ML20237B517
| ML20237B517 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/31/1987 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20237B515 | List: |
| References | |
| WCAP-11639, NUDOCS 8712160319 | |
| Download: ML20237B517 (47) | |
Text
$
WESTINGHOUSE CLASS 3
{
WCAP-11639 UPFLOW CONVERSION SAFETY EVALUATION REPORT BEAVER VALLEY UNIT 1 i
October, 1987
~
Westinghouse Electric Corporation j
Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 j
l 8712160319 871207 PDR ADOCK 05000334 i
P PDR 1030v:1D/103087 7
.______________J
WESTINGHOUSE CLASS 3 TABLE OF CONTENTS SECTION TITLE PAGE 1.0
SUMMARY
1-1
2.0 INTRODUCTION
2-1 3.0 LICENSING APPROACH AND SCOPE 3-1 4.0 00WNFLOW CONFIGURATION 4-1 5.0 UPFLOW CONFIGURATION 5-1 6.0 UPFLOW MODIFICATION 6-1 7.0 UPFLOW DESIGN BASES AND EVALUATIONS 7.1 Fuel Assembly 7-1 7.2 Reactor Internals 7-4 7.3 Core Barrel Plug Design Functional Requirements
'7-5 8.0 ACCIDENT EVALUATIONS / ANALYSES
{
8.1 Loss-of-Coolant (LOCA) 8-1 I
8.2 LOCA-Rolated Evaluations 8-10 8.3 Non-LOCA 8-23
9.0 CONCLUSION
9-1
10.0 REFERENCES
10-1 1030v:10/103087 i
WESTINGHOUSE CLASS 3 f
LIST OF FIGURES FIGURES TITLE PAGE
~
4-1 Downflow Configuration Flow Path 4-2 5-1 Upflow Configuration Flow Path 5-2 1030v:10/103087 ii L___--_-_-____-____.
~
WESTINGHOUSE CLASS 3 LIST OF TABLES TABLE TITLE PAGE 8.1 Input Parameters Used in the LOCA Analyses 8-16 1
8.2 Containment Parameters used in the C0C0 Code 8-17 8.3 Large Break LOCA Results 8-19 8.4 Large Break LOCA Time Sequence of Events 8-20 8.5 Small Break LOCA Results 8-21 8.6 Small Break LOCA Time Sequence of Events 8-22 9
l 1
1030v:10/103087 iii
WESTINGHOUSE CLASS 3 1.0 SUGARY This Safety Evaluation Report demonstrates that no unreviewed safety concern is involved in converting the reactor vessel downflow design to an upflow design, and the elimination of the fuel thimble plugs assuming that the 10%
steam generator plugging has been incorporated in the NSSS design.
The upflow conversion consists of changes to the reactor vessel components, which are to plug the core barrel inlet flow holes and to provide holes in the top former plate. This changes the flow path from being downflow between the core barrel and baffle plate to upflow and has the effect of increasing the i
core bypass flow from 4.5 to 6.5%.
Changing the flow path reduces the pressure differential across the Baffle plates eliminating the jetting of coolant between the joints between the baffle plates. The elimination of the thimble plugs effects a decrease in active core flow since the hydraulic resistance through the bypass region is decreased, thereby also increasing the core bypass flow.
l The effect of these changes have been evaluated, which based on the reactor internals, the fuel assembly integrity, the core barrel plug, the thermal hydraulic design analysis, and the appropriate LOCA, non-LOCA and SGTR postulated accidents. The FSAR revisions resulting from this safety evaluation have been transmitted separately via DLW-87-667. The 10% steam generator tube plugging report (WCAP-11591) has also been transmitted via DLW-87-641.
I l
d 1030v:1o/103087 1-1
1 WESTINGHOUSE CLASS 3
2.0 INTRODUCTION
In an effort to reduce the potential for fuel / fuel assembly damage, Westinghouse has developed a program to reverse the reactor coolant flow pattern in the core barrel and baffle region of the reactor vessel. A reversal of coolant flow in this region will significantly reduce cross-flow jetting through a decrease in the differential pressure across the baifle plate. Significant evidence exists identifying cross flow jetting as a potential contributor in fuel / fuel assembly damage at the periphery of the core. As a result, the Duquesne Light Company (DLC) has decided to pursue this program at their Beaver Valley Unit 1 Nuclear Power Station.
In addition, DLC has also requested that the program include the elimination of the fuel thimble plugs.
It is the fuel thimble plugs that limit bypass flow through rod cluster control guide thimbles in fuel assemblies which do not contain either control rods, source rods or burnable poison rods.
It should.be noted that the primary impact of fuel thimble plug removal is the increase in core bypass flow. However, this increase in core bypass flow does not significantly alter critical plant operating parameters and as a consequence does not represent a config'uration which could compromise the safety of Beaver Valley Unit 1.
w~
1030v:10/103087 2-1
i WESTINGHOUSE CLASS 3 3.0 LICENSING APPROACH AND SCOPE l
The conversion of Beaver Valley Unit 1 from a reactor vessel downflow j
configuration to a reactor vessel uoflow configuration represents a change to j
the plant. Chapter 10 of the Code of Federal Regulations, Section 50.59 (10 CFR 50.59) allows the holder of a license authorizing operation of a
)
\\
nuclear power facility the capacity to initiate certain changes, tests, and experiments not described in the Final Safety Analysis Report (FSAR).
Prior Nuclear Regulatory Commission (NRC) approval is not necessary to implement the modification provided that the proposed change, test, or experiment does not involve an unreviewed safety question or change in the technical
{
specifications incorporated in the license.
It is however, the obligation of j
the licensie to maintain records of changes, tests, and experiments to the facility to the extent that such changes impact the FSAR.
10 CFR 50.'59 further stipulates that these records shall include a written safety i
evaluation which provides the bases for the determination that the change, test, or experiment does not involve an unreviewed safety question.
It is the purpose of this document to support the requirement for a written safety
)
, evaluation providing testimony that an unreviewed safety question will not result as a consequence of the described change to the plant.
1 The scope of this safety evaluation is confined to a brief description of the i
existing reactor vessel configuration with respect to coolant flow paths, the intended upflow design modification, and finally a thorough presentation and evaluation of the affected Beaver Valley Unit 1 reactor vessel lower internals and fuel system Final Safety Analysis Report (FSAR) design bases, including an evaluation of these bases.
In addition, a discussion of the impact upflow has on the current LOCA and NON-LOCA design limits, assuming that both the thimble plug removal and the 10% steam generator tube plugging have been incorporated in the NSSS designr is also included in this' report.
1030v:10/103087 3-1
WESTINGHOUSE CLASS 3 4.0 00WNFLOW CONFIGURATION The Beaver Valley Unit i reactor vessel lower internals assembly consists of a core barrel into which baffle plates are installed, supported by interconnecting former plates. A lower core support structure is provided at the bottom of the core barrel.
The thermal neutron shield hangs from and surrounds the periphery of the core barrel.
The components which comprise the lower internals assembly are precision machined with the baffle and former plates being installed into the core barrel by bolting.
In addition to supporting the core, another function of the reactor vessel lower internals assembly is tg direct coolant flows within the vessel.
While it directs the primary flow through the core, the internals assembly also establishes secondary coolant flow paths for cooling the upper regions of the reactor vessel and for cooling the lower internals structural components.
Some of the parameters influencing the mechanical design of the lower internals assembly are the pressure and temperature differentia'Is across its component parts and the flow rate required to remove the heat which is generated within the structural. components due to radiation (e.g., gamma heating). The configuration of the lower internals provides for flow into the baffle-barrel region to ensure adequate cooling capability. The configuration also maintains the thermal gradients within and between the various structural components (resulting from gamma heating and core coolant temperature changes) within acceptable limits.
The Beaver Valley Unit i reactor vessel lower internals configuration which incorporates downward coolant flow in the region between the core barrel and the baffle plates, is conceptually depicted in Figure 4-1.
In this configuration, the coolant flow exits the reactor vessel inlet nozzle and passes the thermal _ neutron shield, turns and flows up through the core region. A portion of the downflow stream is diverted to cool the barrel / baffle region. This diverted flow passes through holes in the core barrel at an elevation between the top two former plates and flows downward through holes in the lower former plates. At the lower core plate, it 1030v:lo/103087 4-1
WESTINGHOUSE CLASS 3 I
THERMAL NEUTRON
{
SHIELD n
I FLOW i
j; o
{
Am /
E
[f
[
]
a t
5,-
)
j l kj t
. p 5
l' 3
i i
e
-l-SAFFLE gg rLATE I
/N
- FORMER s,s,
.i ruTe
& &s JL i
.is b5 f}f lO j
.s as t
47
&A%-
I
-E
-ts s) 2
&A&A
.J[.
3 f'I Icq.
[O.
h N. N.
FIGURE 4-1 00WNFLOW CONFIGURATION 1030v:10/103087 4-2 s
WESTINGHOUSE CLASS 3 combines with the main coolant stream flowing up to the core region. The distribution of the hydraulic pressure differentials in the downflow configuration is such as to cause any leakage through the baffle joints to flow into the core region and results in a relatively high pressure differential across the baffle plates at the upper level of the core region.
)
1030v:10/103087 4-3
WESTINGHOUSE CLASS 3 5.0 UPFLOW CONFIGURATION Westinghouse has developed and qualified a program for field modifying the reactor vessel lower internals assembly to reduce the potential for fuel rod damage resulting from haffle joint jetting. With this modification the coolant downflow path in the baffle / barrel region, as shown in Figure 4-1, is converted to an upflow path as shown in Figure 5-1.
The objective of this j
conversion is to reduce the hydraulic pressure differentials which exist across tha baffle joints in the downflow cor. figuration. Reducing these pressure differentials, via the upflow conversion, results in a substantial reduction of the coolant jetting through the baffle joints and results jn improved fuel rod reliability.
I l
i 1
1030v:10/103087 5-1
WESTINGHOUSE CLASS 3 THERM AL NEUTRON g SHIELD
\\
=e COOLANT ptow
))
NEW I
f
, f/
j HOLE g FLOW fi
~*
- e
- /
}g PLUG s-J vl 4N CORE
. y n_
{
BARREL
~
en j
Nl h
- s
/N
[f 1 9 ".: *. 5/
l' COOLANT
& &$&AS, FLOW y
.b S*&$$$
\\
d
!& 1;&A&
r/
ta &AL$y; 5
gp Y'd
'y
s.
I l
FIGURE 5-1 UPFLOW CONFIGURATION 1030v:10/103087 5-2
MEST!NGHOUSE CLASS 3 6.0 UPFLOW MODIFICATION The modification of Beaver Valley Unit 1 from downflow to upflow is, conceptually, relatively simple. The hardware changes consist of plugging the existing flow holes in'the core barrel and machining new holes in the top former plate. The tooling for machining the former plate holes has been designed to preclude any machining chips from falling down below the top former plate. Core barrel holes will be sealed by hydraulically expanding plugs via customized installation equipment.
1030v:1o/103087 6-1 i
WESTINGHOUSE CLASS 3 l
7.0 UPFLOW DESIGN BASES AND EVALUATIONS In order to satisfy the general performance and safety criteria, specific fuel system and reactor int.ernals design bases are established and documented in the FSAR. Certain of these design bases are potentially affected by the modification to upflow and fuel thimble plug removal, and are therefore presented and evaluated in the fc11owing subsections of this report.
In addition, the design functional requirements for the core barrel plug are presented and discussed.
7.1 Fuel Assembly 7.1.1 Fuel Assembly Structural Integrity Design Basis The fuel assemblies are designed to perform satisfactorily throughout their lifetime. The combined effects of design basis loads, as specified in FSAR, are considered in the design of the fuel rods and fuel assembly to assure the fuel assembly structural integrity. This assurance is necessary to ensure that the core geometry' remains coolable and the reactor core can be safely shutdown.
Design Evaluation f
The following discussions summarizes an evaluation performed for Beaver Valley concerning upflow modifications made to the reactor internals system.
i The upflow modification has no direct impact on the reactor core system under the earthquake loading condition. Therefore, the fuel assembly structural integrity during a seismic event is not affected by the upflow modifications.
i 1030v:10/103087 7-1
WESTINGHOUSE CLASS 3 i
The potential effects due to the Loss Of Coolant Accident (LOCA) contribution, as a result of upflow modifications, have been evaluated.
The evaluation has demonstrated that the impact of the change in the forces from downflow to upflow are inconsequential.
Therefore, the fuel assembly structural integrity and coolable geometry are maintained during a LOCA.
7.1.2 Fuel Assembly / Fuel Rod Response to Flow Induced Vibration Design Basis The fuel assembly and its associated structural components are designed to accommodate the effects of flow induced vibrations normally encountered during reactor operation without any resultant fuel failures.
Design Evaluation The upflow modifications will reduce the crossflow from the baffle joint gap while maintaining fuel rod structural integrity.
Fuel rod and control rod wear have also been considered in light of the modified flow paths.
The removal of the fuel thimble plugs changes the distribution of core outlet loss coefficients. The core outlet loss coefficient (PFO) distribution shows j
an increase in PF0 mismatch after fuel thimble plug removal.
Therefore, the issue of crossflow induced fuel rod vibration and wear due to this increased PF0 mismatch has been evaluated. Tests have determined that there is no significant difference in fuel rod response during conditions representative of large PF0 mismatch.
It is therefore concluded that fuel thimble plug removal will have an inconsequential effect on fuel rod vibration and wear.
The removal of fuel thimble plugs causes a reduction in flow through the Rod Control Cluster Assembly (RCCA) guide tubes. This tends to reduce control rod However, since the core PF0 distribution changes when the fuel thimble wear.
plugs are removed, the effect of potential control rod vibration due to 1030v.lo/103087 7-2
?
WESTINGHOUSE CLASS 3 interassembly cressflows in the region of the control rod / fuel assembly guide thimble interface has been addressed. However, the maximum PF0 mismatch between an RCC location and an adjacent assembly does not increase with fuel thimble plug removal.
Therefore, the magnitude of the crossflow seen by the control rods and the vfbration of the rods caused by this crossflow will not be increased. As such, fuel thimble plug removal is not forcasted to have an adverse impact on control rod wear.
7.1.3 Thermal Design Flow / Core Bypass Flow Design Basis A minimum of 93.5 percent of the thermal flow rate will pass through the fuel rod region of the core and will be effective for fuel rod cooling.
The remaining 6.5 percent core bypass flow (revised from 4.5 percent to reflect fuel thimble plug elimination); which is comprised of Rod Cluster Control (RCC) guide thimble cooling flow, head cooling flow, baffle / barrel leakage, j
fuel assembly / baffle cavity leakage, and leakage to the vessel outlet nozzle; is not considered effective for heat removal.
' Design Evaluation The modification of the reactor vessel flow paths as a consequence of the upflow conversion has no impact on the established core bypass flow design limit of 6.5 percent.
The only portion of the core bypass flow that is significantly affected by the internals modification is the baffle / barrel l
region flow. For the downflow flow path in the baffle / barrel region, only the flow leaking through the baffle joints is considered bypass since the rest of the flow turns at the bottom of the baffle plate and then passes through the core region. However, for the converted upflow flow path in the baffle / barrel region, all the baffle / barrel region flow is assumed to be bypass. Even though the bypass flow in the baffle / barrel region does increase with'the upflow modification, there is sufficient margin to accommodate this increase
)
without increasing the established bypass flow.
l
\\
l 1030v:10/103087 7-3
s WESTINGHOUSE CLASS 3 7.2 Reactor Internals 7.2.1 Reactor Internals Thermal / Hydraulic Requirements Design Basis The reactor internals in conjunction with the fuel assemblies shall direct reactor coolant through the core to achieve acceptable flow distribution and to restrict bypass flow so that the heat transfer performance requirements are met for all modes of operation.
In addition, required cooling for the p.ressure vessel head shall be provided so that the temperature differences between the vessel flange and head do not result in leakage from the flange during reactor operation.
Design Evaluation As mentioned the upflow conversion does not impact the 6.5 percent core bypass flow limit. That portion of the flow required for cooling of the pressure vessel head is essentially unaffected by the conversion to upflow. Also, since the flow through the baffle / barrel region increases slightly, coolability of the baffle / barrel region is enhanced.
7.2.2 Reactor Internals Structural Response to Seismic and LOCA Loads Design Basis The core internals are designed to withstand mechanical loads arising from the Safe Shutdown Earthquake (SSE), 1/2 SSE, and pipe ruptures and have mechanical provisions sufficient to adequately support the core and internals, and sufficient to assure that the core is intact with acceptable heat transfer a
geometry followi~ng~ transients arising from abnormal operating conditions.
)
Following the Design Basis Accident (DBA), the plant shall be capable of being shutdown and cooled in an orderly fashion so that fuel / clad temperature is kept within specified limits. This implies that the deformation of certain critical reactor internals must be kept sufficiently small to allow core cooling.
1 1030v:1o/103087 7-4 a
i WESTINGHOUSE CLASS 3 Design Evaluation The conversion of the reactor internals to the upflow configuration and removal of fuel thimble plugs has no significant impact on the seismic response of the reactor" internals. An evaluation of the impact of blowdown j
loads on the upflow reactor internals resulting from a LOCA has been performed for Beaver Valley Unit 1.
The evaluation showed that the impact of the change in the forces from downflow to upflow are inconsequential.
7.3 Core Barrel Plug Design Functional Requirements Requirement
\\
The plug must provide continued sealing (100 percent preferred, but some leakage may occur provided that the total core bypass flow is not exceeded).
i Discussion A test program was undertaken to verify and document the ability of the plugging device to seal the core barrel flow holes against coolant flow and to remain retained under limiting conditions (LOCA).
In addition, the tests j
were used to demonstrate critical parameters for plug installation and possibility of removal.
To fulfill the overall program requirements hydrostatic seal, retention, and installation tests were conducted. The 1
hydrostatic seal tests subjected each plug to a 500 psig hydrostatic pressure differential. Retention tests were used to conservatively simulate LOCA loads. These tests were then continued until plug removal occurred, thus allowing information to be gathered for intentional removal purposes.
Finally, plug installation tests were used to determine the loads required to fully insert the mandrel during installation and to observe the installation process.
~
1030v:1o/103087 7-5
t WESTINGHOUSE CLASS 3 Requirement The design and analysis of the core barrel flow hole plugging device shall be to the guidelines of Subsection NG of the ASME Boiler and Pressure Vessel Code 1974 Edition or later.
Discussion l
The core barrel plug has been categorized in compliance with the requirements for the acceptability of internal structures as given in Subsection NG of the ASME B and PV Code,Section III, 1974 Edition or later. Specifically, the plug has been defined, with respect to design, as an internal structure.
However, for the purpose of analysis, the plug has been defined as a support structure and appropriately analyzed to the rules of Subarticle NG-3200 of the ASME B and PV Code. Critical areas of the plug have been analyzed for all appropriate operating conditions.
It is noted that as an internal structure, the plug is not required by the code to meet the requirements of NG-3000.
However, positive margins of safety are demonstrated for all limiting areas.
l Analysis and testing of the core barrel flow hole plug' demonstrate that it is designed to the guidelines of the ASME B & PV code,Section III, Subsection NG.
Requirement l
1 l
The plug and weld must be designed to accommodate radiation and corrosion effects on the materials.
l Discussion The material properties of the 304 austenitic stainless steel and 308 weld metal used in the fabrication and welding of the plugs are well documented both experimentally and in the field. Consequently, compatibility with the general operating environment of the RCS is not considered to be an issue.
1030v.lo/103087 7-6
4 WESTINGHOUSE CLASS 3 3
l 8.0 ACCIDENT EVALUATIONS / ANALYSES
)
8.1 Loss-Of-Coolant Accidents In support of the implementation of the upflow barrel / baffle conversion I
program at the Beaver Valley Unit I nuclear plant, the large and small break ECCS LOCA analyses were performed. The results of these analyses are presented in this section.
I 8.1.1 Identification of Causes and Frequency Classification A Loss-of-Coolant Accident (LOCA) is the result of a pipe rupture of the RCS pressure boundary. For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total cross-sectional area equal to or 2
greater than 1.0. square foot (ft ).
This event is considered an ANS Condition IV event, a limiting fault, in that it is not expected to occur during the lifetime of the plant but is postulated as a conservative design basis.
A minor pipe break (small break), as considered in this section, is defined as a rupture of the reactor coolant pressure boundary with a total 2
cross-sectional area less than 1.0 ft in which the normally operating charging system flow is not sufficient to sustain pressurizer level and pressure.
This is considered a Condition III event, in that it is an I
infrequent fault which may eccur during the life of the plant.
i i
The Acceptance Criteria for the LOCA are described in 10CFR50.46 (Reference 1) as follows:
1.
The calculated. peak fuel element clad temperature is below the requirement of 2200*F.
2.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
1030v:10/103087 8-1
WESTINGHOUSE CLASS 3 3.
The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.
4 4.
The core remains amenable to cooling during and after the break.
5.
The core temperature is reduced and decay heat is removed for an extended period of time. This is required to remove the heat from the long lived radioactivity remaining in the core.
These criteria were established to provide significant margin in Emergency Core Cooling System (ECCS) performance following a LOCA.
2 In all cases, small breaks (less than 1.0 ft ) yield results with more margin to the Acceptance Criteria limits than large breaks.
8.1.2 Sequence of Events and Systems Operations Should a major break occur, depressurization of the RCS results in a pressure l
l decrease in the pressurizer. The reactor trip signiti subsequently occurs.when the pressurizer low pressure trip setpoint is reached. A safety injection signal is generated when the appropriate setpoint is reached.
These countermeasures will limit the consequences of the accident in two ways:
Reactor trip and borated water injection complement void formation in the a.
)
core and cause a rapid reduction of nuclear power to a residual level j
corresponding to the delayed fission and fission product decay heat.
However, no credit is taken during the LOCA blowdown for negative l
reactivity due to boron content of the injection water.
In addition, the l
insertion of control rods to shut down the reactor is neglected in the large break analysis.
i b.
Injection of borated water provides the fluid medium for heat transfer from the core and prevents excessive clad temperatures.
1030v:1o/103087 8-2
)
i
WESTINGHOUSE CLASS 3 Description of a Large Break LOCA Transient Before the break occurs, the unit is in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system.
During blowdown, heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core.by forced convection with some fully developed nucleate boiling. Thereafter, the core heat transfer is based on local conditions with transition boiling, film boiling, and forced convection to steam as the major heat transfer mechanisms.
The heat transfer between the RCS and the secondary system may be in either direction depending on the relative temperatures.
When the RCS depressurizes to approximately 600 psia, the cold leg accumulators begin to inject borated water into the reactor coolant loops.
Since the loss of offsite power is assumed, the reactor coolant pumps are assumed to trip at the time of reacter trip during the accident.
The effects of pump coastdown are included in the blowdown analyses.
The blowdown phase of the transient ends after the RCS pressure (initially
~
assumed at a nominal 2280 psia) falls to a value approaching that of the containment atmosphere. Prior to or at the end of the blowdown, the mechanisms that are responsible for the bypassing of emergency core cooling water injected into the RCS are calculatad not to be effective. At this time (called end of-bypass) refill of the reactor vessel lower plenum begins.
Refill is complete when emergency core cooling water has filled the lower plenum of the reactor vessel which is bourided by the bottom of the fuel rods (called bottom of core recovery time).
The reflood phase of the transient is defined as 'the time period lasting from the end-of-refill until the reactor vessel has been filled with water to the extent that the clad temperature rise has been terminated.
From the later I
stage of blowdown and then the beginning-of-reflood, the safety injection l
1030v:1D/103087 8-3 i
a
WESTINGHOUSE CLASS 3 accumulator tanks rapidly discharge borated cooling water into the
- RCS, contributing to the filling of the reactor vessel downcomer.
The downcomer water elevation head provides the driving force required for the reflooding of the reactor core.
The centrifugal charging and low head safety injection pumps aid in the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the reflooding process.
Continued operation of the ECCS pumps supplies water during long term cooling.
Core temperatures have been reduced to long-term steady state levels associated with dissipation of residual heat generation.
After the water level of the refueling water storage tank reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching of the cold leg recirculation phase of operation in which spilled borated water is drawn from the containment sump by the low head safety injection pumps and returned to the RCS cold legs.
Description of Small Break LOCA Transient Ruptures of small cross section will cause leakage of the coolant at a
rate which can be accommodated by the charging pumps.
These pumps would maintain an operational water level in the pressurizer permitting the operator to execute an orderly shutdown. The coolant which would be released to the containment contains the fission products existing at equilibrium.
The maximum break size for which the normal makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the Reactor Coolant System through the postulated break against the charging pump makeup flow at normal Reactor Coolant System pressure, i.e.,
2250 psia.
A makeup flow rate from one charging pump is adequate to sustain pressurizer level at 2250 psia for a break through a 0.375 inch diameter hole.
This break results in a loss of approximately 17.25 lb/sec.
i 8-4
WESTINGHOUSE CLASS 3 Should a larger break occur, depressurization of the Reactor Coolant System causes fluid to flow into the loops from the pressurizer resulting in a pressure and level decrease in the pressurizer. Reactor trip occurs when the low pressurizer pressure trip setpoint is reached. During the earlier part of the small break transient, the effect of the break flow is not strong enough to overcome the flow maintained by the reactor coolant pumps through the core as they are coasting down following reactor trip. Therefore, upward flow through the core is maintained. The ECCS is actuated when the appropriate setpoint is reached.
Before the break occurs the plant is in an equilibrium condition, i.e., the heat generated in the core is being removed-via the secondary system.
The heat transfer between the Reactor Coolant System and the secordary system may be in either direction depending on the relative temperatures.
In the case of continued heat addition to the secondary, secondary system pressure increases and steam relief via the atmospheric relief and/or safety valves may occur.
Makeup to the secondary side is automatically provided by the auxiliary feedwater pumps. The safety injection signal isolates normal feedwater flow by closing the main feedwater isolation valves and initiates auxiliary feedwater flow by starting the auxiliary feedwater pumps.
The secondary flow aids in the reduction of Reactor Coolant System pressure.
When the RCS depressurizes to approximately 600 psia, the cold leg accumulators begin to inject borated water into the reactor coolant loops.
For all breaks, the accumulator injection provides enough water supply to reverse any core uncovery and to bring the mixture level up to the upper plenum region where it is maintained. Due to the loss of offsite power assumption, the reactor coolant pumps are assumed to-be tripped at the time of i
reacter trip during the accident and the effects of pump coastdown are included in the analyses.
1030v:10/103087 8-5
WESTINGHOUSE CLASS 3 8.1.3 Method of Analysis and Results Large Break LOCA Evaluation Model
~
The current large break LOCA analysis of record forming the licensing basis for Beaver Valley Unit 1 contained in Reference 1, was performed with the NRC Westinghouse Large Break Evaluation Model described in Reference 2.
In order to evaluate the effect of an upflow barrel / baffle design, assuming that the 10% steam generator tube plugging and thimble plug removal have been incorporated in the NSSS design, on the large break LOCA analysis, a complete reanalysis had to be performed.
The reanalysis of a large-break LOCA transient is divided into three phases:
Blowdown, Refill and Reflood. A series of computer codes has been developed to analyze the transient based on the specific phenomena which govern each phase. During the blowdown portion, the SATAN-VI code [ Reference 3] is used I
\\
l to calculate the RCS pressure, enthalpy, density and mass and energy flows in the primary system, as well as the heat transfer between the primary and l
secondary system. At the end of the blowdown, information on the state of the system is transferred to the WREFLOOD code [ Reference 5] which performs the calculation of the refill period to bottom of core (B0C) recovery time.
Once l
the vessel has refilled to the bottom of the core the reflood portion of the transient begins. The BASH code [ Reference 4] is used to calculate the thermal-hydraulic simulation of the RCS for the reflood phase.
Information concerning the core boundary conditions is taken from all of the
- t. cove codes and input to the LOCBART code [ Reference 4] for the purpose of calculating the core fuel rod thermal response for the entire transient.
From the boundary conditions, LOCBART computes the fluid conditions and heat transfer coefficient for the full length of the fuel rod by employing j
mechanistic models appropriate to the actual flow and heat transfer regimes.
Conservative assumptions ensure that the fuel rods modeled in the calculation represent the hottest rods in the entire core.
l l
1030v:10/103087 8-6
WESTINGHOUSE CLASS 3 The containment pressure analysis is performed with the C0CO code
[ Reference 6], which is interactive with the WREFLOOD code (Reference 5]. The transient pressure computed by the C0C0 code is then inpat to the BASH code
[ Reference 4] for the purpose of supplying a backpressure at the break plane while computing the reflood transient. The containment parameters used in the C0CO code to determine the ECCS backpressure are presented in Table 8.2.
Calculations of cold leg double-ended guillotine pipe breaks are performed over a range of Moody discharge coefficients (C ) to identify the case which D
produces the highest peak clad temperature.
For this analysis calculations were performed for discharge coefficients of 0.4, 0.6 and 0.8.
This spectrum of breaks was performed assuming the availability of only minimum safety injection flow capacity (Minimum Safeguards), in accordancs with the single failure criteria of 10CFR50, Appendix 'K.
A break discharge coefficient of 0.4 was found to result in the highest peak clad temperature.
Consistent with the methodology described in reference 7, an additional calculation is performed for the worst break size.
In this calculation, termed Maximum Safeguards, no failures of the safety injection systems are assumed to occur.
This case was found to result in the limiting peak clad 1
temperature of 1918*F, which is acceptably'below the 2200*F limit of 10CFR50.46.
The results of these calculations are summarized in Tables 8.3 and 8.4.
Table 8.1 contains some key plant parameters input to the analyses.
For each break calculation, transients of the following parameters are presented.
For the blowdown portion of the transient:
a.
RCS pressure, b.
Core inlet and outlet flowrates, c.
Cold leg accumulator delivery rate, i
d.
Core pressure drop, e.
Break mass flowrate, f.
Break energy discharge rate, g.
Normalized core power.
1030v:1o/103087 8-7
t! WESTINGHOUSE CLASS 3 For the reflood portion of the transient:
a.
Core and Downcomer liquid levels, b.
The core inlet. fluid velocity, as input to the rod thermal analysis
- code, i
c.
The accumulator and pumped safety injection flowrates.
i From the fuel rod thermal analysis, at the peak temperature location:
1 a.
Fluid mass flux, b.
Rod heat transfer coefficient, c.
The clad peak temperature transient, and d.
The temperature transient at the burst elevation.
e.
Fluid temperature.
Small Break LOCA Evaluation Model The current small break LOCA analysis of record forming the licensing basis for Beaver Valley Unit'I and contained in Reference 8, was performed with the NRC Westinghouse Small Break LOCA ECCS Evaluation Model described in Reference 4.
In order to evaluate the effect of the reactor internals upflow conversion, assuming that 10% steam generator tube plugging and thimble plug removal have been incorporated in the NSSS design, on the small break LOCA analysis, a complete reanalysis was performed.
I For loss of-coolant accidents due to small breaks less than 1 square foot, the NOTRUMP computer code is used to calculate the transient depressurization of the RCS as well as to describe the mass and enthalpy of flow through the break. The NOTRUMP computer code is a state-of-the-art one-dimensional general network code consisting of a number of advanced features. Among these
{
features are the calculation of thermal non equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes and regime-dependent heat transfer correlations.
The NOTRUMP small break LOCA
)
I i
1030v:10/103087 8-8 I
I WESTINGHOUSE CLASS 3 emergency core cooling system (ECCS) evaluation model was developed to determine the RCS response to design basis small break LOCAs and to address the NRC concerns expressed in NUREG-0611
" Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants".
In NOTRUMP, the RCS is nodalized into volumes interconnected by flowpaths.
The broken loop is modelled explicitly, with the intact loops lumped into a second loop. The transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum applied throughout the system. A detailed description of the NOTRUMP code is provided in References 8 and 9.
The use of NOTRUMP in the analysis involves, among other things, the representation. of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculation.
lhe multinode capability of the~ program enables an explicit and detailed spatial representation of various system components.
In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant accident.
A full three break spectrum was reanalyzed modeling the upflow conversion, along with the 10% steam generation tube plugging thimble plug removal modifications.
The limiting case was for the 3 inch equivalent diameter break and resulted in a PCT of 1802*F. This result exhibits considerable margin to the limits defined in 10 CFR 50.46 and Appendix K.
The results of these calculations are summarized in Table 8.5 and 8.6.
1030v:1o/103087 8-9 l
{
i WESTINGHOUSE CLASS 3 8.2 LOCA-Related Evaluations 8.2.1 Containment Long Term Mass and Energy Release and Containment Subcompartment Analyses - FSAR Chapter 14.3 The containment analyses are described in the FSAR sections 14.3.4.2 and 14.3.4.3.
These sections address the long term mass and energy and the containment subcompartment release mass and energy analyses for postulated LOCAs. For the containment subcompartment analyses, the effects of upflow conversion, 10% steam generator tube plugging and thimble plug removal has been determined to be of no adverse impact on the current FSAR analyses.
The critical factors modeled in a short term mass and energy release analyses are the Thermal Design. Parameters. Available information for Beaver Valley Unit 1, indicates that for upflow conversion the Thermal Design Parameters will not change.
In adcition, the subcompartment analysis is run for only a short duration of the blowdown transient (<3 seconds), during which upflow conve'rsion has very little impact.
The modeling of steam generator tube plugging and thimble plug removal is a benefit in subcompartment analyses, each effect will decrease the Reactor Coolant System volume and subsequently, the amount of energy available to be put into the containment, therefore there is no adverse effect on the current analyses.
The long term mass and energy release analysis is performed to calculate the maximum available releases which can be put into the containment following a I
LOCA. An evaluation was completed to determine the effect of upflow conversion, 10% steam generator tube plugging, and thimble plug removal on the current FSAR analysis. As identified for the subcompartment analyses, steam generator tube plugging and thimble plug removal modeling would tend to decrease the available calculated energy releases which would result in a lower peak calculated containment pressure.
Based upon generic sensitivities, it was determined that upflow conversion could have a ootential effect of a 1 psi increase in the Reflood Phase during a long term mass and energy transient. Even though a 1 psi pressure increase is identified, this effect is due to the higher reflooding rate and the total 1030v:1o/103087 8-10
WESTINGHOUSE CLASS 3 energy of this phase is less as compared to the downflow transient.
Therefore, General Design Criterion 38,of NUREG - 0800, pertaining to Subatmospheric Containments will not be affected.
The plant change also does not effect the current limiting break for Beaver Valley Unit 1, because based upon the current FSAR analyses, the limiting peak calculated pressure occurs during the blowdown phase of the transient.
For reasons discussed above, the related effects of upflow conversion, 10%
steam generator tube plugging and, thimble plug removal would not have any additional effects on the FSAR containment analyses results for Beaver Valley Unit 1.
8.2.2 Steam Generator Tube Rupture - FSAR Chapter 14.2.4 i
For the Steam Generator Tube Rupture (SGTR) event, the FSAR analysis was performed to evaluate the radiological consequences resulting from a SGTR accident. The major factors that affect the radiological doses resulting from a SGTR event are the amount of primary coolant transferred to the secondary side of the ruptured steam generator through the ruptured tube after reactor trip, and the steam released from the ruptured steam generator to the atmosphere. The effect of upflow conversion, 10% steam generator tube plugging and thimble plug removal on the FSAR SGTR analysis has been evaluated. The results of the evaluation show that the primary coolant transferred through the tube rupture to the secondary will increase by less than 1% while the steam release via the ruptured steam generator will decrease slightly. The net effect on the radiological consequences has been determined to be an increase of less than 1% to the offsite radiation doses.
Based on.the above evaluation, it is concluded that upflow conversion, 10%
steam generator tube plugging and thimble plug removal will not change the conclusions reporfed in the FSAR SGTR analysis,. Since the offsite doses, with the proposed changes, are still much less than a small fraction of the 10CFR100 limits, no reanalysis or FSAR changes are required.
1030v:1o/103087 8-11
WESTINGHOUSE CLASS 3 8.2.3 Hot Leg Switchover to Prevent Potential Boron Precipitation Post-LOCA hot leg recirculation switchover time is determined for inclusion in emergency procedures to ensure that boron precipitation in the reactor vessel due to boiling in the core is precluded in the long term.
This time is dependent on power level, and water volumes and boron concentrations in the RCS, RWST and accumulator.
The upflow conversion, 10% steam generator tube plugging and thimble plug removal do not affect the power level or the maximum boron concentrations. The volumes assumed for the RCS will be slightly reduced due to a steam generator tube plugging level of 10%. However, the effect of this re.iuction is small and beneficial to the mixed mean boron concentration calculation.
Thereforo, there is no adverse affect on the post-LOCA hot leg switchover time due to the plant changos.
8.2.4 LOCA Hydraulic Vessel and Loop Forces - FSAR Chapter 14.3.3 and Appendix B The blowdown hydraulic loads resulting from a loss of coolant accident are considered in FSAR Section 14.3.3 and Appendix B-3.
As part of the effort to convert Beaver Valley Unit 1 to an upflow barrel / baffle design, a LOCA hydraulic vessel forces analysis was performed. The analysis consisted of a comparison between hydraulic forcing functions for the two plant configurations:
upflow without thimble plugs and downflow with thimble plugs.
In addition, the effect of the NSSS performance information for 10%
steam generator tube plugging was also considered in the selection of parameters.
The effect on vessel forces of thimble plug removal was identified, based on a direct comparison between two runs where the sole difference was thimble plug modelling.
It was, determined that thimble plug removal has a negligible I
effect on the LOCA hydraulic forces, except for the force on the upper core plate which is higher with thimble plugs installed in the core. However, the maximum force on the upper core plate occurs at steady state, and is therefore not a LOCA concern.
1030v:1o/103087 8-12
WESTINGHOUSE CLASS 3 The results of this analysis were used to perform an evaluation to demonstrate that the structural and functional integrity of the reactor internal components is maintained after the upflow conversion is complete. A separate report was issued which includes the hydraulic forces analysis and the i
subsequent structural evaluation.
J 8.2.5 Long Term Core Cooling - Boron Evaluation - FSAR Chapter 14.3.2 4
The Westinghouse licensing position for satisfying the requirements of 10CFR Part 50 Section 50.46 Paragraph (b) Item (5) "Long Term cooling" is defined in WCAP-8339 (pp. 4-22). The Westinghouse commitment is that the reactor will remain shutdown by borated ECCS water residing in the sump post LOCA (reference 1). Since credit for the control rods is not taken for large break LOCA, the borated ECCS water provided by the accumulators and the RWST must
\\
have a concentration that, when mixed with other sources of borated and non-borated water, will result in the reactor core remaining suberitical assuming all control rods out.
The long term core cooling evaluation was performed as part of the upflow conversion / steam generator tube plugging program.
The results of this evaluation show that the upflow conversion, thimble plug removal; and the uniform plugging of up to 10% of the steam generator tubes at Beaver Valley Unit 1, have an insignificant effect on the the mass average boron concentration required to maintain core suberiticality following a large break LOCA. The revised Boron source data will become the N-loop current limit data for use in the evaluation of the cycle 7 reload core design.
i 1030v:10/103087 8-13 3
1 WESTINGHOUSE CLASS 3 FSAR SECTION ACCIDENT DESCRIPTION EFFECT ON RESULTS 14.3.2 Large Break LOCA Full reanalysis performed. Compliance with 10CFR 50.46 shown.
14.3.1 Small Break LOCA Full reanalysis performed. Compliance with 10CFR 50.46 shown.
14.3 Containment Integrity No adverse effect on (Short and Long Term short or long term mass and Energy Release) mass and energy releases 14.2.4 Steam Generator Tube No adverse effect on Rupture mass releases No adverse effect on-offsite radiation doses 14.3.3 &
Blowdown Reactor Vessel Reanalysis performed and f
Appendix B and Loop Forces new LOCA hydraulic forcing functions were generated
~
14.3.2 Post-LOCA Longterm Core No significant effect on Cooling the post-LOCA sump boron concentration E0Ps Hot Leg Switchover to No adverse effect on the Prevent Potential Boron post-LOCA hot leg
~~
Precipitation swi.tchover time 1030v:10/103087 8-14 l
l l
WESTINGHOUSE CLASS 3 TABLE 8.1 PLANT PARAMETERS USED IN THE LOCA ANALYSES Licensed Core Power 102% of 2652 MWt Total Core Peaking Factor 2.40 Peak Linear Power 102% of.
Barrel-Baffle Configuration Upflow Steam Generator Tube Plugging 10% (uniform)
RCS Initial Conditions:
Vessel flowrate 27873 lbm/s Core flowrate 26089 lbm/s Vessel inlet temperature 546.3*F Vessel outlet temperature 614.6'F
[
Pressure 2280 psia Cold Leg Accumulator Conditions:
)
Cover gas pressure 600 ' psia Water volume / Accumulator Large Break 1024.8 ft3 (Minimum plus line volume)
Small Break 980.0 ft -(Nominal minue line volume)
Water temperature 90*F 9
1030v 1o/103087 8-15
WESTINGHOUSECl. ASS 3 TABLE 8.2 CONTAINMENT DATA 0
3 NET FREE VOLUME 1.89 x 10 ft INITIAL CONDITIONS Pressure 9.5 psia Temperature 90*F RWST Temperature 40*F-Outside Temperature 35'F SPRAY SYSTEM i
Number of Pumps Operating 2
{
Runout Flow Rate (each) 2200 gpm per pump Actuation Time.
55 seconds 1030v:10/103087 8-16 l
WESTINGHOUSE CLASS 3 TABLE 8.2 (Cont)
CONTAINMENT DATA l
II)
STRUCTURAL HEAT SINKS.
l Wall j
Number Material Thickness (ft)
Surface Area (sq ft) 1 Concrete 0.5 6,972 2
Concrete 1.0 77,446 3
Concrete 1.5 36,848 4
Concrete 2.0 17,010 5
Concrete 3.0 8,632 6
Carbon Steel 0.03125 18,270 Concrete 4.5 j
7 Carbon Steel 0.03125 32,445 Concrete 4.5 8
Carbon Steel 0.04167 26,250 3
Concrete 2.5 9
Concrete 2.0 13,125 Carbon Steel 0.03125 Concrete 10.0 10 Stainless Steel 0.06875 3,270 11 Carbon Steel 0.0E202 11,750 12 Carbon Steel 0.06242 748 13 Carbon Steel 0.1932 2,132 14 Carbon Steel 0.1833 5,479 15 Carbon Steel 0.0893 3,770
~
16
~ Carbon Steel 0.1041 10,938 17 Carbon Steel 1.020 600 18 Carbon Steel 0.0119 118,091 (1) All walls are painted with the exception of walls 9 and 10.
The thickness of paint is 5.0 mils for all painted walls with the exception of wall 11, which has a paint thickness of 3.75 mils.
1030v:1o/103087 8-17 w
?
WESTINGHOUSE CLASS 3 TABLE 8.3 LARGE BREAK LOCA FUEL CLADDING DATA Break loss coefficient:'
O.40 0.6 0.8 0.4 Max SI Parameter Peak clad temperature (F) 1911 1759 1667 1918 Elevation (ft) 8.50 8.50 8.50 8.50 Maximum Zr/H2O reaction (%)
3.92 2.13 1.50 3.90 Elevation (ft) 8.50 8.50 8.50 8.50 Total Zr/H2O reaction (%)
<0.3
<0.3
<0.3
<0.3 Hot rod burst time (s) 43.0 48.64 52.57 43.0 Elevation (ft) 6.00 6.00 6.00 6.00 i
f 1030v:10/103087 8-18
WESTINGHOUSE CLASS 3
~
TABLE 8.4 LARGE BREAK LOCA TIME SEQUENCE OF EVENTS Break loss coefficient:
O.40 0.6 0.8 0.4 Max SI Parameter l
Reactor trip 0.45 0.44 0.44 0.45 Safety injection signal 2.09 1.71 1.50 2.09 Accumulator injection begins 15.30 11.4 9.22 15.30 End of bypass-31.88 36.46 23.26 31.88 End of blowdown 31.88 36.46 23.26 31.88 Pumped safety injection 29.09 28.71 28.50 29.09 begins Bottom of core recovery 45.88 40.25 37.34 45.20 Accumulators empty 56.90 51.44 48.46 55.21 1
103cv:10/103087 8-19 i
WESTINGHOUSE CLASS 3 TABLE 8.5 SMALL BREAK LOCA FUEL CLADDING DATA 3 INCH BREAK RESULTS Peak Cladding Temperature (*F) 1802 Peak Clad Temperature Location (ft) 11.75 Local Zr/ Water (%) Reaction of Thickness 2.78 Local Zr/ Water Reaction, Location for Maximum Reaction 11.75 Total Zr/ Water Reaction (%) of Mass
<0.3 Hot Rod Burst Time N/A Hot Rod Burst Location N/A 1030v:10/103087 8-20
WESTINGHOUSE CLASS 3 TABLE 8.6 SMALL BREAK LOCA TIME SEQUENCE OF EVENTS 3 INCH BREAK Event Time (sec)
Break 0.0 Reactor Trip Signal 5.98 Safety Injection Signal 11.18 Safety Injection Begins 38,.18 Loop Seal Venting Begins 482.0 Top of Core Uncovered 612.0 Cold Leg Accumulator Injection 1157.0 Peak Clad Temperature Occurs 1383.0 Top of Core Covered 4422.0
-gma 1030v:1o/103087 8-21
c WESTINGHOUSE CLASS 3 8.3 Non-LOCA The main non-LOCA safety analysis parameters that are changed by the conversion of the Beaver Valley Unit 1 plant to upflow configurations are 1) the core bypass flow fraction, which is being increased by 2.0% (from 4.5% to 6.5%), and 2) the reactor vessel component pressure drops.
8.3.1 Increase In Core Bypass Flow Converting the Beaver Valley Unit 1 to upflow configuration does not change the plant Thermal Design Flow (TDF); however, it does cause the core bypass flow (i.e., the fraction of the total Thermal Design Flow which does not go through the core) to increase by 2.0%.
As a result of the core bypass flow l
fraction increasing from the present value of 4.5% to 6.5%, the flow through the reactor core will be reduced.
The core bypass flow and vessel pressure drops were calculated assuming the new upflow configuration and thimble plug l
removal.
The results showed an insignificant change in the vessel pressure drop values assumed in the non-LOCA safety analysis.
Therefore, the Upflow Conversion alone does not invalidate the conclusions of the BVPS-1 FSAR.
However, the core bypass flow increased significantly, the impact of which will be further discussed in the Thimble Plug Removal below.
Thimble Plug Removal Thimble Plug Removal may also impact the vessel pressure drops assumed in the non-LOCA safety analysis. However, as discussed above, the new values represented an insignificant change to these parameters.
The impact of Thimble Plug Removal results in a 2% increase in core bypass flow.
The i
resulting reduction in core flow represents a DNB penalty.
However, the core limits and the corresponding safety analyses setpoints remain unchanged.
Therefore, the DNB~ design bases are met for the following FSAR transients:
14.1.1 Uncontrolled RCCA Bank Withdrawal from a Suberitical Condition 14.1.2 Uncontrolled RCCA Bank Withdrawal at Power 14.1.3 RCCA Misalignment 1
1030v:1o/103087 8-22
i WESTINGHOUSE CLASS 3 l
l 14.1.4 Partial loss of Forced Reactor Coolant Flow
{
14.1.5 Startup of an Inactive Reactor Coolant Loop 14.1.6 Startup of an Inactive Reactor Coolant Loop 14.1.7 Loss of Load / Turbine Trip I
14.1.9 Excessive He'at Removal Due to Feedwater Malfunctions 14.1.10 Excessive Load Increase j
14.1.13 Accidental Depressurization of the Main Steam System 14.1.15 Accidental Depressurization of the Reactor Coolant System I
14.1.16 Spurious Operation of the Safety Injection System 14.2.5 Major Secondary System Pipe Rupture (Main Steamline) 14.2.9 Complete Loss of Forced Reactor Coolant Flow l
14.2.10 Signal RCCA Withdrawal at Full Power In addition to the DNB concern, the following evaluations are presented for those accidents whien are not DNB related or for which DNBR is not the only l
safety criterion to be met. No new analyses have been performed for these evaluations, however, in each transient discussed, it is concluded that all appropriate safety criteria are met.
l Rod Withdrawal from Suberitical. A control rod assembly withdrawal incident l
when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 14.1.1 of the FSAR).
The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative Doppler coefficient.
The power excusion causes a heatup due to the moderator. However, since the power rise is rapid and is followed by an immediate reactor trip, the moderator temperature rise is small. Thus, nuclear power response is primarily a function of the Doppler coefficient. An increase in temperature due to increase core bypass flow would result in more Doppler feedback, thus reducing the nuclear power 1
1 1
excursion as presented in the FSAR which partially compensates for the flow i
l reduction. Results reported in the BVPS-1 FSAR indicate that the peak heat l
flux is less than the nominal full power value.
It is expected that the increase in core bypass flow will result in a slight increase in peak heat flux. However, the value will remain below the nominal full power value.
Therefore, the conclusions of the FSAR remain valid.
i 1030v:10/103087 8-23 m___.___.___________._______
WESTINGHOUSE CLASS 3 J
Boron Dilution. The results of the baron dilution transient will remain unchanged for all modes of operation due to the increase in core bypass flow.
The maximum dilution flow rate and RCS boron concentrations are not impacted by an increase in bypass flow. Therefore, the conclusions presented in the FSAR are not invalidate'd by the Thimble Plug Removal effort.
Loss of Load / Turbine Trip. The Loss of Load accident presented in Section 14.1.7 of the FSAR results in an increase in core power which exceeds the secondary system power extraction, thus, causing an increase in core water temperature. An increase in core bypass flow will result in a more rapid pressure rise then that shown in the FSAR.
The effect will be minor, however, since the. reactor is tripped on high pressurizer pressure for the limiting RCS j
pressure cases.
Thus, the time to trip will decrease, which will result in a lower total energy input to the coolant.
The Loss of Load analysis reported in the FSAR shows that the pressurizer pressure remains below 2550 psia in all cases analyzed. The increase in core 5
bypass flow is expected to have a negligible impact on the pressure i
Therefore, the conclusions stated in the FSAR remain valid.
Loss of Normal Feedwater/ Station Blackout.
This transient is analyzed to demonstrate that the peak RCS pressure does not exceed allowable limits and that the core remains covered with water.
These criteria are assured by applying the more stringent requirement that the pressurizer must not be filled with water. The effect of increasing core bypass flow would result in an initially more rapid heatup of the core outlet temperature. However, the additional water that bypasses the core would serve to lessen the increase in coolant temperature as seen at the vessel outlet. Therefore, the resultant coc,lant density decrease would not significantly increase the peak water volume in the pressurizer.
Results of these analyses reported in the FSAR show that the peak pressurizer 3
3 water volume reaches 1353 ft in the LONF event.
This is 84 ft short of 3
the total pressurizer volume (including surge line) of 1437 ft.
It is expected that the increase in core bypass flow would result in a slight 1030v:1D/103087 8-24
WESTINGHOUSE CLASS 3 increase in peak pressurizer water volume.
However, sufficient margin to the limits exist to accommodate the expected increase.
Therefore, the conclusions of FSAR remain valid.
~
Locked Rotor.
Section 14.2.7 of the FSAR indicates that the most severe locked rotoer event is an instantaneous seizure of a reactor coolant pump rotor. Following the incident, RCS temperature rises until shortly after reactor trip. An increase in core bypass flow will not affect the time to DNB since DNB is conservatively assumed to occur at the beginning of the transient.
In addition, the bypass flow increase will not impact the time of reactor trip.
Therefore, the nuclear power and heat flux transients will not change from those presented in the FSAR.
However, the increase in a core bypass flow will result in slightly higher system pressures and clad temperatures.
The peak pressure reported in the FSAR is 2725 psia and the peak clad temperature is reported as 2031*F. These are below the '.imit values of 2750 and 2700*F respectively.
It is expected that there would be a slight increase in the peak system pressure and' peak clad temperature.
Furthermore, this small clad temperature increase will have j
a negligible impact on the Zirconium-steam reaction results.
Since sufficient margin to the limits exist to accommodate the expect 9d increases, the conclusions of the FSAR remain valid.
Feedline Break.
This transient is analyzed to demonstrate that the peak RCS pressure does not exceed allowable limits and the the core remains covered
~
with water.
The results reported in the reduced Auxiliary Feedwater Report show that the hot leg temperature never reaches saturation temperature, thus the core remains completely covered for the duration of the transient.
An
)
increase in core bypass flow will have a negligible impact on the peak hot leg temperature.
Therefore, the conclusions of the FSAR remain valid.
Rod Ejection.
This event is analyzed at full power and hot standby for both beginning, and end-of-life conditions. An increase in core bypass flow will result in a reduction in heat transfer to the coolant, which could increase 1030v:1D/103087 8-25
~
e WESTINGHOUSE CLASS 3 peak clad and fuel temperatures and peak fuel stored energy. The critical parameters reported in the FSAR analyses are as follows:
peak fuel center temperature (*F):
5080 (BOL HFP) peak clad temperature ('F):
2565 (BOL HFP) peak fuel stored energy (cal /gm):
181 (BOL HFP)
The increase in core bypass flow is expected to have a negligible impact on the fuel temperature and stored energy results. The peak clad temperature is expected to increase slightly, but is well within the criteria limit of 2700*F. Therefore, the conclusions of the FSAR remain valid.
j Thimble Plug Removal will result in more core flow passing through the thimble guide tubes.
This may impact the design value for rod drop time assumed in the safety analyses. However, it has been confirmed that the value for rod drop time used in the safety analyses will not change. Therefore, Thimble Plug Removal does not invalidate the conclusions presented in the FSAR.
4 1030v:1o/103087 8-26
]
m WESTINGHOUSE CLASS 3
9.0 CONCLUSION
This safety evaluation has been prepared utilizing the criteria of 10 CFR 50.59 in addressing the reactor vessel upflow configuration and fuel thimble plug removal scheduled for Unit 1 at the Beaver Valley Nuclear Power Station.
In summary, it has been demonstrated that the various reactor vessel lower internals, fuel system design bases and FSAR accident analyses remain valid, and plant safety uncompromised.
Consequently, the upflow modification and fuel thimble plug removal will have no safety related implications or consequences during all licensed modes of operation.
i 1030v:1D/103087 9-1 N
7 WESTINGHOUSE CLASS 3
- 10. REFERENCES 1.
Beaver Valley Unit 1 (DLW) FSAR-Updated 1/87 Revision 5.
2.
WCAP-9220, (Proprietary), WCAP-9921 (Non Proprietary), " Westinghouse ECCS Evaluation Model February 1978 Version:, February 1978.
3.
F. M. Bordelon, et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant", WCAP-8302 (Proprietary Version), WCAP-8306 (Non-Proprietary Version), Westinghouse Electric Corporation (June 1974).
j 4.
WCAP-10266-P, Rev 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model using BASH".
5.
R. D. Kelly, et al, " Calculation Model for Core Reflooding after a Loss of-Coolant Accident (WREFLOOD Code)", WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Proprietary Version), Westinghouse Electric Corporation (June 1974).*
6.
F. M. Bordelon, and E. T. Murphy, " Containment Pressure Analysis Code (C0CO)", WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Proprietary Version), Westinghouse Electric Corporation (June 1974).*
7.
Rahe, E. P. (Westinghouse) letter to Tedesco, R. L. (USNRC),
NS-EPR-2538, December 1981..
l 1
8.
Lee, H., Rupprecht, S. D., Tauche, W. D., Schwarz, W. R.,
)
" Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP J
Code," WCAP-10054-P-A, August 1985.
9.
Meyer, P.E., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A, August 1985.
10.
NS-TMA-2273; Boron Dilution Concerns at Cold and Hot Shutdown.
1030v:10/103087 10-1 l