ML20003C277
| ML20003C277 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 02/20/1981 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML20003C276 | List: |
| References | |
| NUDOCS 8102270522 | |
| Download: ML20003C277 (107) | |
Text
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AMENDMENT NUMBER 1 l'
CESSAR-F i
DOCKET STN-50-470F i
The following sheets of CESSAR-F are to be removed and inserted:
i REMOVE THE FOLLOWING INSERT THE FOLLOWING i
)
Chapter 3 i
i-Effective Page Listing I
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V via j
3.5-1 3.5-1 3.5-la 3.7-1 3.7-1 Fig. 3.7.1-1 Fig. 3.7.1-1 Fig. 3.7._1-2 Fig. 3.7.1-2 Fig. 3.7.1-3 Fig. 3.7.1-3 l
Fig. 3.7.1-4 Fig. 3.7.1-4
.3.7-3 3.7 ;
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3.7-9 3.7-9
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3.7-10' 3.7-10
. Table 3.7.2-1 Table 3.7.2-1 3.7-11 3.7 i i~
Chapter 4 i
Table 4.4-4 Table 4.4-4 Appendix 4B~
Effective Page Listing i
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4B-3 4B-3 48-5 48-5 Table 4B-1 i
Table 43-6 Table 4B-7 Chapter 6
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I REMOVE THE FOLLOWING INSERT THE FOLLOWING Chapter 8 Effective Page Listing 8.1-3 8.1-3 8.3-3 8.3-3 Table 8.3.1-1 (Sheet 3)
Table 8.3.1-1 (Sheet 3)
Chapter 11 Effective Page Listing 1
1 11.1-13 11.1-13 11.1-13b 11.1-13d Chapter 12 Effective r age Listing i
i 12.1-1 12.1-1 12.1-3 12.2-1 12.2-1 12.2-2 12.2-4 12.2-4 Table 12.2-11 12.3-1 12.3-1 12.3-3 12.3-3 12.3-5 Appendix 15D Effective Page Listing ii ii iia 15D.6-1 Appendix A Effective Page Listing A-34 A-34 A-34a O
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l EFFECTIVE PAGE LISTING I
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Chapter 3 l
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Table of Contents f
Page Amendment l
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iv to via 1
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vii to xvii Text Page Amendment l
3.1-1 to Table 3.2-2 I
- 3. 5-1 1
3.5-la 1
l 3.5-2 to Figure 3.6-1 3.7-1 to Figure 3.7.1-4 1
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- 3. 7-2
- 3. 7-3 1
- 3. 7-4 to 3.7-8 3.7-9 to Table 3.7.2-1 1
1 Table 3.7.2-2 to Figure 3.7.2-3 l
l 3.7-11 to 3.7-11a 1
3.7-12 to 3.11-7 I
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TABLE OF CONTENTS (Cont'd.)
fsg CHAPTER 3 Section Subject Page No.
3.1.35*
CRITERION 39 - INSPECTION OF CONTAINMENT 3.1-22 HEAT REMOVAL SYSTEM 3.1.36*
CRITERION 40 - TESTING 0F CONTAINMENT HEAT 3.1-22 REMOVAL SYSTEMS 3.1.37*
CRITERION 41 - CONTIANMENT ATMOSPHERE CLEANUP 3.1-23 3.1.38*
CRITERION 42 - INSPECTION OF CONTAINMENT 3.1-23 ATMOSPHERE CLEANUP SYSTEMS 3.1.39*
CRITERION 43 - TESTING 0F CONTAINMENT 3.1-23 ATMOSPHERE CLEANUP 3.1.40*
CRITERION 44 - COOLING WATER 3.1-24 3.1.41*
CRITERION 45 - INSPECTION OF COOLING WATER 3.1-24 SYSTEM f'~N 3.1.42*
CRITERION 46 - TESTING OF COOLING WATER SYSTEM 3.1-24 i
3.1.43*
CRITERION 50 - CONTAINMENT DESIGN BASIS 3.1-24 3.1.44*
CRITERION 51 - FRACTURE PREVENTION OF CONTAINMENT 3.1-25 PRESSURE B0UNDARY 3.1.45*
CRITERION 52 - CAPABILITY FOR CONTAINMENT 3.1-25 LEAKAGE RATE TESTING 3.1.46*
CRITERION 53 - PROVISIONS FOR CONTAINMENT 3.1-25 TESTING AND INSPECTION 3.1.47 CRITERION 54 - PIPING SYSTEMS PENETRATING 3.1-25 l
CONTAINMENT i
l 3.1.48 CRITERION 55 - REACTOR COOLANT PRESSURE BOUNDARY 3.1-26 l
PENETRATING CONTAINMENT 3.1.49 CRITERION 56 - PRIMARY CONTAINMENT ISOLATION 3.1-27
-3.1.50 CRITERION 57 - CLOSED SYSTEM ISOLATION VALVES 3.1-27
.h
- See Applicant's SAR
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TABLE OF CONTENTS (Cont'd.)
CHAPTER 3 Section Subject Page No.
3.1.51*
CRITFRION 60 - CONTROL OF RELEASES OF RADI0 ACTIVE 3.1-28 MATERIAL TO THE ENVIRONMENT 3.1.52*
CRITERION 61 - FUEL STORAGE AND HANDLING AND 3.1-28 RADI0 ACTIVITY CONTROL 3.1.53*
CRITERION 62 - PREVENTION OF CRITICALITY IN 3.1-28 FUEL STORAGE AND HANDLING 3.1.54*
CRITERION 63 - MONITORING FUEL AND WASTE STORAGE 3.1-29 3.1.55*
CRITERION 64 - MONITORING RADI0 ACTIVITY RELEASES 3.1-29 3.2 CLASSIFICATION OF STRUCTURES, C0fiPONENTS, AND SYSTEMS 3.2-1 3.2.1 SEISMIC CLASSIFICATION 3.2-1 3.2.2 SYSTEM QUALITY GROUP CLASSIFICATIONS (SAFETY CLASS) 3.2-2 3.3*
WIND AND TORNADO LOADINGS 3.521 3.4*
WATER LEVEL (FLOOD) DESIGN 3.5-1 3.5 MISSILE PROTECTION 3.5-1 3.5.1 MISSILE SELECTION AND DESCRIPTION 3.5-1 3.5.1.l*
Internally Generated Missiles (Outside 3.5-1 Containment) 3.5.1.2 Internally Generated Missiles (Inside 3.5-1 Containment) 3.5.1.3*
Turbine Missiles 3.5-la 1
3.5.1.4*
Missiles Generated by Natural Phemonema 3.5-2 3.5.1.5*
Missles Generated by Events Near the Site 3.5-2 3.5.1.6*
Aircraft Hazard 3.5-2 3.5.2*
SYSTEMS TO BE PROTECTED 3.5-2
- See Applicant's SAR Amendment No. 1 iv February 20, 1981 l
TABLE OF CONTENTS (Cont'd.)
C CHAPTER 3 Section Subject Page No.
3.5.3*
BARRIER DESIGN PROCEDURES 3.5-2 3.5.3.1 Missile Barrier Design Interface Requirements 3.5-2 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED 3.6-1 WITH THE POSTULATED RUPTURE OF PIPING 3.6.1 POSTULATED PIPING FAILURES IN FLUID SYSTEMS 3.6-1 OUTSIDE OF CONTAINMENT 3.6.1.1 Design Bases 3.6-1 3.6.1.1.1 High Energy Piping Systems 3.6-1 3.6.1.1.2 Moderate Energy Piping Systems 3.6-2 3.6.1.2 Description 3.6-2 3.6.2 DETERMINATION OF BREAK LOCATIONS AND DYNAMIC 3.6-2 EFFECTS ASSOCIATED WITH THE POSTULATED RUFTURE
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0F PIPING i
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3.6.2.1 Criteria Used to Define Break and Crack 3.6-2 Location and Configuration 3.6.2.2 Analytical Methods to Define Forcing 3.6-5 Functions and Response Models 3.6.2.3 Dynamic Analysis Methods to Verify Integrity 3.6-5 and Operability 3.6.2.4 Guard Pipe Assembly Design Criteria 3.6-5 3.6.2.5 Material Submitted for the Operating 3.6-5 License Review 3.7 SEISMIC DESIGN 3.7-1 i
3.7.1 SEISMIC INPUT 3.7-1 3.7.1.1 Design Response Spectra 3.7-1 l
3.7.1.2*
Design Time History 3.7-1 3.7.1.3 Critical Damping Values 3.7-1 i
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)
3.7.1.4 Supporting Media for Category I Structures 3.7-1 1
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- See Applicant's SAR Amendment No. 1 v
February 20, 1981
TABLE 0; CONTENTS (Cont'd.
CHAPTER 3 Section Subject Page No.
3.7.2 SEISMIC SYSTEM ANALYSIS 3.7-2 3.7.2.1 Dactor Cool at System 3.7-2 3.7.2.1.1 Introduction 3.7-2 3.7.2.1.2 Mathematical Models 3.7-3 3.7.2.1.3 Calculations 3.7-4 3.7.2.1.4 Results 3.7-9 3.7.2.1.5 Conclusion 3.7-9 3.7.2.2 Natural Frequencies and Response Loads 3.7-9 3.7.2.3 Procedure Used For Modeling 3.7-9 3.7.2.4*
Soil / Structure Interaction 3.7-9 3.7.2.5*
Development of Floor Response Spectra 3.7-9 3.7.2.6 Three Components of Earthquake Motion 3.7-9a 3.7.2.7 Procedure for Combining Modal Responses 3.7-9a 3.7.2.8*
Interaction of Non-Category I Structures 3.7-10 with Seismic Category I Structures 1
3.7.2.9*
Effects of Parameter Variations on Floor 3.7-10 Response Spectra 3.7.2.10 Use of Constant Vertical Static Factors 3.7-10 3.7.2.11 Torsional Effects of Eccentric Masses 3.7-10 3.7.2.12*
Comparison of Responses 3.7-10 3.7.2.13 Methods for Seismic Analysis of Dams 3.7-16 3.7.2.14*
Determination of Seismic Category I 3.7-10 Structure Overturning Moments 3.7.2.15 Analysis Procedure for Dampinq 3.7-10 3.7.3 SEISMIC SUBSYSTEM ANALYSIS 3.7-11
'.>. 7. 3.1 Seismic Analysis Methods 3.7-11
' 7.3.2 Determination of Number of Earthquake Cycles 3.7-11
- See Applicant's SAR Amendment No. I vi February 20, 1981
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L TABLE OF CONTENTS (C.at'd.)
CHAPTER 3 i
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Section Subject Page No.
3.7.3.3 Procedure Used For Modeling 3.7-11
(
f 3.7.3.4 Basis for Selection of Forcing Frequeacies 3.7-11 i
3.7.3.5 Use of Equivalent Static Load Method of 3.7-11a 1
f Analysis j
l 3.7.3.6 Three Components of Earthquaka Motion 3.7-12 I
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Amendment No. 1 via February 20, 1981
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m 3.3 WIND AND TORNADO LOADINGS
" Interface requirements for protection of systems within CESSAR scope from wind and tornado loadings are included in the sections describing those systems.
In summary, those interface requirements specify that the seismic 1
category I components listed in Table 3.2-1 are to be protected.
See Applicant's SAR for infomation on equipment protection."
3.4 WATER LEVEL (FLOOD) DESIGN
" Interface requirements for protection of systems within CESSAR scope from floods are included in the sections describing those systems.
In summary, those interface requirements specify that the Seismic Categorf I components I
listed in Table 3.2-1 are to be protected.
See Applicant's SAR for information on equipment protection.
3.5 MISSILE PROTECTION 3.5.1 MISSILE SELECTION AND DESCRIPTION 3.5.1.1 Internally Generated Missiles (Octside Containment)
" Interface requirements for protection of systems within CESSAR scope from missiles are included in the sections describing those systems.
In summary,
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those interface requirements specify that the Seismic Category I components 3
listed in Table 3.2-1 are to be protected.
See Applicant's SAR for information on equipment protection."
3.5.1.2 Internally Generated Missiles (Inside Containment)
Plant structures, systems and components inside containment within the CESSAR scope whose failure could lead to offsite radiological consequences or which are required for safe plant shutdown are included in Table 1.2-1.
The criteria for protection of the reactor coolant system from postulated missiles are: prevention against the occurrance of a loss of coolant accident due to missile impact, the maintenance of a safe shutdown capability.
Protection criteria for the Containment _and other plant systems will be addressed in particular Applicants SAR. The natural separation in the C-E plant arrangement and inherent structural barriers, such as the primary shield walls, are sufficient to preclude the need for significant additional structures for missile protection.
In general, items inside the Containment are protected from missiles which orginate outside Containment, by the Containment itself, or by measures which protect the Containment.
Structural design criteria, identification of missile barriers provide (if any) and design leading will be addressed in the Applicant's SAR.
'd Amendment No. 1
- 3. 5-1 February 20, 1981
I I
i The selection of potential missiles from equipment in the CESSAR scope is based on the application of a single-failure criteria to the normal retantion l
features of plant equipment for which there is a source of energy capable of creating a missile in the event of the postulated renoval of the normal retention features. Where redundancy is provided by the normal retention l
features such that sufficient retention capability remains to prevent creation of a missile in the event of a postulated failure of a single retention feature, no potential missile is postulated.
Table 3.5-1 presents tne potential missiles postulated to orgir. ate from C-E equipment, and sumnarizes their characteristics.
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3.5.1.3 Turbine Missiles f
See Applicant's SAR.
See Sections 3.5.3.1 and 5.1.4 for CESSAR Interface i
i Criteria.
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3.5.1.4 Missiles Generated by Natural Phenomena See Applicant's SAR.
See Section 3.5.3.1 and 5.1.4 for CESSAR Interface Criteria.
- 3. 5.1. 5 Missiles Generated by Events Near the Sita See Applicant's SAR.
See Sections 3.5.3.1 and 5.1.4 for CESSAR Interface Criteria.
3.5.1.6 Aircraft Hazard See Applicant's SAR.
See Sections 3.5.3.1 and 5.1.4 for CESSAR Interface Criteria.
3.5.2 SYSTE?iS TO BE PROTECTED Protection provided by means of arrangement or separation will be addressed in the Applicant's SAR.
3.5.3 BARRIER DESIGN PROCEDURES See Applicant's SAR.
3.5.3.1 Missile Barrier Design Interface Requirements 1.
For systems and parts of systems located inside containment (Reactor Coolant System and connecting systems, Engineered Safety Features Systems), appropriate design procedures shall be used to insure that the impact of any potential missile will not lead to a loss of coolant accident or preclude systems from carrying out their specified safety functions.
2.
For systems and equipment outside containment, listed in Table 1.2-1, appropriate design procedures (for example, proper turbine orientation, natural separation, or missile barriers) shall be used to insure that the impact of any potential missile does not prevent the system or equipment from carrying out its sp(cified safety functions.
3.
For all systems and equipment, appropriate design procedures shall be I
used to,nsure that the impact of any potentail missile does not prevent the conduct of a safe plant shutdown, or prevent the plant from remaining in a safe shutdown condition.
O 3.5-2
-=
3.7 SEISMIC DESIGN p
3.7.1 SEISMIC If4PUT 3.7.1.1.
Design Response Spectra This section discusses the seismic design of those systems and sob-systems important to safety and classified as Category I in Section 3.2.
j The System 80 Standard Design as defined by CESSAR is not based on a specific site, therefore seismic and geologic infomation cannot be provided.
Seismic response spectra which envelope actual design requiremer,ts for current System 80 plants are provided in Figures 7.3.1-1 through 3.7.1-4.
l These spectra reflects responses for several different building types located throughout the continental United States.
The response spectra shown in Figures 3.7.1-1 thru 3.7.1-3 are applicable to the upper most horizontal support on the component named.
Figure 3.7.1-4 is applicable to the vertical supports of all of the RCS najor components.
The effect of differential seismic displacement on the equipment and supports is included in the site specific analysis of the Reactor Coolant System described in Section 3.7.2.1.
CE provides the following to assist the Ap,.licant in his design of support structures:
a)
A simplified mathematical model which accounts for the mass and stiffness properties of the System 80 system, suitable for coupling with the mathematical model of the supporting structures and foundations.
b)
A set of design basis seismic loads transmitted from the Systen 80 systems to the supporting structures at all support locations.
c)
The set of design basis floor response spectra at each support location, upon which the design basis loads are based.
The final verification of the design basis seismic loads is perfomed as described in Section 3.7.2 and 3.7.3, based on the site specific seismic excitations provided by the Applicant.
3.7.1.2 Design Time History See Applicant's SAR for site specific information.
3.7.1.3 Critical Daraping Values Critical damping values used for Category I Systems and Components are given in Table 3.7.2-1.
3.7.1.4 Supporting Media for Category I Structures See Applicant's SAR.
Amendment tio. 1 3.7-1 February 20, 1981
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RESPONSE SPECTRUM FOR UPPER REACTOR VESSEL SUPPORTS 3.7.1-1
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C-E RESPONSE SPECTRUM FOR UPPER STEAM 8
GENERATOR SUPPORTS 3.7.1-2
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February 20, 1981 c-E RESPONSE SPECTRUM FOR VERTICAL SUPPORTS Figm FOR ALL COMPONENTS 3.7.1-4
pb by the sum of the absolute values method and, in turn, combined with the responses of the remaining significant modes by the square root of the sum i
of the square method.
i Contribution >.~ rom all significant modes of response are retained in the analyses.
l The damping factors used in seismic analysis of Category I structures, l
systems and equipment are selected from Table 3.7.2-1.
Modal damping i
i factors of 2 to 3 percent of critical and 1 to 2 percent of critical for the SSE and OBE, respectively, are used in the seismic analysis of the i
I coupled components of the reactor coolant system. Modal damping factors of 1 to 2 percent of critical and 1/2 to 1 percent of critical for the SSE and OBE, respectively, are used in separate seismic analyses of branch runs of piping for which C-E has responsibility for supply, such as the surge line piping. The damping factors given in Table 3.7.2-1 include and are in 4
i agreement with those recommended in Regulatory Guide 1.61.
l The dynamic analyses of the major components of the reactor coolant system to cor. firm the seismic adequacy of the design are scheduled for completion such that a report of the results will be included in the Applicant's FSAR.
)
3.7.2.1.2 Mathematical Models In the descriptions of the typical mathematical models which follow, the i
spatial orientations are defined by the set of crthogonal axes where Y is t
in the vertical direction ano X and Z are in the horizontal plane, in the directions indicated on the apprcpriate figure. The mathematical representa-tion of the section propertias of the structural elements employs-a 12 x 12 stiffness matrix for the three-dimensional space frame models, and ernploys j
i a 6 x 6 stiffness matrix for the two-dimensional plane frame model.
Elbows in piping runs include the in-plane /out-of-plane bending flexibility factors as specified in the ASf1E Code,Section III.
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The methods used to combine the responses due to the different components
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of Earthquake Motion are described in Section 3.7.2.1.4 for the dynamic i
seismic system analysis, and in Section 3.7.3.6 for the seismic subsystem analysis.
Reactor Coolant System - Coupled Components A schematic diagram of the typical composite mathematical model used in the 4
analyses of the dynamically coupled components of the reactor coolant system is presented in Figure 3.7.2-1.
This model includes 30 mass points with a total of 83 dynamic degrees of freedom. The mass points and corres-ponding dynamic degrees of freedom are distributed to provide appropriate representations of the dynamic characteristics of the components, as follows:
the reactor vessel, with internals, is represented by 4 mass points with a total of 11 dynamic degrees of freedom; each of the two steam generators is represented by 4 mass points with a total of 10 dynamic degrees of. freedom, each of the four reactor coolant pumps is represented by 2 mass points with
- O a total of 6 dynamic degrees of freedom; and each branch of piping is represented by a mass point with 3 dynamic' degrees of freedom. The represen-tation of the reactor vessel internals is formulated in conjunction with Amendment No. 1 3.7-3 February 20, 1981
the analysis of the reactor vessel internals discussed in Section 3.7.3.14, and is designed to simulate the dynamic characteristics of the modelt used in that analysis.
The mathematical model provides a three-dimensional representation of the dynamic response of the coupled components to seismic excitations in both the horizontal and vertical directions.
The mass is distributed at the selected mass points and corresponding translational degrees of freedom are retained to include rotary inertial effects of the components.
The toatl mass of the entire coupled system is dynamically active in each of the three coordinate directions.
Pressurizer The mathematical model employed in the analysis of the pressurizer is shown schematically in Figure 3.7.2-2 This lumped parameter, 3-dimensional model provides a muli.imass representation of the pressurizer which includes 6 mass points with a total of 13 dynamic degrees of freedom.
Surge Line The lumped parameter, multimass mathemat.ical model employed in the analysis of a typical surge line is shown schematically in Figure 3.7.2-3.
The surge line is modeled as a three dimensional piping run with end points anchored at the attachments to the pressurizer and the reactor vessel outlet piping.
In the definition of this particular mathematical model, 10 mass points with a total of 27 dynamic degrees of freedom were selected to provide a three-dimensional representaion of the dynamic response of the surge line.
All supports defined for the surge line assembly are included in the mathematical model.
The total mass of the surge line is dynamically active in each of the three coordinate directions.
3.7.2.1.3 Calculations General The general matrix form of the undamped coupled equations of motion can be written (Reference 8) as follow:
MX + KU = F (1)
Where X represents the absolute acceleration of the mass point dynamic degrees of freedom, and U represents the displacements of the mass and support point dynamic degrees of freedom relative to a datum support which is chosen to eliminate free body motion.
Expanding Equation (1) gives:
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K U" '
m m;
mm ms (2)
=
0Mi X
K K
F V
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O 3.7-4
3.7.2.1.4 Results 3
The reaction (forces and moments) at all design points in the system, obtained from the dynamic seismic analysis, are compared with seismic loads in each component design specification. The results of this comparison are summarized in tabular form for the points of maximum calculated load.
The maximum response due to each of the three components of the earthquake motion are calculated separately on a time history basis and combined by the SRSS method.
When the three components of earthquake motion are statistically independent, the maximum responses are calculated by a simultaneous application of motion resulting from all three components of earthquake.
In either case the maximum seismic loads calculated by the time history techniques are the result of a search and comparison over the entire time domain of each individual component of load. The maximum calculated components of load for each design location do not in general occur at the same time and therefore use results in a conservative worst case.
The maximum seismic loads calculated by the response spectrum techniques are the result of combining the mod::1 reactions due to both horizonatal and vertical excitations. The method of modal combination is discussed in Section 3.7.2.5.
The maximum responses due to each of the three earthquake components are then combined by the SRSS method.
' n) 3.7.2.1.5 Conclusion
(
v It is concluded that the seismic loadings specified for the design of the reactor coolant system components and supports are adequate, when all seismic loads calculated by the dynamic seismic analysis are less tMn the corresponding loads in the component design specification.
3.7.2.2 Natural Frequencies and Response Loads l
This section is provided in the Applicant's SAR.
3.7.2.3 Procedure used for Modeling This procedure used for modeling NSSS comronents and interconnecting piping are described in Section 3.7.2.1.2.
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3.7.2.4 Soil / Structure Interaction i
See Applicant's SAR.
1 3.7.2.5 Development of Floor Response Spectra See Applicant's SAR.
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1{D Amendment No. 1 l
3.7-9 Februtry 20, 1981
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3.7.2.6 Three Components of Earthquake Motion 1
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l The procedures for considering the effects of three components of earthquake motion in determining the seismic response of NSSS systems, components and supports are discussed in Section 3.7.2.1.4.
3.7.2.7 Procedure for Combining Modal Responses 1
r The square root of the sum of the squares method is the procedure normally used to combine the modal responses when the modal analysis response spectum method of analysis is employed. The procedure is modified only in two cases:
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In the analysis of simple system where three or less dynamic degrees-of-freedom are involved, the modal responses are combined by the summation of the absolute values method; b.
In the analysis of complex systems where closely spaced modal frequencies are encountered, the responses of the closely spaced modes are combined by the sumation of the absolute values method and, in turn, combined with the responses of the remaining significant modes by the square root of the sum of the squares method. Modal frequencies are considered closely spaced when their difference is less than +10 percent of the lower frequency.
3.7.2.8 Interaction of Non-Category I Structures with Seismic Category I Structure See Applicant's SAR.
1 3.7.2.9 Effects of Parameter Variations in Floor Response Spectra See Applicant's SAR.
3.7.2.10 Use of Constant Vertical Static Factors A constant seismic vertical load factor is not used for the seismic design of Seismic Category I structures, components and equipment.
1 3.7.2.11 Torsional Effects of Eccentric Masses The mathe.matical
'als used in seismic anclysis of Category I systems, components, and pipir.3
-- include sufficient mass points and corresponding dynamic degrees-of-freedt
'" a three-dimensio 1al representation of N dist :bution of mass and the dynamic characteristics of tiie s., m the selected location of mass points account..
.orsional effects of valves and other eccentric masses.
3.7.2.12 Comparison of Responses 1
See Applicant's SAR.
3.7.2.13 Methods for Seismic Analysis o, Dams _
1 See Applicant's SAR.
3.7.2.14 Determination of Seismic Category I Structure Overturning Moments See Applicant's SAR.
3.7.2.15 Analysis Procedure for Damping I
Uniform modal damping factors given in Section 3.7.2.1.1 are used in the analysis of the coupled components of the reactor coolant system.
Procedures for accounting for system damping are included in Section 3.7.2.1.3.
Amendment No. 1 3.7-10 February 20, 1981
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TABLE 3.7.2-1 i
DAMPING VALUES USED IN ANALYSIS OF CATEGORY l STRUCTURES, SYSTEMS AND COMP 0NENTS j
l i
Maximum Allowable l
Damping percent of critical viscous damping i
Operational Basis Safe Shutdown Item Earthquake Earthquake
(
f Equipment and large diameter piping f
systems, pipe diameter greater
+
1 than 12 inches 2
3 Small diameter piping systems, l
diameter less than or equal 1
to 12 inches 1
2 j
Welded steel structures 2
4 l
Bolted steel structures 4
7
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TABLE 3.7.2-2 (Sheet 1 of 2)
NATURAL FREQUENCIES AND DOMINANT DEGREES OF FREEDOM REACTOR COOLANT SYSTEM Dominant Degrees of Freedom Mode Frequency Number (CPS)
Joint Number Direction Locations 1
1.75 9911 Z
Reactor Internals 2
1.74 9911 X
Reactor Internals 3
12.18 9916 Z
R.V. Top Mass 4
13.06 9916 X
R.V. Top Mass 5
17.43 404, 3404 X
S.G Top Masses 6
17.54 4103, 2103, etc.
X Top Masses of all RCP 7
17.63 2103, 4103 X
Top Masses of RCP IB & 2B 8
17.67 1103, 5103 X
Top Masses of RCP
',A & 2A 9
17.83 3408, 408 Z
S.G. Internals 10 17.83 408, ?408 Z
S.G. Internals 11 17.94 4103, 2103 X
Top Masses of RCP IB & 2B 12 18.00 5103, 1103 X
Top Masses of RCP 1A & 2A 13 19.76 9995 Z
R.V. Lower Mass 14 20.23 9911 Y
Reactor Internals 15 21.02 2103, 4103, etc.
Z Top Masses for all RCP 16 21.02 5103, 4103, etc.
Z Top Masses for all RCP 17 21.02 1103, 5103 Z
Top Masses of RCP 1A & 2A 18 21.02 2103, 4103 Z
Top Masses for RCP 1B & 2B 19 22.34 5103, 1103, etc.
Y All RCP 20 22.36 1103, 5103, etc.
Y All RCP 21 22.36 4103, 2103, etc.
Y All RCP 22 22.36 2103, 4103, etc.
Y All RCP 23 23.11 9905, 9995 X
RV Internals & Externals 24 24.16 404, 3404 Y
S.G. Externals 25 25.23 3404, 404 Y
S.G. Externals 26 24.55 408, 3408 X
S.G. Internals 27 25.89 9905 X
R.V. Internals 28 29.37 404, 3404 Z
S.G. Top Masses 29 29.37 3404, 404 Z
S.G. Top Masses 30 30.05 2580, 4580 Z, Y C.L. Piping 31 32.12 4580, 2580 Z, X C.L. Piping 32 32.40 9911 Y
Reactor Internals 33 32.46 1580, 5580 Z, X C.L. Piping 34 32.53 5580, 1580 Z, X C.L. Piping 35 36.14 5580, 1530 X
C.L. Piping 36 36.40 4580, 2580 X
C.L. Piping 37 36.44 5580, 1580 X
C.L. Piping 38 36.51 2580, 4580 X
C.L. Piping 39 39.41 2580, 4580 X
C.L. Piping Il0 39.79 1580, 5680 X
C.L. Piping
i 3.7.3 SEISMIC SUBSYSTEM ANALYSIS The discussion presented in this section describes procedures employed in seismic subsystem analysis.
3.7.3.1 Seismic Analysis Methods See Section 3.7.2.1 3.7.3.2 Determination of Number of Earthquake Cycles The procedure used to account for the fatigue effect of cyclic motion j
associated with the OBE recognizes that the actual motion experienced during a seismic event consists of a single maximum or peak motion, and some number of cycles of lesser magnitude.
The total or cumulative fatigue effect of all cycles or different magn tude will result in a an equivalent i
r cumulative usage factor.
The equivalent cumulative usage factor can also be specified in terms of a finite number of cycles of the maximum or peak motion.
Based on this consideration, Seismic Category I subsystems, compo-nents, and equipment are designed for total of 200 full-load cycles about a mean value of zero and with an amplitude equal to the maximum response produced during the entire OBE event.
3.7.3.3 Procedure Used For Modeling See Section 3.7.2.3.
3.7.3.4 Basis for Selection of Forcing Frequencies The basis for acceptability of the seismic design of equipment and subsystems is that the stresses and deformations produced by vibratory motion of the postulated seismic events', in combination with other coincident loadings, be within the established limits.
Within practical limitations, the seismic design is accomplished in a manner to maintain the resonant frequencies well above the range which is significantly excited by the forcing frequencies.
If the stresses and deformations resulting from analysis of the preliminary design exceed y
the established acceptable limits the stiffness of the restraint and supports system is modified as required to maintain the fundamental frequen-cies of equipment and subsystems sufficiently removed from the resonant range and, thereby, maintain'the seismic response within the established limits.
The subsystem supports design is sufficiently adaptable that, dependent on the quantitative change in frequency required and the subsystem involved, modifications can be made either by changing the stiffness of existing support assembly components or by adding additional support system restraints to the subsystems or components whose response otherwise y
exceeds the established limits.
If, during the analysis of the preliminary design, frequencies of the reactor coolant system were found tc, be in the range of resonance with those of the O
crease their natural frequencies.
building, the supports for each of the components could be modified to in-gmgndmentNo.g 3.7-11
Specifically, the fundamental frequencies of *.he reactor vessel can be increased in both horizontal directions bs the welding of a set of keys to the RV to further restrain latert.1 motion or rotation of the vessel.
The keys would be laterally restrained by a structure supported by the primary shield wall.
The RCP moves in all three directions when seismically excited in any one direction.
The fundamental frequency of the RCP can be raised y
by relocating the snubber from the top of the motor mount to the top of the motor.
The orientation of the snubber would remain unchanged.
The SG frequency can be raised in the direction parallel to the axis of the RV outlet piping by the addition of a second set of snubbers and levers and in the direction perpendicular to the axis of the RV outlet piping by an additional set of keys above the original set.
See Section 3.9.3.1.3.1 for auxilliary components.
3.7.3.5 Use of Equivalent Static Load Method of Analysis The equivalent static load method involves the multiplication of the total weight of the equipment cr component member by the specific seismic accelera-tion coefficient.
The magnitude of the seismic acceleration coefficient is established on the basis of the expected dynamic response characteristics of the component. Components that can be adequately characterized as a single-degree-of-freedom system are considered to have a modal participation factor of one.
Seismic acceleration coefficients for multi-degree of 9
Amendment No. 1 February 20, 1981
- 3. 7-Il a
c A ASEmam4A.h_ah--u dem mee4A.
M dN..-44.&JJ_N4_h.46
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i freedom systems, which may be in the resonance region of the amplified response spectra curves, are increased by 50% to account conservatively for the increased modal participation.
3.7.3.6 Three Components of Earthquake Motion i
See Section 3.7.2.1.4.
3.7.3.7 Combination of Modal Response See Section 3.7.2.7.
3.7.3.8 Analytical Procedures for Piping See Applicant's SAR.
3.7.3.9 Multiply Supported Equipment Components With Distinct Inputs See Section 3.7.2.1.3.
3.7.3.10 Use of Constant Vertical Static Factors See Section 3.7.2.10.
3.7.3.11 Torisional Effects of Eccentric Masses See Section 3.7.2.1.1.
3.7.3.12 Buried Seismic Category I Piping Systems and Tunnels See Applicant's SAR.
3.7.3.13 Interaction of Other Piping With Category I Piping Dynamic coupling, or interaction effects, between Category I piping systems and piping systems which are not Category I is accounted for in the seismic design of Category I piping.
Where practical, this is accomplished by providing support features at the interfaces between the Category I and the other piping system to dynamically decouple the two systems, otherwise, the mathematical model used for seismic analysis fo the Category I piping is extended to incorporate the pertinent features of the other piping system.
3.7.3.14 Seismic Analysis of Reactor Internals, Core and CEDMs 3.7.3.14.1 Reactor Internals and Core The seismic analyses of the reactor internals and core consists of two phases.
In the first phase, linear lumped parameter models are formulated, natural frequencies and mode shapes for the models are determined, and the response is obtained utilizing the modal analysis response spectrum method.
The response spectra used are based upon the acceleration of the reactor vessel flange.
The response spectrum analysis is used to obtain prelim-inary design seismic loads and displacements in the vertical and horizontal directions.
3.7-I2
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Appendix 4B
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Page Amendment e
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4B-1 to 48-2 48-3 to Table 48-1 1
Figure 4B-1 to 4B-5 Figure 48-6 to Figure 4B-7 1
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TABLE 4.4-4 G
4 REACTOR VESSEL BEST ESTIMATE PRESSilRE LOSSES AND COOLANT TEMPERATURES Pressuregoss Temperature l
Component (lb/in. )
( F)
Inlet nozzle and 90 turn 8.7 565 Downcomer, lower plenum, and support structure 15.4 565 Fuel assembly 15.9 595 1
Fuel assembly outlet to outlet nozzle 16.7 624 Total pressure loss 56.7 l
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TABLE 4.4-5 DESIGN STEADY STATE HYDRAULIC LOADS ON VESSEL INTERNALS AND FUEL ASSEMBLIES Load Values at 500 F Steady State Load Component Description Load Value 2
1.
Core support Radial pressure differential 78 lb/in barrel directed inward opposite inlet duct 6
Uplift load 1.36 x 10 lb Lateral load 0.35 x 10 lb 6
2.
Upper guide Uplift load 0.51 x 10 lb structure Lateral load 0.47 x 10 lb 3.
Flow skirt Radial pressure differential 68 max, psi directed inward 27 avg. psi Axial load directed downward 3050 max. lb/ft. of circ.
1400 avg. Ib/ft. of circ.
4.
Instrumentation Lateral drag load F rected 380 lb, max. support plate supports inward 5.
Instrumentation Uplift load 3300 lb.
support plate 6.
Instrumentation Lateral drag load directed 1200 lb, max.
tube inward tube 7.
Bottom plate Drag load directed upward 70,000 lb 8.
Lower support Drag load directed upward 1,900 lb structure beams Lateral load 8,400 lb 9.
Fuel assembly Uplift load 2480 lb 2
10.
Core shroud Radial pressure differential 40 lb/ip at bottom directed outward 0 lb/in at top
- 11. Fuel alignment Drag load directed upward 184,000 lb plate
- 12. CEA shroud Lateral drag load 2100 lb, max. tube tubes
- 13. Upper guide Lead directed downward 148,000 lb plate
TABLE OF CONTENTS p
_)
CHAPTER 4 APPENDIX 4B Section Subject Page No.
1.0 INTRODUCTION
2.0 DESCRIPTION
OF FLOW fiODEL 48-1 2.1 PRESSURE VESSEL AND CORE SUPPORT STRUCTURES 48-1 2.2 MODEL CORE 4B-1 2.3 MODEL INSTRUMENTATION 4B-1 3.0 DESCRIPTI0fLOFJST__F_ACI_LITY 4B-2 3.1 TEST FACILITY AND OPERATING {0NDITIONS 4B-2 3.2 TEST LOOP 4B-2 3.3 DATA ACQUISITION SYSTEM 4B-2 O\\
\\ j 3.4 CALIBRATION STANDARDS 4B-2 4.0 DATA ANAllSI_S_
4B-2 4.1 MODEL POINT PRESSURES 4B-2 4.2 CORE INLET FLOW DISTRIBUTION 4B-3 4.3 CORE OUTLET PRESSURE DISTRIBUTION 4B-3 4.4 REACTOR VESSEL PRESSURE DROP 4B-4 4.5 COMPONENT HYDRAULIC LOADING 4B-4 t
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LIST OF TABLES l
CHAPTER 4 APPENDIX 4B j
Table Subject 48-1 Reactor Vessel Best Estimate Loss Coefficients & Pressure Drops I
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LIST OF FIGURES f
CHAPTER 4 APPENDIX 4B l
Figure Subject 1
48-1 Reactor Flow Model I
r l
4B-2 Comparison of Reactor and liodel Fuel Assembly Layout 48-3 Pressure Tap Locations in the Reactor Flow Model 4B-4 Test Loop Schematic I
4B-5 Schematic of Data Acquisition System l
4B-6 Core Inlet Flow Distribution Q /Q l
48-7 Core Exit Euler Numbers, E j i
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a.
For the planar pressure distributions at the core inlet and O
core outlet, Pin and Pout:
v Ei = (Pi - Pi) / (Pin - Pout) b.
For other point and spatially averaged pressures:
Ei = (Pi - Pin) / VHref where VH is a reference model inlet velocity head.
ref c.
For point-to-point pressure differentials:
Ei = (P upstream - P downstream) / VHref Euler numbers are readily converted to desired pressure drop loss coefficient and hydraulic loading coefficient forms, considering averaged data from repet.t runs.
4.2 CORE INLET FLOW DISTRIBUTION The System 80 core inlet flow distribution for the normal condition with four operating reactor coolant pumps is provided in Figure 4B-6.
At each fuel assembly location, the inlet flow is expressed as a fraction of the average rm fuel assembly flow rate in the core.
Flow model test data, in the form of L
(V the core inlet pressure distribution, is scaled to reactor conditions and used in a TORC-HERMITE simulation of the System 80 core to determine the I
core inlet flow field. This technique for determining the core inlet flow distribution is discussed further in CENPD-206-P(l), section 3.1.1.
The uncertainty on the core inlet flow distribution includes the test measurement uncertainty and the uncertainty in the TORC-HERMITE calculation of-the flow distribution from the measured pressure distribution. The typical uncertainty on the core inlet flow distribution (AQ /l{} is 0.06 at j
the lo level.
4.3 CORE OUTLET PRESSURE DISTRIBUTION The reactor core outlet pressure distribution is provided in Figure 48-7, for the normal condition with four operating reactor coolant pumps.
Euler numbers at fuel assembly locations express the core outlet pressure distribution in a non-dimensional form which is defined as, Pg - P outlet E ) outlet j
Kcore 9 core I
where: Pg = local static pressure, core outlet F = core average static pressure, core outlet
)
Kcofecore.overall ~ loss coefficient, based on core flow area 5cofeaverage core outlet velocity head, based on core flow area
' Amendment No. 1 48-3
. February 20, 1981
The core outlet pressure distribution is obtained as a result of interfacina two analytical simulations:
the first sinulation is a representation o.
the core region, using the TORC code; the second is a multi-flow-path simulation of the upper plenum region between tne core exit and the outlet nozzles.
Input to the TORC code contains the core inlet flow distribution as determined earlier from flow model test data.
Input to the upper plenun simulation contains the flow resistances found in flow nodel tests for 1
this region.
Matching of the interface conditions between the two siralations provides the core outlet flow and pressure distributions.
Uncertainty in core outlet pressure distribution takes into account the uncertainty in the TORC representation of the core and the uncertainty in the analytical model of the outlet plenum and core exit regions.
The typical uncertainty in core outlet pressure distribution (aE ) is 0.008 et the le level.
O 4.4 REACTOR VESSEL PRESSURE DROP The System 80 reactor vessel incremental loss coefficients and pressure drops, based on test results, are provided in Table 4B-1.
Data for the fuel assem-bly come from full scale flow tests on a typical fuel assembly.
Frictional pressure drops for the fuel assembly and the downcomer are based upon standard friction factor methods. The remaining pressure drops are based upon results from the 3/16 scale flow nodel test.
Uncertainty in incremental loss coefficients from the full scale ficw tests 1
on the fuel assembly is considered to be independent of the un ertainty in the loss coefficients from the 3/16 scale flow model test. The uncertainties are combined by the root-sum-square technique.
The uncertainty in the total reactor vessel pressure drop due to test measurement uncertainty is calculated to be 4.4% at the la level.
4.5 COMPONENT HYDRAULIC LOADING Reactor internal component design steady state hydraulic loads which are verified using scale model flow test data include the following:
Amendment No. 1 February 20, 1981 4S-4
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Core support barrel and upper guide structure uplift forces.
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Differential pressure loadings on the:
Flow skirt Bottom plate Fuel alignment plate Upper guide structure support barrel i
l c.
Steady state drag loading on the cylindrical shroud tubes in
{
the outlet plenum and instrument nozzles in the lower plenum.
Design values for these hydraulic loads based on earlier flow model test results and on analytical methods are in all cases shown to be conservative, on the basis of final model test results.
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..,-,,,-n.__,w,,.
,,n,,,,.,
,,.,,,.n,_
i I
TABLE 4B-1 Reactor Vessel Best Esti;nate Loss Coefficients & Pressure Drops i
Flow Path Stations Loss Coeff., Ki Press. Drop, psi Temp. F l
Inlet Nozzle &90 Turn 1-5 0.69 8.7 565 l
l Downcomer, Lower Plenum f
& Lower Supp. Structure 5-15 1.21 15.4 565 7(
j Fuel Assembly 15-20 1.27 15.9 595 l
I Fuel Assy. Outlet to Outlet Nozzle 20-2s 1.18 16.7 624 L
Total Pressure Drop 56.7 r
Ill!)
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e--,e--mew,-----=-_
~ -.. - - - - - - - - -
--+---.-+----e
a 2
1.11 0.97 1.00 1.05 0.99 0.95 0.93 0.86 0.98 0.87 0.83 0.84 1.04 0.93 0.90 1.02 0.87 0.94 0.93 1.16 1.12 0.97 1.11 1.08 1.13 1.11 1.06 1.09 0.85 0.99 1.10 2A 28 1.07 0.89 0.95 0.93 0.95 0.95 1.01 1.00 1.07 0.96 0.96 0.92 0.88 0.84 0.98 1.10 1.08 0.98 1.08 1.05 1.02 0.95 1.07 0.99 0.95 1.04 1.08 0.96 1.09 1.12 0.90 0.86 1.03 0.89 0.98 0.98 0.95 0.95 0.99 1.15 1.01 1.04 0.98 1.03 0.98 0.94 0.90
< 0.95 1.03 1.05 1.02 1.04 1.07 1.03 1.06 1.08 1.05 1.01 1.05 1.05 1.05 1.12 1.20 0.96 1.02 0.90 1.02 0.92 0.97 0.98 0.99 1.03 0.96 0.98 1.01 0.91 0.93 0.96 0.98 1.00 0.94 1.05 0.88 0.92 1.03 0.87 0.99 0.98 1.03 1.01 0.97 1.02 1.00 0.92 0.95 0.94 0.85 1.16 1.05 1.09 1.08 0.96 0.94 1.04 1.06 1.09 0.97 0.91 0.97 1.11 1.05 0.94 1.07 0.81 0.96
)
0.93 1.09 1.04 1.01 1.16 1.12 1.02 1.06 1.02 1.11 0.96 1.05 1.00 1.08 0.98 1.00 1.01 v
1.02 0.92 1.05 0.95 0.79 1.08 0.98 0.99 0.89 1.01 0.99 1.03 0.93 0.93 1.03 0%7 0.90 1.00 1/4
- J.96 1.03 1.05 0.98 1.08 1.03 1.04 1.14 1.03 1.09 1.05 1.13 1.08 1.01 0.87 0.91 0.86 0.97 1.05 0.97 1.00 0.95 0.90 1.00 0.92 0.93 0.90 1.00 18 1A 1.07 0.92 0.93 1.10 1.11 1.13 1.17 1.03 1.11 1.09 0.98 1.00 1.06 0.94 1.07 0.90 0.99 0.89 1.00 0.88 0.94 0.79 1.07 0.86 0.93 0.96 1.02 1.04 1.02 0.96 1.03 1
9 TN Amendment No. 1
()
February 20, 1981 CORE INLET FLOW DISTRIBUTION Oi/6
^2 O'
O
.001
.004
.004
.004
.001 0
.002
.007
.002
.003
.004
.004
.004
.003.002
.007
.002
.002
.002
.001
.004.004 004
.004
.004
. n04.004
.001
.002
.002 2A 2B
.007
.005 001
.001
.002.002
.002
.002
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.002.002
.001
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.005
.007
.006
.003
.001
.001
.002.002
.002
.002
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.002.002
.001
.001
.003
.006
.004
.007
.001 0
0
.002.002
.002
.002
.002
.002.002 0
0
.001
.007
.004
.006
.003
.001 0
u
.002
.002.002
.002
.002
.002.002
'J 0
.001
.003
.006
.006
.004
.001 0
0
.002
.002
.002
.002
.002
.002.002 0
0
.001
.004
.006
.006
.004
.001 0
0
.002
.002
.002
.002
.002
.002.002 0
0
.001
.004
.006
.006
.004
.001 0
0
.002
.002
.002
.002
.002
.002.002 0
0
.001
.004
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0
.002
.002
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0
.001
.003
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.007
.001 0
0
.002
.002
.002
.002
.002
.002.002 0
0
.001
.007
.004
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.003
.001
.001
.002
.002
.002
.002
.002
.002.002
.001
.001
.003
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.007
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.081
.001
.002
.002.002
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.002.002
.001
.001
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.004
.004 004
.004
.004
.004
.004
.001
.002
.002
.002
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.003
.004
.004
.004
.003
.002
.007
.002 0
.001
.004
.004
.004
.001 0
1 9
(9 Amendment No. 1
\\
)
February 20, 1981 C-E Figure CORE EXIT EULER NUMBERS, EI 48 7
1 1
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6.1-1 to Table 6.3.3.6-1 6.3-40a 1
9 Table 6.3.3.7-1 1
l 6.3-41 l
6.3-42 1
6.3-43 to 6.4-1 t
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TAE'E OF CONTENTS (Cont'd.)
.d CHAPTER 6 Section Subject Page No.
6.3.2.6 Protection Provisions 6.3-20 6.3.2.6.1 Capability to Withstand Design 6.3-20 Bases Environment 6.3.2.6.2 Misrile Protection 6.3-20 6.3.2.6.3 Seismi: Design 6.3-21 6.3.2.7 Required Manual Actions 6.3-21 6.3.3 PERFORMANCE 6.3-22 6.3.3.1 Introduction and Summary 6.3-22 6.3.3.2 Large Break Analysis 6.3-23 6.3.3.2.1 Mathematical Model 6.3-23 6.3.3.2.2 Safety Injection System Assumptions 6.3-23 6.3.3.2.3 Core and System Parameters 6.3-24 6.3.3.2.4 Containment Parameters 6.3-25 6.3.3.2.5 Break Spectrum 6.3-25
(
6.3.3.2.6 Results and Conclusions 6.3-25 6.3.3.3 Small Break Analysis 6.3-27 6.3.3.3.1 Evaluation Model 6.3-27 6.3.3.3.2 Safety Injection System Assumptions 6.3-27 6.3.3.3.3 Core and-System Parameters 6.3-28 6.3.3.3.4 Containment Parameters 6.3-28 6.3.3.3.5 Break Spectrum 6.3-28
- 6.3.3.3.6 Results-6.3-29 6.3.3.3.7 Instrument Tube Rupture 6.3-30 6.3.3.4 Post-LOCA Long Term Cooling 6.3-33 6.3.3.4.1
. General Plan 6.3-33 6.3.3.4.2 Assumptions Used in the Performance 6.3-33 i.
Evaluation of the LTC Plan 6.3.3.4.3 Parameters Used in the Perforamnce 6.3-34 Evaluation of the LTC Plan
-6.3.3.4.4 Results of the LTC Performance Evaluation 6.3-35 6.3.3.5 Sequence of Events and Systems Operation 6.3-36 6.3.3.6
-Radiological Consequences 6.3-39
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6.3.3.7 Chapter 15 Accident' Analysis 6.3-40a 1
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TABLE OF CONTENTS (Cont'd.)
CHAPTER 6 Section Subject Page No.
6.3.4 iESTS AND INPSECTIONS 6.3-41 6.3.4.1 ECCS Performance Tests 6.3-41 6.3.4.2 Reliability Tests and Inspections 6.3-41 6.3.4.2.1 System Level Tests 6.3-41 6.3.4.2.2 Component Testing 6.3-41 6.3.5 INSTRUMENTATION 6.3-41 6.3.5.1 Design Criteria 6.3-41 6.3.5.2 System Actuation Signals 6.3-42 6.3.5.2.1 Safety Injection Actuation Signal (SIAS) 6.3-42 6.3.5.3 Instrumentation During Operation 6.3-42 6.3.5.3.1 Temperature 6.3-43 6.3.5.3.2 Pressure 6.3-43 6.3.5.3.3 Valve Position 6.3-43 6.3.5.3.4 Level 6.3-44 6.3.5.3.5 Flow 6.3-44 6.3.5.4 Post Accident Instrumentation 6.3-44 6.4*
HABITABILITY SYSTEMF 6.4-1 6.5*
FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.4-1 6.6 INSERVICE INSPECTION OF CLASS 2 & 3 COMPONENTS 6.4-1 6.6.l*
COMP 0NENTS SUBJECT TO EXAMINATION 6.4-1 6.6.2 ACCESSIBILITY 6.4-1 APPENDIX 6A CONTAINMENT SPRAY SYSTEM LICENSING REPORT 6A-1 APPENDIX 6B IODINE REMOVAL SYSTEM LICENSING REPORT 6B-1
- See Applicant's SAR O
vi
LIST OF TABLES CHAPTER 6 i
Section Subject J
6.3.3.2-1 Time Sequence of Important Events for a Spectrum of Large l
LOCAs (Seconds After Break) 6.3.3.2-2 General System Parameters and Initial Conditions 6.3.3.2-3 Large Break Spectrum 6.3.3.2-4 Peak Clad Temperatures and 0xidation Percentage for the Large Break Spectrum 6.3.3.2-5 Variables Plotted as a Function of Time For Each Large Break in the Spectrum 6.3.3.2-6 Additional Variables Plotted as a Function of Time For the Worst Large Break 6.3.3.3-1 Safety Injection Pumps Minimum Delivered Flow to RCS (Assuming One Emergency Generator Failed) 6.3.3.3-2 General System Parameters
^#
6.3.3.3-3 Small Break Spectrum 6.3.3.3-4 Variables Plotted as a Function of Time for Each Large Break in the Spectrum 6.3.3.3-5 Fuel Rod Performance Sumary 6.3.3.3-6 Times of Interest for Small Breaks (Seconds) 6.3.3.5-1 Sequence of Events for Representative Large and Small Break LOCAs 6.3.3.5-2 Disposition of Neanally Operating Systems for Large and Small Break L0t;A Analyses 6.3.3.5 Utilization of Safety. Systems for Representative Small 2
j.
Break (0.02 ft )
p 6.3.3.5-4 Utilization of Safety Systems for Representative Large
. Break (0.8 DEG/PD) 6.3.3.6-1 Parameters Used in the Radiological Consequences of a LOCA p) 6.3.3.7-1 Chapter 15 Limiting Events Which Actuate the Safety Injection -
1 System Amendment No. 1 xi-February 20, 1981
'IST OF FIGURES CHAPTER 6 Figure Subject
)
6.2.1-1 Nonnalized Decay Heat Curve 6.2.1-2 Safety Injection Flow Rate vs Time 6.2.1-3 Safety Injection Flow Rate vs Time 6.2.1-4 Safety Injection Flow Rate vs Time 6.2.1-5 Safety Injection Flow Rate vs Time 6.2.1-6 Safety Injection Flow Rate vs Time 6.2.1-7 Safety Injection Flow Rate vs Time 6.2.1-8 Safety Injection Flow Rate vs Time 6.2.1-9 Safety Injection Flow Rate vs Time 6.2.1-10 Steam Line Model 6.2.1-11 102% Power - Slot MSLB, Loss of 1 CHRS, Feedwater Addition vs Time 6.2.1-12 102% Power - Guillotine MSLB, loss of 1 CHRS, Feedwater Addition vs Time 6.2.1-13 75% Power - Slot MSLB, Loss of 1 CHRS, Feedwater Addition vs Time 6.2.1-14 75% Power - Guillotine MSLB, Loss of I CHRS, Feedwater Addition,vs Time 6.2.1-15 50% Power - Slot MSLB, Loss of 1 CHRS, Feedwater Addon vs Time 6.2.1-16 50% Power - Guillotine MSLB, Loss of 1 CHRS, Feedwater Addition vs Time 6.2.1-17 25% Power - Slot MSLB, Loss of 1 CHRS, Feedwater Addition vs Time 6.2.1-18 25% Power - Guillotine MSLB, Loss of 1 CHRS, Feedwater Addition vs Time O
xii
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6.3.3.7 Chapter 15 Accident Analysis l
For the limiting event in Chapter 15, the safety systems actuated are listed i
in tables designated 15.X.X.X-3, " Utilization of Safety Systems". The events 1
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which result in safety injection actuation are identified in Table 6.3.3.7-1.
These limiting events meet the acceptance guidelines of Table 15.0-3.
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TABLE 6.3.3.7-1 j
CHAPTER 15 LIMITING EVENTS WHICH ACTUATE i
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THE SAFETY INJECTION SYSTEM 1
Event Section 15.1.3.2 Increased Main Steam Flow through an ADV 15.1.4.2 Small Steam Line Break, Outside Containment, l
Upstream of the MSIV 15.2.1.1 Turbine Trip i
15.2.2.1 Loss of Condenser Vacuum with Failure to Achieve e
Fast Transfer (FTF) i 15.4.5.1 and l
15.4.5.2 CEA Ejection with FTF and Technical Specification
[
Tube Leakage l
15.6.2.2 Steam Generator Tube Rupture O
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6.3.4 TESTS AND INSPECTIONS During fabrication of the SIS components, tests and inspections are performed and documented in accordance with code requirements to assure high quality construction. As necessary, performance tests of components are performed in the vendor's facility.
The SIS is designed and installed to permit inservice inspections and tests in accordance with ASME Code Section XI.
6.3.4.1 ECCS Performance Tests Prior to initial plant startup, a comprehensive series of system flow tests as detailed in section 14.2, will be performed to verify that the design performance of the system and individual components is attained.
6.3.4.2 Reliability Tests and Inspections 6.3.4.2.1 System Level Tests After the plant is brought into operation, periodic tests and inspections of the SIS components and subsystems are performed to ensure proper operation in the event of an accident.
The scheduled tests and inspections are necessary to verify system operability, since during normal plant operation, SIS components are aligned for emergency operation and serve no other function.
The tests defined permit a complete checkout at the subsystem and component level during normal plant operation.
Satisfactory operability of the complete system can be verified during normal scheduled refueling O
detailed in Chapter 16.
shutdown.
The complete schedule of tests and inspections of the SIS is 6.3.4.2.2 Component Testing In addition to the system level tests described in Section 6.3.4.1, tests to verify proper operation of the SIS components are also conducted.
These tests supplement _the system level tests by verifying acceptable performance of each active component in the SIS.
Pumps and automatic valves will be tested in accordance with ASME Section XI.
6.3.5 INSTRUMENTATION 6.3.5.1 Design Criteria The instruments and controls for the Safety Injection System are designed in accordance with the applicable portions of IEEE 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations." The controls are interlocked to automatically provide the sequence of operations required to initiate Safety Injection System operation.
The instrumentation and controls which actuate and control the Safety Injection System are designed on the following basis.
a.
Redundant insMusiiants are provided for initiation of Safety Injection Systems actions.
Four sensors are used for each of the critical
)
parameters.
A trip from any two of these four sensors initiates the J
6.3-41
appropriate Safety Injection System action.
Circuits are run in separate wiring raceways to assure the availability of Safety Injection Actuation Signals.
b.
Electric power required for Safety Injection System controls and instruments is supplied via two preferred ac buses.
Emergency Generators provide an alternate source of power.
Actuator-operated valves are provided with key-operated control switches where considered necessary to prevent unintentional misalignment of safety injection flow paths during power operation.
All valv'es that are not required to operate on initiation of safety injection or recirculation in the safety injection flow path are locked in the safety injection position during cperation.
Administrative controls ensure that the valves are locked in the correct position.
A further discussion of the instrumentation and associated analog and logic channels employed for safety injection initiation is given in Section 7.3.
6.3.5.2 System Actuation Signals Operation of the Safety Injection System is controlled by two actuation signals. The first of these, the Safety Injection Action Signal (SIAS),
initiates operation of the Safety Injection System in the event of low reactor coolant system pressure or high containment pressure. Both of these parameters provide an indication of a Loss-of-Coolant Accident which requires operation of the Safety Injection System.
SIAS may be manually initiated from the control room. The second control signal is the Recircu-lation Actuation Signal (RAS). This signal changes the operation mode of the Safety Injection System from injection with suction from the refueling water tank to recirculation with suction from the containment sump.
The RAS is initiated by low refueling water tank level: RAS occurs automatically, whether SIAS is initiated manually or automatically. Changing from the injection mode of operation to recirculation permits continuous flow to the core when the RWT water supply is depleted.
6.3.5.2.1 Safety Injection Actuation Signal (SIAS)
Initiation of safety injection is derived from four independent pressurizer pressure sensors and four independent containment pressure sensors. Coinci-dence trip signals from two-out-of-four sensors for either parameter will automatically initiate safety injection. Automatic Safety Injection System operation is acuated at a pressurizer pressure of 1750 psia during power operation or a containment pressure of 5 psig. During startup and shutdown operations, a variable setpoint on the low Pressurizer Pressure is used. A further discussion of the SIAS is given in Section 7.3.
6.3.5.3 Instrumentation During Operation The instrumentation provided for monitoring Safety Injection System components during Safety Injetion System operation is discussed in this section.
See Figures 6.3.2-1A and IB for instrumentation readout locations, and Figures I
6.3.2-lC through lL for component usage during the various modes.
Amendment No. 1 6.3-42 February 20, 1981
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8.1-1 to 8.1-2 8.1-3 1
8.2-1 to 8.3-2 c
8.3-3 1
Table 8.3.1-1 (Sheets 1 & 2) j Table 8.3.1-1 Sheet 3 1
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IEEE Standard 323-1974; Qualifying Class lE Equipment for Nuclear
(
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Power Generating Stations.
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IEEE Standard 336-1971; Installation, Inspection and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations.
aj.
IEEE Standard 338-1971; Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems, ak.
IEEE Standard 344-1971; Guide for Seismic Qualification of Class 1 Electrical Equipment for Nuclear Power Generating Stations.
al.
IEEE Standard 379-1972; Guide for the Application of the Single Failure Criterion to Nuclear Power Generating Station Protection Systems.
am.
IEEE Standard 384-1974; Criteria for Separation of Class lE Equipment and Circuits.
ar.
IEEE Standard 387-1972; Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Stations.
ao.
IEEE Standard 450-1972; Recommended Practice for Maintenance, Testing and Replacement of Large Stationary Type Power Plant and Substation Lead Storage Batteries.
f);
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IEEE Standard 334-1971; Trial Use Guice for Type Tests of Continuous-
\\
Duty Class I Motors Installed Inside the Containment of Nuclear Power Generating Stations.
aq.
IEEE Standard 382-1972; Trial Use Guide for Type Test of Class I Electric Valve Operators for Nuclear Power Generating Stations.
ar.
IEEE Standard 383-1974; Type Test of Class lE Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations.
1 as. Regulatory Guide 1l40; Qualification Tests of Continuous-Duty Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants (Rev. O, 3/73).
at. Regulatory Guide 1.73; Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants (Rev. O,1/74).
l Regulatory Guide 1.89; Qualification of Class lE Equipment for Nuclear au.
Power Plants (Rev. O, 11/74).
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a.
Table 8.3.1-1 Power Requirements for CESSAR Design Scope Safety-g j
Related Equipment.
v b.
Table 8.3.1-2 Power Requirements for CESSAR Design Scope Safety-Related Equipment at Various Operating Conditions.
c.
Table 8.3.1-3 Power Requirements for CESSAR Design Scope Safety-Related Electrical Equipment During Emergency Operation.
de Table 8.3.1-4 Required Standby Generator Loads 9.
The vital instrument buses shall be designed such that the maximum voltage fault shall not exceed 480 VAC + 10% or 325 VDC + 10%.
This y
maximum fault would remove the faulted channel from service.
- 10. Cabling shall meet the requirements specified in Sections 7.1.3, 7.2.3, and 7.3.3 for seperation and independence so that no credible fault in one channel can be propagated.
8.3.2 DC POWER SYSTEM The DC Power System is supplied by the Applicant and will be discussed in his safety Analysis Report. The following interface requirements are imposed to ensure safety-related equipment is provided with adequate and appropriate power under all operating conditions.
Interface Requirements U
l.
The dc power supplied shall be 125 volts with maximum limits of 105 volts to 140 volts.
2.
Four vital buses, one for each safety channel, shall be supplied.
Each shall have a battery and charger separate and independent from the other safety channel batteries and chargers.
1 3.
Each battery shall be capable of supporting the dc loads until ac power is restored, from the standby power supply.
4.
Non-safety loads shall not degrade the vital supplies.
5.
The dc supplies shall meet the same seismic and environmental require-l ments as the loads they supply, consistent with their location in the l
facility.
6.
Maximum credible fault voltage applied to the distribution system shall be 480 Vac. This maximum fault would remove the faulted channel from service.
7.
The maximum voltage transient shall be 75 Vdc for two milliseconds.
8.3.3 FIRE PROTECTION FOR CABLE SYSTEMS
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Refer to the Applicant's Safety Analysis Report.
Amendment No. 1 February 20, 1981 8.3-3
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i TABLE 8.3.1-1 (Cont'd) (Sheet 3 of 3) t POWER REQUIREMENTS FOR CESSAR DESIGN SCOPE SAFETY-RELATED EQUIPMENT i
BUS NOMINAL LOAD /
SYSTEM CESSAR EQUIPMENT QUANTITY V0LTAGE (3)
COMPONENT (6) 4
!~
l-Miscellaneous Loads
~ Vital. Instrument Buses (5) 4' 120 14.5KVA (A) 17.2KVA (B) i 15.9KVA (C) i 13.8KVA (D) i e
Reactor Trip Switchgear Breakers 4
125VDC 5.51KW (A) 1 i
5.51KW (B) 1 t
5.51KW (C)
F i
5.51XW (D) i NOTES:
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. The third charging pump is capable of being shifted to either emergency bus.
l 1
(1) Automatic connection to Emergency Buses required.
I j.
(2) Split between Emergency Buses required.
j
.(3) Specific voltages to be supplied by Applicant, specified voltages are typical.
(4) Manual connection to Emergency Buses capability required, i
(5) Refer to Section.8.3.1 AC Power System. Interface Requirement 4.
(6) - Nominal Load / Component to be supplied by the Applicant, specified values are typical.
(7). Refer to Section 8.3.1 AC-P0wer System, Interface Requirement 6.
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.(8). Powered.such that Section 6.3.1.3.A.6 is satisfied (9) Powered such that Section 5.4.7.1.3.A.2 C is satisfied.
Amendment No. 1 February 20, 1981 i
1
TABLE 8.3.1-2 POWER REQUIREMENTS FOR CESSAR DESIGN SCOPE SAFETY-RELATED EQUIPMENT AT VARIOUS OPERATING CONDITIONS TOTAL POWER IN KILOWATTS (3)
SYSTEM ELECTRICAL SYSTEM QUANTITY START UP SHUTDOWN REFUELING NORMAL 4.16 KV (1)
0 0
1442 0
752 752 752 0
480 V (1)
CVCS Charging Pumps 3
80 160 0
160 120 VAC Vital (2)
Vital Instrument Bus (KVA) 4 61.4 61.4 61.4 61.4 NOTES:
(1) Specific voltages to be supplied by Applicant, specified voltages are typical.
(2) Refer to Section 8.3.1 AC POWER SYSTEM, Interface Requirement 4.
(3) Specific total power to be supplied by the Applicant, specified total power it. typical.
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(s TABLE OF CONTENTS CHAPTER 11 Section Subject Page No.
11.0 RADI0 ACTIVE WASTE MANAGEMENT 11.1-1 11.1 SOURCE 11.1-1 11.1.1 DESIGN BASIS SOURCE TERMS 11.1-1 11.1.1.1 Maximum Fission Product Activities in 11.1-1 Reactor Coolant 11.1.1.2 Normal Operating Source Terms Including 11.1-3 Anticipated Operational Occurrences 11.1.2 DEPOSITED CRUD ACTIVITIES 11.1-4 11.1.3 TRITIUM PRODUCTION IN REACTOR C0OLANT 11.1-7 11.1.3.1 Activation Sources of Tritium 11.1-7 11.1.3.2 Tritium From Fission 11.1-7 3,
(\\_ /)
11.1.4 NEUTRON ACTIVATION PRODUCTS 11.1-9 11.1.4.1 Nitrogen-16 Activity 11.1-9 11.1.4.2 Carbon-14 Production 11.1-9 11.1.5 FUEL EXPERIENCE 11.1-10 11.1.6 LEAKAGE SOURCES 11.1-11 11.1.7 SPENT FUEL P00L FISSION PRODUCT AND CORROSION 11.1-12 PRODUCT ACTIVITIES 11.1.8 STEAM GENERATOR ACTIVITY MODEL 11.1-13 11.1.9 RADWASTE SYSTEMS 11.1-13 1
APPENDIX CORE RESIDENCE TIMES ll.A-1 11.1-A
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N. s Amendment No. 1 i
February 20, 1981
LIST OF TABLES CHAPTER 11 Table Subject 11.1.1-1 Basis for Reactor Coolant Fission oroduct Activities 11.1.1-2 Maximum Activities in the Reactor Coolant Due to Continuous Operation at 4100 Mwt with One Percent Failed Fuel 11.1.1-3 Reactor Coolant System Activities During Nornal Operations Including Anticipated Operational Occurrences 11.1.2-1 Long-Lived Isotopes in Crud 11.1.2-2 Measured Radioactive Crud Activity (dpm/mg-crud) 11.1.2-3 System Parameters 11.1.2-4 System Parameters 11.1.2-5 Average and Minimum Residence Tines, Days 11.1.2-6 Assumed System Parameters, System 80 11.1.2-7 Long-Lived Crud Activity for a Standard 3817 Mwt Plant 11.1.2-8 Average Calculated Reactor Coolant Crud Activity 11.1.2-9 Equilibrium Crud Film Thickness 11.1.3-1 Tritium Activation Reactions 11.1.3-2 Parameters Used in Tritium Production Determination 11.1.3-3 Tritium Production in Reactor Coolant 11.1.3-4 Tritium Production and Release at Operating PWR's 11.1.6-1 Leakage Assumptions From C-E Supplied Equipment 11.1.7-1 Maximum Fission and Corrosion Product Activities in the Spent Fuel Pool 11.1.7-2 Fission and Corrosion Product Activities in the Spent Fuel Pool Under Nonnal Conditions Including Anticipated Operational Occurrences 11.1.8-1 Basis for Steam Generator Liquid Activities O.
ii
11.1.8 STEAM GENERATOR ACTIVITY MODEL The specific activities in the steam generator and secondary systems are to be discussed in the Applicant's SAR.
The bases data for these activities are supplied in Table 11.1.8-1.
11.1.9 RADWASTE SYSTEMS Detailed information, including references to P& ids pressures, temperatures, flow rates, and expected volumes of waste input to each of the radwaste systems is pro <ided below.
Liquid Waste Management Systen (LW'45) 1.
Chemic.il addition package strainer drain (Zone G-5 of Figure 9.3-1).
Flow 0-10 gpm Chemical Nature Primary grade water and 2060 ppm LiOH(max)
Pressure 25 psig Temperature 40-120 F 2.
Supply to LWMS waste condensate tank (Zone D-5 of Figure 9.3-3).
Flow 0-20 gpm i
Chemical Nature Primary grade makeup water Temperature 120-130'F p
Pressure 55 psig 3.
Supply to LWMS waste concentrator (Zone G-7 of Figure 9.3-3).
Flow 0-20 gpm Chemical Nature Pdmary water Temperature 40-90*F Pressure 60 psig 4.
BAC Drains (Zone D-5 of Figure 9.3-3).
Flow 0-20 gpm Chemical Nature Primary water and component cooling water 40-200 F Temperature Pressure ATM Solid Waste Management System (SWMS) 1.
Ion exchanger resin sluicing lines (Zones A-1, C-4, and G-5 of Figures 9.3-4, 9.3-2, and 9.3-3, respectively).
V Amendment No. I 11.1-13 February 20, 1981
Flow 100 gpm (max) water and 100 SCFM (max) air Chemical f(ature Resin, air, reactor makeup water Pressure 75 psig 40-12g/ionexchangersluicing Temperature F
Volume of dewatered resin 36 ft operatjon Volume of resin discharged 180 ft oer year 2.
Strainerblowdown lines (Zones A-1, F-4, and G-5 of Figures 9.3-4, 9.3-2 and 9.3-3, respectively)
Flow 10 gpm Chemical flature Resin slurry Pressure 50 psig Temperature 40-120 F 3.
Boric acid concentrator concentrate discharge to SWMS (Zone B-5 of Figure 9.3-3).
Continuous:
Flow 0-20 gpm 12 wt % boric acid (max)
Temperature 160-180'F Pressure 55 psig Batch:
Flow 20 gpm 12 wt % boric acid (max) 1 Temperature 160-180 F Pressure 55 psig Volume 2000 gallons (max) 4.
The Solid Waste Management System shall be capable of receiving the following quantities of spent filter cartridges or equivalent each year:
3 Seal injection filters 4.18 ft Purification filters 16.72 ft 3 Boric acid filters 2.09 ft Reactor drain filters 2.09 ft 3 Reactor makeup filters 2.09 ft Gas Waste Management System (GWMS) 1.
Purification and deborating ion exchanger vent (Zone D-1 of Figure 9.3-4; one ion exchanger vent rate / year).
Flow 0-20 scfm Chemical flature Air Temperature 40-120 F Pressure 0 psig Volume 170 scf/ year O
Amendment tio. 1 February 20, 1981 11.1-
/
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2.
Refueling Water Tank vent (Zone E-8 of Figure 9.3-1)
Flow 0-1500 scfm (during refueling)/
0-25 scfm (normal) i Chemical Nature Air i
Temperature 60-90 F Pressure 1.5 psig Volume 86,000 scf/ year 3.
Volume Control Tank Gas relief (Zone F-4 of Figure 9.3-1)
Flow 40 scfm Chemical Nature H or N 2
Temperature 130-140F Pressure Built-up & superimposed back-pressure to be supplied by A/E Volume 1200 scf/ year 4.
Volume Control Tank vent (Zone F-3 of Figure 9.3-1)
Flow 0-22 scfm Chemical Nature H and/or N Pressure 4bpsig(mak)
Temperature 120-140 F Volume 3110 scf/ year 5.
Volume Control Tank gas sample line (Zone E-3 of Figure 9.3-1) 1 Flow
<1 scfm (during gas sampling)
Chemical Nature H, and/or N2 Pressure 20 psig Temperature 120-140 F 6.
Reactor Drain Tank vent (Zone G-7 of Figure 9.3-2)
Flow 0-25 scfm (intermittent) */<lscfm (during gas sampling)
Chemical Nature N H Temperature 13bf Pressure 4 psig (max)
Volume 7759 scf/ year 7.
Preholdup ion exchanger vent (Zone D-4 of Figure 9.3-2)
Flow 0-20 scfm Chemical Nature Air Temperature 40-120 F Pressure:
0 psig Volume 60 scf/ year t
- based upon 165 gpm makeup to tank Amendment No. 1 ll.1-13b February 20, 1981 I
8.
Equipment Drain Tank vcr+ (Zone H-5 of Figure 9.3-2)
Flow 0-25 scfm (intermittent)*/<1 scfm (normal)
Chemical Nature N
Temperature 130F Pressure 4 psig (max)
Volume 7759 scf/ year 9.
Equipment Drain Tank gas sample line (Zone G-5 of Figure 9.3-2)
Flow
<1 scfm (during gas sampling)
Chemical Nature N2 Pressure 3 psig Temperature 120 F 10.
Holdup Tank vent (Zone G-2 of Figure 9.3-2)
Flow 0-22 scfm (intermittent)/
0-20 scfm (normal)
Chemical Nature Air Temperature 120 F Pressure 0 psig Volume 160,000 scf/ year
- 11. Holdup Tank gas to analyzer (Zone G-2 of Figure 9.3-2)
Flow
<l scfm (during gas sampling)
Chemical Nature Air Temperature 40-90 F Pressure 0 psig
- 12. Boric Acid Concentrator vent (Zone F-5 of Figure 9.3-3) y Flow 0-5 scfm (startup)/<1 scfm (nonnal)
Chemical Nature Air, N, water vapor 2
Temperatpre 150 F Pressure 6 psig Volume 1400 scf/ year 13.
Boric Acid Condensate ion exchanger vent (Zone G-4 of Figure 9.3-3)
Flow 0-20 scfm Chemical Nature Air Temperature 120 F (max)
Pressure 0 psig Volume 100 scf/ year
- based upon 165 gpm makeup to tank O
Amendnient No. I ll.1-13c February 20, 1981
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14.
Reactor Makeup Water Tank vent (Zone D-3 of Figure 9.3-3) l Flow 0-22 scfm @ 165 gpm fill rate i
Chemical fiature prinary grade water
{
Tempereture 40-90 F Pressu+e O psig Volume 67,920 scf/ year
- 15. Gas Stripper Vent to gas surge header (Zone D-2 of Figure 9.3-2)
Flow 0-20 scfm*/<1 scfm (nonaal) 1 l
Chemical fiature H, noble gases g
Temperature 140 F (max) 4 l
Pressure 9 psig i
j Volume 145,672 scf/ year
- 16. Gas Stripper Gas to analyzer (Zone C-2 of Figure 9.3-2) t Flow
<1 scfm (during gas sampling)
Chemical fiature H, noble gases 2
Pressure 9 psig Temperature 140 F (max)
- Startup, shutdown or purges O
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6 Amendment tio. 1 i
oll.1-13d February 20, 1981
[
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l
REFERENCES FOR SECTION 11.I 1.
M. E. Meek, B. F. Rider, " Summary of Fission Product Yields", NE00-12154, January 1972.
2.
" Chart of Nuclides," USAEC, Modified by Battelle-Northwest, May 1969 and May 1970.
3.
" Neutron Cross Secticas," BNL 325 Supplement No. 2, May 1964.
4.
J. D. Eichenberg, " Effects of Irradiation on Bulk 00," WAPD-183, 2
October 1957.
5.
G. M. Allison and H. K. Rae, "The Release of Fission Gases and Iodines from Defected UO Fuel Elements of Different lengths,' AECL-2206, June 2
1965.
6.
ANSI N237, " Radioactive Materials in Principal Fluid Streams of Light-Water Cooled Nuclear Power Plants.
7.
Connecticut Yankee Monthly Operating Reports, 2/68, 3/68, 6/68, 7/68, 12/68, 1/69, 3/69-5/69, 8/69, 10/69, 12/69, 3/70, 10/70, 11/70.
8.
San Onofre Monthly Operating Reports, 1/71-3/71, 6/71-9/71, 11/71, 12/69, 1/70.
9.
Yankee Rowe Montly Operating Reports, 2/69-6/69, 8/69-12/69, 160-12/70, 1/72, 4/72-7/72.
10.
Large Closed-Cycle Water Reactor Research and Development Program, Progress Report January 1, 1965 - March 31, 1965, WCAP-3620-12.
11.
The Saxton Chemical Shim Experiment, Weisman 1, Bartnoff S, July 1965, WCAP-3269-24.
12.
Large Closed-Cycle Water Reactor Research and Development Program, Progress Report April 1, 1965 - June 30, 1965, WCAP-3269-1.3.
13.
Corrosion Product Behavior in Stainless-Steel-Clad Water Reactor Systems, Nuclear Applications, Vol. I October 1965.
14.
Decontamination of the Shippingport Atomic Power Station, Abrams C.
S.,
Salterelli E. A., January 1966, WAPD-299.
15.
Radiation Buildup on Mechanisms and Thermal Barriers, Weingart E.,
June 1963, WAPD-PWR-TE-145.
16.
Indian Point 1 Semi-Annual Operations Reports, 9/66, 9/67, 3/68, 9/68.
17.
Test Data Sheets, Maine-Yankee Core Crud Removal, CENPD-113, August 23, 1973.
11.1-14
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J EFFECTIVE PAGE LISTING i
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Chapter 12 I
e Table of Contents Page Amendment i to 11 1
j Text i
Page Amendment l
i 12.1-1 to 12.2-la 1
l 12.2-2 to 12.2-3 i
12.2-4 1
l Table 12.2-1 to Table 12.2-10 Table 12.2-11 1
r 12.3-1 1
i 9
12.3-2 l
12.3-3 to 12.3-5 1
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TABLE OF CONTENTS
(
)
CHAPTER 12
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Section Subject Page No.
12.0 RADIATION PROTECTION 12.1-1 12.1*
ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES 12.1-1 ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.l*
POLICY CONSIDERATIONS 12.1-1 12.1.2 DESIGN CONSIDERATIONS 12.1-1 1
12.1.3*
OPERATIONAL CONSIDERATIONS 12.1-3 12.2 RADIATION SOURCES 12.2-1 12.2.1 CONTAINED SOURCES 12.2-1 12.2.1.1 Containment 12.2-1 12.2.1.1.1 Reactor Core 12.2-1 12.2.1.1.2 Reactor Coolant System 12.2-1 12.2.1.1.3 Main Steam Supply System 12.2-1
~s
/
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12.2.1.1.4 Spent Fuel Handling and Transfer 12.2-1
(,_,/
12.2.1.1.5 Processing Systems 12.2-2 12.2.1.1.5.1 Chemical and Volume Control 12.2-2 System (CVCS) 12.2.1.2 Safety Equipment Building 12.2-4 12.2.1.2.1 Shutdown Cooling System 12.2-4 12.2.1.2.2*
Component Cooling Water System 12.2-4 12.2-4 12.2.1.3
. Fuel Building 12.2.1.3.1 Spent Fuel Storage and Transfer 12.2-4 12.2.1.3.2*
Spent Fuel Pool Cooling and Cleanup 12.2-4 System 12.2.1.4*
Turbine Building 12.2-4 12.2.1.5*
Auxiliary Building 12.2-4 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 FACILITY DESIGN FEATURES 12.3-1
/
12.3.1.l*
Radiation Zone Designation 12.3-1
'~~'
12.3.1.2 Equipment and System Design Features 12.3-1 for Control of On-Site Exposure Amendnent No. 1
- See Applicant's SAR i
February 20, 1981
LIST OF TABLES CHAPTER 12 l
Table _
Subject 12.2-1 Maximum Neutron Spectra Outside Reactor Vessel 12.2-2 Maxinum Gamma Spectra Outside Reactor Vessel 12.2-3 Shutdown Gamma Spectra Outside Reactor Vessel 12.2-4 N-16 Activity 12.2-5 Spent Fuel Gamma Source l
t 12.2-6 CVCS Heat Exchanger Soluble Inventories l
l 12.2-7 CVCS Heat Exchanger Activity, flaximum Values 12.2-8 CVCS Ion Exchanger Inventories 12.2-9 CVCS Filter Inventories 12.2-10 CVCS Tank Inventories 12.2-11 Shutdown Cooling System (SDCS) Specific 1
Source Strengths O
Amendment No. 1 ii February 20, 1981
fm}
12.
RADIATION PROTECTION v
This chapter describes the radiation protection measures incorporated in the station design and in the operating procedures to ensure that internal and external radiation exposures to station personnel, contractors, and the general population due to station conditions, including anticipated operational occurrences, will be within all applicable limi+.s, and furthermore, will be as low as is reasonably achievable (ALARA).
12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS IS REASONABLE ACHIEVABLE (ALARA) 12.1.1 Policy Considerations (This subsection to be provided by applicant) 12.1.2 Design Considerations Experience from past designs and operating reactors has been employed in the establishment of radiation protection design guidelines.
A program of data acquisition and retrieval has been employed to establish an equipment and system design bases.
The engineering effort is directed toward characterizing the mechanisms of radiation level buildup, evaluating the performance of plant systems in mitigating radioactive buildup, and establishing the role of operating procedures in reducing rediation level buildup.
N Data and experience gained from this effort is directly applied to all disciplines in the design and development of equipment employed in the System 80 NSSS.
Systems and equipment employed in the System 80 NSSS have been designed with the objective of reducing the need for maintenance within radiation areas. Whenever possible components requiring frequent maintenance are separated for ldpation in low radiation zones or are flanged to facilitate ease of removal to a low radiation zone.
Whenever possible materials are selected to withstand a 40 year service life thus minimizing the need for i
replacement and reducing maintenance frequencies.
Controls are remotely mounted for location in a low radiation zone.
Equipment such as heat exchangers and valves are designed for ease of access during maintenance.
Equipment is environmentally qualified to meet their performance requirements under the environmental l
and operating conditions in which they will be required to function.
The overall objective in equipment design is to insure the occupa-tional radiation exposures are ALARA by ensuring that operators are required to spend a minimum amount of time in a radiation environment.
Experience from past NSSS designs and inservice inspection programs have resulted in design features being incorporated into the System 80 NSSS that reduce occupational radiation exposure.
The most significant improvement for performing inservice inspection is the reduction of linear feet of weld in the major components.
5 4
V 12.1-1 Amendment No. 1 February 20, 1981 L
The reduction in weld footage has been accomplished by component redesign, u p of forged sections versus forged-welded plate sections, and increasing the size of certain sections.
Systems and equipment employed in the Systeu 80 NSS$ have been designed with the objective of ensuring that occupational exposure due to decommissioning pror.edui as will be ALARA.
Decommissioning can be facilitated in the. design stage through features which will minimize the buildap of in-plant radiation and contamination.
It is anticipated that decommissioning will be accomplished terough the application of one of sevaN1 available alternative methods, e.g., mothballing, entombment, immediate or delayed dismantling, etc.
The experience Jained in the continued application of these methods, and any developing variations, will further minimize occupational radiation exposures.
The System 80 NSSS incorporates many of the design features recommended in Regulatory Guide 8.8 in addition to other specific designs and established guidelines to keep in plant exposures ALARA.
The following design features are specifically effective in reducing in plant exposures during decommissioning.
1.
Components containing radioactive material, such as primary coolant, resin and concentrates are provided with connections for flushing with water or decontamination chemicals.
2.
Equipment is designed to minimize crud buildup and facilitate decontamination.
1 3.
Spaces are provided where appropriate to place shielding for the purpose of reducing neutron activation.
4.
Activated corrosion product buildup over the ife of the NS$$
is minimized in the design stage through appropriate selection of corrosion resistant materials, specification of an appro-priate chemistry control program and limitation applied to the cobalt content of materials exiosed to the primary coolant.
Established radiation protection guidelines have been employed to meet the intent of Regulatory uuide 8.8.
These guidelines are provided to design engineers in each discipline to ensure occupa-tional radiation exposures are maintained ALARA.
Radiation protection guidance provided in Regulatory Guide 8.8.
These guidelines are provided to design engineers in each discipline to ensure occupational radiation exposures are maintained ALARA.
Radiation protection design reviews are performed based upon these guidelines and the guidance provided in Regulatory Guide
- 8. 8.
The general design objective for systems and equipment is to reduce exposure to operating personnel as low as reasonably achievable to meet the intent of Regulatory Guide 8.8 and to 12.1-2 Amendment No. 1 February 20, 1981
i j
A operate within the limits of radiation protection in restricted artas given in 10CFR20 and 10CFR50.
Radiation protection design review is an ongoing activity through-out all phases of the design.
The Supervisor of the Radiation i
Control Systems Group is responsible for the establishment of design guidelines to maintain occupational radiation exposrue ALARA and for the performance of radiatinn protection design and review.
Radiation control engineers ensure that these guidelines are provided to engineers working in other disciplines involved i
in the NSSS design and subsequently monitor their implementation.
Since all disciplines involved in the NSSS design are covered by 1
the guidelines, every discipline is involved in the reivew, t
Radiation control engineers work directly with engineers and designers in other disciplines to ensure that all radiation protection considerations are taken into account.
These engineers advise on the most desirable design option for radiation protection when alternate designs are possible in satisfying the process requirements.
Such recommendations involve a study of the exposares which are likely to be derived from alternate designs and the selection of the option resulting in the lowest exposure.
Radiation protection design reviews take place prior to the release of design drawings, system design requirements or component
"g design requirements.
Comments from the engineers performing the 3
review are transmitted by appropriate documentation to the applicable design engineers for resolution.
Followup reviews are conducted to ensure resolution within the established radiation protection guidelines.
12.1.3 Operational Considerations (This subsection to be provided by applicant) i l
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i 12.1-3 Amendment No. 1 February 20, 1981
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O THIS PAGE INTENTIONALLY BLANK 1
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e 12.2 RADIATION SOURCES This section discusses and identifies the sources of radiation that form the basis for shield design calculations for the design of personnel protec-tive measures and for does assessment.
12.2.1 CONTAINED SOURCES The shielding design source terms are based on full ;:ower operation with 1%
fuel cladding defects. Sources in the primary coolanu include fission products released from fuel clad defects, activation and corrosion products.
The sources in the reactor coolant are discussed in Section 11.1.
Throughout most of the reactor coolant system, activation products, principally nitrogen-16 (N-16), are the primary radiation sources for shielding design during power operation.
The design sources are presented in this subsection by system.
Location of the equipment is discussed in the Applicant's SAR.
12.2.1.1 Containment 12.2.1.1.1 Reactor Core The primary radiation emanating from the reactor core during normal operation are neutrons and gamma rays.
Tables 12.2-1 and 12.2-2 list neutron and m
gamma multigroup fluxes in the reactor cavity at the side of the reactor (V
)
vessel; thr le tables are based on nuclear narameters discussed in Chapter 4.
Table i!.2-3 lists core gamma sources after shutdown for shielding requirements during shutdown and inservice inspection.
12.2.1.1.2 Reactor Coolant System Sources of radiation in the reactor coolant system are fission products released from fuel and activation and corrosion prodbcts that are circulated in the reactor coolant. These sources are listed in Section 11.1 and Table 12.2-4 and their bases are discussed in Section 11.1.
Tables 11.1.2-7 and 11.1.2-9 list maximum expected activities due to crud i
deposits on steam generator tubing and primary system pipina.
l The activation product nitrogen-16 is predominant activity in the tacLcr coolant pumps, steam generators, and reactor coolant piping. The N-16 activity in each of the components depends on the total transit time to the component. The derivation of N-16 activity is shown in Section 11.1.3.
12.2.1.1.3 Main Steam Supply Syrtem (This section shall be presented in the Applicant's SAR).
I n
l Amendment No. 1 12.2-1 February 20, 1981 L
~
12.2.1.1.4 Spent Fuel Handling and Transfer The spent fuel assemblies are the predominant lon', tern source of radiation in the containment after plant shutdown for refeeling.
A reactor operating time necessary to establish near-equilibirum fission product buildup for i
the reactor at rated power is used in determining the source strength. The I
initial fuel corrposition that produced the maximum decay source is used.
The spent fuel decay gama source is given in Table 12.2-5.
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e Amendment No. 1 12.2-la February 20, 1981
O h
THIS PAGE INTENTIONALLY BLANI' O
12.2.1.1.5 Processing Systems 12.2.1.1.5.1 Chemical and Volume Control System (CVCS).
The shielding design is based on the maximum expected activity in each component.
These sources are listed in Tables 12.2-6 through 12.2-10 and their bases are discussed below.
A.
Heat Exchangers (Table 12.2-6 and l?.2-7) 1.
Regenerative Heat Exchanger Letdown side volume is based on 6 gallons of water with reactor coolant specific activity.
Charging side volume is based on 42 gallons of water with volume control tank specific activity.
2.
Letdown Heat Exchanger Total tube volume is based on 69 gallons of water with reactor coolant specific activity.
3.
Seal Injection Heat Exchanger Total tube volume is based on 15 gallons of water with volume control tank specific activity.
B.
Ion Exchangers (Table 12.2-8) 1.
Purification Ion Exchanger Total curie inventory is based on a resin buildup of 1.2 effective years.
This ion exchanger is used for lithium removal and normal nurification of RCS letdown. When it is used for lithium remi" al it is on line an average of 58 days prior to placing it in service as a purification ion exchanger for 292 days.
All nuclides except Xe, Kr, Rb and Cs have a decontamination factor (DF) of 10 and efficiency of 90%, Xe and Kr have a DF of 1.0 and efficiency of 0%, Rb and Cs have a DF of 2.0, and efficiency af 50%.
2.
Preholdup Ion Exchanger Total curie inventory is based on resin buildup of 1.0 effective year (292 days).
All nuclides except Xe, Kr, Rb, Cs, have a decontamination factor (DF) of 10 and an efficiency of 90%, Rb and Cs have a DF of 100 and efficiency Of 99%, Xe and Kr have a decontamination factor of 1 and an efficiency of 0%.
Sourcesgrocessedbytheprehold-uplon-exchangerinclude:
1.1 x 10 gallons of letdown previously pr0 cessed through the purification Ion exchanger and purification filter, 200 gpa from the Reactor Drain Tank (RDT) and 50 gpd from the Equipment Drain Tank (EDT).
12.2-2
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3.
Bnric Acid Condensate Ion Exchanger
}
Total curie inventory is based on resin buildup of 1.0 effective year (292 days). Anton decontamination factors of 10, and efficien-cy of 90% were used.
All other ions have a decontamination factor of 1 and an efficiency of OL Total liquid processed is 5
1.83 x 10 gallons.
t C.
Filters (Table 12.2-9)
{
Total curie inventories on all CVCS filters are based on crud buildup i
l of 292 days. All CVCS filters remove crud with a decontamination factor of 10 and an efficiency of 90L 4
-0.
Tanks (Table 12.2-10) 7 1.
Reactor Drain Tank (RDT) i i
The total curie inventory in the RDT is based on a water volume 3
of 2565 gallons and an equivalent vapor volume of 217 ft.
The tank vapor-liquid phases are in equilibrium and the tank liquid i
activity fraction is 1.0 of the RCS.
l 2.
Equipment Drain Tank (EDT) i The total curie inventory in the EDT is based on a water volume 3
(
of 5102 gallons and an equivalent vapor volume of 895 ft.
The i
tank vapor-liquid phases are in equilibrium and the tank liquid activity fraction is 0.1 of the RCS.
3.
Volune Control Tank (VCT) e i
The total curie inventory in the VCT is based on the average 2-water volume in the tank of 3170 gallons of RCS letdown and an 3
effective vapor volume of 292 ft.
Gas stripping was considered and the VCT vapor gas is in equilibrium with the liquid.
4.
Hold-up Water Tank (HT) l The total curie inventory in the HT is based on the average
. volume of 246,930 gallens.
Gas stripping was considered and the tank vapor-liquid pha',es are assumed not in equilibrium.
Activity l
4 the tank is based on holdup of 200 gpd from the RDT, 50 gpd from the EDT and 3,767 gpd from RCS letdown.
5.
Reactor Make-up Water Tank (RMWT) i j
The total curie inventory in the RMWT is based on a water volup l
of 492,954 gallons.
Activity in the tank is based on 9.9 x 10
.ga11cas processed by the' Boric Acid Concentrator.
i.
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12.2-3 l>
t,
.,.,_...,_.-.m.
.... _.. _,... _. -, _ - _,, ~., _.. _,. -. - -, _.. _,. - - -..,,-
6.
Refueling Water Tank (RWT)
The total curie inventory in the RWT is based on a water volump, of 697,818 qallons. Activity in the tank is based on 1.8 x 10 gallons processed by the Boric Acid Concentrator.
12.2.1.2 Safety Equipment Building 12.c.1.2.1 Shutdown Cooling System The pumps, heat exchangers, and associated piping of the shutdown cooling system (SDCS) are potential carriers of radioactive materials.
For plant shutdown, the SDCS pumps and heat exchanger sources of radioactivity result from the radioactive isotopes carried in the reactor coolant, discussed in Section 12.2.1.1.2, after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of decay following shutdown and dilution.
Table 12.2-11 Provides a listing of the maximum specific source strengths 1
(MeV/gm-sec) in the SDCS.
12.2.1.2.2 Component Cooling Water Syr' ?m This section shall be presented in ti.e Applicant's SAR.
12.2.1.3 Fuel Building 12.2.1.3.1 Spent Fuel Storage and Transfer lhe predominant radioactivity sources in the spent fuel storaae and transfer areas in the fuel building are the spent fuel assemblies.
Spent Fuel assembly sources are discussed in Section 12.2.1.1.4.
The spent fuel decay gamma source to be used in shielding design is given in Table 12.2-5.
12.2.1.3.2 Spent Fuel Pool Cooling and Cleanup System This section shall be presented in the Applicant's SAR 12.2.1.4 Turbine Building This section shall be presented in the Applicant's SAR 12.2.1.5 Auxiliary Building This section shall be presented in the Applicant's SAR.
O Amendment No. I 12.2-4 February 20, 1981
l i
TABLE 12.2-10 (Cont'd. ) (Sheet 2 of 2)
CVCS TANK INVENTORIES Maxi num Values
(,uries) k Reactor Reactor Equipment Volume Makeup Refueling Nuclide Orain Orain Control Hol6ca Water Water CR-51 1.6(-02)*
1.8(-03) 1.9(-04) 7.l(-04) 1.5(-06) 1.8(-02) l MN-54 3.0(-03) 5.3(-04) 3.1(-05) 1.3(-04) 6.2(-07) 3.7(-03) i FE-55 1.5(-02) 2.9(-03) 1.6(-04) 6.6(-04) 3.5(-06) 2.0(-02)
FE-59 9.0(-03) 1.1(-03) 1.0(-04) 3.9(-04) 1.1(-06) 9.9(-03)
CO-58 1.5(-01) 2.1(-02) 1.6(-03) 6.3(-03) 2.2(-05) 1.7(-01) l C0-60 1.9(-02) 3.7(-03) 2.0(-04) 8.3(-04) 4.5(-06) 2.6(-02)
\\
- Number in parentheses denote powers of ten.
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i TABLE 12.2-11 I
i SHUTD0' ail C00LIf4G SYSTEM (SDCS) SPECIFIC SOURCE STRENGTHS r
Maxinum Values (MeV/ gram-sec)
Y l
Energy (MeV)
(hr) 0.3 0.63 1.10 1.55 1.99 2.38 2.75 3.25 3.70 l
1 3.3(+4)* 2.4(+5) 6.7(+4) 1.9(+4) 4.7(+3) 3.4(+2) 1.6(+2) 9.9(+1) 1.2(+2) l l
l 10 2.5(+4) 1.2(+5) 2.9(+4) 7.5(+3) 2.2(+3) 2.9(+1) 6.7(-1) 6.2(-1) 8.9(-3) 100 1.8(+4) 4.4(+4) 6.3(+3) 2.4(+3) 3.5(+2) 2.2(+1) 2.7(-2) 8.7(-3) i l
l
- nurter in parentheses denotes power of ten l
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Anendnent tio. 1
{
February 20, 1981 l
1
12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES 12.3.1.1 Radiation Zone Designation See Applicants SAR 12.3.1.2 Equipment and System Design Features for Control of On-Site Exposure Following are some of the specific design features which are used to assure that occupational radiation exposure due to operations and maintenance of the System 80 NSSS will be ALARA.
Demineralizers are addressed under the heading of ion exchangers.
Sample stations are addressed in the applicant's 1
FSAR.
A.
famps 1.
Most pumps and associated piping are flanged to facilitate ease of removal to a low radiation area for maintenance or repair.
Pump internals can be removed to a low radiation area for maintenance.
V) t 2.
All pump casings are provided with drain connections to facilitate decontamination.
B.
Ion Exchangers 1.
Ion exchangers are designed for complete drainage.
2.
Spent resin removal is designed to be done remotely by hydrauli-cally flushing the resin from the vessel to the Solid Waste Management System.
3.
The fresh resin inlet is designed to extend into a low radiation area above the shielded compartment housing the ion exchanger.
4.
Ion exchangers are designed with a minimum M crevices in order not to accumulate radioactive crud.
l C.
Liquid Filters 1.
Filter housings are provided with vent connections and designed for complete drainage.
2.
Filter housings are designed with a minimum of crevices in order i
not to accumulate radioactive crud.
/
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(#)
3.
Filter housings and cartridges are designed to permit remote l
l removal of the filter elements.
Amendment No. 1 12.3-1 February 20, 1981
D.
Tanks 1.
Tanks are designed to be isolated for maintenance and provisions will be made for complete drainage.
2.
Tanks are provided with at least one of the following means of cleaning the tank internals for decontamination purposes:
a.
Ample space is provided to facilitate cleaning from the tank manway.
b.
Internal spray nozzles are provided on potentially highly contaminated tanks for internal decontamination.
c.
The ability to back flush or drain inlet screens hydraulically will be provided (on tanks or vessels with these screens) to facilitate decc7tamination.
3.
All tanks are vented to either the gas collection header or the gas surge header which will facilitate removal of potentially radioactive gases during maintenance.
4.
Non pressurized tanks are provided with overflows, routed to a floor drain or other suitab1ve collection point to avoid radioactive fluids spilling to the floor or ground.
5.
Tanks are designed with a minimum of crevices in order not to accumulate radioactive crud.
E.
Package Units 1.
Each package unit is skid mounted with all motors and pumps located on the periphery of the skid for free access and for quick removal to a low radiation area for maintenance or repair.
2.
Space is provided on the skid for placement of portable shielding.
3.
All package components are provided with provisions for flushing, drawing and chemical cleaning.
4.
Heat exchangers are readily accessible for maintenance.
5.
Controls are remotely mounted and the package will be able to be remotely monitored.
As many control elements as possible are mounted remotely from the components.
6.
Components are designed with a minimum of crevices in order not to accumulate radioactive crud.
7.
Radioactive gas is collected and sent to the Gaseous Waste Manage-ment System.
O 12.3-2
l l
F.
Valves 1.
Radiation resistant seals, gaskets and elastomers are employed when practical to extend the design life and reduce maintenance requirements.
2.
Power operated valves in the primary system are provided with double packing, a lantern gland and stem leakoffs to collect leakage and to direct radioactive fluid away from access areas.
All valve packing glands have provisions to adjust packing com-pression to reduce leakage.
3.
Valves are designed so that they may be repacked without removing the yoke or topworks.
4.
Remotely operated valves are utilized where practical and necessary.
5.
Valve wetted parts are made of austenitic stainless steel or other corrosion resistant material.
6.
Low leakage valves with backseats are employed wherever possible.
Packless diaphragm valves are employed in highly contaminated systems.
G.
Piping 1
)
See Applicant's SAR for specific design considerations.
O Even though balance of plant piping is not within CESSAR's scope, guidelines provided to the Applicant contain design recommendations and information for keeping in plant personnel exposures ALARA.
The following information and. recommendations are provided to the Applicant:
1.
Interface criteria and radiation source terms are provided to insure that field-run piping carrying radioactive material is either run in shielded pipe chases or within shielded cubicles.
2.
Whenever possible, pipe runs should be sloped to prevent accumulation and to assist in the removal of radioactive corrosion deposits.
3.
The number of elbows, tees, deadlegs, etc., should be minimized to reduce corrosion deposits. Where elbows are required, they should be a large radius type for minimization of deposits.
H.
Heat Exchangers 1.
Heat exchangers are designed to accommodate the requirements of inservice inspection and for ease of access during maintenance to reduce the time operators are required to spend in a radiation environment.
2.
' Materials are selected to minimize the need for replacement and
\\
to reduce maintenance -frequencies corrosion resistant materials are employed.
Amendnent No.1 12.3-3 February _20,1981
I.
Material Selection 1
Material is selected as described below to reduce exposures by reducing maintenance frequencies and by providing less circulating crud as a source of exposure where maintenance will be necessary.
1.
Materials of construction for components containing radioactive materials will be selected with consideration of potential release of activated corrosion products from these materials.
2.
Radiation exposure levels were considered when selecting materials for 40 year service.
3.
Material selection was made with consideration given to other fluid conditions which could lead to premature material failure.
4.
Other materials considerations are discussed in Section 5.2.3.
J.
Reactor Vessel Head Vent A vent nozzle and line is provided on the reactor vessel head.
Utiliza-tion of this design feature will allow a reduction of exposure during the head removal process by minimizing the gases discharged directly to the containment atmosphere while the head is being removed.
K.
Reactor Coolant System Leskage Control 1
Exposures from airborne radionuclides to personnel entering the contain-ment will be minimized by controlling the amount of reactor coolant leakage to the containment atmosphere.
Examples of such controlled leakage are listed below:
1.
Primary pressurizer safety valve leakage is directed to the Reactor Drain tank, as discussed in Section 5.2.2.
2.
Valves larger than 2" in diameter are provided with a double-packed stem with an intermediate lantern ring with a leak-off connection to the Reactor Drain Tank.
3.
Instrumentation is provided to detect abnormal reactor coolant pump seal leakage.
The reactor coolant pumps are equipped with two stages of seals plus a vapor or 'oackup seal as described in Section 5.5.
The vapor or backup seal will prevent leakage to the containment atmosphere and allow sufficient pressure to be maintained to direct the controlled seal leakage to the Volume Control and Reactor Drain Tanks.
The vapor seal is designed to withstand full Reactor Coolant System pressure in the event of failure of any or all of the two primary seals.
L.
Refueling Equipment 1
1.
All spent fuel transfer and storage operations are designed to be conducted underwater to insure adequate shielding and to limit the maximum continuous radiation levels in working areas.
Amendment No. I 12.3-4 February 20, 1981
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2.
The equipment is designed to prevent the fuel from being lif ted s./
above the minimum safe water depth, thereby limiting personnel exposures and avoiding fuel damage.
3.
The equipment design limits the possiblity of inadvertent fuel drops which could cause fuel damage and personnel exposures.
4.
The refueling equipment design will facilitate the transfer of new and spent fuel at the same time to reduce overall fuel handling time; and, therefore, personnel exposures during refueling.
5.
Underwater cameras are used to facilitate safe handling and visual control, thus minimizing errors and potential exposures.
6.
Portable hydraulic cutters are provided to cut expended Control Element Assemblies and in-core instrumentation leads.
The cutters allow underwater handling of these items.
7.
Equipment is provided to allow for the underwater determination of leaking fuel elements.
M.
Inservice Inspection Equipment 1
Inspection of the reactor coolant pressure boundary can ce done with remote equipment to keep personnel exposures to a minimum.
A detailed
/~N discussion of the Inservice Inspection Program is provided in the Applicant's SAR.
N.
Remote Instrumentation 1
All systems containing radioactive fluids are designed to be controlled remotely to the maximum extent practical.
This will allow personnel radiation exposures from the normal operation o*f these systems to be minimized.
O.
Inservice Inspection of Reactor Vessel Nozzle Welds The design of welds joining the reactor vessel nozzle to reactor coolant pipe permits inservice inspection to be accomplished from the I.D. of the reactor vessel.
Automated equipment normally used for i
reactor vessel pressure boundary inspections can be utilized in this area.
In the event that inservice inspection of this area is performed from the outside, insulation for the reactor vessel and reactor coolant piping utilizes removable sections for access.
These removable sections are lightweight and are held in place mainly by quick actuation type buckle fasteners.
After-the necessary panels are removed, remote equipment can be utilized to perform the required inspections.
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Amendment No. I 12.3-5 February 20, 1981
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- V CHAPTER 15 APPENDIX 15D Section Subject Page No.
15D.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 150.1-1 15D.l.3 Limiting Fault 1 Events 150.1-1 150.1.3.1 Increased Main Steam Flow Through Turbine Bypass 15D.1-1 15D.1.3.1.1 Identification of Event and Causes 150.1-1 15D.1. 3.1. 2 Sequence of Events and Systems Operation 15D.1-1 15D.l.3.1.3 Analysis of Effects and Consequences 15D.1-1 15D.l.3.1.4 Conclusion 150.1-2 15D.l.3.2 Decrease in Main Feedwater Temperature 15D.1-3 If-)
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15D.1.3.2.1 Identification of Event and Causes 150.1-3 s
150.1.3.2.2 Sequence of Events and Systems Operation 150.1-3 15D.l.3.2.3 Analysis of Effects and Consequences 15D.1-3 150.1.3.2.4 Conclusion 15D.1-4 15D.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15D.2-1 l
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150.2.2 Infrequent Events 15D.2-1 Loss of Offsite Power 15D.2-1 150.2.2.1 l
150.2.2.1.1 Identification of Event and Causes 15D.2-1 15D.2.2.1.2 Sequence of Events and System Operation 150.2-1 15D.2.2.1.3 Analysis of Effects and Consequences 15D.2-1 15D.2.2.1.4 Conclusions 15D.2-2
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TABLE OF CONTENTS (Cont.)
CHAPTER 15 APPENDIX 15D Secti-a Subject Page No.
15D.2.2.2 Loss of Instrument Air 150.2-3 150.2.2.3 Total Loss of Normal Feedwater Flow 150.2-3 15D.2.2.3.1 Identification of Event and Causes 150.2-3 150.2.2.3.2 Sequence of Events and Systems Operation 15D.2-3 150.2.2.3.3 Analysis of Effects and Consequences 150.2-3 150.2.2.3.4 Conclusions 15D.2-4 15D.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15D.4-1 15D.4.2 Infrequent Events 15D.4-1 150.4.2.1 Inadvertent Deboration 15D.4-1 150.4.2.1.1 Identification of Event and Causes 15D.4-1 150.4.2.1.2 Sequence of Events and Systems Operation 15D.4-1 15D.4.2.1.3 Analysis of Effects and Consequences 150.4-150.4.2.1.4 Conclusions 150.4-3 150.6 DECREASE IN REACTOR COOLANT INVENTORY 15D.6-1 15D.6.5 Limiting Fault 3 Events 15D.6-1 15D.6.5.1 Steam Generator Tube Rupture with Loss of Offsite Power on Turbine Trip 150.6-1 1
15D.6.5.1.1 Identification of Event and Causes 150.6-1 15D.6.5.1.2 Sequence of Events and System Operation 15D.6-1 15D.6.5.1.3 Analysis of Effects and Consequences 150.6-1 150.6.5.1.4 Conclusions 150.6-2 Amendment No. 1 ii February 20, 1981
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15D.8.2.1 Startup of an Inactive Reactor Coolant Pump 150.8-1 l
i 15D.8.'.1.1 Identification of Event and Causes 15D.8-1 t
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i 15D.8.2.1.4 Conclusions 15D.8-3 i
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g 150.6 DECREASE IN REACTOR COOLANT INVENTORY 150.6.5 LIMITING FAULf 3 EVENTS 150.6.5.1 Steam Generator Tube Rupture With Loss of Offsite Power on Turbine Trip 150.6.5.1.1 Identification of Event and Causes The Steam Generator Tube Rupture (SGTR) with Loss of Offsite Power (LOP) on turbine trip is in the Limiting Fault 3 (LF-3) frequency category of the decrease in reactor coolant inventory category.
With respect to radiological releases, no event resulting in the decrease in reactor coolant inventory category is as severe as the Double-Ended Break of the Letdown Line Outside Containment Upstream of the Letdown Line Control Valve (DBLLOCUS) descr.ibed in Section 15.6.5.2.
The limiting steam generator tube rupture event was determined to result from the flow associated with a double er.ded rupture of a steam generator U-tube at full power conditions causing reactor coolant to leak into the secondary system.
The SGTR is in the Loss of Primary System Fluid to the Secondary System event group and is described in Section 15.6.2.
LOP may occur due to a failure of-the grid system to accommodate a turbine-generator trip and results in a loss of normal feedwater flow, forced
-reactor coolant flow, and condenser availability.
15D.6.5.1.2 Sequence of Events and System Operation The sequence of events for SGTR with' LOP on turbine trip is initially 1
similar to SGTR which is described in Section 15.6.2.2.2.
At the time of turbine trip, LOP occurs causing coastdown of the reactor coolant pumps (RCP), loss of the main feedwater pumps, and unavailability of the condenser.
The loss of offsite power is followed by automatic startup of the standby diese17 generators, the power output of which is sufficient to supply elec-trical power to all necessary engineered safety features systems and to provide the capability.for maintaining.the plant in a safe shutdown condition.
Subsequent to the reactor trip, stored and fission product decay energy must be dissipated by.the reactor coolant system and main steam system.
In the absence of forced reactor coolant flow, convective heat transfer into and out of the RCS is supported by natural circulation of the reactor coolant.
Initially, the residual water inventory in the steam generators is used as a heat sink, and the resultant steam is-released to atmosphere by the main steam safety ~ valves. With the availability of standby diesel power, emergency feedwater is automatically initiated on a low steam gen-erator water level signal.
Plant cooldown is. operator controlled via the atmospheric steam dump' valves until offsite power is restored, at which time.the steam bypass control system and the condenser are utilized for the remainder of the cooldown.
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Amendment No. 1 150.6-1 February 20, 1981
l 15D.6.5.1.3 Analysis of Effects and Consequences During the SGTR with LOP, primary coolant is leaked into the damaged steam generator.
Only safety valves are utilized for steam release to atmosp:.ere until 1800 seconds.
At this time, the operator identifies and iso!ates the damaged steam generator by closing the main steam isolation valves.
Using the intact steam generator, the operator then initiates an orderly cooldown via the atmospheric dump valves and, if offsite power becomes available, through the steam bypass system.
This event results in a smaller primary coolant release into the damaged steam generator before the main steam safety valves close than the release to outside containment reported for the DBLLOCUS, Section 15.6.5.2.
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combined with the favorable decontamination factor associated with releases from the steam generator (i.e., DF=10) versus releases from the auxiliary building (i.e., DF=2) during DBLLOCUS results in radiological releases for SGTR with LOP which are less than those for DBLLOCUS.
l 15D.6.5.1.4 Conclusions The Steam Generator Tube Rupture With Loss of Offsite Power is less severe than the Letdown Line Break Outside Containment Upstream of the Letdown Control Valve presented in Section 15.6.5.2 with respect to radiological releases.
Therefore, it is within the 10CFR100 guidelines and meets the acceptance guidelines presented in Table 15.0-3 for LF-3 events.
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- I Regulatory Guide 1.130, Revision 0 DESIGN LIMITS AND LOADING COMBINATIONS FOR l9 CLASS 1 PLATE AND SHELL-TYPE COMPONENTS SUPPORTS I
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SUMMARY
The Regulatory Guide addresses design limits and loading combinations for i
Class 1 plate and shell type component supports.
POSITION i
Most Class 1 supports designed by C-E are of the linear type and, therefore, l
are not affected by this guide.
The only exception is the steam generator l
sliding base support which is subjected to biaxial stress fields.
The current design conforms to Regulatory Guide 1.130.
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A-33 r
Regulatory Guide 8.8, Revision 2 IfFORMATION RELEVANT TO ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AT NUCLEAR POWER STATIONS WILL BE AS LOW AS IS REASONABLY ACHIEVABLE (NUCLEAR POWER REACTORS)
SUMMARY
The purpose of this Regulatory Guide is to provide guidance to be used in the design process so as to maintain occupational radiation exposures "As Low As Reascnably Achievable" (ALARA).
Recommendations for a utility ALARA organization structure, as well as a specific list of design considerations are included. General recommendations for cost / benefit analysis are also provided.
POSITION It is C-E's position that System 80, as designed, is in compliance with the intent of Regulatory Guide 8.8.
In several presentations, C-E outlined those features which have been included in the System 80 design so as to minimize exposures received by plant personnel.
The following specific commitments are provided to support this position:
a)
Primary water chemistry controls have been recognized as indispensible in maintinaing the System 80 NSSS occupational radiation exposures ALARA.
Corrosion control begins in the early stages of construction when the chemistry effort includes cleanliness practices, quality control in material procurement, and flushing for the removal of foreign contaminants.
During the conmissioning period, which corres-ponds to the pre-core through low power operation, corrosion control is enhanced through the development of a passive oxide film on the system surface prior to critical operation and the protection of the system from localized corrosion induced by oxygen or halides. The passive oxide film serves to slow the corrosion reaction to steady I
state values. Hydrazine is added to the primary coolant as the inhibitor against halide-induced attack at low temperatures and as an oxygen scavenger. Hydrogen is added af ter criticality when there is a gamma radiation flux present to enhance its reaction with oxygen.
Corrosion products produced within the reactor primary coolant system deposit on the reactor core heat transfer surfaces.
In order to minimize the effect of crud deposition on these surfaces lithium hydroxide is recommended as a base additive in the primary coolant. This additive reduces the solubility of corrosion products so that less will be in circulation in the reactor coolant system.
The additive further reduces corrosion products in the coolant by increasing the coolant conductivity and enhancing agglomeration of colloidal particles. These colloidal particles are removed by the purification system.
O Amendment No. 1 A-34 February 20, 1981
Specifically the water chemistry parameters controlled throughout plant p
operation include oxygen to less than 0.1 ppm, chlorides and fluorides to less than 0.15 ppm. The ranges for pH, hydrazine, lithium, and hydrogen will vary as appropriate for a given plant condition.
b)
The cobalt content of the materials employed in fabrication of the wetted surfaces of the reactor coolant system is controlled at a low level to insure that occupational radiation exposure is main-tained ALARA.
Specifically the specified cobalt content of these materials is less than 0.2 maximum weight percent with the exception of the small quantity of material which is required to be hardfaced.
Metallurgical testing is psrformed during the procurement and fabri-cation stages of construction to ensure that this limit is maintained.
c)
Systems and equipment employed in the System 80 NSSS, with surfaces that are wetted by primary coolant, have been designed with a minimum use of stellite to ensure that occupational radiation exposure will be ALARA.
Specifically the use of stellite has been minimized to those wetted surfaces which could be subjected to a high level of wear or erosion. These surfaces include valve seats, reactor coolant pump seals, reactor vessel internal hardfacing, and control element drive mechanism pins and latches. Stellite has been utilized for these surfaces only after design evaluations eliminated the use of alternative materials.
d)
Systems and equipment employed in the System 80 NSSS have been designed A
with the objective of mini 'izing the release of potentially activated
('V) material to the reactor cc lant by minimizing the erosion of surface I
metals.
Favorable geometrics and fluid flow configuration are employed to eliminate vibration which may cause erosion or wear of surface materials. Materials are appropriately selected to function for the design life of the plant in the required environment with minimal erosion.
Those surfaces which may be subjected to a high level of erosion are hard-faced.
e)
The Chemical 'and Volume Corarol System functions to remove activated corrosion products from the reactor coolant system during operation.
A description of this system and a definition of its functional require-ments are provided in Section 9.3.4.
f)
Several design features have been incorporated into the System 80 NSSS which provide for more efficient inspection of primary system welds.
The most significant improvement for performing inservice inspection is' the reduction of linear feet of weld in the major components. The
-reduction in weld footage has been accomplished by component redesign, use of forged sections versus forged-welded plate sections, and in-creasing the size of certain sections.
Inspection is facilitated by eliminating the use of safe end welds on reactor vessel nozzles.
Locating the welds further away from the nozzle contours reduces the inspection time in the high radiation area.
Inspection of the pressure boundary is performed with remote equipment to further redace occupational exposure.
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.,n) bu) t Amendment No. 1 A-34a February 20, 1981'
g)
Capabilities for the use of a multistud tens.oning/detensioning I
device to minimize the time, and censequer.tly the exposure, associated with tightening / loosening the reactor vessel closure bolts.
l In addition, C-E in cooperation with some of its utility customers, has an on-going program to reduce radiation exposures received by operating personnel.
l Data obtained from this program has been and will continue to be utilized to improve design and operations from an ALARA standpoint.
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knendment No. 1 A-34b February 20, 1981 1
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