ML20064H450
| ML20064H450 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/11/1994 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML19304B842 | List: |
| References | |
| NUDOCS 9403170303 | |
| Download: ML20064H450 (72) | |
Text
_ - _ - _ _ _ - _ _ _ _ - _ _ _
,1 Attachment to Document Control Desk letter
~ TSP 930019~
STEAM GENERATOR REPLACEMENT TECHNICAL SPECIFICATION REVISION-
.3 l
[
(
l l
9403170303.940311 PDR ADOCK 05000395 p
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continuedi h.
By performing a flow balance test, during shutdown, following cot.pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
1)
For centrifugal charging pump lines, with a single pump running and with recirculation flow; a)
The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 338 gpm, and b)
The total pump flow rate is less than or equal to 700'gpm.
'l 1.
By performing a flow test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
1)
For residual heat removal pump lines, with a single pump running the sum of the injection line flow rates is greater tt n or equal-to 3663 gpm.
f SUMMER - UNIT 1 3/4 5-6 Amendment No. 75
SUBMITTAL SCh10ULE AND FORMAT TO SUPPORT STEAM GENERATOR REPLACEMENT TECIINICAL SPECIFICATION CIIANGES FOR TIIE VIRGIL C. SUMMER NUCLEAR STATION Submittal
=
Title 1
2 3
4 List of Tables X
X X
l List of Figures X
X l
List Acronyms and Abbreviations X
X Executive Summary X'
X X
1.0 Introduction-Description of License Amendment Request X
X 1.1 Purpose for Change 1.2 Current License Basis and Function of Identified Technical Specifications j
1.3 Description of Proposed Change l
2.0 Basis for Evaluations / Analyses Performed X
2.1 Design Power Capability Parameters X
2.1.1 Discussion of Parameters 2.1.2 References 2.2 NSSS Design Transients 2.3 Control System Setpoints 2.4 Reactor Protection System / Engineered Safety Features Actuation System Setpoints 3.0 Safety Evaluations / Analyses l
3.1 less of Coolant Accident Analyses X
3.1.1 Large Break LOCA X
l 3.1.2 Small Break LOCA 3.1.3 Post-LOCA Long Term Core Cooling Suberiticality X
4 3.1.4 Hot leg Switchover to Prevent Potential Boron Precipitation X
3.1.5 References X
3.2 LOCA Hydraulic Forces X
3.2 1 Introduction 3.2.2 Method of Analysis 3.2.3 Results 3.2.4 References Target Subminal Deter 1; August 31,1992 2: April 30.1993 1
3: October 29,1993 l 4: March 11,1994 RSO-TOC 2.COE: 3/7/94 i
~
l Submittal J
Title 1
2 3
4 3.3 Non-LOCA Analyses X
3.4 High Energy Line Break Analyses X
3.4.1 LOCA Mass & Energy Releases X
X 3.4.1.1 1.ong Term LOCA Mass and Energy Releases 3.4.1.2 Short Term LOCA Mass and Energy Releases 3.4.2 Short Term Containment Analysis - LOCA Reactor Building X
X Subcompartment Analysis 3.4.3 Main Steamline Break Mass / Energy Releases X
3.4.3.1 Inside Containment 3.4.3.2 Outside Containment 3.4.4 Imng Term Containment Analysis X
F 3.4.4.1 Main Steamline Break Containment Integrity Analysis 3.4.4.2 LOCA Reactor Building
- s grity Analysis X
3.4.5 Environmental Conditions - Steam Line Break (SLB) Outside X
Containment 3.4.6 Equipment Qualification X
3.4.7 References X
X 3.5 Steam Generator Tube Rupture Accident Analysis X
X 3.6 Reactor Cavity Pressure Evaluation X
3.6.1 Introduction 3.6.2 Evaluation Results 3.6.3 References 3.7 Radiological Analysis X
X 3.7.1 Introduction 3.7.2 Source Terms 3.7.3 Radiological Consequences 3.7.3.1 Loss of Offsite Power 3.7.3.2 Waste Gas Decay Tank Rupture 3.7.3.3 Break in a CVCS Line 3.7.3.4 Large Break LOCA 3.7.3.5 Main Steam Line Break 3.7.3.6 Steam Generator Tube Rupture 3.7,3.7 Locked Rotor Tarcet Submittal Dates:
1: August 31,1992 2: April 30,1993 3: October 29.1993 l 4: March 11.1994 RSO-TOC 2 COE: 3r//94 ii i
~
i i
1 1 '
Submittal Title -
1 2
3 4
3.7.3.8 Fuel Handling Accident 3.7.3.9 RCCA Ejection 3.7.4 References 3.8 Primary Components Evaluations 3.8.1 Reactor Vessel X
3.8.1.1 Reactor Vessel Structural Evaluation 3.8.1.2 Reactor Vessel Brittle Fracture Integrity 3.8.2 Reactor Internals X-3.8.'2.1 Thermal-Hydraulic Performance 3.8.2.2 Bypass Flow Analysis l
3.8.2.3 Hydraulic Lift Force Analysis 3.8.2.4 RCCA Scram Performance Evaluation 3.8.3 Steam Generators X
3.8.3.1 Thermal-Hyaraulic Performance Evaluation 3.8.3.2 Structural Evaluation 3.8.4 Pressurizer X
3.8.5 Reactor Coolant Pumps (RCPs) and RCP Motors X
3.8.6 Control Rod Drive Mechanism X-3.8.7 Reactor Coolant Piping and Supports X
3.8.8 Application of Leak-Before-Break Methodology X
X-3.8.9 Conclusions X
3.8.10 References X
3.9 Fluid and Auxiliary Systems Evaluations X
3.9.1 Introduction 3.9.2 Discussion of Evaluations Performed 3.9.2.1 Fluid Systems Evaluation 3.9.2.2 Auxiliary Equipment Evaluation 3.9.2.3 NSSS/ Balance of Plant Interface 3.9.3 Conclusions 3.10 Fuel Structural Evaluation X
3.10.1 General Considerations 3.10.2 Fuel Assembly Structural Evaluation l
3.11 Summary of Technical Specification Changes X-X Terret Submitta! Dates:
7 1: August 31,1992 2: April 30,1993 3: October 29.1993 l 4: March 11,1994 RSG-TOC 2.COE: 3/7/94 iii
. =..
Submittal Title 1
2 3
4 l
4.0 Conclusions X
X Appendix 1 10 CFR 50.59 (Assessment of Unreviewed Safety Questions)
X X
Appendix 2 10 CFR 50.92 (No Significant llanrds Determination)
X X
Appendix 3 Proposed Technical Specification Changes X
X l
Appendix 4 WCAP-13480 " Westinghouse Delta 75 Steam Generator Design X
X X
l and Fabrication hiformation for the VCSNS" Appendix 5 WCAP-13605 " Primary Loop Lenk-Before-Break Reconciliation X
X to Account for the EITects of SGR/Uprating" Appendix 6 VCSNS FSAR Chapter 15 Write Ups X
X Appendix 7 VCSNS S1 Pump Test Report X
t J
Terret Submittal Dates:
)
1: August 31,1992
)
2: April 30,1993 3: October 29,1993 l 4: March II,1994 RsG-TOC 2.COE: 30/94 iv 1
LIST OF TABLES TABLE 2.1-1 DESIGN PERFORMANCE CAPABILITY PARAMETERS FOR VCSNS A-75 REPLACEMENT STEAM GENERATORS ANALYSES TABLE 2.4-1
SUMMARY
OF THE REACTOR PROTECTION SYSTEM / ENGINEERED SAFETY FEATURES ACTUATION SYSTEM SETPOINT CHANGES TABLE 3.1-1 INPUT ASSUMPTIONS FOR LBLOCA REPLACEMENT SG ANALYSIS l TABLE 3.1-2 INPUT ASSUMPTIONS FOR SDLOCA ANALYSIS TABLE 3.4.1-1 SYSTEM INITIAL CONDITIONS TABLE 3.4.1-2 BLOWDOWN MASS AND ENERGY RELEASES DOUBLE-ENDED PUMP SUCTION GUILLOTINE TABLE 3.4.1-3 BLOWDOWN MASS AND ENERGY RELEASES DOUBLE-ENDED HOT LEG GUILLOTINE TABLE 3.4.1-4 REFLOOD MASS.AND ENERGY RELEASES DOUBLE-ENDED PUMP SUCTION - MIN Si TABLE 3.4.1-5 REFLOOD MASS AND ENERGY RELEASES DOUBLE-ENDED PUMP SUCTION - MAX SI TABLE 3.4.1-6 PRINCIPAL PARAMETERS DURING REFLOOD DOUBLE-ENDED PUMP SUCTION - MIN SI TABLE 3.4.1-7 PRINCIPAL PARAMETERS DURING REFLOOD DOUBLE-ENDED PUMP SUCTION - MAX SI TABLE 3.4.1-8 POST-REFLOOD MASS AND ENERGY RELEASES DOUBLE-ENDED PUMP SUCTION - MIN SI TABLE 3.4.1-9 POST-REFLOOD MASS AND ENERGY RELEASES DOUBLE-ENDED PUMP SUCTION - M AX SI TABLE 3.4.1-10 MASS BALA'NCE DOUBLE-ENDED PUMP SUCTION - MIN SI 4
TABLE 3.4.1-11 MASS BALANCE DOUBLE-ENDED PUMP SUCTION - MAX Si TABLE 3.4.1-12 MASS fiALANCE DOUBLE-ENDED HOT LEG GUILLOTINE TABLE 3.4.1-13 ENERGY BALANCE DOUBLE-ENDED PUMP SUCTION - MIN SI TABLE 3.4.1-14 ENERGY BALANCE DOUBLE-ENDED PUMP SUCTION - MAX Si TABLE 3.4.1-15 ENERGY BALANCE DOUBLE-ENDED HOT LEG GUILLOTINE V
EXECUTIVE SUALTIARY This document contains the safety analysis and evaluation results to support the Replacement Steam Generator (RSG) Technical Specification changes for the Virgil C. Summer Nuclear Station (VCSNS).
l The Table of Contents lists the topics that are addressed to support the RSG Technical Specification l changes and reflects the schedule of all planned submittals (Note: The results of the Small Break LOCA l analysis were submitted as replacement pages to the 10/29/93 submittal). As indicated in the Table of l Contents, the supporting documentation for the license amendment request is contained in this document.
l The topics addressed in this submittal are as follows:
Section 1.0 Introduction - Description of License Amendment e
Section 2.0 Basis for Evaluations / Analyses Performed l
e Section 3.1 LOCA Analyses Section 3.2 LOCA Hydraulic Forces e
Section 3.3 Non-LOCA Safety Analyses e
Section 3.4 High Energy Line Break Analyses Section 3.5 Steam Generator Tube Rupture Accident Analysis Section 3.6 Reactor Cavity Pressure Evaluation Section 3.7 Radiological Analysis e
Section 3.8 Primary Components Evaluations.
Section 3.9 Fluid and Auxiliary Systems Evaluations e
Section 3.10 Fuel Structural Evaluation e
Section 3.11 Summary of Technical Specification Changes In addition to the above analysis sections, several WCAPs are submitted to support the RSG Technical Specification changes. The WCAPs are contained in Appendix 4 (WCAP-13480, " Westinghouse A75 Steam Generator Design and Fabrication Information for the VCSNS") and Appendix 5 (WCAP-13605,
" Primary Loop Leak-Before-Break Reconciliation to Account for the Effects of Steam Generator Replacement /Uprating"). Appendix 6 contains the VCSNS Chapter 15 FSAR writeups to support this l submittal. Appendix 7 provides the VCSNS SI Pump Test Report. Appendices 1,2, and 3 contain the Assessment of Unreviewed Safety Questions (10CFR50.59), No Significant Hazards Determination (10CFR50.92), and marked up Technical Specification changes, respectively.
As noted in a previous letter from SCE&G [ John L. Skolds to DCD "VCSNS Proposed Schedule for Submittal of Information Supporting Steam Generator Replacement (REM 6000)", dated June 4,1992],
it is expected that early submittal of discrete packages of analyses will assist the NRC with meeting SCE&G's Fall 1994 SG Replacement (SGR) schedule. Two previous submittals have been made to the NRC to support the RSG Technical Specification changes (Letters dated September 3,1992 & April 30, 1993).
It should be noted that, where possible, the analyses and evaluations performed to support the RSGs incorporate the Engineered Safeguards Design Rating (" stretch" power rating) of 2900 MWt core power.
This conservatively bounds the current licensed core power of 2775 MWt and is used to minimin. future reanalysis for a potential stretch power application. However, it should be emphasized that approval is not being sought at this time for operation at stretch power, i
xiii
)
A brief summary of the results of each analysis, evaluation, and supporting documentation contained in this submittal is as follows:
Basis of Evaluations / Analyses:
The analyses and evaluations performed to support the RSGs bound a range of operating conditions for VCSNS. Four cases are presented which define a range of primary operating temperatures from 572*F to 587.4*F and a range of steam generator tube plugging levels from 0% to 10%. This will provide SCE&G with the flexibility to select the appropriate primary temperatures on a cycle-by-cycle basis necessary to achieve full megawatt electric output and to adjust the temperature as necessary to compensate for steam generator tube plugging or to perform end-of-cycle T, coastdown.
Larce Break LOCA:
The Large Break (LB) LOCA analysis was performed at the current power level of 2775 htWt (core power) with the NRC approved ECCS Evaluation hiodel using the BASH code. In addition to the 2775 h1Wt power level assumed, the analysis assumed a total peaking factor (F ) of 2.45 and a hot channel n
enthalpy rise factor (Fw) of 1.62. An initial RCS pressure of 2300 psia and a TDF of 277,800 gpm was also assumed. A complete spectrum of breaks was analyzed for the Large Break LOCA along with a maximum safety injection case for a vessel average temperature range of 572 F to 587.4 F. Additional input assumptions for the Large Break LOCA analysis are listed in Table 3.1-1. All LOCA acceptance criteria as described in 10CFR50.46 were met and it was concluded that operation of the VCSNS with the A75 steam generators is acceptable with respect to LB LOCA. The detail and results of the Large Break LOCA analysis can be found in Section 3.1 and Appendix 6 (Chapter 15 FSAR writeups).
l Small Break LOCA1 l The Small Break LOCA analysis was performed using the NOTRUMP code and the small break version l of the LOCTA code. Analyses were performed for the 1.5,2, and 3 inch break sizes. The Small Break l LOCA analysis was performed for a core power level of 2900 MWt with a total core peaking factor (F )
n l of 2.45, a hot channel enthalpy rise factor (Fw) of 1.62, and a hot assembly average power (P ) of y
l 1.443. The A75 replacement steam generators were modeled in the analysis, assuming a 10% steam l generator tube plugging level with a thermal design flow of 277,800 gpm. Analyses were performed for l the range of reactor coolant average temperatures from 572.0 to 587.4'F. The maior input assumptions l for the analysis are summarized in Table 3.1-2. The results of the Small Break lot
.malysis indicate
~\\
l that the LOCA acceptance criteria in 10CFR50.46 will continue to be met. The detaih and results of the l Small Break LOCA analysis can be found in Section 3.1 and Appendix 6 (Chapter 15 FSAR writeups).
Other LOCA Analyses:
Post-LOCA Long Term Core Cooling Suberiticality (Section 3.1.3) and Hot Leg Switchover to Prevent Boron Precipitation (Section 3.1.4) analyses were performed, incorporating revised operating conditions and the A75 steam generators. The analysis for the post-LOCA long term core cooling showed that the reactor core remains subcritical assuming all control rods out. The analysis for the hot leg switchover to prevent potential boron precipitation showed that switchover to hot leg recirculation within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of a LOCA will prevent boron precipitation in the reactor vessel. Both analyses concluded that operation of the VCSNS with the RSGs and a power level up to 2912 MWt is acceptable.
The LOCA hydraulic forces analysis performed for the RSG conditions (Table 2.1-1) postulated auxiliary line breaks and used the NRC approved Leak-Before-Break (LBB) methodology. The LOCA hydraulic forces were generated for the vessel, loop, and the RSG (A75). The peak lateral LOCA hydraulic load
{
l on the core barrel was determined to be 3.55 x 10" lbf, which is approximately 43 % lower in magnitude 1
l than the previous analysis, which considered a 150 in' rupture of the reactor vessel inlet nozzle. The analysis, thus, yielded considerable margin to the analyses contained in Sections 3.6.2.2.1 and 3.9.3.5 xiv
E of the FSAR. The Model D3 steam generator structural analysis was performed with double-ended guillotine inlet and outlet nozzle breaks, which yield far larger loads than auxiliary line breaks analyzed for this submittal. The new LOCA hydraulic forcing functions were used for qualification of components -
in combination with other applicable design basis loads.
Non-LOCA Analyses; The non-LOCA licensing basis safety analyses have been reanalyzed for the VCSNS plant operation with the A75 steam generator design and a power level up to 2912 MWt. These analyses bound full power nominal RCS average temperatures in the range of 572.0 *F to 587.4 *F. The thermal design parameters assumed in these analyses are provided in Table 2.1-1. The computer codes and methods utilized for these analyses have all been previously approved by the NRC unless otherwise noted. Where appropriate, the Revised Thermal Design Procedure (Reference 1) has been used for the plant specific uncertainties for the evaluation of those transients which are DNB limiting. A detailed description of each of the non-LOCA licensing basis analyses may be found in the revised FSAR writeups in Appendix 6. ' These analyses demonstrate that all licensing basis criteria continue to be met.
HidJncrev Line Break Analyses:
The results of the high energy line break analyses demonstrate that:
1)
For the short term containment response, consisting of the reactor building subcompartment analyses, the current analyses for the steam generator compartment and reactor cavity are shown to remain bounding. The surge line and spray line mass and energy releases are shown to increase; however, large margins continue to be maintained between the calculated and design differential compartment pressures. In summary, the structural integrity of the Reactor Building subcompartments will be maintained for the RSG and associated changes in plant operating conditions.
2)
For the long term analysis of the Reactor Building integrity, it was determined that the RSGs, when analyzed at conditions corresponding to the stretch power leve) of 2912 MWt NSSS result in increases in the Reactor Building pressure and temperature following a design basis LOCA and/or SLB. The calculated peak pressure, however, remains below the Reactor Building design pressure.
Environmental conditions from high energy line breaks, both inside and outside containment, are presented.
Impacts are currently being reconciled to ensure applicable equipment quali0 cation requirements are met (see Section 3.4.6).
Strain Generator Tube Rupture Analysis:
The Steam Generator Tube Rupture (SGTR) Analysis results are summarized in Table 3.5-2.
The parameters in Table 3.5-2 are for the primary to secondary break flow and the atmospheric steam release via the faulted steam generator and are based on the VCSNS SGTR sensitivity analysis. These results can be used to determine the radiological consequences on SGTR for VCSNS with replacement steam generators when operated within the bounds of the design power capability parameters. Note that these results account for Steam Generator replacement, increasing power to the stretch power limit, hot leg temperature reduction, and Steam Generator Tube Plugging programs and are bounding for operation within the ranges of parameters listed in Table 3.5-1.
Ernetor Cavity Pressure Ryalpation:
The reactor cavity pressure evaluation is provided in Section 3.6. Use of the previously approved Leak-Before Break methodology eliminates the dynamic effects of postulated primary loop pipe ruptures from xv
the design basis. This means that the current reactor cavity breaks no longer need to be considered for the short term effects. Since the RCS piping has been eliminated from consideration, the large branch-nozzles must be considered for design verification. This includes the surge line, accumulator line, and the RHR line. These smaller breaks, which are outside the cavity region, result in minimal asymmetrical pressurization in the reactor cavity region. 'Ite decrease in mass and energy releases associated with the smaller piping breaks (as compared to the larger piping breaks) more than offsets the penalties associated with the revised conditions for the replacement steam generators. Therefore, the current mass and energy releases for the RCS nozzle breaks remain bounding for the revised conditions for the reactor cavity r:gion.
Radiolozical Analysis:
The reactor core and reactor coolant iodine and noble gas fission product activities were recalculated to support the radiological consequence analyses in FS AR Chapter 15 with the RSGs, revised design power capability parameters, and transition to VANTAGE + fuel. These fission product activities are utilized in the calculation of offsite doses presented in Section 3.7.3.
Section 3.7.2 provides the ' specific VANTAGE + core, coolant, and fuel handling accident source terms with a comparison to the VANTAGE 5 core and to a generic 2900 MWt core.
NSSS Components:
The NSSS components were evaluated to support the RSG, incorporating an NSSS power level up to 2912 MWt, with up to 10% steam generator tube plugging. The NSSS design transients were reviewed and revised as necessary to incorporate RSG parameters and associated conditions as reflected in Table 2.1-1.
For the NSSS component evaluations in Section 3.8, the most limiting transients for each component were used. The components that were evaluated are as follows:
Reactor Vessel Reactor Internals Steam Generators Pressurizer Reactor Coolant Pumps (RCPs) and RCP Motors Control Rod Drive Mechanism Reactor Coolant Piping and Supports Each primary component evaluation is discussed in more detail in Section 3.8.
The results were compared to the allowable stress and fatigue limits defined by the ASME Code Editions to which the components were originally designed and evaluated. In almost all cases, the revised conditions and transient loadings resulted in stresses and fatigue usage factors below the Code allowable limits. The exceptions are the reactor vessel inlet and outlet nozzles and the reactor vessel head flange, for which the combined primary, secondary, and bending stresses were determined to exceed the Code limit of 3 times the membrane stress (3S.) as determined by the original reactor vessel stress report prepared by Chicago Bridge and Iron Co. As part of the original qualification of these components, simplified elastic-plastic analysis was performed and the results indicated that these components are acceptable. The conservative assumptions of the orig'nal stress report remain bounding for the revised conditions reflected in Table 2.1-1, and therefore tl e original results remain unchanged. Therefore, it has been determined that the NSSS component.e Mi not be adversely affected by the replacement steam generators for an NSSS power level of up to 2912 MWt.
Fluid & Auxiliary Systems Evalualions:
The impact of the RSGs on the integrity and operation of the NSSS fluid systems, the NSSS auxiliary equipment and the NSSS/ Balance-of-Plant interface systems was evaluated. For the NSSS fluid systems xvi
and the NSSS/ Balance-of-Plant interface systems, operation with the RSGs at the current licensed power level of 2787 MWt, for a range of temperatures from T, of 572*F to 587.4*F, was considered. For the auxiliary equipment evaluations, the assumed power level was 2912 MWt, consistent with the conditions of Table 2.1-1. The evaluations discussed in Section 3.9 regarding the fluid and auxiliary systems and NSSS/ BOP interface systems concluded that the design requirements of these systems are met for the RSG and associated primary temperature conditions at the NSSS rated power level evaluated.
Fuel Structural fivalpation; Section 3.10, Fuel Structural Evaluation, summarizes.the evaluation of the impact of the revised conditions associated with the RSGs on the nuclear fuel The design power capability parameters used for the evaluations are those shown in Table 2.1-1. The revised conditions associated with the RSGs have no significant impact on the operating and faulted condition loads, or on the fuel assembly function 11 requirements. Fuel assembly structural integrity is not adversely affected by the average primary temperature reduction, nor by the associated increased lift and buoyancy forces and limiting LOCA pipe break event. Core coolable geometry is maintained under the revised conditions outlined in Section 2.1.
Summary of Technical Specification Chances:
The proposed changes to the VCSNS Technical Specifications are summarized in Section 3.11. These changes reflect the impact of the design, analytical methodology, and safety analysis assumptions outlined in this license amendment request. Marked up Technical Specifications are provided in Appendix 3 of this report. The following areas of the Technical Specifications are affected b) 'he RSG, incorporating a power level up to 2912 MWt, where appropriate.
Core Safety. Limits RCS Flow Negative Flux Rate Trip Overpower AT & Overtemperature AT trip setpoints Steam Generator Water Level Setpoints Shutdown Margin for Modes 3,4, and 5 DNB Parameters Steam Generator Surveillance Reactor Coolant System Volume l
Charging Pump Flow Rates Containment Pressure Sunnortine Documentatiom The 10CFR50.59 Assessment of Unreviewed Safety Questions and 10CFR50.92 No Significant Hazards Determination to support the proposed Technical Specification changes is provided in Appendices 1 and 2, respectively. It was determined that these changes do not involve an unreviewed safety question nor a significant hazards consideration.
Appendix 4 contains WCAP-13480, " Westinghouse A75 Steam Generator Design and Fabrication' information for the VCSNS" (Proprietary Class 2). This WCAP was submitted to the NRC for information on September 3,- 1992. The WCAP provides the following information on the 475 and the original Model D3 steam generator:
475 and Model D3 General Design Features A75 Materials of Construction Material and Manufacturing Processes for Tubes, Tubesheets, and Support Plates Welding Processes Used in Fabrication xvii
. ~. --
f t
Heat Treatment Process A75and Model D3 Performance and Operational Characteristics A reconciliation was performed for the previously approved Leak-Before-Break (LBB) methodology, WCAP-132%, to incorporate the effects of hardware changes and a potential stretch power application.
The hardware changes include removal of several SG support snubbers, removal of crossover leg whip-restraints, and the replacement of the SGs, The reconciliation of the LBB is contained in WCAP-13605 which is included as Appendix 5 of this document.. The results of the calculations performed to reconcile the elimination of the RCS primary loop breaks for the VCSNS under the new loop configuration and potential stretch power application demonstrate that the conclusions reached in WCAP-13206 remain unchanged. Thus, it was concluded that dynamic effects of RCS primary loop pipe breaks need not be -
considered in the structural design basis for VCSNS. This WCAP was submitted to the NRC with the April 30,1993 submittal.
1 The Ch9pter 15 FS AR writeups are provided in Appendix 6.
l The evaluation, results, and supporting test data of the VCSNS runout flow verification testing of the l centrifugal charging / safety injection pumps are documented in Appendix 7, "Si Pump Test Report".
In summary, the safety analyses and evaluations provided with this submittal demonstrate acceptable results in each case, incorporating the revised operating conditions associated with the RSGs.
xviii
l I
1.0 INTRODUCTION
- DESCRIPTION OF LICENSE AMENDMFNr REQUEST 1.1 PURPOSE FOR CIIANGE j
South Carolina Electric & Gas (SCE&G) plans to replace the current Model D3 steam generators with the A75 model (Appendix 4, WCAP-13480, provides a detailed description of the design and fabrication information for the A75 SG). The V.C. Summer Nuclear Station (VCSNS) currently operates at a Rated Thermal Power of 2775 MWt. Accident analyses have been performed, where possible, at a core power level up to 2900 MWt (a 4.5% increase). The accident analyses results are presented in Section 3 of this document. Additionally, in order to permit more flexible plant operation, a range of full power nominal T., values from a maximum value of 587.4 "F to a minimum of 572.0 "F has been analyzed. Normal plant operation is expected to be at 587.4 'F.
Thermal Design flow will be reduced to 92,600 gpm/ loop, to support up to 10% steam generator tube 1
plugging. Minimum Measured Flow will be 283,500 gpm. Table 2.1-1 further delineates the design performance capability parameters for VCSNS with the 475 SGs.
This amendment request reflects the impact of the design, analytical methodology, and safety analysis assumptions on the VCSNS Technical Specifications for the A75 steam generators. These modifications result in several changes to the VCSNS Technical Specifications including:
Core Safety Limits RCS Flow Negative Flux Rate Trip Overpower AT & Overtemperature AT trip setpoints Steam Generator Water Level Setpoints Shutdown Margin for Modes 3,4, and 5 DNB Parameters Steam Generator Surveillance Reactor Coolant System Volume Containment Pressure l
Charging Pump Flow Rates This report supports the requirement for a written safety evaluation and explicitly addresses the regulatory screening criteria of both 10CFR50.59 and 10CFR50.92 in Appendices 1 and 2, respectively. The determination that the subject changes do not involve an unreviewed safety question was made based on the individual evaluations in Section 3 performed in accordance with pertinent licensing basis acceptance criteria for the VCSNS. The applicable acceptance criteria for each evaluation is described in the respective section.
1.2 CURRENT LICENSE BASIS AND FUNCTION OF IDENTIFIED TECIINICAL I
SPECIFICATIONS Provided below is the current license basis for each Technical Specification affected by the 475 SGs and revised operating conditions along with their identified function. A summary of each change is provided in Section 3.11 and the proposed marked up Technical Specification changes are contained in Appendix 3.
l-1 1
Core Safety Limih Technical Specification Figure 2.1-1, Reactor Core Safety Limits, shows the loci of points of thermal power, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less :han the enthalpy of saturated liquid. The figure is based on the enthalpy hot channel factor.
Reactor Trips The reactor trip setpoint limits specified in Table 2.2-1 are the nominal values at which the reactor trips are set for each functional unit. A number of functional units are affected by the RSG and revised operating conditions and are listed below.
Overpower AT Overtemperature AT Pressurizer Pressure - High Steam Generator Level Low-Low Steam /Feedwater Flow Mismatch Coincident with Steam Generator Water Level Low Loss of Flow Neutron Flux High Negative Rate The OPAT trip function provides assurance of fuel integrity (e.g., no fuel melting and less than 1 percent cladding strain) under all possible overpower conditions, limits the required range for OTAT protection, and provides a backup to the High Neutron Flux trip.
The OTAT trip function provides sufficient core protection to preclude DNB over a range of operating and transient conditions. The setpoint is automatically varied with temperature, pressure, and the axial power distribution. The F(AI) penalty function adjusts the trip setpoint for axial peaks greater than design.
The pressurizer pressure high setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the RCS against system overpressure.
The SG water level low-low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater How mismatch resulting from loss of normal feedwater. The specified setpoint is used to start the emergency feedwater system.
The Steam /Feedwater Flow Mismatch Coincident with Steam Generator Water Level Low trip is used to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the RPS. This trip is redundant to the SG water level low-low trip.
The footnote to the loss of flow reactor trip values is used to identify, in absolute terms, the flow values at which the trip function should actuate. Since the Trip Setpoint and Allowable Value are given in terms of relative flow (90% and 88.9%, respectively), the footnote can be used to determine actual flowrates which preserve the basis of the accident analysis. The revision to the minimum measured flow value is used as input to the RTDP DNBR analyses for the loss of flow event.
The deletion of the Power Range, Neutron Flux High Negative Rate is proposed since this reactor trip function is currently not credited in the dropped rod analysis (Technical Specification Table 3.3-2).
The dropped rod analysis is consistent with the EOG program developed in WCAP-1-2 i
.~,
11394," Methodology for the Analysis of the Dropped Rod Event" and the Technical Specification changes are consistent with WCAP-12282, " Implementation Guidelines for WCAP-11394 (Methodology for the Analysis for the Dropped Rod Event)".
Shutdown Marcin Figure 3.1-3 of the Technict.] Specification defines shutdown margin requirements as a function of average RCS boron concentration during Modes 3,4, and 5. In these modes, the most limiting accident is a boron dilution accident. The shutdown margin is varied as a function of average boron concentration in order to provide adequate protection in these modes.
DNB Parameters The DNB parameters (T.,, and Pressurizer Pressure) are found in Technical Specification 3.2.5 (Table 3.2-1) and assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.
Steam Generators TS 4.4.5.3 (Surveillance Requirement) describes the requirement for the first inservice inspection of the steam generators. Tube inspection, sleeve plugging and repair limits and F* and L* criteria are e
discussed in Surveillance Requirements 4.4.5.2, 4.4.5.4, and 4.4.5.5.
Desien Features The' combined water and steam volume of the RCS at an indicated Tu condition is found in Technical Specification 5.4.2.
Containment Pressure Primary containment integrity ensures that the release of radioactive material from the containment atmosphere will be restricted to those leak paths and associated leak rates assumed in the accident analyses. Surveillance requirements for leak rate testing are defined based on the muimum peak pressure expected during an accident.
l Chareine Pumo Flow Rates l The surveillance requirements provided for each ECCS subsystem ensure that at a minimum, the l assumptions used in the safety analyses are met and that subsystem operability is maintained. The l requirements for the flow balance testing provide assurance that proper ECCS flows will be l maintained in the event of a LOCA.
1.3 DESCRIPTION
OF PROPOSED CHANGE The proposed changes to the VCSNS Technical Specifications are summarized in Section 3.11. These changes reflect the impact of the A75 SGs and revised operating conditions. When implemented, the proposed changes will preserve the design, analytical methodology, and safety analysis assumptions outlined in this amendment request.
1-3
3.0 Safety Evaluations / Analyses i
i l
l
.~
3.1 LOSS OF COOLANT ACCIDENT ANALYSFE This section presents the results of Loss of Coolant Accident (LOCA) analyses performed to support the VCSNS replacement steam generators (RSGs). The irr. pact of the RSGs and associated RCS conditions (as described in Section 2.1) was evaluated and the results are presented in the following subsections:
Subsection:
Title:
3.1.1 Large Break (LB) LOCA analysis l
3.1.2 Small Break (SB) LOCA analysis 3.1.3 Post-LOCA Long Term Core Cooling Suberiticality 3.1.4 Hot Leg Switchover to Prevent Potential Boron Precipitation 3.1.1 Imrce Break LOCA
==
Introduction:==
The RSGs affect the LB LOCA analysis in two ways. First, the steam generator performance parameters at steady-state, full-power operation determine the temperature, pressure, and flow conditions which ultimately determine the initial conditions for the LB LOCA calculations, including those for the fuel rods. In addition, the hydraulic and heat transfer characteristics of the steam
. generators are modeled directly in the calculational model, which depicts the specific plant in the analysis. These characteristics are especially important to the NRC approved 1981 Evaluation Model plus BASD leference 1) that was used to perform the VCSNS licensing calculation, since this model s
is composed of a detailed nodal network which models each of the components in the RCS loops.
The A75 steam generators to be installed at VCSNS incorporate new design features and, most significantly, differences in tube length, number of tubes, and tube thickness (Appendix 4). These changes, in turn, lead to differences in flow and heat transfer areas. A complete spectrum of breaks was analyzed for the large break LOCA licensing basis.
Analysis inpal:
The analysis assumed a core power level of 2775 MWt, with a total peaking factor (F ) of 2.45, and n
a hot channel enthalpy rise factor (Fm) of 1.62. An initial RCS pressure of 2300 psia and a Thermal' Design Flow of 277,800 gpm were assumed. A full break spectrum was performed along with a maximum safety injection case for a vessel average temperature range of 572"F to 587.4*F. Both Integral Fuel Burnable Absorbers (IFBA) fuel and non-IFBA fuel were analyzed. Additional input assumptions for the LB LOCA analysis are listed in Table 3.1-1.
R nt[is:
3 All LOCA Acceptance Criteria as described in 10 CFR 50.46 were met. The limiting fuel type was determined to be IFBA, with a 100 psig initial fill pressure. The limiting safety injection case was minimum safety injection. The limiting break size is a Moody Discharge Coefficient of 0.4; this is i
the same limiting break size as the last LB LOCA spectrum performed for VCSNS in 1987. For the j
3.1-1 i
l i
w :
non-IFBA fuel, the low vessel average temperature (T,y) case resulted in a calculated peak fuel j
cladding temperature of 1924*F, which is 3*F higher than the high T., non-IFBA case. For the IFBA fuel, the low T, case resulted in a calculated peak fuel cladding temperature of 2007 F, which is 2*F higher than the high T,, IFBA case.
In conclusion, operation of VCSNS with the A75 steam generators is acceptable with respect to LB.
LOCA, with the analysis described above as the licensing basis LB LOCA analysis (further described in Appendix 6 - Chapter 15 FSAR changes).
3.1.2 Small Break LOCA l
Introduction:
I l Installation of the A75 steam generators and use of 2900 MWt for core power affects the small break l LOCA analysis in several ways. First, the combination of increased power and RSG performance l parameters for steady-state, full-power operation are used to determine the temperature, pressure and l flow conditions which are used as the initial conditions for the small break LOCA analysis.
j Secondly, the increased power level results in higher decay heat levels after reactor trip which j
l directly affects the peak cladding temperature calculations. The design features of the replacement j
l steam generators also result in differences in flow and heat transfer areas, which affects the hydraulic l and heat transfer characteristics during the small break LOCA transient calculations. The different i
l conditions resulting from utilizing a core power of 2900 MWt and replacement of the steam l generators makes it necessary to reanalyze a spectrum of breaks for the small break LOCA.
I l Analysis Inpg i
I d
l The small break LOCA analysis was performed using the NOTRUMP code (References 4 and 5) and
~
l the small break version of the LOCTA code (Reference 6). Analyses were performed for the 1.5,2, j and 3 inch break sizes. The small break LOCA analysis was performed for a core power level of l 2900 MWt with a total core peaking factor _(F ) of 2.45, a hot channel enthalpy rise factor (FJ of n
l 1.62, and a hot assembly average power (Fu ) of 1.443. The A75 replacement steam generators were l modeled in the analysis, assuming a 10% steam generator tube plugging level with a thermal design l flow of 277,800 gpm. Analyses were performed for the range of reactor coolant average l temperatures from 572.0 to 587.4*F. The major input assumptioas for the analysis are summarized l in Table 3.1-2.
l l Results:
l l A peak cladding temperature of 1840"F was calculated for the limiting 2 inch break for operation at l an average coolant temperature of 587.4*F. The peak cladding temperature of 1840*F is based on l the results of a calculation performed for beginning of core life conditions plus a 32*F penalty to l account for fuel rod burst / blockage effects which are projected for later in life, a 16*F benefit for l previous permanent small break LOCA model assessments, and a 1*F penalty for a fuel reconstitution l safety evaluation. T he results of the small break LOCA analysis indicate that the peak cladding.
l temperature will be significantly less than the 10CFR50.46 limit.
I 3.!-2 3:
+
.r..
l In conclusion, operation of VCSNS with the A75 steam generators and with a core power up to 2900 l MWt is acceptable with respect to small break LOCA. Further details on the small break LOCA l analyses are provided in Appendix 6 - Chapter 15 FSAR changes.
3.1.3 Post-LOCA Lonz Term Core Coolire Suberiticality References 2 and 3 present the Westinghouse lice'nsing position for satisfying the requirements of 10 CFR 50.46 Paragraph (b), item (5), "Long Term Cooling" The Westinghouse position concludes that the core will remain shut down by borated ECCS water residing in the RCS/ sump after a LOCA.
Since credit for the control rods is not taken for a LB LOCA, the borated ECCS water provided by the accumulators and the RWST must have a boron concentration that, when mixed with other water sources, will result in the reactor core remaining suberitical assuming all control rods out. The calculation is based upon the steady state conditions at the initiation of a LOCA and considers sources of both borated and unborated fluid in the post-LOCA containment sump. The steady state conditions are obtained from the LB LOCA analysis. A post-LOCA long term core cooling subcriticality analysis was performed, incorporating the 475 steam generators and an NSSS power level of 2912 MWt The analysis showed that the reactor core remains suberitical assuming all control rods out.
Therefore, it is concluded that operation of VCSNS with the A75 steam generators and higher, bounding power level of 2912 MWt is acceptable with respect to post-LOCA long term core cooling subcriticality.
3.1.4 Ilot In Switchover to Prevent Potential Boron Precinitation Post-LOCA hot leg recirculation time is determined for inclusion in emergency operating procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. This time is dependent on power level and on the RCS, RWST, and accumulator water volumes with their associated boron concentrations. An analysis for hot leg switchover to prevent potential boron precipitation was performed, incorporating the A75 steam generators and an NSSS power level of 2912 MWt. The analysis showed that switchover to hot leg recirculation within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of a LOCA will prevent boron precipitation in the reactor vessel. Thereafter, the operator will alternate between cold and hot leg recirculation every 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. In conclusion, operation of VCSNS with the 475 steam generators and the higher, bounding power level of 2912 MWt is acceptable with respect to hot leg switchover to prevent potential boron precipitation.
l 1
3.1-3
l
)
3.1.5 References 1.
WCAP-10266, Rev. 2, with Addenda (Proprietary), "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code" August,1986.
2.
WCAP-8339 (Non-Proprietary), " Westinghouse Emergency Core Cooling System Evaluation Model - Summary", June 1974.
3.
Westinghouse Technical Bulletin NSID-TB-86-03, " Post-LOCA Long-Term Cooling: Boron Requirements", October 31, 1986.
l 4.
WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model Using the l
NOTRUMP Code", Lee, N. et al., August 1985.
1 l S.
WCAP-lll45, " Westinghouse Small break LOCA ECCS Evaluation Model Generic Study l
With the NOTRUMP Code", Rupprecht, S.D., et al.
I l
6.
WCAP-8301, "LOCTA-IV Program: Loss of Coolant Transient Analysis", Bordelon, F.M.,
l et al., June 1974.
e I
i 3.1-4 4
+._ _
0'
\\
TABLE 3.1-1 INPUT ASSUMPTIONS FOR LBLOCA REPLACEMENT SG ANALYSIS Type of break Double-Ended Cold Leg l
Guillotine (DECLG)
Core Power Level 2775 MWt Steam Generator Tube Plugging 10 %
l Fuel Assembly Array 17X17 Vantage +/ Vantage 5 Target Core Peaking Factor, F 2.45 q
Hot Channel Enthalpy Rise Factor, F, 1.62 RCS Pressure (including uncertainties) 2300 psia Thermal Design Flow (Reactor) 277800 gpm Vessel Average Temperature 585.4/569.9"P" Hot Leg Temperature 619.3/604.6'Fm Cold Leg Temperature 551.5/535.2*PU Steam Temperature 537.6/520.9"Pu Steam pressure 944.0/818.8 psia
- Steam Flow (total) 12.411X106/12.348X10' lb/hrm Feedwater Temperature 437.0"F Minimal Accumulator Gas Pressure 600 psia Accumulator Temperature 120"F Intact Accumulator Tank Volume 2975.6 ft'/acc.m
-)
Broken Accumulator Tank Volume 1491.2 ft'/ ace.*
'l Intact Accumulator Water Volume 2078.0 ft'/acc.*
Broken Acenmulator Water Volume 1042.4 ft'/acc.m Power Shapi Chopped Cosine l
Fuel Back Pi ssure 275, 200,100 psig*
j Reactor Trip Setpoint 1845 psia Safe.ty injection Signal Setpoint 1715 psia Safety injection Delay Time 27.0 sec S1 Spilling Containment Press.
0.0 psig j
i 1.
Nominal / Reduced RCS temperature program. These values are based on design power capability parameters but differ slightly from them because bounding assumptions are used to
)
calculate the values used in the large break LOCA analysis.
2.
The accumulator line volume was added to the accumulator tank and water volume in the large break LOCA analysis.
l 3.
Analysis and results for 275 psig fuel back pressure are also applicable for 200 psig and 100 l
psig fuel back pressure.
H 3.1-5 l
i
l
' 'I ABLE 3.1-2 i
j INPUT ASSUMPTIONS FOR SBI,0CA ANALYSIS I
l Type of Break 1.5,2, and 3 inch cold leg l
Core Power Level 2900 MWt l
Steam Generator Type A75 l
Steam Generator Tube Plugging 10 %
l Fuel Assembly Array 17X17 Vantage +/ Vantage 5 l-Target Core Peaking Factor, F 2.45 q
l Hot Channel Enthalpy Rise Factor, F 1.62 w
l Hot Assembly Average Power, Pm 1.443 l
Axial Offset
+ 13 %,-20 %
l RCS Pressure (including uncertainties) 2300 psia l
Thermal Design Flow (Reactor) 277,800 gpm l
Vessel Average Temperature 587.4/572.00 F l
Hot Leg Temperature 622.6/608.08"F l
Cold Leg Temperature 552.2/535.92"F l
Steam Temperature 537.86/521.11
- F l
Steam pressure 945.75/820.37 psia l
Steam Flow (total) 13.065X10^/12.996X10' lb/hr l
Feedwater Temperature 442.0*F l
Pumped Safety Injection Flow Rate See FSAR Figure 15.3-1 l
Nominal Acetimulator Water Volume 1014 ft'/ accumulator l
Accumulator Tank Volume 1450 ft*/ accumulator l
Minimal Accumulator Gas Pressure 600 psia l
Fuel Back Pressure 275,200,100 psig"'
l Reactor Trip Setpoint 1845 psia l
Safety injection Signal Setpoint 1715 psia l
Safety injection Delay Time 27.0 sec l
Auxiliary Feedwater Delivery Delay Time 60.0 sec l
Signal Processing Delay and Rod Drop Time 5.7sec l
Reactor Coolant Pump Delay Time 0.0 sec I
1 I
l l
1.
Analysis and results for 275 psig fuel back pressure are also applicable for 200 psig and l
100 psig fuel back pressure.
-l 3.1-6
~
3.11
SUMMARY
OF TECHNICAL SPECIFICATIONS CilANGES The proposed changes to the Virgil C. Summer Nuclear Statien Technical Specifications are summarized in Table 3.11-1. These changes reflect the impact of the design, analytical methodology, and safety analysis assumptions outlined in this amendment request and are provided in the proposed Technical Specification page changes (See Appendix 3 of this report). A brief overview of the significant changes follows.
Core Safety Limits Core safety limits and associated bases for 3-loop operation during modes 1 & 2 (Figure 2.1-1) are revised to reflect the impact of the A75 RSGs and a core power level up to 2900 MWt through 1.
The increase in the Reactor Core lleat Transfer Rate up to 2900 MWt.
2.
The reduced Steam Generator Tube Plugging,0% to 10%.
3.
The application of the Revised Thermal Design Procedure.
J.tCS Flow The revision proposed for Table 2.2-1 corresponds to the minimum measured flow value used as input to the RTDP DNBR analyses for the loss of flow event.
The RCS Total Flow Rate as shown in the Figure 12 of the COLR has been reduced. Indicated RCS flow is derived from the thermal design flow based on the flow measurement uncertainty of 2.1 %
including a 0.1 % uncertainty for Feedwater Venturi fculing.
These changes to the RCS flow rate are proposed to accommodate the following:
1.
The differences between the old and the new steam generators.
2.
Up to 10% Steam Generator tube plugging in all three Steam Generators.
Negatiltflux Rate Trip The deletion of the Power Range, Neutron Flux High Negative Rate is proposed since this reactor trip function is currently not credited in the dropped rod analysis. The dropped rod analysis is consistent with the yj0G program developed in WCAP-11391," Methodology for the Analysis of the Dropped Rod Event" and the Technical Specification changes are consistent with WCAP-12282,
" Implementation Guidelines for WCAP-11394 (Methodology for the Analysis for the Dropped Rod Event)"
3~.1 1 -1 h
QPAT/OTAT Setpoints T,
Revisions to the limiting safety system settings for the Tncrmal Overpower AT and Thermal Overtemperature AT trip functions are proposed to maintain consistency with the non-LOCA Accident Analysis. These trip functions provide primary protection against departure from nucleate boiling and fuel centerline melting (excessive kw/ft) during postulated transients. The proposed settings have been based on the new core safety limits and account for instrument uncertainties. The reference temperatures are now indicated values and the temperature range that was analyzed is specified.
Steam Generator Water Level Setpoints Revisions to the Steam Genera >or Water Level Low-Low setpoints are proposed to incorporate the.
differences between the current Model D3 Steam Generators and the A75 Replacement Steam Generators (e.g., constant level control). The setpoints are given for both Barton transmitters and Rosemount transmitters. The bases also show an increase to the steam /feedwater flow mismatch activation setpoint.
Similar changes are also proposed for the Steam Generator Water Level High-High for Turbine Trip and Feedwater Isolation wi h Engineered Safety Features. Setpoint changes have been provided for t
both Barton and Rosemount transmitters.
Shutdown Marnin for Modes 3. 4. and_5 i
Figure 3.1-3 of the Technical Specifications defines shutdown margin requirements as a function of average P.CS boron concentration during Modes 3,4, and 5. The proposed revisions are driven by both the steam generator replacement and the increase in the NSSS power limit and are required to maintain the current bases of the Technical Specifications.
1 DNB Parameters The proposed changes to the DNB related parameters (T, and Pressurizer Pressure) assure that each parameter is maintained within the normal steady state envelope of operation assumed in both the i
transient and the accident analysis. The proposed changes are consistent with the accident analysis, l
Both parameters now represent indicated values by including allowance for reading and averaging
- three control board indications. Pressurizer pre-ssure has been changed to gauge pressure since these units are read from the control boards.
Steam Generator Surveillance The replacement of steam generators requires that a first inservice inspection of the steam generators be performed again as required in Technical Specification 4.4.5.3. To add clarity to this surveillance, the first sentence of the surveillance which describes when the "first inservice inspection" should take place has been modified. The surveillance must take place after at least 6 Effective Full Power c
3.11-2
Mor.ths (EFPM) from the time of the replacement but within 24 calendar months of initial criticality after the steam generator replacement.
The F*and the L* criteria have been deleted from the Technical Specifications since these criteria are not applicable to the A75 replacement steam generators. The Model D3 SG sleeving process reports have also been removed since these reports are not applicable to the A75 replacement steam generators.
Reactor Coolant System Volume The combined water and steam volume of the Reactor Coolant System at an indicated Tu condition has been increased to address the differences between the old and new steam generators and includes up to 10% steam generator tube plugging.
Containment Pressure The limiting conditions for operation and surveillance requirements for the containment systems are deendent in part on the maximum peak pressure during an accident. The proposed values for P, and P affect the maximum containment pressure calculated for the SLB accident with the A75 SGs.
Charnine Pump Flow Rates The revision to the maximum allowed flow for the centrifugal charging pump is proposed to increase the system's operating margin by providing a larger difference between minimum / maximum flow limits. This revision is supported by the S1 Pump Test Report provided in Appendix 7. No change in minimum required pump flow or developed head is necessary to support the revised safety _
analyses.
l 3.11-3
TABLE 3.11-1 VCSNS TECIINICAL SPECIFICATION CIIANGES FOR STEAA1 GENERATOR REPLACEh1ENT PAGE SECTION DESCRIPTION OF CH ANC6 JUSTIFICATION 2-2 Figure 2.1-1 Core Limits are revised Core limits are revised due to analysis at higher core power.
2-5, Table 2.2-1 Various Reactor Trip System Setpoints are consistent with the 2-6, Setpoints are changed new safety limits, instrument 2-8, uncertainty and flow. These 2-9, changes are necessary to support 2-10 both the steam generator replacement and stretch power conditions.
2-5 Table 2.2-1 Delete Negative Flux Rate Trip This trip is not credited in current B 2-4 2.2.1 Bases analysis, removal is consistent with 3/4 3-2 Table 3.3-1 WCAP-12282.
3/4 3-9 Table 3.3-2 3/4 3-11 Table 4.3-1 2-6 Table 2.2-1
" Steam Generator Water Level This change is to clarify the name of Low-Low" changed to " Steam the reactor trip to be consistent with Generator Water Level Low" the Bases, with the Tables in section 3.3.1 and with the functional diagrams.
B 2-1 2.1.1 Bases Reference to specific correlations Future changes to the correlation -
used in DNB analysis removed.
used in analysis will not require a change to the Bases.
B 2-6 2.2.1 Bases Steam /Feedwater Flow hiismatch This change is necessary to support activation setpoint is increased the steam generator replacement.
3/4 1-3a Figure 3.1-3 Shutdown hiargin limits are revised This change is associated with both.
the Steam Generator Replacement and stretch power conditions.
3'.11-4
. TABLE 3.11-1 VCSNS TECIINICAL SPECIFICATION CIIANGES FOR STEAM GENERATOR REPLACEMENT PAGE SECTION DESCRIPTION OF CHANGE JUSTIFICATION 3/4 2-8 3.2.3 Change to uncertainty to include This uncertainty must be taken into B 3/4 2-4 3/4.2.3 Bases 0.1% uncertainty for feedwater account per NRC request.
venturi fouling 3/4 2-9 4.2.3.5 RCS flowrate method IV Methodology for RCS flow measurement.
3/4 2-16 Table 3.2-1 Change to DNB Parameters to This change is associated with both B 3/4 2-5 3/4.2.5 Bases include indication uncertainties the steam generator replacement and stretch power conditions.
3/4 3-28 Table 3.3-4 Change to ESFAS Steam Generator This change is associated with the Water Level High-High and Low-steam generator replacement.
Low setpoints 3/4 4-11 3/4.4.5 Removal of F*and L* criteria and This change is associated with the 3/4 4-12 the licensed sleeving process steam generator replacement.
3/4 4-14 reports.
3/4 4-14a 3/4 4-15 3/4 4-15a B 3/4 4-3 3/4.4.5 Bases 3/4 4-13 4.4.5.3.a First inservice inspection will have This c6nge is associated with the to be performed again for the new steam ge.,erator replacement.
Steam Generators. The wording of this SR is changed to redect this.
3/4 5-6 4.5.2.h.I Change to the maximum allowed This change reDects the results of Cow for the SI/ Charging Pumps.
onsite pump runout testing and is -
proposed to increase the system's operating margin by providing a larger difference between minimum / maximum flow limits.
3.Il-5
TABLE 3.11-1 VCSNS TECIINICAL SPECIFIC ATION CIIANGES FOR STEAM GENERATOR REPLACEMENT PAGE SECTION DESCRIPTION OF CH ANGE JUSTIFICATION 3/4 6-1 4.6.1.1.c Revise reference values for P, and This change reflects the maximum 3/4 6-2 3.6.1.2.a.1 P,.
containment pressure during a SLB.-
3.6.1.2.a.2 4.6.1.2.a 3/4 6-3 4.6.1.2.c.3 4.6.1.2.d 3/4 6-4 3.6.1.3.b 3/4-6-5 4.6.1.3.b B3/4 6-2 3/4.6.1.4 Bases 3/4.6.1.6 Bases 5-6 5.4.2 Increase in total RCS water / steam This change is associated with the volume.
steam generator replacement.
5-6 5.4.2 T,, is changed from a nominal This change is for clarity and value to an indicated value.
consistency throughout the Technical Specifications.
3.11-6 j
4.0 CONCLUSION
S Provided in this document are the results and conclusions of the safety analyses and evaluations to support the 475 SGs and revised operating conditions Technical Specification changes for the VCSNS. As str.ted previously, all supporting documentation for the license amendment request is l contained in this document. The results of the Small 13reat LOCA analysis and associated Technical l Specification changes with supporting documentation (e.g., Safety Injection pump test report) were l submitted as replacement pages to the 10/29/93 submittal.
The safety analyses, evaluations, and supporting documentation provided with this submittal demonstrate acceptable results in each case, incorporating the revised operating conditions associated with the RSGs. A brief summary of the results of each analysis, evaluation, and supporting documentation is provided in the Executive Summary.
4-1
APPENDIX 1 Assessment of Unreviewed Safety Questions 10CFR50.59 Operation of VCSNS with the proposed A75 SGs, revised operating conditions, and requested changes to the Technical Specifications has been evaluated using the guidance of NSAC-125 and 10CFR50.59 i
and been determined to not represent an unreviewed safety question. A summary justification for this determination follows:
1.
Will the probability of an accident previously evaluated in the FSAR be increased?
No.
Implementation of the A75 SGs and revised operating conditions do not contribute to the initiation of any accident evaluated in the FSAR. Supporting factors are as follows:
The 475 SG is designed in accordance with ASME Code Section 111,1986 edition and other applicable federal, state, and local laws, codes and regulations and meets the original interfaces for the Model D3 SGs with the exception that provisions for a larger blowdown nozzle have been made and the feedwater inlet nozzle is located in the upper shell.
All NSSS components (i.e., reactor vessel, RC Pumps, pressurizer, CRDMs, A75 SGs, and RCS piping) are compatible with the revised operating conditions. Their structural integrity is maintained during all proposed plant conditions through compliance with the ASME code.
l Fluid and auxiliary systems which are important to safety, including the CHG/SI l
system with maximum pump flows up to 700 gpm, are not adversely impacted and l-will continue to perform their design function.
i Overall plant performance and operation are not significantly altered by the proposed changes.
Therefore, since the reactor coolant pressure boundary integrity and system functions are not adversely impacted, the probability of occurrence of an accident evaluated in the VCSNS FSAR will be no greater than the original design basis of the plant.
l i
2.
Will the consequences of an accident previously evaluated in the FSAR be increased?
l l
No.
An extensive analysis has been performed to evaluate the co1 sequences of the following accident types currently evaluated in the VCSNS FSAR:
Non-LOCA Events l
Large Break and Small Break LOCA Steam Generator Tube Rupture r,
{
r
= -..
With the A75 SGs and revised operating conditions, the calculated results (i.e., DNBR, Primary and Secondary System Pressure, Peak Clad Temperature, Metal Water Reaction, Challenge to Long Term Cooling, Environmental Conditions Inside and Outside Containment, etc.) for the accidents are similar to those currently reported in the VCSNS FSAR. Select results (i.e., Containment Pressure during a Steam Line Break, Minimum DNBR for Rod Withdrawal from Suberitical, etc.) are slightly more limiting than those currently reported in the FSAR due to the use of the assumed operating conditions with the new A75 SGs, and in some cases, use of an uprated core power of 2900 MWt. However, in all cases, the calculated results do not challenge the integrity of the primary / secondary / containment pressure boundary and remain within the regulatory acceptance criteria applied to VCSNS's current licensing basis. The assumptions utilized in the radiological evaluations, described in Section 3.7, are thus appropriate and are judged to provide a conservative estimate of the radiological consequences during accident conditions. Given that calculated radiological consequences are not significantly higher than current FSAR results and remain well within 10CFR100 limits, it is concluded that the consequences of an accident previously evaluated in the FSAR are not increased.
3.
May the possibility of an accident which is different than any already evaluated in the FSAR be created?
No.
l The A75 SGs, revised operating conditions, and higher allowable CllO/SI pump flows will l
not introduce any new accident initiator mechanisms. Structural integrity of the RCS is maintained during all plant conditions through compliance with the ASME code. No new failure modes or limiting single failures have been identified. Design iequirements of auxiliary systems are met with the RSGs. Since the safety and design ' equirements continue c
to be met and the integrity of the reactor coolant system pressure boundary is not challenged, no new accident scenarios have been created. Therefore, the types of accidents defined in the FSAR continue to represent the credible spectrum of events to be analyzed which determine safe plant operation.
4.
Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
F No.
The 475 SGs and revised operating conditions will not adversely affect the operation of the Reactor Protechon System, Engineering Safety Features, or other systems, components, or devices regi ed for accident mitigation. These systems will remain qualified and capable to perform tFeir design function for the revised operating conditions during normal operation and conditions which can evolve during accident conditions. In addition, the NSSS components are compatible with the revised operating conditions and will continue to meet their original design requirement. The integrity of the primary / secondary / containment pressure boundary l
during normal operation and accident conditions will not be challenged. The CHG/SI system l
will also continue to perform its design function with pump flows up to 700 gpm. Based on the above, it is concluded that the probability of a malfunction of equipment important to safety currently evaluated in the VCSNS FSAR will not be increased.
e
5.
Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR he increased?
No.
An extensive analysis has been performed to evaluate the consequences of the following accident types currently evaluated in the VCSNS FSAR:
Non-LOCA Events l
Large Break and Small Break LOCA Steam Generator Tube Rupture Consistent with VCSNS's current licensing basis, the effects of a single failure of equipment important to safety have been considered when evaluating the accident consequences. With the A75 SGs and revised operating conditions, the calculated results for the accidents are similar to those currently reported in the VCSNS FSAR, with select results being slightly more limiting than those currently reported in the FSAR due to the use of the revised operating conditions with the new A75 SGs and, in some cases, use of an uprated core power of 2900 MWt. However, in all cases, the calculated results do not challenge the integrity of the primary / secondary / containment pressure boundary and remain within the regulatory acceptance criteria applied to VCSNS's current licensing basis. Systems and components responses to accident scenarios are not affected. The assumptions utilized in the radiological evaluations, described in Section 3.7, are thus appropriate and are judged to provide a conservative estimate of the radiological consequences during accident conditions Given that calculated radiological consequences are not significantly higher than current FSAR results i
and remain well within 10CFR100 limits, it is conc 1"<'
that the consequences of a malfunction of equipment important to safety previouf., evaluated in the FSAR are not increased.
6.
May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?
No.
l The A75 SGs, revised operating conditions, and higher allowable CHG/SI pump flows will l
l not adversely affect Qe operation of the Reactor Protection System, Engineering Safety Features, or other systems, components, or devices required for accident mitigation. These i
systems will remain qualified and capable to perform their design function for the revised operating conditions durhw normal operation and conditions which can evolve during accident conditions. No new failure modes are created.
The A75 SGs also meet the original interfaces for the Model D3 SGs with the exception that j
provisions for a larger blowdown nozzle have been made and the feedwater inlet nozzle is j
located in the upper shell. These exceptions are judged to not create the possibility of a new equipment malfunction.
The NSSS components, including the new SGs, are also compatible with the revised operating conditions and will continue to meet their original design requirements. Furthermore, analyses demonstrate that the integrity of the primary / secondary / containment pressure boundary will not be challenged during normal operation or accident conditions,
Based on the above, it is concluded that the possibility of a malfunction of equipment important to safety different than any already esaluated in the VCSNS FSAR will not be created.
7.
Will the margin of safety as defined in the BASES to any technical specification he reduced?
No.
l Although the A75 SGs, revised operating conditions, and higher allowable CHG/SI pump l
flows will require changes to the VCSNS Technical Specifications, it.will not invalidate the l
LOCA, non-LOCA, or SGTR conclusions presented in the FSAR accident analyses. ~ For all the FSAE non-LOCA transients, the DNB design basis, primary and secondary pressure limits, r.nd dose limits continue to be met. The LOCA peak cladding temperatures remain below the limits specified in 10CFR50.46. The calculated doses resulting from a SGTR event will wantinue to remain within a small fraction of the 10CFR100 permissible releases.
Environmental conditions associated with High Energy Line Breaks (HELB) both inside and outside containment have been evaluated. The containment design pressure will not be violated as a result of the HELB. Eqiiipment qualification will be updated, as necessary, to reflect the revised conditions multing from tiELB. The margin of safety with respect to primary pressure Soundary '.s provided, in part, by the safety factors' included in the ASME Code, Since the esmponeuts remain in compliance with the codes and standards in effect when VCSNS was originally licensed (with the exception of the A75 RSGs which use the 1986 ASME Code Section III Edition), the margin of safety is not reduced. Thus, there is no reduction in the margin to safety as defined in the bases of the VCSNS Technical Specifications. As stated above, changes will be required to the Technical Specifications in order to implement the proposed modification.
l APPENDIX 2 No Significant llazards Determination 10CIM50.92 i
Several changes will be made to the VCSNS Technical Specifications to support the A75 RSG at a power level up to 2912 MWt. Pursuant to 10 CFR 50.92 each application for amendment to an operating license must be reviewed to determine if the proposed change involves a significant hazards consideration. The amendment describing the technical specification changes, as defined in this report, has been reviewed and deemed not to involve significant hazards considerations. As discussed in Appendix 1, all applicable acceptance criteria are satisfied. Thus, the proposed Technical Specification changes do not constitute an unreviewed safety question and the accident analyses support the changes as discussed in Section 3.0. A description of the changes is provided in Section 1.0. The proposed changes are outlined in Section 3.11. The basis for this determination follows.
]
1)
Operation of VCSNS in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Implementation of the A75 SGs and revised operating conditioris do not contribute to the initiation of any accident evaluated in the FSAR. Supporting factors are as follows:
The A75 SG is designd in accordance with ASME Code Section 111,1986 edition and other applicable federal, state, and local laws, codes and regulations and meets the original interfaces for the Model D3 SGs with the exception that provisions for a larger blowdown nozzle have been made and the feedwater inlet nozzle is located in the upper shell._
All NSSS components (i.e., reactor vessel, RC Pumps, pressurizer, CRDMs, A75 SGs, and RCS piping) are compatible with the revised operating conditions. Their structural integrity is maintained during all proposed plant conditions through compliance with the ASME code.
l Fluid and auxiliary systems which are important to safety, including the CHG/SI l
system with maximum pump flows up to 700 gpm, are not adversely impacted and I
will continue to perform their design function.
Overall plant performance and operation are not significetly altered by the proposed changes.
Therefore, since the reactor coolant pressure boundary integrity and system functions are not adversely impacted, the probability of occurrence of an accident evaluated in the VCSNS FS AR will be no greater than the original design basis of the plant.
An extensive analysis has been performed to evaluate the consequences of the following accident types currently evaluated in the VCSNS FSAR:
Non-LOCA Events l
Large Break and Small Break LOCA Steam Generator Tube Rupture
.~
With the A75 SGs and revised operating conditions, the calculated results (i.e., DNBR, Primary and Secondary System Pressure, Peak Clad Temperature, Metal Water Reaction, Challenge to Long Term Cooling, Environmental Conditions Inside and Outside Containment, etc.) for the accidents are similar to those currently reported in the VCSNS FSAR. Select results (i.e., Containment Pressure during a Steam Line Break, Minimum DNBR for Rod Withdrawal from Suberitical, etc.) are slightly more limiting than those currently reported in the FSAR due to the use of the assumed operating conditions with the new A75_ SGs, and in some cases, use of an uprated core power of 2900 MWt. However, in all cases, the calculated results do not challenge the integrity of the primary / secondary / containment pressure boundary and remain within the regulatory acceptance criteria applied to VCSNS's current -
licensing basis. The assumptions utilized in the radiological evaluations, described in Section 3.7, are thus appropriate and are jud, i to provide a conservative estimate of the radiological consequences during accident conditions. Given that calculated radiological consequences are not significantly higher than current FSAR results and remain well within 10CFR100 limits, it is concluded that the consequences of an accident previously evaluated in the FSAR are not increased.
f 2)
The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
l The 475 SGs, revised operating conditions, and higher allowable CHG/SI pump flows will l
not introduce any new accident initiator mechanisms. Structural integrity of the RCS is maintained during all plant conditions through compliance with the ASME code. No new failure modes or limiting single failures have been identified. Design requirements of auxiliary systems are met with the RSGs. Since the safety and design reqeirements continue to be met and the integrity of the reactor coolant system pressure baundary is not challenged, no new accident scenarios have been created. Therefore, the types of accidents defined in the FSAR continue to represent the credible spectrum of events to be analyzed which determine -
safe plant operation, 3)
The proposed license amendment does not involve a significant reduction in a margin of safety.
l Although the A75 SGs, revised operating conditions, and higher allowable CHG/S! pump. _
l flows will require changes to the VCSNS Technical Specifications, it will not invalidate the l
LOCA, non-LOCA, or SGTR conclusions presented in the FSAR accident analyses. ' For all the FSAR non-LOCA transients, the DNB design basis, primary and secondary pressure-limits, and dose limits continue to be met.. The LOCA peak cladding temperatures remain '
below the limits specified in 10CFR50.46. The calculated doses resulting from a SGTR event will continue to remain within a small fraction of the 10CFR100 permissible releases.
Environmental conditions associated with High Energy Line Breaks (HELB) both inside and outside containment have been evaluated. The containment design pressure will not be -
violated as a result of the HELB. Equipment qualification will be updated, as necessary', to 6
redect the revised conditions resulting from HELB. The margin of safety with respect to primary pressure boundary is provided, in part, by the safety factors included in the ASME Code. Since the components remain in compliance with the codes and standards in effect when VCSNS was originally licensed (with the exception of the A75 RSGs which use the 1986 ASME Code Section til Edition), the margin of safety is not reduced. Thus, there is no reduction in the margin to safety as defined in the bases of the VCSNS Technical Speci0 cations.
a
i
-l i
I APPENDIX 3 l
PROPOSED TECIINICAL SPECIFICATION CIIANGES i
I i
l I
l H
1 l
I
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) h.
By performing a flow balance test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:
1)
For centrifugal charging pump lines, with a single pump running and with recirculation flow; a)
The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 338 gpm,
.l_
and 700 b)
Thetotalpumpflowrateislessthanorequaltop86gpm.
i.
By perforneing a flow test, during shutdown, following completion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying.that:
1)
For residual heat removal pump lines, with a single pump running the sum of the injection line flow rates is greater than or equal to 3663 gpm.
9 SUMMER - UNIT 1 3/4 5-6 Amendment No.
75
SMALL BREAK LOCA FSA.R CIIAPTER 15 WRITE-UPS
15.3 CONDITION III - INFREQUENT FAUI.TS By definition Condition III occurrences are faults which may occur very infrequently during the life of the plant. They will be accomodated with the failure of only a small fraction of the fuel rods although sufficient fuel damage might occur to preclude resumption of the operation for a considerable outage time.
The release of radioactivity will not be sufficient to interrupt or restrict public use of those areas beyond the exclusion radius.
A Condition III fault will not, by itself, generate a Condition IV fault or result in a consequential loss of function of the reactor coolant system or containment barriers.
For the purposes of this report the following faults have been grouped into this categoryt 1.
Loss of reactor coolant, from small ruptured pipes or from cracks in large pipes, which actuates the emergency core cooling system.
2.
Minor secondary system pipe breaks.
3.
Inadvertent loading of fuel assembly into an improper position.
4 Complete loss of forced reactor coolant flow.
5.
Waste gas decay tank rupture.
6.
Single rod cluster control assembly withdrawal at full power.
Each of these infrequent faults are analyzed in this section.
In general, each analysis includes an identification of causes and description of the accident, an analysis of effects and consequences, a presentation of results, and relevant conclusions.
The time sequence of events during applicable Condition III faults 1 and 4 above is shown in Table 15.3-1.
15.3.1 LOSS OF REACIOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES WHICH ACTUATES THE EMERGENCY CORE COOLING SYSTEM 15.3.1.1 Identification of Causes and Accident Description A loss of coolant accident is defined as a rupture of the reactor coolant system piping or of any line connected to the system.
See Section 5.2 for a more detailed description of the loss of reactor coolant accident boundary limits. Ruptures of small cross section will cause exrulsion of the coolant at a rate which can be accommodated by the charging rumps which would maintain an operational water level in the pressurir,er, permitting the operator to execute an orderly shutdown. The coolant which would be released to the containment contains the fission products existing in it.
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 centrifugal charging pump is typically adequate AMENDMENT 6 15.3-1 AUGUST, 1990
to sustain pressurizer level at 2250 psia for a break through a 3/8 inch diameter hole. This break results in a loss of approximately 17.5 lb/sec.
Should a larger break occur, depressurization of the reactor coolant system causes fluid to flow to the reactor coolant system from the pressurizer, resulting in a pressure and level decrease in the pressurizer.
Reactor trip c; curs when the pressurizer low pressure trip setpoint is reached. The safety injection system is actuated when the appropriate setpoint is reached. The consequences of the accident are limited in two ways:
1.
Reactor trip and borated water injection complement void formation in causing rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay.
2.
Injection of borated water ensures sufficient flooding of the core to prevent excessive clad temperatures.
Before-the treak occurs, the plant is in an equilibriun condition, i.e.,
the heat genertted in the core is being removed via the secondary system. During blowdown, heat from decay, hot internals and the vessel continues to be transferred to the reactor coolant system.
The heat transfer between the reactor coolant system and the secondary system may be in either direction, depending on the relative temperatures.
In the case of continued heat addition to the secondary, system pressure increases and steam dump may occur. Makeup to the secondary side is automatically provided by the emergency feedwater pumps. The safety injection signal stops normal feedwater flow by closing the main feedwater line isolation valves and initiates emergency feedwater flow by starting the emergency feedwater pumps. The secondary flow aids in the reduction of reactor coolant system pressure. When the reactor coolant system depressurizes to 600 psia, the accumulators begin to inject water into the reactor coolant loops.
The reactor coolant pumps are assumed to be tripped at the initialization of the accident, and effects of pump coastdown are included in the blowdown analyses.
15.3.1.2 Analysis of Effects and Consecuences 15.3.1.2.1 Method of Analysis For loss-of-coolant accidents due to small breaks less than 1 square foot, the NOTRUMP computer code {l3.14] 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 incorporating a number of advanced features. Among these are calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with i
counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes and regime-dependent heat transfer correlations. The NOTRUMP small-break LOCA emergency core cooling (ECCS) evaluation model was developed to determine the RCS response to design basis I
small break LOCAs, and to address NRC concerns expressed in NUREC-0611, "Ceneric Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants".
\\
AMENDMENT 6 15.3-2 AUCUST, 1990
i The mactor coolant system model is nodalized into volumes interconnected by flowpaths.
1 The broken loop is modelled explicitly, while the two intact loops are lumped into a second loop. Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum. The multinode capability of the progmm enables explicit, detailed spatial representation of various system components; which, among other j
capabilities, enables a proper calculation of the behavior of the loop seal during a loss-of-coolaut accident. The reactor core is represented n heated control volumes with associated phase separation models to permit transient mixture height calculations.. Detailed descriptions of the NOTRUMP code and the evaluation model are provided in References 13 and 14.
Safety injection systems consist of gas pressurized accumulator tanks and pumped injection systems. Minimum emergency core cooling system availability is assumed for the analysis.
Assumed pumped safety injection characteristics as a function of RCS pressure used as boundary conditions in the analysis are shown in Figure 15.3.-l. The injection rate is based upon the pump performance curves, but degmded for conservatism and to account for possible reduced injection rates due to pump cooling recirculation miniflow operation.
Injection is delayed after the occurrence of the injection signal as indicated in Table 15.3-1 to account for diesel generator startup and emergency power bus loading in case of a loss of offsite power coincident with an accident.
Peak cladding temperature calculations are perfonned with the LOCTA-IVNeode using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture height as boundary conditions. Figure 15.3-2 depicts the hot rod axial power shape used to perform the small break analysis. This shape was chosen because it repn:sents a distribution with power concentrated in the upper regions of the core. Such a distribution is i
limiting for small-break LOCAs because it minimizes coolant level swell, while maximizing vapor superheating and fuel rod heat generation in the uncovered elevations. Figure 15.3-3 presents the normalized core power curve as a function of time after reactor trip. The scram delay times denoted in Table 15.3-1 reflect the assumption that the core is assumed to continue to operate at full rated power until the control rods are completely inserted.
15.3.1.2.2 Results This section presents results of the limiting break size analysis as detennined by the highest peak fuel rod cladding temperature for a range of break sizes. The limiting break size was found to be a 2-inch diameter cold leg break. The maximum temperature attained during the transient was 1839'F.* Important parameters are summarized in Table 15.3-2, while the key transient event times are listed in Table 15.3-1. Figures 15.3-4 through 9 show for the two-inch break transient, respectively:
RCS pressure Core mixture level
- The peak cladding temperature of 1839*F is based on the results of the analysis calculation (1823 F) perfonned for beginning of core life conditions plus a 32 F penalty to account for fuel rod burst / blockage effects which are projected for later in life and a 16 F benefit for previous SBLOCA model assessments.
15.3-3
Peak cladding temperature, Core outlet steam flow, Hot spot rod surface heat transfer coefficient, and Hot spot fluid temperature.
During the initial period of the small-break transient, the effect of the break flow is not strong enough to overcome the flow maintained by the reactor recirculation cooling pumps as they coast down. Nonnal upward flow is maintained through the core and core heat is adequately removed. At the low heat generation rates f,llowing shutdown the fuel rods continue to be well cooled as long as the core is covenxi by a two-phase mixture level.
From the cladding temperature transient for the 2-inch break calculation shown in Figure 15.3-6, it is seen that the peak cladding temperature occurs near the time at which the core is most deeply uncovered when the top of the core is steam cooled. This time is also accompar.ied by the highest vapor superheating above the mixture level.
15.3.1.2.3 Additional Break Sizes Studies documented in References 9 and 10 determined that the limiting small-break size occurred for breaks less than 10 inches in diameter. To insure that the 2-inch diameter break was limiting, calculations were run with breaks of 1.5 inches and 3 irches. The msults of these calculations are shown in the Sequences of Events Table 15.3-1, and the Results Table 15.3-2. Plots of the following parameters are shown in Figures 15.3-10 through 15 for the 1.5-inch break, and Figures 15.3-16 through 21 for the 3-inch break.
RCS pressum, Core mixture level, Peak cladding temperature, Com ortlet steam flow, Hot spot rod surface heat transfer coefficient, and Hot spot fluid temperature.
As seen in Table 15.3-2, the maximum cladding temperatures were calculated to be less than that for the 2-inch break.
15.3.1.2.4 Additional Analysis NUREG-0737t"I,Section II.K.3.31, required plant-specific small break LOCA analysis using an Evaluation Model revised per Section II.K 3.30. In accordance with NRC Generic Letter 83-85n23, generic analyses using NOTRUMPI""I were perfonned and are presented in WCAP-11145f"I. Those results demonstrate that in a comparison of cold leg, hot leg and pump section leg break locations, the cold leg break location is limiting.
Analyses of a LOCA in the pmssurizer vapor space such as that caused by opening a pressurizer relief valve or a safety valve were provided in WCAP-960001 The conclusion presented in WCAP-9600 is that these breaks are not limiting since little or no core uncovery will take place. WCAIL9600 states that the analyses reported therein apply to all Westinghouse designed plants.
15.3-4
1 Calculations were also performed for the Virgil Summer plant with the NOTRUMPl""I and LOCTA-IVU3 codes to examine the influence of initial loop fluid operating temperatures on small break LOCA peak cladding temperature. The results showed that peak cladding temperature decreased as loop operating temperature decreased.
15.3.1.2.5 Impact of ECCS Evaluation Model Changes The October 17,1988 revision to 10CFR50.46 requires applicants and holders of operating licenses or constmction permits to notify the NRC of crmrs and changes in the ECCS Evaluation models, which are not significant, on an annual basis. Reference 18 defines a significant error or changes as one which results in a calculated peak fuel cladding temperature (PCT) different by more than 50 from the temperatum calculated for the limiting transient using the last acceptable model, or is an accumulation of changes and ermrs such that the sum of the absolute temperature change is greater than 50*F. There are currently no ECCS Evaluation Model changes for Virgil C. Summer Nuclear Station identined which affect the calculated PCT.
15.3.1.3 Conclusions Analyses presented in this section show that the high head portion of the emergency com cooling system, together with accumulators, provide sufficient core Gooding to keep the calculated peak cladding temperatures below required limits of 10 CFR 50.46. Hence, adequate protection is afforded by the emergency core cooling system in the event of a small break loss of coolant accident.
3.2 MINOR SECONDARY SYSTEM PIPE BREAKS 15.3.2.1 It entification of Causes and Accident Description Included in this groupin mptures of secondary system lines which would result in steam release rates equivalent to a h diameter break or smaller.
15.3.2.2 Analysis of Effects and Const _ - nces Minor secondary system pipe breaks must be accomn ated with the failure of only a small fraction of the fuel elements in the reactor. Since the resu gf analysis presented in Section 15.4.2 for a major secondary system pipe rupture also meet thisW 'ria, separate analysis for minor secondary system pipe breaks is not required.
The analysis of the more probable accidental opening of a secondary system steam u p, relief or safety valve is presented in Section 15.2.13. These analyses are illustrative of ;
pipe break equivalent in size to a single valve opening.
15.3-5 l
TABLE 15.3-1 IIME SEOUENCE OF EVENTS FOR SMALL BREAK LOCA Small-break' Loss of Coolant Accident Event Time (s)
Break Size:
.L11nch 2-Inch 3-Inch Break -occurs 0
0 0
Reactor trip signal 84.7 43.7 18.7 Core power shutdown 90.4 49.4 24.4 Safety injection signal 103.2 58.5 27.5 Safety injection begins 130.2 85.5 54.5 Top of core uncovered 2217 1620 673 Accumulator injection begins N/A N/A 1348 Peak cladding temperature occurs 2228 2973 1410 i
15.3-16
... =.
TABLE 15.3-2 SMALL-BREAK LOSS OF COOLANT ACCIDENT CALCULATION l
RESULTS Parameter Value Break Size:
L1-I.n_st 2-Inct 3-Inch Peak Cladding Temperature 588 1839*
1817
(*F) 12.00 11.75 11.75 Elevation (ft)
Zr-H O cumulative reaction 2
Maximum Local (%)
0.02 4.12 4.01 Elevation (ft) 9.50 11.75 11.75 Total core (%)
<1
<1
<1 Rod Burst no no no SIGNIFICANT INPUT PARAMETERS Core power 2900 MWt Peak linear heat generation rate 13.68 kw/ft l
Accumulator Tank water volume 1014 ft Pressure 600 psia l
Includes 32*F penalty to account for fuel rod burst / blockage effects and a 16*F benefit for previous SBLOCA model assessments.
15.3-17 4
I
50 N
n 40 N
k
\\ *\\
Eo O 30 N
2 cE
~
g 20 E
E
~
O 2 10 o
O 500 1000 1500 2000 2500 Pressure (psig)-
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Pumped Safety injection Flow Rate Figure 15.3-1
2.5
\\
2 m
E
[
o 1.5 bao O_
s 1
~e a:
0.5 7
0 0
2 4
6 8
10 12 Core Elevation (ft)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Peak Rod Axial Power Shape Figure 15.3-2 I
8 10 8 =
~~
4 TOTAL RE11 DUAL HEAT (WITH 4% SHUTDOWN) 2
{io-1 =-
2 C
s l
4 5
2
- 2 0 to-2 E
a =
b 4
2 l ! l illf tl 1 1 1 11111l 1 i l l!!ll!
I l 11111ll' 1 I I l ? ill 3,.2 2
3 4
8 2 4 s a to' 2 4 8 810 2
4 s a 10 2
4 s a 10 10-1 2
4 s s 10 TIME AFTER SHUTDOWN (SECOND$1 AMENDMENT 6 AUGUST,1990 SOUTH CAROLINA ELECTRIC & G AS CO.
VIRGIL C. SUMMER NUCLEAR STATION Nortnalized Core Heat Generation Rate Following Shutdown (full rod insertion)
Figure 15.3 3
2500 2000
~
n en CL v
1500 wm a
en en
~
w m
u-1000 N
N N~
i iiii i i i i iiii 500 0
1000 2000 3000 4000 5000 6000 TIME (SEC) l i
I SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Reactor Coolant System Pressure j
(2-inch break)
Figure 15.3-4
35 9
30 m
n Lt.
v
~
_J w
l
>w 25 F
_2 w
a::
a TOP OF CORE
[
- E 20 M
15 O
1000 2000 3000 4000 5000 6000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS C6-
)
VIRGIL C. SUMMER NUCLEAR STATION.
Core Mixture Height (2-inch break) l Figure 15.3-5 I
l 2000 1800
^
1600 n
u-1400 l
w m
U 1200 g
e w
/
1000 s
w
~
800 600 3
1
\\
400 1000 2000 3000 4000 5000 6000 TIME (S)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION.
Clad Temperature Transient at Peak Temperature Elevation (2-inch break)
Figure 15.3-6
L 200 m
(n 150 N
E cc v
w l-
< 100 f
ce se o
La.
(n en 50 0
O 1000 2000 3000 4000 5000 6000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Steam Mass Flow Rate Out Top of Core (2-inch break)
Figure 15.3-7
5 10 n
u_
I N
W 4 u 10 I
c:
I N
m
~
w 03 3 v10 tu u_
wo
~
O 2 10 w
N A
Z 1
H 10 H
w I
j[
t i i i e i e i e i i i e i e i i i e i 1000 2000 3000 4000 5000 6000 TIME (S)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Clad Surface Heat Transfer Coefficient at Peak Temperature Elevation (2-inch break)
Figure 15.3-8
1800 1600 1400
^
u_
v 1200 w
x D
N 1000 w
r o_
1 w
w
~
800 600 j
400 1000 2000 3000 4000 5000 6000 TIME (S)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Fluid Temperature at Peak Clad Temperature Elevation (2-inch break)
Figure 15.3-9
2400 2200 2000 m
m CL 1800 v
w D
1600 m
cn
)
W l
1400 1200 i
N 1000 0
5000 10000 15000 20000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Reactor Coolant System Pressure.
(1.5-inch break)
Figure 15.3-10
34 r
32
^
30 i--
LL.
ss J 28 w>
~
w J
g w 26 -
-W l
oc p
f\\
/
- 24
[
22 c
5000 10000 15000 20000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Core Mixture Height (1.5-inch break)
Figure 15.3-11 4
0
~.
585 580
[
t 1
u_
575 v
w m
[ 570
<c m.....
m Illy
[ [
i w
CL E
W 565 F-560
\\
i 555 2000 4000 6000 8000 10000 12000 14000 16000 TIME (S)'
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Clad Temperature Transient at Peak Temperature Elevation (1.5-inch break)
Figure 15.312
-~
i i
200 l
l m
u) 150 N
=E CD a
w k
- < 100 o
!l
_J LL.
%g 50-j 0
0 5000 10000 15000 20000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Steam Mass Flow Rate Out Top of Core (1.5-inch break)
Figure 15.3-13
vCwaNxEl u_ W N l.u n Iw<H H%<zn_we Oowuu c
eu D
j 1
1 1
1 1
2f 0
0 0
0 0
2 3
4 5
6 0
0
~ - _ : :
~
~ __::_
0 4
0 0
0 i
6 i
0 0
0 i
T8 i
I 0 0
M 0
V E
i SO I
C RU al GT i
1 ad I
H (0
L 0
t Pu C C S
S 0
A
)0 i
er f aa SR kc UO i
(. Te ML F
1 I
1 i
gi 5eH MN 2
u i me EA 0
re np a cet R
0 E
0 i
1 hr T NL 5
a r
bt a UE i
3 r u er n C C a es L T 1
1 f
4 4
k Ee E RI 0
)
l r
eC AC 0
v R
ao 0
i ie S
t of TG f
ni ei AA c
1 TS n
I 6
OC 0
t N.O 0
0
575 570
~
C I;
y l11tll l
l
~%
v W
m
[ 565 4
eW CL
- E w
560
\\
555 2000 4000 6000 8000 10000 12000 14000 16000 TIME (S)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION-Fluid Temperature at Peak Clad Temperature Elevation (1.5-inch break)
Figure 15.3-15
.. ~.
2500 2000
~
1500 n.
v w
m o
M 1000 g
g w
E o_
500 n
c- -
0 1000 2000 3000 4000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Reactor Coolant System Pressure (3-inch break)
Figure 15.3-16
35 1
30 m
H La_
v l
J w
>w 25
_a w
m x
s
_ TOP _OF CORE _
=s 20
\\
15 O
1000 2000 3000 4000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION
.i Core Mixture Height (3-inch break)
Figure 15,3-17 i
r
1 1
2000
^
1800 s
1600 h--
m u.
1400 w
x
[
1200 p
x w
1000 w
w 800 600 400 500 1000 1500 2000 2500 3000 3500 4000 TIME (S)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Clad Temperature Transient at Peak Temperature Elevation (3-inch break)
Figure 15.3-18
350 300 n
cn N 250 2
cn
_.2 v
200 bJ W
E
' l 3c:
150
_)
LL.
100
~
k 2
50--
0, 0
1000 2000 3000 4000 TIME (SEC)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Steam Mass Flow Rate Out Top of Core (3-inch break)
Figure 15.3-19 l
l
5 10 m
u.
l
^
N H 4 u_10 Ier No H
~
CD 3 v10 u_
u_
uJo
~
1 0 2
.)
10 i
e V
W z
1 H 10 H
ua I
jf iiie ie i i iiie iie i iiia i e i e i i i i 500 1000 1500 2000 2500 3000 3500 4000 TIME (S)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Ciad Surface Heat Transfer Coefficient at Peak Temperature Elevation (3-inch break)
Figure 15.3-20
- s..
V 1800 1600 1400 m
u_
v 1200 wx o
W
-I m
1000 W
p
=E w
w 800 j
600 400 500 1000 1500 2000 2500 3000 3500 4000 TIME (S)
SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Fluid Temperature at Peak Clad Temperature Elevation (3-inch break)
Figure 15.3-21
f R
P N
N L
K J
H G
F.
E D
C.
8 A
-19*
I
-20%
2-
-81
- 15'.
-87%
-21%
3
-81
-16'
- llt 4
+0i
-55
-lif.
-20 %
-215
-221 5-
- 1,
- 11 '
-is%
6 s'
-s f.
-115
-135
-2i f 7
16:
151 98
-125
-19' - 20 1 3
175
- 55,
-95
-20; 9
1 351 104
-lai
-174 10
- 1%
6%
-51
-si 11-711 151
-9%'
~ 12'
.[2'
\\
N
/
g sF
-11 13 t
225 45 lq 10!
15 CASE A AMENDMENT 6 AUGUST,1990 SOUTH CAROLINA ELECTRIC & GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION Inadvertent Fuel Misloading interchange of Region 1 and Region 3 Assembly -
Figure 15.3 22
REVISED RADIOLOGICAL FSAR CIIAPTER 15 WRITE-UPS l
i l
the moderator temperature coefficient and the Doppler power coefficient. These reactivity coefficients and their values are discussed in detail in FSAR Chapter 4 In the analysis of certain events, conservatism requires the use of large reactivity coefficient values whereas in the analysis of other events, conservatism requires the use of small reactivity coefficient values. Some analyses such as loss of reactor coolant from cracks or ruptures in the reactor coolant system do not depend on reactivity feedback effects. He values used are given in Table 15.1-4 Reference is made in that table to Figure 15.1-5 which shows the upper and lower bound Doppler power coefficients as a function of power, used in the transient analysis. The justification for use of conservatively large versus small reactivity coefficient values are treated on an event by event basis.
15.1.7 FISSION PRODUCT INVENTORIES 15.1.7.1 Radioactivity in the Core The fission product inventories for all isotopes which are important from a health hazards standpoint were calculated using the ORIGEN code" and are given in Table 15.1-5. This code uses a data base of fission product yields, cross sections, and decay constants taken fron. the ENDFB-IWV fission product library. The calculation of the core iodine fission product inventory is consistent with TID-14844". The ORIGEN code takes into account fuel burnup '
cIl as fission product buildup ano decay. Continuous operation at full power is assumed during the fuel residence time to provide an upper limit estimate of the fission product inventory. The isotopes included in Table 15.1-5 are the isotopes controlling from considerations of inhalation dose (iodines) and from direct dose due to immersion (noble gases).
The isotopic yields used in the calculations r.re from the data of APED-5398, utilizing the isotopic yield data for thermal fissioning of U-235 as the sole fissioning source. He change in fission product inventory resulting from the fissioning of other fissionable atoms has been reviewed. He results of this review indicated that inclusion of all fission source data would result in small (less than 10 percent) change in the isotopic inventories due to the overall conservatism.
1
.7.2 Radioactivity in the Fuel Pellet Clad Gan The comput ap activities (Table 15.1-5) are based on buildup in the fuel from the fission pro:ess and diffusion to t.
. at rates dependent on the operating temperature. He temperature dependence is accounted for by dete ing the core fuel fraction operating within each of 10 temperature regions (Table 15.1-6), each with a re e rate to the gap dependent on the mean fuel temperature witnin that region. Since the temperature distri
'on changes during cote life, the highest expected values are used.
1 The temperature dependence of the diffusion coefficie
- for Xe and Kr in UO, follows the 2
Arrhenius law:
~
1 D'( T) =D'(167 3 ) exp -S 1 R\\ T 1673
\\
\\
Where:
4 Diffusion coefficient at temperature T. sec D' (T)
=
diffusion coefficient at 1673 *K = 1 x 10-11 sec-
'(1673)
=
temperature. 'K T
=
3 gas constant.1.99 x 10 kilocalories/ mole 'K R
=
The above exp ssion is valid for temperatures above 1100'C. Below 1100 *C fission gas release occurs mainly by wo temperature independent mechanisms, recoil and knock out, and is predicted by using D' at 1100
- The value used for D'(1673), based on data at bumups greater than 10*
fissions /cc. accounts possible fission gas release by other mechanisms as well as pellet cracking during irradiation.
He diffusion coefficient for ine isotopes was conservatively assumed to be the same as for Xe and Kr. Toner and Scott"' observed iodine diffuses in UO at about the same rate as Xe and Kr and 2
has about the same activation ene
. Data reponed by Belle'5>lndicates that the iodine diffuses at slightly slower rates than Xe and '
With the diffusion coefficient detennine or the fuel temperature region of interest the fraction of radioactive fission gas which crosses the fu boundary into the fuel rod gap is found from; f=3 co h g
Where:
fraction of a given radioactive fission gas in fu rod gap f
=
fission gas decay constant, sec'8 A
=
4 diffusion coefficient, sec D*
=
The above expression is the steady state solution of the diffusion equation in herical geometry as given by Booth *,
Table 15.1-5 lists the total core activities as well as the activities present in the gap r each peninent isotope obtained using the above equations and the fuel temperature distribution given ' Table 15.1-6.
This fuel temperature distnbution is conservatively bounding for VANTAGE + fuel.
l The activities in the mactor coolant as well as in the volume control tank, pressurizer and w gas decay tanks are given in FSAR Chapter 11 including the data on which the computation of the activities are based.
15.1.8 RESIDUAL DECAY IIEAT Residual heat in a subcritical core consists of:
1.
Fission pmduct decay energy, 2.
Decay of neutron capture products, and
TABLE 15.1-5 CORE AND CAP ACTIVITIES (IODINE AND NOBLE CASES)
BASED ON FULL POWER OPERATION FOR 444-DAYS Curies in Core rcent of Curies Cap Isotope (x 107)
ActiqtyinGap (x
5)
I-131 1 T ?.t.
1.
0.92 I-132 M 6 2..o 0.1 1.68 I-133 Frr8 L'.. ?
0.46 7.268 I-134 4-he LE.o 0.095 1.615 I-135 M + is 4 0.26 3.796 Xe-131m W 0 o86 1.7 0.0918 Xe-133 4he i7 o 1.1 16.5 Xe-133m 2-6 2 9 0.74 1.7 Xe-135 M37 0.31 1.02 Xe-135m M39 0.52 1.61 Xe-138 L3,0 it.o 0.54 7'.02 Kr-83m OW o.96 0.14 0.127 Kr-85 Gre%- c.o 83 21.2 1.36 s.42 Kr-85m 2T 2.1
- 0. 1 o
Kr-87 4+ 3 8 0.11 0.407 Kr-88 5-d S.y
.17
- 0. 01 Kr-89
-frd-4. C 0.023 0.1 mot E. - TN pcdy of %c;N;ti4y in Ag d k k EfNt4 %
of M Nil J rt.
- 1. e n 4% gcc t hk.
hg A t-tiqik7
-fo r e_d g et (M i s.
t a cois.1
',w h q p\\i c Ab.0 a se c46 fe< %b. %cti M.
AMENDMENT 6 15.1-23 AUGUST, 1990
APPENDIX 7 SI PUMP TEST REPORT
>