ML15238A804
| ML15238A804 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/20/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15238A803 | List: |
| References | |
| NUDOCS 8305310089 | |
| Download: ML15238A804 (74) | |
Text
- SAFETY EVALUATION REPORT ON ASYMMETRIC LOCA LOADS OCONEE 1, 2 and 3 I.
7NTROJUCT ION On May 7, 1975, the Nuclear Regulatory Commission 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 origin.al design of the reactor vessel support for North Anna Units 1 and 2. It hac been identified that in the event of a postulated, instantaneous, aouole-ended offset shear pipe breaK at the vessel nozzle, asymmetric loading could result from forces induced on the reactor internals by transient differential pressures across the core barrel ano by forces cn the vessel due to transient differential pressures in the reactor cavity.
Witn tne advent of more sophisticated computer codes and the development of more detailed analytical models, it became apparent that such differential pressures, although of short duration, could place a significant load on the reactor vessel supports and on other components, thereby possioly affecting tneir integrity. Although this potential safety concern was first identified during tne review of the North Ana racilities, it was detarminee to nave generic implications fcr all pressurized water reactors.
In October of 1975, the NRC staff notifiec eac-operating ?-WR licensee of a potential safety problem', concerning he desicn of their reac:or pressure vessel support system. From this survey it was discovered that these asymmetric loads had not been consiCered in tne design of any PwR primary system. Suosequently, in June 1975, the NRC staff informed each PWR licensee that a reassessment of the reactor vessel support design for each of its facilities was required.
Althouch the NRC staff's original emcnasis and concerns were focused primarily on the integrity of the reactor vessel support system with 1
8305310089 830520 PDR ADOCK 05000269 P
respect to Postulated breaKs inside the reactor cavity (i.e., at a nozzle),
it became apparent tanat significant asymmetric forces could also be generated by postulated pipe breaks outside the cavity and that the scope of tne problem was not limited to the vessel support system, itself.
Tne staff, after reviewing this proolem, aetermined that a re-evaluation of the primary system integrity of all PWR plants to withstand these loaas was necessary.
Therefore in January of 1978, the NRC staff requested eacn PWR licensee to submit additional information in accordance with the expanded scope of the problem. Those letters outline the present scope of the problem specifying a minimum number of pipe break locations to be assessea and the reactor system components to be evaluated.
Since the identification of the asymmetric load problem in May 1975, EG&G Idaho, Inc. has performed a number of independent audit analyses to verify licensee submittals on this problem. A total of six analyses have been completed (one linear elastic and one nonlinear-inelastic analysis of a 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 for asymmetric LOCA loads were developed (Reference 3).
In July of 1980, Babcock and Wilcox, on behalf of the Duke Power Company, a member of the B&W 177-FA Owners Group, submitted a final asymmetric LOCA loads evaluation report, BAW-1621 1, applicable to the Oconee Units 1, 2 and 3 nuclear power plants. - This material,. suomitted in response to the January 1978 letter of request, 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 January 9, 1981, the NRC staff notified B&W of the additional requests, and the response, BAW-l621 Supplement 12, was submitted in June of 1981.
The Oconee final submittal 1 and supplement2 represent the limiting cases for the asymmetric LOCA evaluation and have been reviewed in conjunction with the criteria outlined in NUREG-06093.
Subsequent 2
sections of this safety evaluation report summarize the evaluations performed by the licensee for subcooled blowdown loads, cavity pressurization, and structural response.
Following this is the staff evaluation of these same analyses whicn includes the assessment of the licensee's compliance with acceptance criteria.
3
- 2.
LICENSEE EVALUATIONS 2.1 Subcooled Loads Analysis The CRAFT2 computer code was used to predict the transient hydraulic response of the reactor primary coolant system for four postulated break locations. The guillotine pipe breaks considered for analysis are as follows:
- 1.
O.6OAa break in 0.018 seconds at the reactor vessel outlet
- nozzle, 2..
1.00A break in 0.022 seconds at the hot leg elbow closest to the reactor vessel,
- 3. 2.OA break in 0.027 seconds at the reactor vessel inlet nozzle, and
- 4.
2.OOA break in 0.027 seconds at the cold leg elbow closest to the reactor vessel.
The breaK locations are illustrated in Figure 1. The CRAFT2 code assumes the fluid boundaries to be rigid and at rest, thereby excluding fluid-structure interaction effects of the core barrel-reactor vessel relative motion on the downcomer pressure transients in the subcooled loads hydraulic analysis. The hydraulic model input into CRAFT2 is discussed quantitatively in a very synoptic fashion. The loads throughout the RCS are computed by the FORCE2 computer pragram, a subroutine in the CRAFT2 code. The methodology employed in CRAFT2 and in this conversion of pressure to applied loads is described in detail in the B&W topical report 8AW-10132P-A 4
- a. A break opening area will be referred to as a multiple of the cross-sectional flow area (A) of the pipe at the specified location. A full guillotine off-set break is defined as 2A.
4
In the evaluation of asymmetric loacings for pipe breaks in-the steam generator subcompartment the CRAFT2 code i.s also utilized in obtaining pipe reaction forces for four additional postulated pipe ruptures.
All cases assumed a oreak opening area of 2A witn an opening time of 0.01 seconds.
The break locations considered in the analysis are as follows:
- 1.
at tne steam generator inlet nozzle,
- 2.
at the steam generator outlet nozzle,
- 3.
at the Pump suction nozzle, ano
- 4.
at the pump discharge nozzle.
Resulting thermal-hydraulic loads were used in the evaluation of primary loop piping (Section 2.4.7) and the steam generator supports (Section 2.4.11).
2.2 Cavity Pressurization Analysis Analysis of the reactor cavity pressure was performed by the licensee using the CRAFT2 computer code.
This code employs the Moody choked flow correlation with a discnarge coefficient of 0.6.
Blowout calculations were performed in the CRAFT-2 code for each venting devic3 using Newton's secona law of motion. Due to the aosense of experimental Gata, frictional effects in the form of drag coefficients were assumed to be zero.
Compensation for this assumption was accounted for by douoling the vent cevice weights in the blowout calculations.
The suomittal includes data, born taoular ano pictorial, that descrioe the model used to calculate cavity pressure transients for a postulated LOCA. The breaks considered for analysis are as follows:
- 1. 0.60A break in 0.018 seconds at the reactor vessel outlet nozzle and
- 2. 2.00A break in 0.027 seconds at the reactor vessel inlet nozzle.
The reactor cavity has an approximate diameter of 23 feet and an approximate height of 38 feet. Since the nominal outer diameter of the reactor vessel is approximately 15.7 feet, the cavity contains a net free volume of approximately 8000 cubic feet.
Figures.2 and 3 illustrate the reactor cavity in an elevation and plan view, respectively. The nodalization indicated in the figures refers to the CRAFT2 modeling scheme of the cavity. The CRAFT2 model used contained 70-volume-nodes, determined from a sensitivity study to be detailed enough to provide a convergent solution. The study involved the comparison of four cavity models, containing 42, 50, 70, and 82 nodes each.
Volumes were divided differently in each model in both the axial and circumferential directions..Resulting peak pressurization forces and moments were shown to converge as the numoer of nodes in the model increased, and the 70- and 82-node models converged to within one percent.
Peak integrated loads on the reactor vessel resulting from the cavity pressurization analysis are shown in Taoles 1 and 2 for the hot leg and cold leg breaks, respectively.
Figure 4 represents the force time histories summarized in Table 1, and Figure 5 contains the corresponding force time histories for Table 2.
The steam generator-cavity pressure analysis was performed using available data.
Force time histories from a previous analysis of a simil.ar plant were scaled to be made applicable to Oconee. Available data for three pipe breaks was used to determine cavity pressurization and jet impingement on the steam generator and pump. The breaks considered were
- 1. at the upper end of the hot leg,
- 2.
at the pump suction nozzle, and
- 3.
at the pump discharge nozzle.
O
The analysis assumed full 2A pipe breaks, and resulting pressure loadings provided the applied loads to the half-loop finite element structural model described in Section 2.3.5. Figure 6 illustrates the steam generator subcompartment in a plan view.
2.3 Structural Analysis The Licensee's structural analysis was performed utilizing essentially five primary finite element models.
Additional component and support finite element models were used to develop input to the pri-mary models and to calculate component/support loads and stresses for detailed evaluations.
The primary mathematical models to which asymrretric LOCA loads were applied are described in the following subsections. The general plant layout is shown by the illustrations of Figures 7 and 8.
2.3.1 Reactor Vessel Isolated Model The reactor vessel isolated model was used for the majority of the loading analyses and for the evaluation, directly and indirectly, of a majority of the major components and supports comprising the primary coolant system. As shown in Figures 9 and 10, the model consists of the reactor vessel, simplified reactor internal components, SSS, CROMs, cold leg piping extenoing to the pumps, and hot leg piping extending to the steam generator.
Stiffness matrices provice appropriate Counaary conditions to the pipe ends and represent the characteristics of the structures to which they are attached. Responses of the modeled components were calculated by the STALUM computer code using applied loads consisting of reactor and piping internal hydraulic olowdown forces (Section 2.1),
loop mechanical loads caused by the release of normal operation static equiliorium forces at the postulated oreak, and asymmetric cavity pressurization forces (Section 2.2) for oreaks within the reactor cavity.
Included in the modeling were the effects of hydrodynamic mass coupling. A linear elastic analysis was performed with the model to evaluate loads and stresses in the piping and reactor vessel internals. Using resulting response time histories in smaller, more detailed models, this system 7
dynamic analysis was used indirectly in the evaluation of tie vessel supports, support embedments, fuel assemolies, SSS, CROMs, and core flood line piping.
2.3.2 Core Bounce Model Due to the propagation of the suocooled blowdown decompression wave, differential pressures result vertically across the core, causing a vertical response from the fuel assemblies, called core bounce. A nonlinear concentric beam model was developed using the ANSYS computer code and is schematically shown in Figure 11.
Results from applied reactor internals pressure differentials on the core oounce model were used to supply additional vertical loadings.on the internals of the system dynamic model of Section 2.3.1. The flow chart given in Figure 12 depicts the applied loads for the system dynamic response analysis,.and the relationship of the core bounce model to the reactor vessel isolated mocel.
2.3.3 Cavity Wall Model The ability of the primary shield wall to-sustain the worst case pipe rupture loads was determined from a linear elastic, tnree cimensional continuum model of the wall using the EDS-SNAP computer code.
The mocel, shown in Figure 13, represents a 180-degree segment of the shield wall using isotropic elements. Included in the modeling are the hot and cold leg penetrations, the fuel canal floor, the fuel canal walls, a recess in the wall directly beneath the hot leg penetration, and the core flood line shield wall stiffness. Loads on the model consisted of reactor cavity pressurization loads produced by the analysis.described in Section 2.2.
2.3.4 Pipe Whip Model The break opening areas used in the calculation of the internal (Section 2.1) and external (Section 2.2) pressure loadings for hot leg breaks were determined utilizing the hot leg nonlinear pipe whip model illustrated in Figure 14. A restraint also exists on the cold leg and was analyzed with the cold leg nonlinear pipe whip model shown in Figure 15.
8
The restraint loading time histories resulting from the two analyses were used to qualify the restraints for the postulated LOCA.
The ANSYS computer code was used in the development of the pipe wnip models, utilizing three dimensional plastic straicht and curved piping elements for the reactor coolant system piping ano a combination of spring and gap elements for the restraints. The reactor vessel centerline and steam generator base were considered fixed points in the analysis, providing appropriate boundary con-ditions. Applied forcing functions in the dynamic pipe whip analysis involved an iterative process that converged to a final break opening area and rate. At the location of the break the system pressure times the pipe cross-sectional area, P X A, was applied to the pipe end as a very conservative initial driving force. The resulting break area and time provided the initial conditions for obtaining thermal-hydraulic load time histories that were applied.to the structural model in the second interation. The iterative steps were continued until a representative break area and rate resulted.
2.3.5 Half-Loop Model Component and support responses due to postulated pipe ruptures within the steam generator compartment were evaluated utilizing a linear elastic finite element half-loop model, schematically shown in Figure 16.
The unmodeled loop stiffnesses are applied to the reactor vessel at the inlet and-outlet nozzle elevation, and adequately represent the influence of the missing loop.
Existing loadings considered in the analysis include internal hydraulic blowdown loads (Section 2.1), any significant asymmetric compartmental pressure loads (Section 2.2), release mecnanical loads at the pipe severance, and any jet impingement loads. Althougn force time history data was recorced for each joint in the model, only the steam generator support reactions were presented and evaluated.
2.3.6 Subsystem Models Numerous smaller, more detailed mathematical models were used in the LOCA analysis to provide representative and meaningful responses to the applied loadings. A few models were developed to determine component 9
support stiffnesses to be applied to the system analyses, and also to qualify the supports once system responses were obtained. The following computer models were developed and utilized in this manner:
- 1. Two pipe whip restraints along the hot leg piping were active during the considered loacings: the bumper restraint at the lower eloow and the collar restraint located above the lower elbow, shown as Restraints 1 and 2, respectively, in Figure 14.
The bumper restraint consists of an assemoly of linked plates as illustrated in Figure 17 and was modeled using the nonlinear finite element computer code ANSYS. The math model is shown in Figure 18, and the associated spring rate data is provided in
- Table 3.. During a postulated hot leg break, movement of the pipe causes the bumper restraint to bear against the steam generator, transmitting normal and transverse forces. The second hot leg restraint consists of a curved channel section to form a collar around the pipe. The ends of the collar are welded to a hollow box section, which is welded to an embedment plate. Tias is illustrated in Figure 19.
Since the restraint is capable of resisting some lateral pipe motion, as well as tension and compression in the axial direction, two spring rates were developed from the analysis:
(1) an axial spring rate, shown in Figure 20, and (2) a transverse.spring rate, shown in Figure 21.
Located at the first elbow from the reactor vessel on the cold leg is the cold leg shield.restraint, shown in Figure 22. The restraint consists of an assembly of plates and wide flange sections. The computer model using the ANSYS coae is shown in Figure 22, and the developed spring rate data is provided in Table 4.
- 2. The reactor vessel skirt support, shown in Figure 23, is the primary supporting system for the reactor vessel.
Using the ANSYS computer program, a nonlinear finite element model was developed to adequately represent the stiffness and dynamic response of the simplified modeling scheme in the system dynamic
-model described in Section 2.3.1.
The model consists of plastic shell elements representing the skirt, its flange, and the 10
portion-of the reactor vessel adjacent to the skirt. seam elements represented the remaining portion of the reactor vessel, and nonlinear axial springs represented the bolts, sole plate an.
shear keys at the skirt connection to the concrete pedestal.
The 180-degree model is snown in Figure 24.
The resulting spring rates for the vessel skirt are given in Figures 25 and 25 for the force-deflection and moment-rotation relationships, respectively. The linear portion of the curves represent the stiffness used in the system analysis, and the nonlinear curves demonstrate the ultimate load carrying ability of the skirt.
- 3.
The evaluation of the reactor vessel supporting system includes the static analysis of the support skirt embedment. Figure 27 shows.a schematic of the connection between the Skirt and concrete pedestal. A nonlinear embeament analysis was performed using the EDS-SNAP computer program to assess the nonlinear material behavior of the pedestal reinforced concrete and the effects of partial lift-off due to the overturning moment.
A 180-degree model, as shown in Figure 28, was developed that consists of nonlinear elastic truss elements, representing the tension properties of the anchor bolts and the compressive strength of the concrete. Figure 29 illustrates a second nonlinear model, the embedment suostructure model, developec to perform a detailed nonlinear analysis of the reinforced concrete in the compression zone of the pedestal.
This model provided two uses:
(1) it defined the stiffness properties of the nonlinear concrete truss elements includea in the nonlinear embedment model (Figure 23),
and (2) it was used for component qualification purposes by utilizing resultant loads from the analysis of the nonlinear emoedment model.
The two-dimensional model of Figure 29 represents a typical cross-section through the pedestal, utilizing nonlinear two-dimensional elements to model the concrete, and beam and truss elements to model the embeaded steel and reinforcement.
The analyses perfor-med utilizing the two models described above resulted in the RPV support skirt 11
embedment lateral force-translation and moment-rotation spring rates snown in Figures 30 and 31, respectively.
These spring rates were supplied to the system dynamic model of Section 2.3.1.
S veral other models.are utilized for detailed qualifications of particular components.
Applied loadings or motions to these models were resultant responses Afrom the system dynamic analysis.
The detailed models are as follows:
- 1. The fuel was analyzed with the STARS computer Code, which was described 5
in the B&W report BAW-10133 and has been approved for licensing 6
calculations by the NRC.
The mathematical model is shown in Figure
- 32.
Displacement time histories of the upper and lower grid plates provided the input motions to this nonlinear core model.
- 2.
The integrity of the core flood line piping was evaluated for asymmetric LCCA loadings.
A linear elastic analysis was performed with the SUPERPIPE computer code using straight and curved pipe elements.
Input excitation to the analysis was provided by the time history motions of the reactor vessel, resulting from the system LOCA response.
Since the motions were calculated at the vessel centerline, slight modifications were made to provide their applicability to the core flood line model (at the reactor vessel/nozzle interface).
The models are shown in Figures 33 and 34 for the Loop A and Loop B flood lines, respectively. Loop A was analyzed for both hot and cold leg breaks, and Loop B was analyzed for the cold leg break, only.
- 3. A detailed finite element model was developed for the control rod drive SSS to obtain a realistic losa distribution. The NASTRAN code was used in the modeling, utilizing plate, beam and spring elements. Since the primary function of the SSS is-to provide lateral support to the CROMs, the resultant force of the control rod drives was applied symmetrically to the model, shown in the planar view of Figure 35.
12
- 4.
-The CROMs were rigorously evaluated by a nonlinear finite element model using the ANSYS computer code. A schematic of the model is represented in Figure 36 and was developed using concentric pipe elements. Displacement time histories were imposed at three
-different elevations of the model to accurately describe the considered loadings.
2.4 Summary of Licensee's Evaluations and Conclusions The basic criteria for acceptaoility 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 trat overall acceptability of the plant for the postulated LOCA was met. This was demonstrated oy 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 Taole 5.
2.4.1 Reactor Vessel Supports The supporting system for the reactor vessel consists of a short, wide, cylindrical steel skirt located at the base of the vessel.
Using the linear system model previously described in Section 2.3.1, the resulting loads at the base of the support skirt from the postulated breaks at the reactor vessel inlet and outlet nozzles are summarized in Table 6. The analysis of the vessel skirt included the combining of seismic and LOCA loads by the square root of the sum of the squares technique. The dynamic response of the detailed mathematical model described in Section 2.3.5 to the applied LOCA loads provided the results presented in Table 3. The applied moment exceeds the skirt's capacity for the two given locations.
In an effort to demonstrate reactor vessel stability and, thereby, qualify the support skirt, a simplified model was developed to perform a dynamic nonlinear analysis. The model consisted of nonlinear beams and trusses to represent the reactor vessel, control roo drives, service support structure, vessel skirt, and skirt embedment. The modeling technique was verified by performing a linear analysis using the simplified model.
The RPV support was considered acceptable since the nonlinear analysis reduced the overstressed condition to an acceptable level.
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2.4.2 Reactor Vessel Suoport Eimbedments The two nonlinear support skirt emoedment models previously descrioed in Section 2.3.6 were used to evaluate the reactor vessel support embedment integrity. System dynamic analysis loadings applied to this model producea the results shown in Table 5..
Stresses resulting from the nonlinear elastic analyses were determined acceptable although acceptance criteria was exceeded at two locations.
The concrete against the vertical bearing plate is overstressed by seven percent. This condition is considered to produce localized yielding and redistribution of stress was not considered. Therefore, based on conservatisms of the analysis the slight overstress is regarded as acceptable. The compressive strain in the concrete directly beneath the sole plate exceeds the allowable strain by 29 percent. However, the condition is considered acceptable based on its localized effect and the strain, corresponding to the maximum stress, is not exceeded.
2.4.3 Service Support Structure A detailed analysis of the SSS was performed in an effort to obtain a realistic load distribution through the structure.
The results of the evaluation using the mathematical model described in Section 2.3.6 indicated that the SSS is well within the allowables (see Table 5).
After completion of the analysis, it was realized that a connection existed' between the 555 platform and cavity wall, rigid enough to cause significant loading in SSS during the postulated LOCA.
The effect of the connection was investigated by the licensee by perfor-ming a detailed stucy.
The study indicated that reduction of the SSS flange load to an acceptable level would require a clearance in the connection of at least 1.5 inches.
2.4.4 Control Rod Drive Mechanism An evaluation of the CROM was based on an explicit analysis for the Oconee plants.
The analysis utilized the previously descriced computer model in Section 2.3.6. Three loading cases were performed on A and C type 14
drives, and the controlling case was the cold leg break at the RPV nozzle.
Since some of the original-motor tubes may have been-honed out, the smaller dimensions were used in the evaluaticn..
From Table 5 it can be seen that the ri tical stress in the modified motor tuce exceeds the allowable by 4.5 percent.
The licensee catermined this to be acceptable since it falls within the accuracy of the LOCA calculations described in the submittal.
The Oconee CROM housings were also evaluated, and the calculated bending moment is well below the allowable.
2.4.5 Reactor Internals Critical components of the reactor internals considering asymmetric LOCA loads-comprise the core support assemoly.
The evaluation of these components was based on a pre-existing detailed LOCA loads stress analysis.
olicaoility to Oconee was provided oy slight modifications to the analysis to reflect structural differences in the plenum cover, plenum cylinder, and column weldment3. By ratioing LOCA loads and pplying t ratios to this base analysis, the limiting stresses were calculated and are shown in Taole S.
All results for reactor internals were detennioed-to be.
acequate.
2.4.5 Fuel Assemblies The evaluation of the reactor core was dynamically analyzed for vertical and horizontal responses, utilizing the core ocunce (vertical) modEl described in Section 2.3.2 and the horizontal fuels model briefly described in Section 2.3.6..The fuel assembly stress analysis was based on the worst case LOCA loads from all S&W 177-FA owners group plant system analyses, thereby enveloping Oconee. Table 5 expresses the critical results of the analysis. Maximum grid motion applied to the fuels model was slightly greater than one inch, resulting in fuel and guide tube stresses considerably below allowables. However, an evaluation of grid impact loads showed that the maximum spacer grid load of 12,500 lbs exceeded the measured critical load (P.)
of 9,135 which corresponds to crit the onset of significant permanent deformation.
A core coolability calculations, i.e., in accordance with Appendix K, was therefore performed to invesitgate the effects of distorted grids on peak clad temperature. A fully collapsed grid was assumed and resulted in a net decrease in core flow area of 41 percent. An increase of 120 F in peak clad temperature was predicted for this reduced flow area. Since this small temperature increase is confined to the lower power peripheral assemblies, the effect of grid crushing on core coolability is concluded to be insignificant9 2.4.7 Reactor Coolant Pipina Baseo on the system dynamic analysis, using the reactor vessel isolateac mcel described in Section 2.3.1, stress results in loop piping were calculated to be less than the 3Sm allowable. Stress results presented in Table 5 were obtained by considering worst case.loadings from all o&W 177-FA Owners Uroup plant system analyses, thereby encompassing the Uconee RCS piping. The lcop piping was also analyzed using the half-loop model cescribed in Section 2.3.5 to evaluate effects of postulated pipe breaks within the steam generator compartment.
Unbroken primary piping stresses met the allowable stress criteria for all considered load cases.
These resultant stresses are also shown in Table 5.
2.4.3 Pipe hio Restraints As previously described, the Oconee LOCA evaluation produced two active hot. leg pipe whip restraints:
the bumper restraint located at the hot leg lower elbow and the collar. restraint positioned just above the elbow. The stress results from the detailed models described in Section 2.3.5, using the responses obtained from the pipe wnip analysis briefly described in Section 2.3.4, were determined to be acceptable as shown in Table 5.
Loads on the cold leg restraint were determined to exceed its ultimate strength of 2400 kips; therefore, the restraint was considered failed, supporting the assumed 2A guillotine pipe break in the internal and external system pressure analyses.
16
2.4.9 Core Flood Line Piping The core flood line piping for both system loops was evaluated based on the results of a.dynamic analysis using the suostructural models describeG in Section 2.3.6.
Motions from the system dynamic analysis provided the input excitations to the models, and the calculated pipe stresses were all shown to meet the acceptance criteria. The licensee supplied maximum load ratios-only, as indicated by the percent margin values given in Table 5.
2.4.10 Primary Shield Wall The. reactor cavity wall was statically evaluated for transient and steady-state cavity pressurization loadings resulting from pipe breaks considered at the vessel inlet and outlet nozzles. The analysis was performed with the structural model described in Section 2.3.3, and the applied cavity loads were developea from the analysis described in Section 2.2. A dynamic load factor of 1.15 was applied to the transient analysis, i.e., at the time of the peak asymmetric pressure load on the cavity wall.
The resulting membrane and oending loads in the hoop direction and shear loads were within the shield wall capacity.
Critical results from the analysis of the primary shield wall are given in Table 5.
The functionability of the wall was maintained for all loading conditions.
2.4.11 Steam Generator Suoports The steam generator supports were evaluated using the half-loop model described in Section 2.3.5. Applied loadings in the analysis were determined by the assumed 2A pipe breaks within the steam generator compartment.
Support components evaluated consist of the upper restraint trunnions and the lower support assembly (support skirt). Acceptability of the support components was based on a comparison of the presently evaluated LOCA loads to design basis LOCA loads determined from a previous analysis.
The design loads are greater in all cases for both upper and lower supports as shown in Table 5, indicating the structural integrity of the steam generator and its supports for the LOCA induced loads.
17
- 3. STAFF EVALUATION The staff evaluation of the effect of asymmetric loss-of-coolant accident loads due to postulated pipe breaks in the primary coolant piping, was accorplished by reviewing the licensee's submittal and using the independent audit calculations performed by the staff or their consultants.
The staff has concluded that the licensee's.assessment of the effects asymmetric LOCA loads is acceptable.
Since seismic loads on the primary system components and supports are substantially less than the asymmetric loads, the staff has further concluded that the components and supports are structurally adequate for both seismic and LOCA loads.
The
.staff evaluation of each specific analysis phase is addressed in subsequent paragraphs following the guidelines set forth by NUREG-06093 3.1 Subcooled 8lowdown Loads The subcooled blowdown.calculdtion portion of the Oconee asymmetric LUCA load submittal has been reviewed and is considered to be acceptable to Lte staff.
The basis of this acceptance is the staff's review and approval of the CrAFT2 computer code used for the internal hydraulic loads calculations.
Independent audit calculations for a 177-FA plant by the staff's consultant established approval for CRAFT2 application to subcooled blow:down.0 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 breaks consisted of reasonable and conservative values.
Nodalization and modeling were also developed in an applicable manner to the existing system.
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3.2 Cavity Pressurization Analysis The reactor cavity pressurization analysis of the Oconee plants for postulated breaks at the reactor vessel inlet and outlet nozzles has been reviewed and is considered to be acceptable to the staff. The primary basis of the acceptance is the staff's review and approval of the CRAFT2 computer code used for calculating LOCA cavity pressure loadings. Review of the input model indicated the licensee's analysis determined junction loss coefficients that were approximately twenty percent lower than if calculated using standard methods. However, the staff has concluded that the loss coefficients utilized are acceptable since the peak asymmetric loads are demonstrated to occur during critical flow, rendering the twenty percent difference insignificant.
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 assumption made by the licensee concerning the gravity multiplier of 2.0 and zero drag force used in the vent device blowout calculations is also acceptable to the staff. This procedure represents a conservative approach to the initial movement of the devices through out the time required to fully open the vents. The existence of a narrow vent space around the shield plugs, separating them from the main structure will reduce to a minimum any possible friction forces during a blowout, supporting the conservatism of the assumption.
The steam generator subcompartment pressurization analysis of the Oconee plants for postulated breaks at the pump suction and discharge nozzles and at the upper end of the hot leg has been reviewed and is considered acceptable. Acceptance is based on the staff's and their consultant's review of the data provided by the licensee and audit calculations 11 performed by the staff's consultant, EG&G Idaho.
The audit evaluated cavity pressurization for a generic B&W steam generator subcompartment. The audit results indicated no significant differential pressure on the steam generator except in one case: a full guillotine
(2A) break at the steam generator outlet nozzle, which-discnarges all tne escaping primary coolant into the support skirt. However, jet forces on the piping would eject the crossover leg from the support skirt, and the primary coolant discharge would expel into the steam generator compartment, resulting in insignificant asymmetric forces on the steam generator.
3.3 Structural Evaluation 3.3.1 Evaluation of -Methods and Models The structural computer codes cited in the licensee's report are considered to be acceptable to-the staff.
The many codes (STALUM, A NSYS, EDS-SNAP,
- STARS, NASTRAN, and SUPERPIPE) utilized in the presented LOCA analyses have been bench marked in a satisfactory manner to the staff.
The methods used in performing the-required structural analyses are acceptable to the staff in as much as they conform to the accepted state-of-the-art, standards, and regulatory codes. Based on the submittall'2 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.
Loads on the reactor.vessel support skirt resulting from the linear elastic system dynamic analysis exceeded the acceptable capacity of the skirt flange.. Acceptability of the component is based on the re-analysi-s of the support skirt considering the nonlinear behavior of the suppcrt ano adjacent components. Stress results -from the nonlinear analysis were determined to be within acceptable limits.
Seismic loads accounted for approximately four percent of the skirt load.
20
Results of the COM analysis indicated that stresses in the motor tubes, considering modified dimensions, exceeded the allowable by 4.5 percent. Based on the conservatism of the structural model and appied loads, the analysis is considered acceptable.
Analysis of the core flood line piping is acceptable based on the bounding analysis performed by the licensee. This analysis consisted of dynamic analyses of both core flood lines for RPV motion as determined frcm the RPV isolated dynamic system analysis.
3.3.2 Compliance with Acceptance Criteria The licensee's stress and/or load evaluation of the reactor system components is considered acceptable to the staff.. The criteria used in the evaluation are, in general, in agreement with industry standards and concur with the acceptance criteria outlined in NUREG-009.
Although some exceptions to the outlined criteria occur,. functionaoility of each analyzed reactor system component is demonstrated.
The licensee's stress and/or load evaluation of the reactor vessel internals, primary piping, and core flood line piping is acceptable since the appropriate ASME Code, Subsection NS and Appendix F, criteria are met.
The reactor vessel support skirt was shown to be acceptaole compared to ASME Code,Section III, Appendix F criteria baseC on a less conservative re-analysis of the skirt and adjacent components and the less stringent policy of excluding seismic loads. Portions of the skirt embedment configuration exceed allowables, but due to the localized nature of the 21
stresses and the-conservatism of the anafysis, the support embedment results are consi'dered acceptable, based on American Concrete Institute standard ACI 349-76 criteria.
For a postulated hot leg break, the hot leg piping is restrained from excessive movement by the bumper and collar restraints located around the lower elbow. A nonlinear analysis showed the pipe whip restraints to compare acceptably to allowable stresses from the ASiME Code,Section III, Appendix F for the restraints, themselves, and to allowable stresses from ACI 349-75 for the restraint embedments in the subcompartment wall.
Acceptability of the steam generator support evaluation is based on the comparison of the calculated support loads to design LOCA loads from a previous analysis. Upper and lower supports are adequate since the calculated loads are less.than the design loads.
As stated in NUREG-0609, Appendix E, two principal. acceptance criteria apply for the asymmetric LOCA:
(1) fuel rod fragmentation must not occur as a direct result of the blowdown loads, and (2) 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 analysis.
Maximum stress levels associated with the fuel rods and fuel assembly components other than grids are deternined by the licensee to be below AS-E Code, Suosection NG-3000 and Appendix F, allowaole values. Fuel rod fragmentation will., therefore, not occur.
Although a small number of spacer grids are predicted to experience some permanent deformation following an asymmetric LOCA, the effect of this grid distortion was conservatively incorporated into an appropriate ECCS analysis. The peak clad temperature predicted for this distorted geometry 22
r increased less than 120 F for peripheral bundles.
We do not believe it would be appropriate to apply the reduced flow area to the hot channel to demonstrate compliance. The evaluation provided for the PCT increase for the lower power 9
peripheral bundle adequately addresses 10 CFR 50.46 Acceptaoility of the CROM pressure boundary evaluation was previously shown by ccmparing the computed principal stresses to ASiE Code, Section 111, Appendix F criteria.
Based on the fuel system analysis, control rod insertion should not be significantly impaired.
Acceptance of the shield wall stress evaluation is based on compliance with Anerican Concrete Institute standard ACI 349-76 criteria.
In conclusion, there is reasonable evidence that the Oconee Units 1, 2 and 3 reactor systems would withstand the effects of an asymmetric LOCA event and still provide hich assurance that the reactor can be brought safely to a cold shutdown condition) 23
TABLE 1. HOT LEG BREAK PEAK CAVITY FORCE SUIMARY Resultant Vertical at Resultant Break Horizontal Peak Horizontal
- Moment, Area
-Force, lb Force pt, lb in.-1b 0.60A 5.0 x 106 0.20 x 106 2.9 x.10 8 TABLE 2. COLD LEG BREAK PEAK CAVITY FORCE
SUMMARY
Resultant Vertical at Resultant Break Horizontal Peak Horizontal
- Moment, Area Force, lb Force pt, lb in.-lb 65 2.OA 5.8 x 10~
5.0 x 10 3.2 x 108
TABLE 3. HOT LEG BUMPER RESTRAINT SPRING RATE DATA (a)
Stiffness, Coefficient
- Gap, Element Kios/inch of friction inches 1
437,500 0
0 2
41,000
.4.67 0
3 283,800 0
0 4
24,400
.5.05 0
608,000 0
4.0 6
67,500 4.00 4.0 7
608,000 0
4.0 8
67,500 4.00 4.0
- Mass, Node KiD-sec 2 1
0.081 2
0.081 3
0.081 4
0.081 E
.0.081 6
0.081 (a) For element and node numbers, see Figure 18.
25
TABLE 4. COLD LEG SHIELD RESTRAINT SPRING RATE.DATA (a)
Stiffness, Coefficient Gap Element Kips/inch of friction inches 1
190,000 (axial) 0.0 0.0 54,600 (in strong direction) 4,900 (in weak direction) 2 120,000 (all directions) 0.0 0.0 3
assumed rigid 0.40 5.0
- Mass, Node Kio-sec 2/ft 1
2 0.0308 3
0.0171 4
0.0089.
(a) For element and node numbers, see Figure 22.
26
iOLE 5.
UCUNEE SIUuCIUlAL RESUtSE Ali) SIRESS SUM4ARY Calculated Allowable Percent Com Location Value Value Margin Oasis of Allowable RPV support Support skirt flange 196400000 (t-lb 189000000 ft-lb
-4 ASME Code. Appendix F RIV support Dear ing plate (compression In 5949 psi 5560 psi
-7 ACI 349-16 emtbedinents concrete)
Uelow sole plate (cumpressive 0.0061 In/in 0.004/ In/in
-29 strain in concrete)
Anchor bolt pull-out above trench 5969000 lb 6350000 lb 16 (reinforced concrete in shear)
Service support Mounting flange 8670000 ft-lb 16400000 ft-lb 447 ASME Code. Appendix f structure CiUM Motor tube 63100 psi 60300 psi
-4.5 ASME Code. Appendix F fX.
4CR1 hous ing 1330000 ft-lb 1/70000 It-lb
- 25 Reactor Internals CSS upper flange 10200000 lb 191100000 lb 40 ASME Codle, Appendix f 1 lenuin cylitnder 2/514 psi 39120 psi
$30 Control rod guide tubes 14115 psi 21500 psi f34 Upper goide pad Joint 24615 psI 39l11 psI 131 Fuel assemblies fuel rod 4290 psi 46900 psi 49 ASHE Code, Subsection UG-3000 and Appendix F Lower end fitting 33000 psi 61050 psI
- 51 Spacer gr id gil4aifgd)
-Q4LA'Jo.
4Jr0
,4tufdt-Jio0.46 1.A1ppendix.6.
11sb ot -a11 e-
-37 Primary coo lant IOt leg pipIng Straight 26040 psi 50920 psi t56 ASME Code, Subsection 110-3650 Elbow 44960 psi 55920 psi, 20 and Appeondix F
IAIIIE 5.
(continued)
Calculated Allowable Vercent Comanent LocaL Ion Value alue Margin liasis of Allowabsle Cold leg Straight 24623 psi 61300 psI
'60 Elbow 30)02 psi 5U260 psI 141 Saf e-end 153) psi 52200 ps I iI Cold leg (half-loop nodel) at elbow near i(l1V 66130 ps 1 10100 ps 16 at IWV 65560 psI 135M psi 124 Vi'pe whip iot leg bumper restraint restraint SLeel member (shear stress) 12400 psI 239(0 psi 148 ASKE Code, Appendix F hot leg collar restraint inshedinent concrete (punching 634000 lb 150/000 lh 464 skear) co Core flood line Piling stress--Loop A 425 ASME Code, Subsection NO-3650 piping and Appendix F Support load--Loop A
-f-9 Piping stress--Loop a 44 ilological shield Delow hot leg penetration 5130l psi 63000 psI 9
ACI 349-16 wall (huoop stress)
Unlow core flod thne pnetration 51200 psI 6J00) psi
$19 (hoop stress)
Pedestal/cavity wall Interface 2/5 psi
- 31.
A's I
- 20 (shear stress)
Stelnl genleratoir Lower support sk irt at base Slupport Later al force 611110 h
11439000 lb 120)
Previous anialIysis and thinent 4?/1000 ft-lb 9111133010 ft-lb ASHEC Cundo.
AppendiLx F
TABLE 6. RPV SUPPORT SKIRT LOA05 Forces (Kios)
Mom.ents (Ft-Kios)
F F
F.
Mm areatk: Location x
y z
x y
z Hot leg at RPV 183 5493 3015 1921c2 1168 4786 Cold leg at RPV 3177 4774 4128 92303 1584 186374 29
Figure 1.
Pipe 3reak Locatians for Reactor Vessel Cavity Evaluation HOT LES 30
3' -1" 8'---"'
7' -7"
-9'-7" 6LEV_.t3 S. REACTOR.
I.VESE L1YEL1-2A P-AR 71-7" L Zv. 1 71 E 9 Figure 2.
Reactor Cavity, Elevation View 31
HOT L.:G 74s*
315*
.155.
520 Ca.0 I T 57 59 85 66 Fiur.3 e'to a105'la Ve 323
/
11 Figure 3.
Reactor Cavit-y, Plan View 32
X Direction 7.5 5.0 2.5 U)
O x
0.0N a'2 5
-5.0
-7.5 T
iI II IrI vIr iI 0.0 0.1 0.2 0.3 0.4 0.5 Time (seconds)
Figure 4.
Ilot Leg Break Time History.
1.0 Y Direction 0.8 0.6 C)
X 0.4 o 0.2 LA 0.0
-0.2-
-r Illi Irll II I
I 0.0 0.1 0.2 0.3 0.4 0.5 Time (seconds)
Figure 4. (continued)
Z Direction 0.0
-0.2 C- -0.6 0I O
LA.
0 0.0 0.1 0.2 0.3 0.4 0.5 Time (seconds)
Figure 4. (continued)
X Rotation
- 6.
00 2
C) 0 r
UJQJ O
0~0
-2 0.0 0.1 0.2 0.3 0.4 0.5 Time (seconds)
Figure 4. (continued)
1.0 Z Rotauion 0.5 C
0.0 4 J E
-0.5 0.0 0.1 0.2 0.3 0.4 0.5 Time (seconds)
Figure 4. (continued)
X Direction _
CD X
COm
-2 0
o Li.. -6 0.0 0.1
.O2 0.3 0.4 0.5 Time (seconds)
Figure 5. Cold Leg Break Time History
6-Y Direction __
5 4
In 2
0 0.0 0.1 0.2 0.3 0.J 0.5 Time (seconds)
Figure 5.
(continued)
1 Z Direction C0
-0 Figur 5.(otiud I
O
-2 0.0 fl.1 0.2 0.3 0.4 0.5 Timie (seconds)
Figure 5. (continued)
5X Rotation 0o 0.5-CD
-a 7
0.0
-I I
E
-0.5
-1.0-
-1.5-T iUFi T
i 0.0 0.1 0.2 0.3 0.4 0.5 Time (seconds)
Figure 5. (continued)
3-Z R~otationi 2
0 0
r C
0
-2
-3
-4 0.0 0.1 0.2 0.3 0.4 0.5 Time (seconds)
Figure 5. (continued)
prTeA GSSItifoot 4C'tW WaS.t I:igulre 6.
Steam Generator Compartment, Plan View
0**>.
Figure 7. Primary.Cootan -t
-System Schematic 44
Steam Generator IA Pumo 1A2 Pump Reactor Vessel Pump 132 Of P ump18 Steam Generator 1B Figure 8.
Primary Coolant System, Plan View 45
SGIB SGIA Pumpm 18 Pumip IA2 Node 51 Reactor Vessel (see Figure 10) figure 9.
HPV Isolated Nodel
4 1 41 sIi 00 t
0 U
-- / 7
~Al Li tipj 41,
- 04.
L.
I I' II II
-.t.s s
ii-----------------i9----'
-i E
UI--*-
U to I
0 t~E In
)
0 5,
(I-U
figure I].
Core Bounce Model PLIUNNUM CVIL.
WE ID u
O NL
.2 20 4/7
@1~MI GXu?
RID coat C)- MAS3 JOIt
_O
-mn1 6-6 5
-LlatAl
$FAIMS
-FUEL 30(&I2 loWg eago 31N L
Nu
-SAtf stalag
Figure 12.
FlwChart -for Generation o" Apolied Fcrming Functions on Reactor Ilessal and Int-aral s L40? AX40 li47RNALS CEMly SYS.
FU 10 Ft~4CALLC7IS~
PV C~iAR.LCMZJIS7IC,
- 57C, ANALYSIS QVT PuzuR AAAuxc1.
USNG"AL31SIX J.YS SNG23A~
.1 TrMALS PRESURE C ORE 3OWNCE FO AJ TiME~I~I~.TIME HISTORY I.
_x 4
R SPESR us IXG 0; x cs TIXE-HISI0R1E3
.S. c~vr 49
Figure 13.
Cavity Wall Model 50
Steam Generator Restraint 2 RPV Nozzle Restraint I Q
Figure 14.
Hot Leg Pipe Whip Model 51
Pump RPV Steam Generator I
t>
Shield Restraint Figure 15.
Cold Leg Pipe Whip Model 52
-Pump TA2 Pump TAl 400 Reactor Vessel Figure 16.
Half-Loop Model 53
S team Generator Outar Wall Insulation Gap Figure 17.
Hot Leg Bumper Restraint 54
Hot Leg Steam Generator
-gid beams Reactor Vessel Figure 18.
Hot Leg Bumper Restraint Math Model 55
.... 4 Section A-A Figure 19.
Hot Leg Collar Restraint 56
1500.(-IEN5 10N1 0.-
I.
CONTROL POINTS HUNDER L00D DEft.
thlPGI lilt)
I
-12004.0
-0.0613 10000.1-2 0.0 0.0000 3
6166.0 0.0630 06 074.0 1.7300
- 2. GPS 6000.0-
=Olt 1.040 I1 OCt 0.600 111
- 3. DYNf-IhlC NASS Im 0.070 1182/Fl
-1.00
-1.20
-0.00
- .60 1.20 1.00 DISPUICEllEHI (INI
--6000.0 C0tif0lEGO101 i50001.o Figure 20.
Axial Spring Rate for Collar Restraint
CONTROL POINTS NUMBER LOD DIEf.
cz
-2119.0
-0.2000 0
2 0.0 0.0000 3
2148.0 0.2000
- 2. ORPS 01 1.040 IN 1000.0-OCG 0.600 IN
- 3. DYNAMIC MASS 1i1 0.016 1102/F
-0 21
-0.14
-0.07 4 07 0.14 0.21 OISPLOCEltENII (1II)
-- 2000.0 00tWHESS10N
-3000.0 Figure 21.
Transverse spring Rate for Collar Restraint
Figure 22.
Cold Leg Shield Restraint A
Section A-A (D
gap
- 2.
3 Pipe Whip Model 59
iV Shall
.**.1.
CI.
Skirt
- ecleo bolts Coaicroto igure 23.
Reactor Vessel Skirt Support
Figure 24.
Reactor Vessel Support Skirt Model 61
8
.0O -
LINEAR MGDEL ASSJMPTCNS RESLTS FRCH Li)
NCNLINEAR CC..
g 4.0 Q
/
2.0 (K = 2.0 X 108 L./IN.)
0.0cc
.0,d
.0478
.072
.096
.120.
CELECTICN (INCHES)
Figure 25.
Reactor Vessel Skirt Force-Oeflection Relationship 62
4.00 LINEAR.00EL ASSUMPTIONS 3.00 RESULTS FROM NONLINEAR MODE S2.00 -
1.00
(= 1.8 X 1o1 IN L /RAG)
S 0.00 a
G4 0.0000&
.00107
.00213
.00330
.00427 Rattaion (Raai ins)
Figure 26.
Reactor Vessel Skirt Moment-Rotation Relationship 63
Figure
- 27.
Reactor Vessel Skirt Emoedment Detail 64
Figure 28.
Reactor Vessel Skirt Nonlinear Embedment Model It Nlonl Inear Coperaei Trpaq Force Force a
diedi.
Figure 29.
Reactor Vessel Skirt Embedment Substructure Model 1
I f1 66,
12.5 10.0 7.5 5.0 2.5 0.02 0.04 0.06 Translation (inches)
Figure 30.
RPV Skirt Embedment Lateral Force-Translation Relationship 67
200 150 OC) x coc 100 50 0.25 0.50 0.75 1.00 Rotation (radians) x 10-3 Figure 31.
RPV Skirt Embedment Moment-Rotation Relationship
gil T IA
-9:
CT Core affle Pate T
I T
I 1
ona b
(
us
/
I-I-*
Ie I-e Io4-
/
/
- 4 I
I -11
/ ~
T*-
Core Baffle PlateLLLL
RPV Nozzl e Core Flood Tank Containment Penetration Figure 33.
Loop A Core Flood Line Math Model 70
- RPV Nozzl e Anchor Figure 34.
Loop 8 Core Flood Line Math Model 71
Figure 35.
Service Support Structure Model m.7 72
Fiore
- 36. LCtt'I Dynamiic Hadel of Type A D~rive 1-25 LIAOSCII[U 51-.5 LIADICUEN SUPPORT lull 44-41 SPACII, INDUH I IARING Uv
)
oa 101-100 IONQUI lull 0oi(j 1 161-311 Mool ug(
.910.
..list
.061 a A.sa,.8136 all6 lea 61I 2
91i I
oi I
al11 1 *I11 0
$Is am IIs polC oas Vd oAblrAI11 NilNIIAA isAffll Taa uljm all smI Id~
_i oi 1141.
1
,drl
- 4. REFERENCES
- 1. -"B&W 177-FA Owners Group-Effe-ts of Asymmetric LOCA Loadings-Phase II Analysis", BAW-1621, Babcock and Wilcox, July 1980.
- 2.
"B&W 177-FA Owners Group-Effects of Asymmetric LOCA Loadings-Phase.II Analysis--Supplement 1, Responses to NRC Questions",
BAW-1621, Supplement 1, Babcock and Wilcox, June 1981.
- 3. "Asymmetric Blowdown Loads on PWR Primary Systems", NUREG-0609, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, January 1981.
- 4. "Reactor Coolant System Hydrodynamic Loadings During a Loss-of Coolant Accident", BAW-10132P-A, Babcock and Wilcox, May 1979.
- 5. "Mark C Fuel Assembly LOCA-Seismic Analyses", BAW-10133P, Rev. 1, Babcock and Wilcox, May 1979 (Proprietary).
- 6. T.L. Bridges and R.W. Macek, "Review of Babcock and Wilcox's Fuel Assembly Structural Analysis Topical Report BAW-10133P, Rev. 1,"
EGG-EA-5933, EG&G Idaho, Inc. July 1982.
- 9. B.W. Sheron (NRC) memo to C. H. Berlinger, "PCT Increase due to Asymmetric LOCA Loads in Oconee Units 1, 2, and 3", dated January 25, 1983.
- 10. J. C. Watkins, et al., "Subcooled Blowdown Analysis for a Babcock and Wilcox BSAR 177 Pressurized Water.Reactor", RE-A-78-130 (Rev.
1), EG&G Idaho, Inc., September 1978.
- 11.
0..W. Golden, "Cavity Pressurization Analysis of a B&W Plant Steam Generator and Pump Subcompartment", EG&G-EA-5102, EG&G Idaho, Inc.,
February 1980.