ML20055C291
| ML20055C291 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 02/20/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20055C290 | List: |
| References | |
| NUDOCS 9003020004 | |
| Download: ML20055C291 (12) | |
Text
W,-
n q
1 EHCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIO ASYMMETRIC LOCA LOADS - MAINE YANKEE NUCLEAR POWER PLANT MECHANICAL ENGINEERING BRANCH DIVISION OF ENGINEERING TECHNOLOGY
1.0 INTRODUCTION
On May 7, 1975, the Nuclear Regulatory Comision (NRC) was informed that asymmetric loading on the reactor vessel supports resulting from a postulated reactor coolant pipe rupture at a specific location (vir'., 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 effset shear pipe break at the vessel nozzle,'asymetric loading could-result from forces induced on the reactor internals by transient differential pressure across the core barrel and by forces on the vessel
-due to transient differential pressure in the reactor cavity.
With the 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 l
the reactor vessel supports and other components, thereby possibly affecting-their integrity. Although this potential safety concern was first identified ~during the review of the North Anna facilities, it was determined to have generic implications for all 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 asymetric loads had not been considered in the design of any PWR primary system.. In' June 1976, the.NRC requested i
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 j
a supported no analyses due to the low probability of the pipe breaks at a
=particular location. Although'the NRC staff recognized some merit in these proposals, they determined that.the more fundamental questions still remaired unanswered. Therefore, licensees of.PWR plants were notified by letter dated January 20, 1978 that the evaluation of'their primary systems for asymetric LOCA loads would be required.
U Although the NRC staff's original emphasis and concerns were focused
/.
primarily on the integrity of the reactor vessel support system with respect to postulated breaks inside the reactor cavity (i.e., at a nozzle), it became apparent that significant asymetric forces could also i
be generated by postulated pipe breaks outside the cavity and that the scope of the problem was not limited to the vessel support system itself. The staff, after reviewing this problem, determined that a
- 88R888e388Sjjgv h
l
Vs, a y
' w
\\
.r 1
reevaluation of the primary system integrity of all PWR plants to withstand these loads was' necessary. Therefore in January of 1978, the NRC staff
. requested each PWR licensee to submit additional information in accordance i
with the expanded scope of the problem specifying a minimum number of pipe break locations to be addressed and the reactor system components to be evaluated.
The asymetric-loading that was determined by the NRC to have generic I
implications for all PWRs was formerly identified in Task Action Plan A-2 Unresolved Safety Issue (USI), "Asymetric Blowdown Loads on Reactor '
i Primary Coolant System," as published in NUREG-0371, " Task Action Plans for Generic Activities (Category A)," USNRC, November 1978. Since the
-identification of the asymetric 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 reactor coolant loop (RCL) for each of the three Based on these analyses' and additional NRC staff. major reactor vendors).
investigations, criterie and guidance for conducting an evaluation of asymmetric loss-of-coolant-accident (LOCA) loads were developed. USI A-2 was resolved in January-~
1981 with the publication of NUREG-0609 (Reference 1). This document provided an acceptable basis for performing and reviewing plant analyses for asymetric LOCA loads and affected all operating and future PWP.s.
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 l
new technical work it was demonstrated that the new-techniques for the analysis of piping failures assured adequate protection against-failures in primary system piping in Pressurized Water Reactors. This was reflected in a revision to GDC-4 published in the Federal Register in J.
final form on April 11,-1986, and in a further revision to GDC-4 published in the Federal Register on July 23, 1986.
In addition, it has also been n
satisfactorily demonstrated in the course of the A-2 effort that:there is a very low likelihood of simultaneous pipe loading with both LOCA and SSE loads.
l
-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 asymetric loading on the core structure resulting from primary system piping LOCA, need not be considered in the design of the core structure.
~
In June of 1980, Combustion En LOCA loads evaluation report (gineering (CE) submitted a final asymetric Reference'2), applicable to the Maine Yankee Power Plant. This material, submitted in response to the January 1978 letter from NRC, was reviewed by the NRC staff and its consultants.
Upon
,o4
\\'
.? *
['
. 1 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 3) was submitted in August of 1981.
4 x
Maine Yankee Atomic Power Company's submittal of September 26,1 1984 (Reference 5)inres earlier submittals (ponse to NRC concerns (Reference 4) in addition to References 6,7) represent the' limiting cases asymmetric LOCA loads evaluation and have been reviewed in conjucfor the tion with the criteria outlined in NUREG-0609. Subsequent sections of this l
safety evaluation report summarize the evaluations performed by the licensee for subcooled blowdown loads, cavity pressurization, and struc '
tural res analyses.ponse. ' Following this is the staff's evaluation of the licensee's The staff's evaluation includes the assessment of the licensee's compliance with acceptance criteria.
2.0 DISCUSSION Maine Yankee's approach toward resolving the asymmetric LOCA loads issue-is based on two parallel programs.
1.
Installation of pipe rupture restraints to limit the blowdown flow area..
~
2.
Conduct an analysis program to address both classical and asymmetric LOCA load effects based upon original design allowable limits, using existing analyses and simplified comparative studies wherever possible.
1 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 i
pressures, and (c). Calculation of the loads and stresses on the various components'and supports of the RCS~which include the vessel and steam-generator supports, vessel internals, fuel assemblies, control element drive mechanisms (CEDM) and emergency core cooling system -(ECCS) piping.
s 2.1 Thermal-Hydraulic Loads Analysis L
L The generic thermal hydraulic analyses and internals model used for the L
Paine Yankee analyses is fully described in Reference 2.
The generic model is based on the St. Lucie design. The major differences between the internal configuration of St. Lucie and Maine Yankee are the preser.ce of a thermal shield, not originally modeled for St. Lucie, and the fact that in Maine Yankee there are three hotleg nozzles obstructing passage of rarefaction and compression waves around the downc er, instead of two T
n-
d E.., (
y x
g 4
i k
in St. Lucie. Consideration of acoustic impedance presented by the three smaller nozzles of Maine Yankee instead of two larger ones, indicated that this effect is small-and can be neglected. Acoustic impedance effects were considered by using rarefaction and compression waves of constant amplitude. The acceptability of the manual procedure was verified by comparing calculated differential pressures with data from RELAP-4 and WHAM 6-computer codes. A later thermal hydraulic model, modified to account for the presence of the thermal shield, had produced asymmetric loads across the barrel and vessel which are less than 10 percent higher than those predicted without the thermal shield The Maine Yankee internal asymmetric loads computed from the generic model have been increased by 10 percent to 3
account for this effect.
I The break locations, break areas and oreak opening times for the design basis breaks utilized for computing the internal system asymetric loads are the same as thoge used for the cavity analyses (external asymmetric loads), i.e.188 in ; 8.2 msec opening time; any one of the three cold leg nozzle safe ends.
2.2. Cavity Pressurization Analysis r-Breaks resulting in cavity pressurization, which were utilized in asymetric load analyses, are cold leg 2 guillotine bgeaks having the 2
following break areas; 100 in, 144 in, and 188 in. Break location is that of any of the three. cold leg nozzles, since the cavity is asymmetric.
Cold leg breaks'are chosen since they produce _ the largest asymmetric loads, b
Break areas and opening times for guillotine breaks are inter-dependent, l
since opening-time relates to the stiffness and mass of the pipe which is constant and area relates to-separation distance between the two ends which is in turn related to time and the stiffness. For cavity analysis purposes, the maximum pressure differentials are achieved for larger-breaks, even though their opening time would be proportionately longer.
This is because the opening time for any size break is shorter than the time at which peak' differential pressures are achieved.
For this 2
reason the 188 in break was chosen as the-design basis break. Opening i
time does_ not significantly affect the internal-asymmetric force as long as it is properly correlated to the break area, i.e., a full area break requiring 35msecfor_ full'openingwouldproduceintgrnalasymmetricloads approximately equal to-those resulting from a 188 in break requiring 0.2 nsec to open. Similarly, opening area and time have no effect on thurst forces frog the break (tension release force). Thus, it is proper to uso the-188 in break, opening in 8.2 msec, as the design basis break, since k
it is the largest break area for a given opening time.
Aftercompletionoftheanalyses,itwasdetegminedthatthemaximum achievable break size is between 60 and 90 in for both hot and discharge legs. Results of studies for similar cavities indicate that peak horizontal forces acting across the vessel from cavity pressurization are directly w
wcvm-
-r I
proportional to the break area for small breaks and proportional to the square root of the break area for larger breaks. The horizontal force is one of the dominant loads in determining the response of the sy". tem to asymmetric loads.
In actuality, the cavity pressure asymetric load at Maine Yankee would be approximately 1/2 of the value used in the analysis.
Thus, the analyses performed are quite conservative.
The' cavity analysis (specific to Maine Yankee) was perforced with the RELAP-4 computer code utilizing the containment option to account for air pressure. To determine the critical flow for air-steam-water mixtures, the. thermal homogeneous equilibrium model was used. Other inputs to the RELAP-4 are mass and energy releases, volumes of the nodes and. junction
. inertias. Additional information on cavity modeling and input parameters can be found-in Reference 2 which used the same approach for other CE Owners Group plants. Mass and energy release data was calculated with the RELAP-4/M003 computer program. The licensee checked the validity of this data against that of breaks of comparable size calculated by CE for other-plants.
.i The nodalization scheme for the subcompartments was chosen so as to consider all physical flow restrictions as_ divisions between subcompartments. This scheme had been shown by previous sensitivity studies conducted on. several different cavities, to result in conservative calculation of forces and.
i moments'on the reactor vessel. Such studies had suggested that there.is
.no more than a 10% uncertainty in the results. This uncertainty is con-L sidered insignificant when compared to the conservatism of utilizing a a
break area which is more than twice that which can actually occur as a l
result of the hardware installed to prevent large breaks. Whenever suffi-L cient structural details were unavailable, volumes and vent areas were.
reduced by 10 percent to account for the presence of small structures, equipment, etc.
Several vent areas (those separating the cavity f f
.the piping penetrations, and the access tunnell. rom the re ueling pool; have been opened as a function of time. The neutron shield which exits at the flange and in.
L the penetrations was assumed to conservatively begin blowing out 15 msec after the start of the accident and provide total flow area by 65 msec
~ fter the event. Opening was assumed to be linear. This time is based L
a l.
upon results obtained for a preliminary neutron sheild design for St.
Lucie I which was similar to Maine Yankee's design. The calculated L
complete blowout time was shorter than 65 msec, hence the calculated L
cavity pressures and pressure differentials are higher than those likely H
to occur. The hatch over the access tunnel was assumed to open at 1.0 psi differential with a nominal 0.1 msec' linear delay.. This value was predicated on the ability of the hatch to withstand a uniform live load of o
150 lb/ft' and is not critical to the results of the analysis.
~ '
]
y
~
~2.3 Structural Analysis The licensee's structural analysis was performed utilizing two primary.
finite element models and several detailed component and support 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 asymmetric LOCA loads were applied are described in the following subsections.
P.3.1 Reactor Coolant Piping Pipe rupture restraints had " hot gaps" ranging from 3/16" to 1/2".
" Cold" gaps and piping thermal deflections were used to estimate " hot" gaps.
These were field verified using "Go-No-Go" guages when the plant her.ted up.
-i Piping transient deflections and resulting pipe restraints were computed on the basis of analytical gaps corresponding to those which were set in the field.- Finite element models were developed to simulate the composite stiffness of restraints, foundation walls and the local pipe including deflection effects..
' Reactor coolant piping.and stiffness characteristics were modeled on the STARDYNE and ANSYS codes. Stresses from RPV responses to LOCA, as well as differential pressure loads throughout the RCS piping were computed using
. dynamic time history analysis techniques. RCS pipe stress results were i
within original design stress limits as given in the Maine Yankee FSAR.
L 2.3.2 Steam Generator And Reactor Coolant Pump Supports l '
Six breaks were analyzed outside the cavity.
Guillotinebreaks(full size) were postulated to occur at the discharge nozzle of the pump and l
1 the inlet nozzle of the pump, the steam generator inlet and outlet nozzles, and the pressure surge and spray nozzles.
Steam generator and RCP' support-loads were determined from transient displacements at the support points. The transient displacements were obtained by applying the trancient motions of the RPV to models simulating l;
piping between the RPV and the steam generator and/or RCP supports. Peak D
displacements of the steam generator and RPV supports were converted to forces by multiplying the. displacements times the. stiffness matrix' values
'which simulated the~ support. The STARDYNE computer program was used to i~
. simulate RCS piping and the RCP support stiffness. The sum of the normal
. operating, seismic and LOCA loads were found to be less than 90% yield.
All stresses'were within the allowable limits.
2.3.3 Steam Generator Subcompartment Walls Local peak differential pressures across steam generator (SG) sub-compartment walls were estimated from the subcompartment break calcula-tions,. Peak differential pressures were compared with subcompartment wall differential pressure capacity given in the Maine Yankee FSAR and the walls were judged to be acceptable.
1
~
, t 2.3.4 ECCS Piping And Supports' ECCS piping and supports were simulated using the STARDYNE and ANSYS computer programs. Transient deflections from the RPV, transmitted to 1
ECCS piping attachments points on RCS piping were applied to the ECCS 1
piping models and'the response was determined by dynamic time history analysis. Supports were evaluated to original design criteria and included the IE Bulletin 79-02 concerns over the base plate flexibility prying action effects.
ECCS piping and supports were fcund.to be~within the stress limits-of the ASME Code, Subsection NC-3611.2 for service levels C and D.
2.3.5. Reactor Internals And Fuel The nodalization model of the reactor vessel and internals was very similar to that employed by Combustion Engineering in Reference 2, except that the Maine Yankee model included representation of'the thermal shield.
Thrust forces and cavity forces was applied.to the node points on the vessel.
Internal asymmetric forces were applied to the node points on the barrel and vessel. The applied loads to the core. supports, barrel-and internals from the asymmetric LOCA analysis were compared with plant specific analyses conducted prior to the asymmetric LOCA analyses. The prior analyses demonstrated the acceptability of the internals and fuel for loads greater in magnitude than those from the asymmetric LOCA analysis. - On the basis of this comparison, it was concluded that internals and fuel would be structurally acceptable under the action of asymmetric LOCA loadings.
2.',6 Control Element Assemblies (CEAs)
The RPV transient displacement history was applied to the CEAs and dynamic time history response of the CEAs were computed.
Resulting stresses were within.the original allowable limits specified in the Maine Yankee FSAR.
L 1:
l L
L I'
w
,y
- l* :
t 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 submittals (References 4, 5, 6, 7, 9, 10, 11, 12, 13 and 14) and using the independent audit calculations performed by the staff or their consultants (References 8 15 16).
In general, the staff has concluded that the
.. licensee's ass,essm,ent of the problem is acceptable. The staff evaluation of each specific analysis phase is addressed in subsequent paragraphs, following the guidelines set forth by NUREG-0609.
~
3.1 Thermal Hydraulic Blowdown Loads The thermal hydraulic blowdown calculation portion of the Maine Yankee 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 RELAP-4 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 RELAP-4 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 pres-sures and loads. A significant number and location-of postulated pipe breaks were. analyzed to determine worst case loadings on the primary i.
coolant system. Size and length of break openings consisted of reasonable I
and realistic values. Nodalization and modeling were also developed in a
. manner that provided reasonable representation of the existing system.
3.2 Cavity Pressurization Analysis L
The licensee's reactor cavity pressurization analysis for postulated E
breaks at the reactor vessel inlet and outlet nozzles has been reviewed L
and is considered acceptable to 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 SG subcompartment pressurization _
analysis of the Maine Yankee 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 I
its previous review and approval of similar codes for calculating LOCA t
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.
L 1
s
Ej b
. l 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, STARDYNE, NASTRAN, and ANSYS utilized in the LOCA analyses have been bench marked in a satisfactory l
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 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 safe shutdown earthquake (SSE) loads in the design of the RCL iping. The staff believes that there is sufficient technical evidence p(Reference 17) which demonstrates that the SSE and LOCA for the main l piping in PWR plants may be considered as independent events in determining the appropriate combination of the effects of. accident t
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.
t The results of the LLNL investigations-indicate that the probability of a direct seismically induced DEGB is extremely small. The best estimate
.i probabilities of direct DEGB using the med modelinguncertaintiesrangesfrom5~x10~gnsofthedgtributionofthe to 7 x 10~
per plant year for both Westinahouse and Combustion' Engineering plants.
From the uncertainty analysis, considering the whole range of modging uncertainty, it.is concluded that a direct DEGB probabilty 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 e plant,themedianprobabilityofDEGBis3.3x10~gpacityWestinghouse per plant year. The 10"gespondingindirectDEGBprobabilityatthe90thpercentileis2.3x cor 1
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 Combustion Engineering plants are also very low. The upper bound probability values for the Combustion Engineering
9 n plants are comparable with those of the Westinghouse plants. Based on the analyses submitted by the licensee and independent assessments by the staff and its consultants, the staff has concluded that the licensee has provided adequate justification for the documented deviations from the requirements of the Standard Review Plan 3.9.3.
The results of the above probability studies provides added assurance that these deviations are acceptable. The instability approach in the analyses of the RCS supports, CEDM, and ECCS piping is acceptable since it complies with ASME Code, Section Ill, 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) 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, EG&G Idaho, Inc. 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 assemblies. Determination of the total stresses in the core barrel resulting from the asymmetric downcomer depressurization using decoupled beam and shell modes is acceptable since this procedure has been shown to be exact for linear analyses.
3.3.2. Compliance with Acceptance 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 accept-ante criteria outline in NUREG-0609. Although some exceptions to the outlined criteria occur, functionality of each analyzed reactor system component is demonstrated. Since the pipe rupture restraints limit the actual loads to approximately fifty percent of the loads used in the analysis, it is determined:that the original allowable limits, per the Maine Yankee FSAR, have been met.
The licensee's stress and/or load evaluations of the reactor vessel, supports, internals, primary piping, CEDMs, and ECCS piping is acceptable since ASME Code, Appendix F criteria are met.-
Acceptability of the steam generator support evaluation is based on the comparison of the calculated support loads to design loads, yield capaci-ties, 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.
CONCLUSION
-In conclusion, there is reasonable evidence that the Maine Yankee reactor system would withstand the effects of asymmetric LOCA loads and that the reactor could be brought to a cold shutdown condition safely.
l
r r.
REFERENCES i
l ~
1.
"Asymetric Blowdown Loads on PWR Primary Systems " NUREG-0609, U. S.
Nuclear Regulatory Comission, Office of Nuclear Reactor Regulation, January 1981..
2.
" Reactor Coolant System Asymmetric Loads Evaluation Program--Final Report--Calvert Cliffs 1 and 2, Fort Calhoun, Millstone 2, Palisades,"
Vol. 1,2 and 3, Combusion Engineering, Inc., June 30, 1980.
3.
" Response to Questions on the Reactor Coolant System Asymetric Loads Evaluation Program--Final Report," Vol. 1,2, and 3, Combusion Engineering, Inc., June 1981.
(Volume 3 Proprietary).
4.
USNRC Letter to'MYAPCo dated May 31, 1984, " Asymmetric LOCA Loads Operating Reactor Licensing Actions - Sumary of Concerns of Independent Plant Safety Evaluations" 5.
MYAPC Letter to USNRC, " Maine Yankee Asymetric LOCA Loading Evaluation",
dated September 26, 1984-6.
MYAPC Letter to USNRC, " Maine Yankee Asymmetric LOCA Loading Evaluation",
dated May 14, 1980 7.
MYAPC Letter. to USNRC, " Maine Yankee Asymetric LOCA l'aading Evaluation",
dated July 1, 1980 8.
" Reactor. Coolant System Asymetric LOCA Load Evaluation" St. Lucie 1 Docket 50-355, Feb. 1980.
Rev. 1, 8, 1980.
9.
" Method for the Analysis of Blowdown Induced Forces in a Reactor Vessel,"
CENPD-252-P,' Combusion Engineering, Inc., December 1977. (Proprietary).
10.
" Method for the Analysis of Blowdown Induced Forces in a Reactor Vessel,"
CENPD-252-P,) Amendment 1-P,CombustionEngineering,Inc., August 1978.
(Proprietary.
- 11. " Topical Report on Dynamic Analysis of Reactor Vessel Internals Under loss-of-Coolant Accident Conditions with Applic: tion of Analysis to CE 800 MWe Class Reactors," CENPD-42, Combusion Engineering, Inc.
(Proprietary).
- 12. Hydrodynamic mass presentation to the NRC on March 14, 1979. Joint effort between EBASCO and B&W owner group.
Report B&W 177-FA owners b,
9roup.
13-Shaaban, S. H., " Implicit Three Dimensional Finite Element Solution to the Fluid-Structure Interaction Induced by Hydrodynamic Accidents".
PVP-39, ASME June 1979.
. 14. Shaaban, S. H. "ADMASS USERS MANUAL". EBASCO Services Inc. program 2361, August 1979.
.t.
,.7
?
~ C~
2 15.
'J. C. Watkins, "Subcooled Blowdown Analysis for a Combustion Engineering 2570 MW Pressurized Water Reactor," RE-A-78-248, EG&G Idaho, Inc.,
November 1978.
16.
T. L. Bridges, " Review of Combustion Engineering. Inc., Fuel Assembly Structural Analysis Topical Report CENPD-178-P, Rev. 1-P," EGG-EA-5824, j
EG Idaho, Inc., March 1982.
J
- 17. Lo, Woo, Holman, and Chou, Lawrence Livermore National Laboratory.
" Failure Probability of PWR Reactor Coolant Loop Piping" (UCRL-86249) presented at the ASME Pressure Vessel and Piping Technology Conference, San Antonio, Texas, June 1984 i
j-
,