ML17223A262
| ML17223A262 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 08/09/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17223A261 | List: |
| References | |
| REF-GTECI-A-02, REF-GTECI-B-06, REF-GTECI-PI, REF-GTECI-RV, TASK-A-02, TASK-A-2, TASK-B-06, TASK-B-6, TASK-OR NUDOCS 8908160134 | |
| Download: ML17223A262 (35) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REPORT ON ASYMMETRIC LOCA LOADS ST.
LUGIE UNIT 1 FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-335
- 1. 0 INTRODUCTION On May 7, 1975, the Nuclear Regulatory Commission (NRC) was informed that asymmetric loading on the reactor vessel supports resulting from a postulated reactor coolant pipe rupture at a specific location (e.g., the vessel nozzle) had not been considered in the original design of the reactor vessel support for North Anna Units 1 and 2.
It had been identified that in the event of a postulated, instantaneous double-ended offset shear pipe break at the vessel nozzle, asyIIIIIetric loading could result from forces induced on the reactor internals by transient differential pressures across the core barrel and by forces on the vessel due to transient differential pressures in the reactor cavity.
With the advent of more sophisticated computer codes and the develop-ment of more detailed analytical models, it became apparent that such differen-tial pressures, although of short duration, could place a significant load on the reactor vessel supports and other components, thereby possibly affecting their integrity.
Although this potential safety concern was first identified during the review of North Anna facilities, it was determined to have generic imp 1 icati ons for a 1 1 pressurized water reactors.
In October of 1975, the NRC staff notified each operating pressurized water reactor (PWR) licensee of a potential safety problem concerning the design of their reactor pressure vessel support system.
From this survey it was discovered that these asymmetric loads had not been considered in the design of any PWR primary system.
In June 1976, the NRC requested all operating PWR licensees to evaluate the adequacy of reactor system components and supports at their facilities, with respect to these newly identified loads.
Licensee and vendor responses to this request were proposals to augment inservice inspection and/or probability studies that supported no analyses due to the low probability of the pipe breaks at a particular location.
Although the NRC recognized some merit in these proposals, they determined that. the more fundaIIIental questions still remained unanswered.
Therefore, licensees of PWR plants were notified by letter dated January 20, 1978 that the evaluation of their primary systems for asymmetric LOCA loads would be required.
Although the NRC staff's original emphasis and concerns were focused primarily on the integrity of the reactor vessel cavity, it became apparent that SOOOSS5 PDR ADOCK 0500 P'
significant asymmetric forces could also be generated by postulated pipe breaks outside the cavity and that the scope of the problems was not limited to the vessel support system itself.
The. staff, after reviewing this problem, deter-mined that a reevaluation of the primary system integrity of all PWR plants to withstand these loads was necessary.
Therefore, in January 1978, the NRC staff requested each PWR licensee to submit additional information in accordance with the expanded scope of the problem.
Those letters outlined the present scope of the problem specifying a minimum number of pipe break locations to be addressed and the reactor system components to be evaluated.
The asyrmetric loading on the primary system that was determined by NRC to have generic implications was formerly identified in Task Action Plan A-2, Unresolved Safety Issue (USI), "Asymetric Blowdown Loads on Reactor Primary Coolant System as published in NUREG-0371, Task Action Plans for Generic Activities (Category A],
- USNRC, November 1978.
Since the identification of the asyometric load problem in May 1975, EG8G Idaho, Inc. has performed a
number of independent audit analy-ses to verify licensee submittals on this problem.
A total of six analyses have been completed (one linear elastic and one non-linear-inelastic analysis of reactor coolant loop (RCL) for each of the three major reactor vendors).
Based on these analyses and additional NRC staff investigations, criteria and guidance for conducting an evaluation of asymetric loss of coolant accident (LOCA) loads were developed.
USI A-2 was resolved in January 1981 with the publication of NUREG-0609, "Asymmetric Blowdown Loads on PWR Primary Systems."
This document provided an acceptable basis for performing and reviewing plant analyses for asymnetric LOCA loads and affected all operating and future PWRs.
During the course of the work on USI A-2, it was demonstrated that there were only a very limited number of break locations which could give rise to significant loads.
Subsequently, after the development of substantial new technical work, it was demonstrated that the new techniques for the analysis of piping failure assured adequate protection against failures in primary system piping in PWRs.
This was reflected in a revision of General Design Criterion-4 (GDC-4) published in the Federal Re ister in final form on April ll, 1986, and in a further revision to GDC-~polis e
1n the Federal Re ister on July 23, 1986.
In addition, it has also been satisfactorily demonstrate 1n the course of the A-2 effort that there is a very low likelihood of simultaneous pipe loading with both LOCA and safe shutdown earthquake (SSE) loads.
For Combustion Engineering plants of the pre-CESSAR vintage without the SSE-LOCA load combination, the loads on primary system piping would not result in pipe breaks which could lead to significant loads on the core structure.
Accordingly, for these facilities the staff had concluded that the potential for asymmetric loading on the core structure resulting from primary system piping LOCA need not be considered in the dgsign of the core structure.
In June
- 1980, Combustion Engineering (CE), on behalf of the Baltimore Gas and Electric Company, a merrber of the CE Owners Group, submitted a final asymetric LOCA loads evaluation report (Reference 1), applicable to the St. Lucie Unit 1 power plant.
The material submitted in response to the January 1978 letter from NRC was reviewed by the NRC staff and its consultants.
Upon review of the submittal, it was determined that additional information was required to satisfy the established guidelines and acceptance criteria.
On February 23, 1981, the NRC staff notified CE of the additional requests, and the response (Reference
- 2) was submitted in June 1981.
In addition, the licensee submitted a report entitled "Reactor Coolant System Asyometric LOCA Load Evaluation," in July 1981 specifically for the St. Lucie Unit I plant.
This submittal and supplement represent the limiting cases for the asymnetric LOCA loads evaluation and have been reviewed in conjunction with the criteria outlined in NUREG-0609.
Subsequent sections of this Safety Evaluation summarize the evaluations performed by the licensee for subcooled blowdown loads, cavity pressurization, and struc-tural response.
The staff's evaluation includes the assessment of the licensee's compliance with acceptance criteria.
2.0 DISCUSSION The licensee's analysis procedure, including analytical
- models, computer methods and analytical results, are discussed in the following paragraphs.
The analytical methodology primarily consists of development of (a) thermal hydraulic loads for the reactor coolant system (RCS) structural analysis (b) calculation of the steam generator and reactor cavity pressures, and (cI calculation of'the loads and stresses on various components and supports of the RCS which include the vessel and steam generator
- supports, vessel internals, fuel asseahlies, control element drive mechanisms (CEDMs) and emergency core cooling system (ECCS) piping.
2.1 Thermal H draulic Loads Anal sis The model used to determine the pressure field at every point in the primary system following the postulated primary system breaks, from which the internal asymmetric forces on the vessel and core support barrel are deduced, is shown in Figure 1.
This model is used for the vessel
Reduced models of the reactor internals and the reactor coolant system ate provided in Figures la and Ib, respectively.
A reduced internals model of the generic plan't is depicted in Figure 2.
The RELAP-4 thermal hydraulic code (Reference
- 3) is used to compute the thermo-dynamic properties in the model volumes and junctions.
Results of the RELAP-4 model have been compared to results achieved by modeling the system with WHAM-6 (Reference 4).
The results of the two models are in good agreement, with RELAP-4 predicting a larger pressure differential across the core support barrel.
Results of the internal asymmetric loads analysis indicate that the peak forces across the core support barrel and the vessel are virtually insensitive to break opening times.
For instance, a change in area from I square foot (with an open-ing time of 8 msec.) to 9.81 square feet full break (opening time of 36 msec.),
only results in a 2 tg-3 percent increase in peak internal asyametric
- loads, whereas a decrease in opening time from 36 msec. to I msec. for the full break brings about a threefold increase in the internal asyranetric load.
The pressure fields, flow rates and density distribution in the case of the hot leg breaks and the full size (1414 square inches) cold leg break, used for the assessment of the internals and fuel, have been computed with the CEFLASH-4B computer code according to the methods documented in Reference 5.
The results of the thermal-hydraulic analysis provide the time history, forcing functions applied to the reactor vessel and internals in the reactor coolant system (RCS) structural analysis.
2.2 Cavit Pressurization Anal sis The reactor cavity pressurization analysis for St. Lucie Unit 1-for stipulated LOCA conditions, including 1414 square inches cold leg guillotine break, was provided in the St. Lucie Unit 1 Final Safety Analysis Report (FSAR) and approved by the NRC during the course of the operating license review.
The results for the controlling 1414 square inches cold leg guillotine break, as reported FSAR Amendment 44, have been directly used for the assessment of the adequacy of the vessel
The internal asymmetric loads on the fuel and reactor internals were determined by using the results obtained for a full size break in the inlet line of a generic plant (Calvert Cliffs) representing the St. Lucie 1
plant.
The FSAR analysis overestimates the resultant parameters for this break since the cavity response had been predicated on a break opening time of 10 msec.;
whereas 23 msec.
is needed to achieve this size break.
Mass and energy release rates in the generic plant were calculated using the modified CEFLASH-4 computer code based on four design basis postulated pipe ruptures which were determined to be the most limiting breaks.
1.
2.0A (or 1414 square inches) break in 0.023 sec at the reactor vessel inlet nozzle.
2.
0.01A (or 135 square inches) break in 0.020 sec at the reactor vessel outlet nozzle.
3.
0.07A (or 1000 square inches) break in 0.024 sec at the steam generator inlet nozzle.
4.
2.0A (or 1414 square inches) break in 0.020 sec at the steam generator outlet nozzle.
Calculation of reactor pressure vessel (RPV) cavity pressures was performed using the RELAP-4-MOD6 computer code.
Pressures were determined with the mass and energy releases from the design breaks at the reactor vessel inlet and outlet nozz les.
The reactor cavity has a net free volume of about 3000 cubic feet.
The input model contains 36 volume-nodes, determined from sensitivity studies to be detailed enough to provide a convergent solution.
The 2.0A double-ended cold leg break results are provided in Figures 3 and 4.
2.3 Structural Anal sis The licensee's structural analysis was performed utilizing two primary finite element models and several component and support detailed finite element models.
The subsystem models were used to develop input to the primary models and to calculate component and support loads and stresses for detailed evaluations.
The mathematical models to which asyrrroetric LOCA loads were applied are described in the following subsections.
2.3.1 Reactor Coolant S stem Anal sis A non-linear elastic time history dynamic analysis of a three-dimensional mathematical model of the reactor coolant system, including details of the reactor internals, pressure
- vessel, supports, and piping, was performed for the postulated pipe break to provide reactor vessel support reaction forces.
This model has 981 total static degrees of freedom and 77 mass degrees of freedom.
The reactor vessel and all internal components are modeled at internal and support interfaces.
The STRUDL (Reference
- 6) computer code generates the condensed stiffness matrix used in the dynamic analysis from the physical definition of the structure.
Qdro-dynamic effects, including both virtual mass and annular effects, are accounted for in the coupling between the reactor pressure vessel (RPV) and the core support barrel (CSB) and between the CSB and the core shroud.
The hydrodynamic (added) mass matrix is evaluated using the ADMAS computer code (Reference 7).
The dynamic analysis to determine the system response was performed with the computer codes QAGS (Reference
- 8) and QFORCE (Reference 9).
2.3.2 Primar Shield Wall Anal sis The ability of the primary shield wall to sustain the worst-case pipe rupture loads was determined from a linear elastic, three-dimensional model of the wall using the NASTRAN computer code (Reference 10).
Loads consisted of reactor vessel support reaction loads and reactor cavity pressurization loads.
2.3.3 Subs stems Anal sis Numerous smaller, more detailed mathematical models were used in the LOCA analysis to provide representative and meaningful responses to the applied loadings.
One model was developed to determine the component support stiffness to be used in the system analyses, as well as to qualify the supports once system responses were obtained.
The analysis of some of the subsystems is discussed in the following paragraphs.
1.
The reactor vessel internals were evaluated for guillotine pipe breaks at the reactor vessel sine~
anB outlet nozzles, employing lateral and axial mathematical models.
Nonlinear analyses were performed in accordance with established procedures (Reference
- 7) using beam elements, gap elements, and linear and nonlinear spring elements.
Hydrodynamic coupling effects were included in the horizontal model.
Both models were subject to a
.combination of applied forces and excitations.
The time history forces applied to vessel internals resulted from the LOCA blowdown analysis, and the time history motions of the reactor vessel resulted from the RCS analysis.
The lateral and axial models are shown in Figures 5 and 6, respectively.
Results of the analyses were time dependent member loads and nodal displacements, velocities, and accelerations.
In addition to the horizontal and vertical responses of the vessel internals, vibration and stability analyses were performed on the CSB to determine possible contributing barrel stresses.
The shell mode response of the barrel due to LOCA pressure loads applied to the barrel from the break at the RPV inlet nozzle was analyzed.
Axial type loadings on the CSB from a pipe break at the RPV outlet nozzle were investigated with the aid of the SAHHSOR/-DYNASOR computer code (References 14 and 15) and the buckling potential of the barrel was determined.
2.
The fuel was analyzed with the CESHOCK computer code (Reference 12).
Disp~acement time histories of the fuel alignment plate, core shroud, and core support plate from the internals analysis provided the input motions for the fuel computer model.
Stability in the fuel assemblies" was determined from a dynamic beam-column analysis using the finite element code ANSYS (Reference 16).
The model was subject to concurrent lateral and axial LOCA loadings.
Effects of adjacent structures were included in the modeling.
3.
The CEDHs were evaluated with an elastic-plastic finite element mode~ime history motions of the reactor vessel
- head, determined from the RCS analysis, were applied to the base of the CEDHs.
The controlling section of the component is the CEDH nozzle, near the interface with the RPV head.
4.
The integrity of the ECCS i in was evaluated for asymnetric LOCA loadings by performing an e ast~c analysis using the STRUDL and DAGS computer codes (References 6 and 8).
Input excitation to the analysis was provided by the time history motions of the ECCS nozzle, resulting from the system LOCA response.
The motions were directly computed at the appropriate location on the reactor coolant pump (RCP) discharge leg.
2.4 Summar of Licensee's Anal tical Results The basic criteria for acceptability of the plant for the postulated faulted condition is to provide high assurance that the reactor can be brought safely to a cold shutdown condition.
The licensee concluded that overall acceptability of the plant for the postulated LOCA was met.
This was demonstrated by the following component and structure evaluations believed by the licensee to be the worst or limiting cases.
A summary of load and stress results from the LOCA analyses is presented in Table 1.
A comparison of peak calculated and design loads at representative locations is provided in Table 2.
2.4.1 Reactor Vessel Su ort The loads calculated for each reactor vessel support are reported in Table 3 for the limiting break, i.e., the 1414 square inches cold leg break at the inlet nozzle, for a range of reactor vessel support stiffnesses.
This range covers the possible values of the overall stiffness of the individual reactor vessel
- supports, the real value being somewhere between the two extremes.
The wide variation in the embedded steel makes the determination of a more precise value very difficult.
However, in the support analyses the higher loads resulting from the use of the highest stiffness have been utilized.
This insures that the absolute maximum load per support is computed.
In reality, lower values are expected.
The capability of the reactor vessel supports is provided in Figure 7 and Table 4.
Since the capabi lity of the supports exceed the maximum loads computed for the given break, it is concluded that the existing support system is adequate for that break.
2.4.2 Steam Generator Su orts Results of the analyses of the loads imposed on the steam generator supports from both hot and cold leg breaks in the system in combination with seismic loads indicate that, with exception of the loads on the four hold down bolts at the vessel end of the steam generator sliding base, none of the design loads have been exceeded.
The computed and design loads are shown in Table 5.
However, individual examination of the sliding base, the bolts, and bolt anchorages indicates that all can acceptably withstand the applied loads.
Therefore, it is concluded by the licensee that the existing steam generator (SG) supports design is adequate.
2.4.3 Reactor Coolant Pi in The primary coolant piping was expected to be most highly stressed at component nozzles.
Considering the four design basis pipe breaks, resu ltant loads on the RPV nozzles, SG nozzles, and RCP nozzles were determined from the RCS analysis.
Table 6 provides the elastically calculated pipe rupture and seismic loads on intact reactor coolant piping associated with the broken loop for the worst break.
Examination of this table reveals that all loads fall within the allowable loads with the exception of the load at the RCP discharge
- nozzle, which exceed the allowable by about 13 percent, on an elastic basis.
An elasto-plastic
- analysis, however, indicates that the functional capability of the discharge nozzle cari be maintained.
2.4.4 Reactor Internals The core support structure components are analyzed for the loads resulting from a
LOCA, both inlet or outlet break, in combination with the mechanical loads associated with normal operating conditions.
The calculated stresses are combined to determine the maximum stress intensity which is compared with the membrane allowable of 2.4S or the membrane plus bending allowable of 3.5S, as defined in the ASNE Boiler and Pressure Vessel Code,Section III, Appendix F.
The elastic material properties and stress allowables are conservatively taken at the reactor internals design temperature of 650'F.
The vertical loads, from the CESHOCK Code (Reference
- 12) derived for the generic plant, were used for the St. Lucie plant-specific stress analysis.
The stresses from the vertical 1oads were combined with the stresses from the horizontal shears and moments to obtain the stress intensity for the reactor internals.
Additional dynamic analyses were performed to compute the horizontal core support structure member loads using different values of friction resulting in different values of constraint at the CSB flange to reactor vessel ledge interface.
Resu,its of the analyses indicate that the structural integrity of the reactor internals is maintained.
2.4.5 Fuel Assemblies Calculations have demonstrated acceptable performance for all end fittings, hold down springs, and fuel rods in the generic plant representing St. Lucie Unit 1.
A limited number of spacer grids in the plant have calculated impact loads in excess of the allowable values in a few peripheral core locations for the inlet break condition.
The number of locations at which this condition exists is reduced as a function of time if credit is taken for spacer grid irradiated mechanical properties.
By taking into consideration the plant-specific differences between the generic and St. Lucie Unit I plants, the licensee has demonstrated acceptable performance of the fuel asseahlies.
2.4.6 Control Element Drive Mechanism The control element assemblies are not required to operate from safe shutdown after a loss of coolant event resulting from breaks which are larger than 0.5 square feet.
However, in order to comply with existing ECCS ana1yses
- methods, the integrity of the elements must be maintained and leakage must be prevented.
The capability of the elements to withstand the effects of the controlling pipe break have been evaluated by determining the response of an individual CEDM to the motion of the reactor vessel head.
Elastic analyses of a cantilevered beam model indicated that elastic level ASME limits would be exceeded for the input motion.
Therefore, an elasto-plastic analysis was performed.
The displacements of the reactor vessel head computed by the system structural analysis performed specifically for St. Lucie were applied to the base of the CEDM.
The results of the elasto-plastic analyses satisfy the 70K plastic instability load criterion of ASME Code, Appendix F, thus assuring the integrity of the pressure boundary.
2.4.7 ECCS and Connected Pi in The stresses in the piping are within 10 percent of the allowable stress, and hence it is concluded that the ECCS piping and other piping connected to the primary loop will maintain functional capability during and after the postulated event.
Table 7 compares the peak computed stresses which, in this case, include normal and seismic loads to the allowable stresses.
The margin existing between peak stresses calculated on an elastic basis and stresses that would be allowed with an elasto-plastic analysis further indicates that this attached piping would be able to withstand the imposed loads resulting from the controlling guillotine break.
2.4.8 Primar Shield Wall The reactor cavity wall was evaluated for cavity pressurization loadings and vessel support reaction loads resulting from pipe breaks at the vessel inlet and outlet nozzles.
Worst-case reaction loads were applied to each support, and peak cavity loads for each model element were applied as a continuous static pressure.
The results indicate that the integrity of the wall is maintained for all loading conditions.
3.0 STAFF EVALUATION The licensee's calculation procedures, including analytical models, computer
- methods, and acceptance criteria for the assessment of the asymmetric LOCA loads problem, have been evaluated by the staff.
The staff evaluation was accomplished by reviewing the licensee's submittal and using the independent audit calculations performed by the staff or their consultants.
In general, the staff has concluded that the licensee's assessment of the problem is acceptable.
The staff evaluation of each specific ana1ysis phase is addressed in subsequent paragraphs, following the guidelines set forth by NUREG-0609.
3.1 Thermal H draulic Blowdown Loads The thermal hydraulic blowdown calculation portion of the St. Lucie Unit 1 asymmetric LOCA load submittal has been reviewed and is considered to be acceptable to the staff.
The basis of this acceptance is the staff's review and approval of the CEFLASH-4B computer code used for the internal hydraulic loads calculations.
Independent audit calculations for CE 2570 MW plant by the staff's consultant aided in approval of the CEFLASH-4B application to subcooled blowdown.
The code does not consider fluid-structure interaction, and the structural boundaries are assumed rigid and at rest.
Such conditions normally give rise to conservative pressures and loads.
A significant number and location of postulated pipe breaks were analyzed to determine worst-case loadings on the primary coolant system.
Size and length of break openings consisted of reasonable and realistic values.
Nodalization and modeling were also developed in a manner that provided reasonable representation of the existing system.
3.2 Cavit Pressurization Anal sis The licensee's reactor cavity pressurization analysis of St. Lucie Unit 1 for postulated breaks at the reactor vessel inlet and outlet nozzles has been reviewed and is considered to be acceptable by the staff.
The basis of this acceptance is the staff's review and approval of the CEFLASH-4 and RELAP-4 computer codes used for calculating LOCA mass and energy release rates and cavity pressure loadings, respectively.
The licensee, used a flow multiplier of 0.7 instead of the recommended value of 1.0 in the CEFLASH-4 calculations.
The 0.7 value was justified by comparing test data with the critical flow correlations.
The nodalization of the input model is acceptable based on the staff's review of input data and sensitivity studies performed by the licensee.
The SG subcompartment pressurization analysis of the St. Lucie Unit 1 plant for postulated breaks at the SG inlet and outlet nozzles has been reviewed and is considered acceptable.
Acceptance is based on the staff's review of the data provided by the licensee and its previous review and approval of the applicable computer code for calculating LOCA cavity pressure loadings.
The nodalization of the input mode is acceptable based on review of the input data and sensitivity studies performed by the licensee.
3.3 Structural Evaluation 3.3.1 Evaluation of Methods and Models The structural computer codes cited in the licensee's report are found to be acceptable to the staff.
The codes (STRUDL,
- DAGS, NASTRAN, MARC, CESCHOCK,
- ASHSD, SSMSOR/DYNASOR, and A%'YS) utilized in the LOCA analyses have been benchmarked in a satisfactory manner to the staff.
The methods used in performing the required structural analyses are acceptable to the staff inasaach as they conform to the accepted state-of-the-art, standards, and regulatory codes.
Based on the submittal reviews, the detail employed in the system and subsystem structural finite element models is considered acceptable by the NRC staff for predicting the mechanical response.
The staff evaluation in this report has considered the need to combine LOCA and SSE loads in the design of the RCL piping.
The staff believes that there is sufficient technical evidence (Reference
- 18) which demonstrates that the SSE and LOCA for the main loop piping in PMR plants may be considered as
independent events in determining the appropriate combination of the effects of accident conditions and natural phenomena as required by GDC-2.
In its load combination program, as a part of Generic Issue,B-6, Lawrence Livermore National Laboratory (LLNL) conducted a program to estimate the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping of PWRs.
The results of the LLNL investigations indicate that the probability of a direct seismically induced DEGB is extremely small.
The best estimate probabilities of direct DEGB using t/g medians of1$ he distribution of the modeling uncertainties ranges from 5 x 10 to 7 x 10 per plant year for both Westinghouse and CE plants.
From the uncertainty analysis, considering the whole range ofyodeling uncertainty, it is concluded that a direct DEGB probability of 3 x 10 per plant year can be considered as the absolute upper bound for Westinghouse and CE plants.
Indirectly induced DEGB in the reactor coolant loop piping (defined as a
DEGB in the reactor coolant loop piping as a result of an impact with a large component or structure, e.g.,
a falling polar crane) is a more likely event compared to direct DEGB; however, the probability of indirect DEGB is also very small.
For the lowest seismic capacity Westinghouse plant, the median probability of DEGB is 3.3 x 10 per plant yeag.
The corresponding indirect DEGB probability at the 90th percentile is 2.3 x 10 Even for this lowest capacity plant, these probability values are still very small.
For all 46 Westinghouse units east of the Rockies as a whole, the median probability is more than one order of magnitude lower.
The probability values for the CE plants are also very low.
The upper bound probability values for the CE plants are comparable with those of the Westinghouse plants.
Based on the analyses submitted by the licensee and inde-pendent assessments by the staff and its consultants, the staff has concluded that the licensee has provided adequate justification for the documented devia-tions from the requirements of the Standard Review Plan 3.9.3.
The results of the above probability studies provide added assurance that these deviations are acceptable.
The instability approach in the analyses of the RCS supports,
- CEDN, and ECCS piping is acceptable since it complies with ASME Code,Section III, Appendix F guidelines.
The determination of fuel deformation and spacer grid impact loads is accepted as the appropriate internals motion (upper and lower grid plate and core shroud) and is adequately incorporated as the fuel assembly forcing functions.
The acceptability of the fuel analysis is also based on audit calculations performed by the staff's consultant, EGSG
- Idaho, Inc.
(Reference 19).
The audit determined that the CE modeling schemes utilize a dual load path for the spacer grids and, therefore, provide an adequate response of fuel asseahlies.
Determination of the total stresses in the core barrel resulting from the asymmetric downcomer depressurization using.<ecoupled beam and shell modes is acceptable since this procedure has been shown to be exact for linear analyses.
Analysis of the ECCS piping is acceptable based on the bounding analysis performed by the licensee.
This analysis consisted of a dynamic analysis of the most highly stressed ECCS lines for motion of the ECCS nozzles on the cold leg piping as determined from the RCS dynamic system analysis.
3.3.2 Com liance with Acce tance Criteria The licensee's stress and/or load evaluation of the reactor system components is acceptable to the staff.
The criteria used in the evaluation are, in general, in agreement with industry standards and meet the acceptance criteria outlined in NUREG-0609.
Although some exceptions to the outlined criteria occur, func-tionality of each analyzed reactor system component is demonstrated.
The reactor vessel supports exceed the ASME Code, Appendix F criteria based on 705 of the instability load.
However, this computation is based on the use of minimum material properties.
If actual values are used, the code limits will be met.
Therefore, the component is acceptab le.
The licensee's stress and/or load evaluations of the reactor vessel internals, primary piping, CEDMs, and ECCS piping is acceptable since ASME Code, Appendix F
criteria are met.
Acceptabi lity of the steam generator support evaluation is based on the compar-ison of the calculated support loads to design loads, yield capacities, and test loads.
The LOCA results for the supports are well within their allowable limits.
Acceptance of the shield wall stress evaluation is based on the use of standard industry practices for determining load criteria and the use of conservative material properties.
Two principal acceptance criteria apply for the LOCA, which includes the asymmetric effects:
(a) fuel rod fragmentation must not occur as a direct result of the blowdown loads, and (b) the 10 CFR 50.46 temperature and oxidation limits must not be exceeded.
The first criterion is satisfied if the calculated loads on the fuel rods and components other than grids remain below designated allowable values.
The second criterion is shown to be satisfied by an ECCS ana lys i s.
The acceptance criteria for the reactor internals following a LOCA is based upon maintaining the core in place and assuring that the adequate core cooling is preserved.
This can be accomplished if the following criteria are met.
For the core support components, the stress intensities must be less than those listed in the ASME Boiler and Pressure Vessel Code,Section III, Division I, Appendix F.
Meeting the stress criteria for core support components assures that the core will be held in place during a
LOCA.
for the internal structures, the component deflection is limited so that the core is held in place, adequate core cooling is preserved, and the resulting loads do not adversely affect the core support components.
Except for loads on peripheral fuel assemblies, the grid strength is greater than the loads imposed on the grids during the postulated LOCA event.
The beneficial effect of irradiation on spacer grid strength is discussed.
Irradiation will increase grid strength sufficiently to provide the required load capability.
An evaluation of the effects of reduced channel flow area in peripheral spacer grids is presented.
Fuel temperature calculations are presented to demonstrate core coolability following the grid loadings induced by the postulated pipe break at the vesesl inlet nozzle.
- 4. 0 CONCLUS ION In conclusion, there is reasonable evidence that the St.
Lucie Unit 1 reactor system would withstand the effects of asymmetric LOCA loads and that the reactor could be safely brought to a cold shutdown condition.
Date:
August 9, 1989 Princi al Contributor:
J.
Rajan
Attachment:
Tab les
- 5. 0 REFEREN CES 1.
"Reactor Coolant System Asymmetric Loads Evaluation
- Program, Final Report, Calvert Cliffs 1 and 2, Fort Calhoun, Millstone 2, Palisades,"
Vol. 1, 2, and 3, Combustion Engineering, !nc., June 30, 1980.
2.
"Response to guestions on the Reactor Coolant System Asymmetric Loads Evaluation
- Program, Final Report," Vol. 1, 2, and 3, Combustion Engineering, Inc., June 1981.
(Volume 3 Proprietary).
3.
"RELAP A Computer Program for Transient Thermal Hydraulic Analysis of Nuclear Reactors and Related Systems,"
User's
- Manual, ANCR-NUREG-1335.
4.
Fabic, S.,
"Computer Program WHAM for Calculating Pressure, Velocity and Force Transients in Liquid Filled Piping Network,". Kaiser Engineering Report No. 67-49-R, November 1967.
5.
Combustion Engineering, Inc., "Method for the Analysis of Blowdown Induced Forces in a Reactor Vessel,"
CENPD-252-P, December 1977 (Proprietary).
6.
" ICES STRUDL II - The Structural Design Language Engineers User's Manual," HIT Press, Cambridge, Hassachusetts, 1968.
7.
"ADHAS -
A Computer Code for Fluid Structure Interaction Using the Finite Element Technique,"
Ebasco
- Services, Incorporated, 1979.
8.
"DAGS-CENDP 168, Revision 1 - Design Basis Pipe Breaks,"
September 1976.
9.
"DFORCE - Design Basis Pipe Breaks,"
Septerrber 1976.
10.
"HSC/NASTRAN - User's Manual," McNeal Scwandler Corporation, Los Angeles, California (1978).
11.
"Topical Report on Dynamic Analysis of Reactor Vessel Internals Under Loss of Coolant Accident Conditions with Application of Analysis to CE 800 HWe Class Reactors,"
CENPD-42, 1971.
12.
"CESHOCK - A Computer Code to Solve the Dynamic Response of Lumped Mass System,"
Described and Verified in above Reference 1.
13.
"ASHSD - A Dynamic Stress Analysis Code of Axisymmetric Structures Under Arbitrary Loading," Described and Verified in above Reference 1.
14.
"SAHMSOR - A Finite Element Program to Determine the Stiffness and Mass Matrices of Shells of Revolution," Described and Verified in above Reference 1.
15.
"DYNASOR - A Finite Element Program for Dynamic Non-Linear Analysis of Shells of Revolution," Described and Verified in above Reference 1.
16.
"ANSYS" Swanson Analyses
- Systems, G. J.
- DeSalvo, J.
A. Swanson, 1975.
17.
"PLAST -
An Elasto-Plastic Computer Program for Stress Analysis of 3-D Piping Systems and Components Subject to Dynamic Effects," submitted to the NRC as ETR-1001 - Ebasco Topical Report.
18.
L. Woo, Holman, and
- Chou, Lawrence Livermore National Laboratory, "Failure Probability of PWR Reactor Coolant Loop Piping" (UCRL-86249) presented at the ASHE Pressure Vessel and Piping Technology Conference, San Antonio, Texas, June 1984.
19.
T. L. Bridges, "Review of Combustion Engineering, Inc., Fuel Assembly Structural Analysis Topical Report CENPD-178-P, Rev. 1-P,"
EGG-EA-5824, EG8G, Idaho, March 1982.
~ I I\\
IW TABLES AND FIGURES
FIGURE 1 RELAP4 hlGDEL FOR ST. LUCIE PRIMARY COOLANT SYSTEi"8 1$
5C 15 1$
9 Z
9 11 e/4 5
TT s QQ~l O(g, 0 1%
056 9
ID a
62 SZ 4
O~
54 IZ IZ
.0
{~~
L'J Zl p
cO 19 JO Jt 45 ZC Cl G'1 Ck
~5 CC 24
+
~
0 4
(81I 71 7O 45 CZ C1 1.,
QO,,
a~t ln5>a
.';~c"s' rt Qa.'
l~
(>D
$ 5 yggy5 5
CI ~&C I C9
<BZ)
TO G.G. IIO. 1' r
I o
9919 AEACTOAVESSEL IIOTLEO TO S.O. N0.2
)w X
INTEANALS SEE FIGunE 3a FOA ETAILSOF INTEANALS)
A fT 991$
m CDn C7 CD ID m.
I CD C) m mn g
CD R7 P)
CD o
0913 LUMPEO hfASS I'OINTOF APPLIEO FOAI;E
FIGURE la REDUCED MODEL OF REACTOR INTERNALS 233 UGS 0 32 X-DIR IS /ITO 3U i =T iYQ~I.ES 28:
10'2S FUEi CORE SHROUD CSB
'!3 l4 I
Q ghtQ H,
e w p' ~ ~ ~ i7 + aa G&?
/I/ <<Qg
~ +
W AAa
~ r% ~
~ Q$ ta
~
~
< ~ ~'~.Wa ~ ~'~4 w o
<<G 5 P pulpy~
SEE FIGURE 3b FOR DETAILS OF REACTOR VESSE'
@NO PIPING
FIG UR E 2
GENERIC PLANT R E D U C E D INTERNALS lA 0 D E 4.
0- 1R)
X e~
DIR. OF INLET SQ - 270 BRMcK UPPER GUIDE STR UCTUR E 1
~a X DIR. 1S P RALLELTO
~
THE OUTL=T NOZZLcS L'-CORE SUPPORT RSL 23 SUPPORTS CORE SHROUD FUEL 16 R EACTCR VESSEL 17 22 10 LOW'ER SUPPORT STRUCTURE LEGS. O:
A >
AXIALCAP' YORIZ NTAI.GAP 1NDICATES AN 'IALLY pRCl 6 ~ f Cp @+I ill sa>g.
L LATERALMCDEL ONLY
AXIA'ODELONLY
30 20 10
-10
-20
-30
-40
-30 0.0 0.2 0.4 0.6 0.8 Time (seconds)
Figure 3.
Cold leg break force time history
28 16 12 0
-4 0.0 0.2 0.4 0.6 Time fseconds) 0.8 Figure 4.
Cold leg break moment time history.
0 4J J
-I
FIG UR E 5
GENERlC PLANT DETAlLED LATERAL !NTERNALS MODEL
~ ~
r'0' SIR. oF IP4L g 7 PO-Z'FOl ha E.ax UQS SUPPORT op PLA E~
FR ICT IOh NC."4 LIi'JEER HY5i c.nE>IS ROTATIGHAL 17 I~
RV LEDGE C~
CZ p
$HRCUDS S7 S2 97 9g 90 96 89 94 CSB L4 ER FLANGE GV!DE LUGS'4 RY 101 NOZZLES 27 24 C1, I~
51 49 47 58 E7
~
~
76 28 CORE SHROUD 13 2S 25 12 RV SUPPORTS 23 37
'75 19 0 52 6>,
. r~l 10 21 20 LOVE"R SUF PORT S RVCTURE CS3 RV SNVBSKRS
D +A I L E D v E R T I c A i.
M 0 lOL LINcAR SPRING FIG VR E 6
NON LINEAR SPRING AXIALGAP.
EXPANSION COPAPENS TING ~
R ING UGS FLANGc FFl ICTI~~4 ELEMENT' FLANGE T 13 12 RV LcQQP 34 CM 10 SUl.t't Ucr SPRINGS 22 27
- FUEL, RODS 2S rl GUIDc TUBERS 20 T9 18 CORE SH ROUQ 30 17 16 15 29 LSS SEA.'AS 5 LSS Pi~ TE T1E ROO KX.TS tT ECTION ONLY)
LSS CYL.
REACTOR VESSEL CSB L~ER FLANGE
tl
~5,o
TROLE 1 ST. LUCiE 1 Nonriini. nnD srtsraic supponT Lonos (xloo. LD.)
IIIIIIMAL QI'EIIAIINII lllEIILIALI COIIOI'IIOH DEAD Ucho Looo ooiooo ouoor III 0
.020 VI
.GGO 1.16$
IIVI LI96 I.3r50 iX
.009
.032 DIIE 5EIDMIC
.002
.3N
.644
.040
~Oll
.004 DDK IEIILIIC
~ av
.009
.070
-I.zao
.092 I
~V lol
~I 11 oT
/
~o ~
4 rVI II2 0
-.091 VZ
..CG4
.720 OV2 I.199 I.216.
II3 0
.070 V3
.034
.741 IIV3 f.I96 t.2IO 1.220
.017 1,139
.387
.001
.253
.010
-.OGO
.023 359
-P04
~
.3DO
.270
.349
.743 2.452 2.270
.000
.003
,507
.030
~120
.500 4.710
.520
.70 I
.540
.09I
.740
<r oVI Vl iIiIIoosVI1111
~WITOO~
OOI 4C IOOOrt log Zil 2 lz Yl Yz Y3 Y4 X
IIY 0
0 0
0
,300 0
.300
.ZI9
.300 1.009
.300
.320 0
0 4.376 l.300 0
0
.OIO
.057
.ODO
-.051 0
0" 0
'.OI9
.155
.4ll l52 9
%.195 1.197 0
.OGO
.07I
.Ori1 0
0 0
.033
.114
.173
.IOD po 0
0
.039
.311
.039
.305 0
%.3DO f.394 0
.121
.143
.I IO 0
I~
IO~ OOO tllVINO IOVI Ioooo ~ ~ rIooo II
Table 2
Co~a-.isoa of Peak Calculated z=d Desi~ Seismic (QBc.) Loads zt Representative Locations Desi~
E,'Zips)
Calculated (~ps)
Eo. =o-.."zl Vertical Fozizontal V'e"tical Cold Leg Spt Hot Leg Spt 2,455 Apwc3 1,268 762 522.6 515.0 354.6 429.4
Table 3
S" Luc'e Unit ~1 RV S'~POi'~L~Z PBS ~"-4"T:QHS (&PS) - LCCA
- SiTSLLLC (SRSS) 4:.
C G
BP~~
AT hCZZL:-
1.A OR 2
LQCATZHiS
~~1A SPPT RV SP T ST~inc.SS VALU:-S K
~ 64.62 x 10 lb/in 5
K
~ 59.71 x 10 lb/in K
77 54 x.10 lb/'e 5
K ~ 75.83 x 10 lb/in Vert'l Hori 1397 m02 2317 1587 vs SPPT Vertical 1-:or zon al 2800 5331 2251 5473 V
ical Ho zontal'/
58 74 93 3048 7777
- For L2"='. : i~op " ac"'c=s add these values to the vertical results:
-::1A SP~i.
S2 T Hot Leg SPPT 710.
K 726.
K 1157 K
For bra k at nozz e
13 o" 23, the loads on the cold 1eg supports wou d
be reversed
~
~
~
AEACTOIR PIREBGUIIE VESSI=L ~APPOINT PAD CJ(PABILITY
'V fe)llI" iC (C ~ Sleet~ gg l)OT LEG SUPPORT a.o SO. t-T. Ct.r; COMPV1 I:t) MAX.l.OntiS j (LOCA 4 SEISMIC t' t))
pA;l(OT Leo supponT uNnnol(LLN COLD LEG s UPPon T pC nflol(l'.NCOLD LEG SUPPOBT~
pA O
zON fC0x-COLO LEG SUPPOATS PC n
~r G.
tt g
. 10 11 12
.13
.14 15 10 By x 10
Table 4
St.
L cie 1 Reactor Pressure Vessel Support Capac'tf SteeL suppor" stoic u".e -
horizontal 8400 PXps+
(conc"ete is Li-it'-.g)
Ste 1 suppcrt struct. re Reactor Cavity VaLL Reac" ox Cavity
$?aLL Reactor Suoport.
Pads Peactor Su porc Pads ve"tica1 do~mua"d horizontaL ve"tical hc izontal ve"tical 12000 kips
-L3000 kips*"-
not limiting See E'iguze 4
See Figure 4
T.oad = ~-div.'dual girde:
c'=ab e "esuLtar.t asy=etric rechanicaL load trans.Zttcd along gi=d =s to conc@'ete,=based on rebar uean axial st"ess being within yie'.
TABLE 5 STEAM G"N:ROTOR LOVJFR SUPPORT CAt CULATFD AND DKStGN 10ADS IREFER
~ 0 iABI~ 1 FOR SYBIEOLSl EFFORT C>> COILL, Na. 1
~ OBE (RSSl t
727 HL, COILL, NO. 2
OESIOV LOAO LOCAi CSE 3.6CO Z"i2 805 FRaiVT Y1 I
~5.9 SICE Y2 I
-756~
5" C!C Y3
'I 605.0 SiOE Y4 X-STOP 78 194
-2487.7 1.868 1,770 1176.4
)
-1,737
+1249.0 I
,691 1175 9 I
1,734 519iL9 I
5.63 273 I
.301 40 I
-1.574 1,8QO OC ANO FT lCIFSI 0 C Si cd CKVERA.OR/
s Ia:"Na BAsE r~zmaTraRT INTERFACE
.Fx Fy Fr tH I CIIIUJ LOCA 5205 LOCA 5 OEE, 5205 3582 4184 SLIOINQ BAS>>
OESICN LOAQS 5653
-2,471.0 11.0 32.0 ills
~ o'i -GATIV:- SIQiV i'ifEAVS~iVSII3iV 65 I
33.6
~809 46'l4 24.0
Table 6
Luclc Untt ttl Reactor C(Io).ant System Reactor Prcssure Vessf 1, anII Ilcacl:or Cool.anf:
Pump Nozzle Loads Duc to a
l ft Reactor Vessel lh Inlet Nozzle Guillotine Break PIPP.
RUPTURR RS.'l HIHR~HT In-Ki~n Nozzle RCP Disci>artcc RCP Suction RV Inlet RV Outlet RCP Snubber Act'in 109,300 50,500 71,750 50,150 RCP Snubber Hnn bc~tin 109,600 54,550 71,910 50,170 Saiainlc kfonicnt 5,910 7,256',272 2,535 t
hllc Mabl c Hoaicnt In-Ki 96,010 70,965 78,965 279,340
Taole 7
ECCS Piping Stresses St. Luc'e Unit Fo.
Cot~ected Piping Stresses Calculated vs. Allc~~a<le 4.0 sq.
Et.
CLG En3.et
- Bzeak, Des ig Poi".."
(Refer to:'ru~re 5)
Calculat d St=ess (Fau.
10 AS.K)
Allo~~a'ale Stress 2
7 8
19 39,070 psi 75, 152*
75,030 41,835 43,475 47,690 33,430 20,171 48,500 psi 481500 48,600 48,600 48,600 48,600 48,600
~
'48,600
- inc"it=ah -; an" integrity are assured i Level B (upset condit'ons) li its of tho 'S:.= Boi~e-and Pressure Vessel
- Code, Section ZII, Division 1 are not
=;"a de*:c..ctionab~ty is i=portent at points 5 and 6 '~here the v"lve
's..';- pof.-ts 2 and 3
zhese 1'm'ts are ezcee8ed.
Ho'ever, L'evel D (=suited 1.ilats) are not. e:cc e" d at these t o points.
Level D 1'mits aze used to de onst".te that '- e'er'ty is m
ntained.
Equation (9) at those t-o points t-ou'" '-e'd 45,043 psi and 44,479 psi respectively with an alla~able of 48,600 psi
l'~~
~S Vl r