary, inasmuch as there are apparently no LWR fuel failures occurring because of external hydriding.
AsdescribedinReference47,themoisturalevelsintheuraniumdio'id[ fuel x
are limited by Westinghouse to less than or equal to 20 ppm, and these specifications are compatible with the ASTri specification (Ref. 48), which allows 2, ug hydrogen per gram of uranium -(i.e., 2 ppm), and they are the same as the limits provided in the Standard Review Plan; they are therefore acceptable.
CEA Ejection As discussed in the BSR, the limiting criterion for the CEA ejection accident is that the average fuel pellet enthalpy at the hotspot will be less than 200 cal /g for unirradiated and irradiated fuel.
This criterion is more conservative than the 280 cal /g limit given in Regulatory Guide 1.77 (Ref. 49) and is therefore acceptable.
Fuel Assembly Design Stress For normal operation and anticipated operational occurrences, the design bases for stress of the Westinghouse reload fuel (except for grids, which are analyzed separately under the Seismic and LOCA Mechanical Response Section) for Millstone, Unit 2 are that (a) the calculated primary membrane stress (P,)
will be less than or equal to the design stress intensity (S ) as defined by m
the ASME code (Ref. 50) and (b) the sum of P and the calculated primary m
bending stress (P ) will be 1bss than or equal to the product of S b
m times 1.5.
The value of S, for Ercaloy components is the lesser value of one-third of the ultimate or two-thirds of the yield strength as corrected for temperature e,ffects.
For stainless-steel components, Sm is taken directly from Reference 50.
These criteria and material limits are generally accepted by the industry at large, and we find'their applicati'on to Millstone, Unit 2 acceptable.
The acceptability of the design relative to various loading conditions was subsequently determined by comparing the maximum result of each individual test or analysis with allowable limits.
Component analyses are performe'd with the Westinghouse computer code WECAN (Ref. 51). All calculated values were found to be less than allowables; consequently, we conclude that fuel assembly design stress has been treated adequately..
Cladding Swelling and Rupture During LOCA The BSR does not provide LOCA analysis; hence, the evaluation of fuel cladding swelling and rupture during LOCA will be addressed elsewhere.
2.4 Evaluation Conclusions We conclude that the Westinghouse reload fuel for Millstone, Unit 2 has been designed so that (a)'the fuel system will not be damaged as a result of normal operation and anticipated operational occurrence'., (b) fuel damage during the postulated accidents that are specifically anal; zed in the BSR would not be severe enough to prevent control rod insertion when it is required, and (c) core coolability will always be maintained, even after the severe postulated accidents that are specifically analyzed in the BSR.
(These accidents do not include the LOCA.) This conclusion is based on the appl'i-cant's documented evidence that the stipulated (and acceptable) design objectives will be met based on Westinghouse operating experience and analytical predictions.
Consequently, on the basis of our review of the BSR, we conclude that (a) the Westinghouse reload fuel design has met all the requirements of the applicable regulations, regulatory guides, and regulatory positions that are applicable to Millstone, Unit 2, and (b) the BSR may be utilized as an acceptable reference in NNECo licensing submittals that address the operation of Millstone, Unit 2.
The following constitute the only limitations on our approval:
1.
Individual reload applications referencing the BSR should confirm that the cladding collapse analysis has been performed in accordance with the conditions of approval given on the generic methods.
2.
Until the issues of fuel assembly guide tube and CEA cladding fretting wear are resolved generically or specifically for Millstone, Unit 2, l
individual reload applications referencing the BSR should propose (or.
provide justification for lack of) surveillance plans that would provide continuing assurance o,f structural integrity.
l.
3.
Themal-Hydraulic Design 3.1 Introduction The reactor for the Millstone, Unit 2 power plant is designed by Combustion Engineering.
This section describes the hydraulic compatibility, design basis and design methodology for the thermal analysis of the Millstone, Unit 2 core when using Westinghouse fuel assemblies. The thermal-hydraulic design of the Westinghouse fuel assemblies is shown to be equivalent to the original Combustion Engineering fuel assemblies because of hydraulic compatibility. The hydraulic compatibility of the Westinghouse and Millstone, Unit 2 reference design (Cycle 3) fuel assemblies is ' based on the similarity of the fuel assemblies and on test results. Westinghouse Reload Safety Evaluation Methodology (Ref. 52) is applied to the fuel assembly design.
3.2 Hydraulic Compatibility 4
The Westinghouse reload fuel assembly for Millstone, Unit 2 is designed to t'e dimensionally and hydraulically compatible with the Combustion Engineering reference fuel assembly (Cycle 3).
As shown in Table 1, the fuel rod diameter (0.440 inch), fuel rod pitch (0.580 inch), and fuel assembly pitch (8.18 inches) are the same for both types of assemblies.
The hydraulic effects of the different configurations used by the Westinghouse and the Millstone, Unit 2 reference cycle in the upper nozzle, lower nozzle, and the grids have been minimized since the Westinghouse components have, as closely as possible, the same blockage as the Millstone, Unit 2 reference cycle design as shown in Table 1.
Therefore, the rod bundle axial and lateral flow areas, the axial frictional pressure drop, and the lateral flow (croscflow) resistance will be the same for both designs.
The pressure drop through these components consists primarily of fom (expansions and contractions) rather than frictional losses, therefore, matching the blocked area results in matching pressure drop. -
The Westinghouse fuel assembly was tested in the Fuel Assembly Test System (FATS) hydraulic loop to confinn that the resi. stance was the same where The two areas of physical dissimilarity are:
physical differences exist.
The grid--The Westinghouse and Millstone, Unit 2 reference cycle grids 1.
have different hold-down spring and dimple arrangements.
Location of fuel rods off bottom--The rods for the Westinghouse 2.
design are from 0.17 to 0.20 inches above the top of the bottom nozzle.
The Millstone, Unit 2 reference cycle fuel rods touch the bottom nozzle.
The results of the FATS test analysis show that the. grids can be treated as having identical resistance and that the effects on pressure drop of the differences between the fuel rods cn and off the bottom nozzle are negligible.
Also, fuel assembly liftoff tests were run which confirmed calculated values.
(See Section 2.3 for further discussion on the fuel assembly lif toff.)
The similarities in dimensions and blockage area and the test results, showing insignificant differences in resistances, indicate that the Westinghouse and Millstone, Unit 2 reference cycle fuel assemblies can be treated as being hydraulically identical. This hydraulic compatibility is assumed in the BSR and the staff finds it acceptable.
3.3 Performance and Safety Criteria The performance and safety criteria of the Millstone, Unit 2 core design as stated in Section 3.2 of the BSR are:
" Fuel damage (defined as penetration of the fission product barrier, i.e.,
1.
the fuel rod clad) is not expected during normal operation and operational transients (Condition I) or any transient conditions arising from faults of moderate frequency (Condition II).
It is not possible, however, to These will be within the preclude a very small number of rod failures.
capability of the plant cleanup system and are consistent with the plant design bases.",
e
2.
"The reactor can be brought to a safe state following a Condition III event with only a small fraction of fuel rods damaged (see above definition) although sufficient fuel damage might occur to preclude immediate resumption of operation."
3.
"The reactor can be brought to a safe state and the core can be kept subtritical with acceptable heat transfer geometry following transients arising from Condition IV events."
3.4 Design Bases The performance and safety cr.iteria' listed above are implemented through the following bases.
3.4.1 Departure from Nucleate Boiling The margin to departure from nucleate boiling at any point in the core is expressed in terms of the departure from nucleate boiling ratio (DNBR).
The DNBR is defined as the ratio of the heat flux required to produce departure from nucleate boiling at the calculated local coolant conditions to the actual local heat flux.
The thermal-hydraulic design basis, as stated in Section 3.2 of the BSR l
is:
"There will be a 95 percent probability that departure from nucleate l
boiling (DNB) will not occur on the limiting fuel rods during normal operation and operational transients and any transient arising from faults of moderate frequency (Condition I and II events) at a 95 percent confidence level."
3.4.2 Core Flow Section 3.2 of the BSR has the following core flow design basis:
"A minimum of 96.3 percent of the primary coolant flow will pass through i
the fuel rod region of the core and be effective for fuel rod coolino."
l t.
Coolant flow through the thimble tubes as well as leakage from the core barrel-baffle region into the core are not considered effective' for heat' removal.
Core cooling evaluations are based on the design flow rate (minimum flow) entering the reactor vessel.
A maximum of 3.7 percent of this value is bypass flow (external leakage).
3.4.2 Other Considerations As given in the fuel rod design bases (Section 2.1 of the BSR) the peak linear heat rate must be below that which will cause incipient centerline melting (i.e., 4700 F).
Also, the average coolant temperature at the exit of the core must be less than the saturation temperature to assure meaningful thermal power measurements.
3.5 Thermal-hydraulic Design Methodology 3.5.1 Departure From Nucleate Boiling The thermal-hydraulic analysis uses the W-3 Critical Heat Flux (CHF) correlation (Ref. 53) in conjunction with the THINC codes (Refs. 54 and 55).
The W-3 CHF correlation was previously used in the Millstone, Unit 2 FSAR (Ref. 56) analysis.
The THINC codes are three-dimensional matrix models which account for hydraulic and nuclear effects on the enthalpy rise in the core. The behavior of the hot assembly is determined by superimposing the power distribution among the assemblies'.upon the inlet flow distr'ibution while allowing for flow mixing and flow distribution between assemblies. The average flow and enthalpy in the hottest assembly is obtained from the core-wide, assembly-by-assembly analysis. The local variations in power, fuel rod and pellet fabrication, (engineering hot channel factors)
I within the hottest assembly are then s::perimposed on the average conditi.ons of the hottest assembly in order to determine the conditions in the hot i
channel.
i l l
i The THINC codes are used to calculate the flow, enthalpy, pressure and DNBR distribution in the core for all expected operating conditions as discussed in Reference 52.
A transient version of the THINC I code is used for transient DNB analyses (e.g., Loss of Flow) as described in Reference 1.
Verification of the THINC codes is contained in the above references. Application to the Millstone, Unit 2 core is made by modeling the particular geometry of the core. A comparison of the core design parameters for DNB analysis for Westinghouse and Millstone, Unit 2 reference cycle is given in Table 2.
Based on the thermal hydraulic similarity of the Westinghouse and Millstone, Unit 2 reference cycle fuel assemblies and on finding that the CHF correlation and thermal hydraulic computer code used by the applicant has been previously approved by the staff, we conclude that the DNBR design methodology used in the Millstone, Unit 2 BSR is acceptable.
3.5.2 Core Flow 1he core flow design basis requires that the minimum flow which will pass through the fuel rod region and be effective for fuel rod cooling is 96.3 percent of the primary coolant flow rate. The amount of bypass flow (external leakage) is determined by a series of hydraulic resistance calculations on the core and vessel internals and verified by model flow tests.
Since the amount of bypass is consistent with the Millstone, Unit 2 FSAR and with approved plants of similar design, the staff concludes that the core flow given in the Millstone, Unit 2 BSR, 96.3 percent of the primary coolant flow rate, is acceptable.
3.6 Hot Channel Factors The total hot channel factors for heat flux and enthalpy rise are defined as j
the maximum-to-core average ratios of these quantities. The heat flux hot channel factor considers the local maximum linear heat generation rate at a point (the hot spot), and the enthalpy rise hot channel factor involves the maximum integrated value along a channel (the hot channel).
l l
TABLE 2 CORE DESIGN PARAMETERS FOR DNB ANALYSIS
. Millstone, Unit 2 Reload Reference Cycle 102 percent Core Power (MWt) 2754 2754 i
System pressure (psia) 2200 2200 Inlet temperature ( F) 551 551 Vessel flow rate, (gpm) 370000 370000
- These parameters include uncertainties.
0 w
.9
Each of the total hot channel factors considers a nuclear hot channel factor describing the neutron power distribution and an engineering hot channel factor, which allows for variations'in flow conditions and fabrication tolerances.
3.6.1 Heat Flux Engineering Hot Channel Factor The heat flux engineering hot channel factor is used to evaluate the maximum heat flux.
This subfactor is determined by statistically combining the tolerances for the fuel pellet diameter, density, enrichment, and the fuel rod diameter, and has a value of 1.03, based on the manufacturing variations expected for Westinghouse fuel in the Millstone, Unit 2 core.
3.6.2 Enthalpy Rise Engineering Hot Channel Factor The effect of variations in flow conditions and fabrication tolerances on the hot channel enthalpy rise is directly considered in the THINC code thermal subchannel analysis under a'ny reactor operating condition.
The items considered contributing the enthalpy rise engineering hot channel factor are discussed below:
1.
Pellet diameter, density, and enrichment; and fuel rod diameter:
Design values employed in 'the THINC analysis related to the above fabrication variations are based on applicable limiting tolerances such that these design values are met for 95 percent of the limiting channels at a 95 percent confidence level. The tolerances used in this evaluation are based on variations expected to be conservative for Westinghouse fuel in the Millstone, Unit 2 core. The effect of variations in pellet diameter and enrichment is employed in the THINC analysis as a direct multiplier on the hot channel enthalpy rise.
~
-24.
~
2.
Inlet Flow Maldistribution:
The effect of in1.et flow maldistribution,in core thermal performances is considered by using a design basi.s of 5 percent reduction in coolant flow (Ref. 2) to the hot assembly in the THINC analysis.
3.
Flow Redistribution:
The flow redistribution accounts for the reduction in flow in the hot channel resulting from the high flow resistance in the channel due to the local or bulk boiling.
The effect of the nonuniform power distribution is inherentif considered in the THINC analysis for every operating condition which is evaluated.
4.
Flow. Mixing:
The subchannel mixing mode incorporated in the THIN' code and used in reactor design is based on tests conducted with spacer grids (no mixing vanes) as described in Reference 55.
A conservative value of the thennal diffusion coefficient of 0.019, determined from these tests, is used in the Millstone, Unit 2 THINC analysis.
The staff has by review and comparison with previourly approved designs found the application of heat' flux engineering hot channel factor and enthalpy rise hot channel factor acceptable.
3.7 Fuel Rod Bowing As discussed in Section 2.3, the fuel rod bow analysis is' acceptable and no fuel r'od bowing penalty is requirea for Millstone, Unit 2.
'3. 8 Thermal Hydraulic Stability Because of the hydraulic compatibility of the Westinghouse fuel assemblies to the Millstone, Unit 2 fuel assemblies, the thermal-hydraulic stability of the core will behave as stated in the Millstone, Unit 2 FSAR and is acceptable..
O
3;9 Crud Buildup Crud buildup is a strong function of the purity of the reactor coolant.
Because of the hydraulic compatibility of the Westinghouse fuel assemblies to the Millstone, Unit 2 fuel assemblies and because there are no other changes in the reactor coolant system, the considerations regarding crud buildup will be unchanged by the inclusion ot Westinghouse fuel in Millstone, Unit 2.
3.10 Uncertainties in Measured Parameter _s_
As indicated in Table 1 there are uncertainties in measured parameters.
In Millstone, Unit 2, Cycle 4, we requested information relative to measurement uncertainties.
This information has not been provided in the Millstone BSR and, therefore, the effects of measurement uncertainties will need to be addressed and accounted for in the reload submittals or Technical Specifications.
3.11 Evaluation Conclusions We conclude that the thermal-hydraulic design of the Westinghouse reload fuel for Millstone, Unit 2 meets the requirements of General Design Criterion 10, 10 CFR Part 50 and is acceptable for final design approval.
We also conclude that the Westinghouse reload fuel for Millstone, Unit 2 has been designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during steady-state operation or anticipated operational occurrences.
This conclusion is based on the applicant's analyses of the core thermal-hydraulic perfonnance which was reviewed by the staff and found to be acceptable.
Consequently, the BSR'may be utilized as an acceptable reference in NNECo licensing submittals that address the operation of Millstone, Unit 2.
The following constitutes the only limitation on our approval-1.
Information on measurement uncertainties has not been provided in _
the Millstone, Unit 2 BSR. Therefore, the effects of measurement uncertainties will need to be addressed and accounted for in the reload submittals or Technical Specifications.
4 REFERENCES 1.
Letter from W. Counsil (NNECo) to R. Reid (NRC), March 6, 1980.
2.
" Basic Safety Report," Westinghouse proprietary report for Millstone, Unit 2, Docket Number 50-336.
3.
Government Printing Office, " Reactor Design," Criterion 10, Appendix A, " General Design Criteria for Nuclear Power Plants," Part 50, Title 10 Energy, Code of Federal Regulations.'
4.
Ibid., " Reactor Site Criteria," Part 100.
5.
Ibid., " Combined Reactivity Control Systems Capability," Criterion 27, Appendix A, Part 50.
6.
Ibid., " Emergency Core Cooling," Criterion 35, Appendix A, Part 50.
7.
Ibid., " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," Part 50.46.
8.
"RESAR-414, Reference Safety Analysis Report for the Westinghouse 3820 Mwt NSSS," Docket Number 50-572, October 8,1976.
9.
" Preliminary Design of the Standard Reference System RESAR-414, Westinghouse Electric' Corporation.
Docket No. STN 50-572," NRC report NUREG-0491, ' November, 1978.
l 10.
Letter from R. S. Boyd (NRC) to T. M. Anderson (Westinghouse), Docket Number 50-572,
Subject:
Preliminary Design Approval: RESAR-414, November 14, 1978.
11.
" Standard Review Plan for the Review of Safety Analysis Reports for' Nuclear Power Plants-LWR Edition," NRC report NUREG-0800, Section 4.2,
" Fuel System Design," Rev. 2, July 1981.
0
12.
R. A. George, et al., " Revised Clad Flattening Model, Westinghouse report WCAP-8377, July 1974.
13.
Memorandum from V. Stello (NRC) to R. DeYoung, " Evaluation of Westinghouse Report WCAP-8377, Revised Clad Flattening Model,"
January 14, 1975.
14.
J. Skaritka, et al., " Fuel Rod Bow Evaluation," Westinghouse report WCAP-8691, Rev. 1. July 1979.
15.
Memorandum from R. O. Meyer (NRC) to D. F. Ross, " Revised Coefficients for Interim Rod Bowing Analysis," March 2, 1978, 16.
Letter from J. F. Stolz (NRC) to T. M. Anderson (Westinghouse),
Subject:
Staff Review of WCAP-8691, April 5, 1979.
17.
Memorandum from D. F. Ross and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller, " Interim Safety Evaluation Report on the Effects of Fuel-Rod Bowing in Thermal Calculations for Light Water Reactors,"-December 8, 1976.
- 18. Memorandum from D. F. Ross and D. G. Eisenhut (NRC) to D. B. Vassallo and K. R. Goller, " Revised Ir-y Evaluation Report on the Effects of Fuel Rod Bowing in Thermal Calculations for Light Water Reactors,"
February 16, 1977.
l I
19.
" Cracks in Control Element Assembly Guide Tubes," preliminary notification of bvent or unusual occurrence PN0-77-221, Docket Number 50-336, December 14, 1978.
20.
Letter from A. E. Scherer (Combustion Engineering) to V. Stello (NRC),
Number LD-77-122, December 23, 1977.
2L Letter from W. P. Johnson (Maine Yankee Atomic Power) to V. Stello (NRC),
February 14, 1978.
22.
Letter from A. E. Lundvall (Baltimore Gas and Electric) to V. Stello (NRC), February 17, 1978.
23.
Letter from R. A. Clark (NRC) to W. G. Council (NNECo),
Subject:
Amendment No. 61 to Facility Operating License No. DPR-65 for Millstone Nuclear Power Station, Unit 2, October 6, 1980.
24.
Letter from R. A. Clark (NRC) to A. E. Lundwall, Jr. (Baltimore Gas and Electric),
Subject:
Amendment No. 48 to Facility Operating License No. DPR-53 for Calvert Cliffs, Unit 1 December 12, 1980.
25.
Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr. (Baltimore Gas and Electric),
Subject:
Amendment No. 31 to Facility Operating License No. DPR-69 for Calvert Cliffs, Unit 2, February 10, 1981.
26.
" Safety Evaluation Report Related to the Operation of Arkansas Nuclear One, Unit'2, Section 4.2, NRC report.NUREG-0308, Supplement 2, September l
1978.
l l
i
, l 9
27.
Letter from W. G. Council (NNECo) to R. Reid (NRC),
Subject:
Sleeved CEA Guide Tube Inspection Program, March 29, 1979.
28.
Letter from W. G. Council (NNEco) to R. Reid (NRC),
Subject:
CEA Guide Tube Inspection Program. April 17, 1979.
29.
Letter from W. G. Council (NNECo) to R. Reid (NRC),
Subject:
Resolution of Cycle 3 Startup Commitments, April 15, 1980.
30.
Letter from W. F. Lee and W. G. Council (NNECo) to R. A. Clark (NRC),
Subject:
Resolution of Cycle 3 Startup Commitments, August 14, 1980.
31.
" Improved Analytical Methods Used in Westinghouse Fuel Rod Design Computations," Westinghouse report WCAP-8720, October 1976.
32.
Letter from J. F. Stolz (NRC) to T. M. Anderson (Westinghouse),
February 9, 1979.
33.
" Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations - Application for Transient Analysis," Westinghouse report WCAP-8720, Addendum 1 September 1979.
34.
Letter from T. M. Anderson (Westinghouse) to J. R. Miller (NRC),
Subject:
Responses to " Request for Additional Information on Addendum 1 to WCAP-8720," Number NS-TMA-2303.
35.
Letter from V. Stello (NRC) to all PWR licenses, January 25, 1978.
~
e.
36.
" Task Action Plan for Generic Activities, Category A " NRC report NUREG-0371, November 1978.
37.
Letter from B. Grimes (NRC) to all PWR operating reactor licenses, January 16, 1979.
38.
Letter from W. G. Council (NNECo) to R. A. Clark (NRC). April 1, 1981.
39.
Letter from W. G. Council (NNECo) to R. A. Clark (NRC), June 8, 1981.
40.
LetterfromW.G. Council (NNECo) tor.A.Chark(NRC), June 11, 1980.
41.
L. T. Gesinski and D. Chiang, " Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident,"
Westinghouse report WCAP-8??6. December 1973.
42.
Letter from J. F. Stolz (NRC) to T. M. Anderson (Westinghouse),
February 6, 1979.
43.
"MATPRO-Version 11 (Revision 10: A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior," NRC Report NUREG/CR-0497, Rev.1. Feb aary 1980.
- 44. Telecon from M. Cass (NNECo) to M. Conner (NRC)
December 4,1981.
45.
" Millstone Nuclear Plant Station, Unit 2 Final Safety Analysis Report,"
NNECo report, page 3.2-2, Docket Number 50-336.
46.
Letter from R. Salvatori (Westinghouse) to D. Knuth (NRC),
Subject:
Core Coolant and Rod Surface Temperature, Number NS-SL-521, Attachment P.
January 4,1973.
47.
Letter from T. M. Anderson (Westinghouse) to J. R. Miller (NRC), Number NS-TMA-2436, April 21, 1981.
l 48.
" Standard Specifications for Sintered Uranium ~ Dioxide Pellets, " ASTM Standard C776-76, Part 45 (1977). '
1
O e
49.
" Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," NRC report Regulatory Guide 1.77.
50.
" Rules for Construction of Nuclear Power Plant Components,"Section III, ASME Boiler and Pressure Vessel Code (1977).
51.
" Benchmark Problem Solutions Employed for Verification of WECAN Computer Program," Westinghouse report WCAP-8920. April 1977.
52.
F. M. Bordelon, et al., " Westinghouse Reload Safety Evaluation l
Methodology," Westinghouse report WCAP-9272, March 1978.
l 53.
L. S. Tong, " Boiling Crisis and Critical Heat Flux " Atomic Energy Commission report, 1972.
54.
H. Chelemer, J. Weisman, and L. S. Tong, "Subchannel Thermal Analysis of Rod Bundle Cores," Westinghouse, report WCAP-7015, Rev. 1, January 1969.
55.
J. Shefcheck, " Application of the THINC Program to PWR Design,"
Westinghouse report WCAP-7838, January 1972.
56.
" Millstone Nuclear Plant Station, Unit 2 Final Safety Analysis Report," NNECo report, page 3.2-9, Docket Number 50-336.
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