ML19340D566
| ML19340D566 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/30/1980 |
| From: | Ehrensperger W GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19340D565 | List: |
| References | |
| NUDOCS 8012310298 | |
| Download: ML19340D566 (45) | |
Text
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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION i NRC Docket Nos. 50-424, 50-425 .
i j In the Matter of GEORGIA POWER COMPANY l
SUPPLEMENT 8 TO l
APPLICATION FOR LICENSE 1
l UNDER THE ATOMIC ENERGY ACT OF 1954 AS AMENDED
- FOR ALVIN W. V0GTLE NUCLEAR PLANT UNITS 1, 2 t
i c i The Applicant, Georgia Power Company, hereby supplements its Application for a Construc-tion Permit and Operating License, originally submitted on August 1,1972, by the addition of supplementary material attached hereto.
i.
By:
hb W. E. Ehrensperger Senior Vice President Power Supply l Georgia Power Company i
Sworn to and subscribed before me, this d[] day of December,1980.
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Notary Public Notary Pub % Georca. State at Large
. My Comm:$Ston Opir85 Seot 20.1984 801z31o g
Ge:rgia Power Comeany Post Of* ice Bon 4545 230 Peachtree Street, N W Atlanta. Georg.a 30302 Telephone 404 522-6060 O $,';,'v$'c'e"UsTe'ni December 30, 1980 Power Sacply O
Director cf Nuclear Reactor Regulation Attn: Darrell C. Eisenhut, Director O Division of Project Managenent U.S. Nuclear Regulatory Comissica Washington, D.C. 20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 ALVIN W. V0GTLE NUCLEAR PLANT-UNITS 1 AND 2 SUPPLEMENT 8 TO APPLICATION
Dear Mr. Eisenhut:
Georgia Power Company hereby files three (3) signed and sixty (60) conformed copies of Supplement 8 to its application for a Construction O, Permit and Operating License for the Alvin W. Vogtle Nuclear Plant, Units 1 and 2.
This supplement consists primarily of revised analysis of the radio-logical consequences of a postulated loss-of-coolant accident (Regulatory Guide 1.4 assumptions). The revised analysis is based on a more restrictive primary contairinent leakage commitment, which together with on-site meteorological data ensures more reliably attained offsite post-accident doses which are less than, or comparable to, doses judged to be acceptable for design now and at the time construction was initially authorized.
If you have any questions, please advise.
Yours truly, gQ,n W. E. Ehrensperpjer /
WEE /caa cc: J. A. Bailey O D. E. Dutton R. A. Thomas J. L. Vota B. L. Lex L. T. Gucwa G. F. Trowbridge, Esq.
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BEFORE THE l UNITED STATES NUCLEAR REGULATORY COMMISSION NRC Docket Nos. 50-424, 50-425 In the Matter of GEORGIA POWER COMPANY SUPPLEMENT 8 TO APPLICATION FOR LICENSE UNDER THE ATOMIC ENERGY ACT OF 1954 AS AMENDED FOR ALVIN W. V0GTLE NUCLEAR PLANT UNITS 1, 2 O The Applicant, Georgia Power Company, hereby supplements its Application for a Construc-tion Permit and Operating License, originally submitted on August 1,1972, by the, addition of supplement.ary material attached hereto.
l i
By:
t#E y W. E. Ehrenspp' ger Senior Vice President Power Supply Georgia ~ Power Company O
Sworn to and subscribed before me, this d O day of December,1980.
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No tary Publ i c % Putsc, Georps, same at Largt-O W DWWS Sept. 20.1983
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TABLE OF CONTENTS (Continued)
Section Title, Pu r e 2.
5.6 REFERENCES
2.5-52 7
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2.5.7 BIBLIOGRAPHY 2.5-53 g
r 2.5.7.1 Seismolocy 2.5-53 2.5-55 F. ^% 2.5.7.2 Geolocy I E r
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TABLE OF CONTENTS (Continued) )
LIST OF TABLES Table Title Page 2.1-1 Commercial and Sport Fish of the Savannah River 2.1-7 2.2-1 Location and Types of Major Industries Within 25-Mile Radius of VNP 2.2-5 2.3-1 Precipitation Data for Augusta Area, )
1902 - 1950 2.3-3 2.3-2 Precipitation Data for Augusta Area, 1940 - 1950 2.3-4 2.3-3 SRP Tower Weather Sensors Used for This Report 2.3-9 2.3-4 Normals, Means, and Extremes 2.3-11 1
2.3-5 Vogtle Onsite Weather Instruments 2.3-15 2.3-6A Atmospheric Diffusion Estimates for use in j Accident Evaluation using 2 years of SRL Data 2.3-17a S8 2.3-6 Summary of Accident X/O Values (sec/m3 )
Based on 3 Years Vogtle Site Data S8 2.3-17 i 2.3-7 Joint ..aquency Tables of Wind Speed and Direction by Stability Group (SRL Data) 2.3-18 2.3-7A Joint Frequency of Wind Speed and Direction by Stability Group SRL Tower (3/66-3/67) 2.3-21a 2.3-7B Joint Frequency of Wind' Speed and Direction by Stability Group SRL Tower (3/67-3/68) 2.3-21d )
2.3-8 Temperature Difference Groups for Determining Pasquill Stability Categories 2.3-24 2.4-1 Gaging Station Records, Savannah River Basin Savannah River at Augusta, Georgia, Annual b 2.4-2 P Flood Peaks 2.4-1A Makeup Water Requirements Per Unit 2.4-14a o
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SUMMARY
OF ACCIDENT X/Q VALUES (sec/m ) 3
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@ Averaging : Boundary Area Area Low Low Time Period 0.5% Boundary Boundary Population Population j
$ After Probable 5% Probable 50% Probable Zone Zone ;
g Accident Direction- Direction- Direction- 0.5%* 50%** Worst j H
e (hours) Dependent Independent Independent Probable Probable Condition *** j y 1 1.8E-4 1.5E-4 2.8E-5 7.2E-5 1.0E-5 4.7E-4 j m
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- Based on log-log interpolation. 5% probable direction independent approach gives j p about the same results. '
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- Based on running-mean calculations over appropriate averaging periods using ,
8 direction-independent approach. j U :
$ *** Values shown are based on running-mean calculations and are the highest calculated ?
Q during any of the three 1-year periods of record. The fact that values are lower 8 than reported for the 0.5% and 5% probable cases for certain averaging periods is y due to differences in methodology introduced by the log-log interpolation scheme
,, of Regulatory Guide 1.145.
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VNP TABLE OF CONTENTS (Continued)
( 'k Section Title Page 3.2.2.1 Instrument Quality Group Classification 3.2-9 3.3 WIND AND TORNADO LOADINGS 3.3-1 e
WIND LOADINGS
! 3.3.1 3.3-1 e
3.3.1.1 Design Wind Velocity 3.3-1 3.3.1.2 Basis for Wind Velocity Selection 3.3-1 3.3.1.3 Vertical Velocity Distribution and
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Gust Factors 3.3-1 t
t 3.3.1.4 Determination of Applied Forces 3.3-2 t- 3.3.2 TORNADO LOADINGS 3.3-2 i
3.3.2.1 Applicable Design Parameters 3.3-2 3.3.2.2 Determination of Forces on Structures 3.3-3 3.3.2.3 Ability of Category I Structures to Perform Despite Failure of Structures Not r sq Designed for Tornado Loads 3.3-3 i c.
3.3.2.4 Exceptions to General Tornado Criteria S8 3.3-3 lS8 3.3.2.4 Exceptions to General Tornado Criteria 3.3-3 3.3.2.5 References 3.3-5 3.4 WATER LEVEL (FLOOD) DE.3IGN CRITERIA 3.4-1 3.5 _
MISSILE PROTECTION .
3.5-1 3.5.1 MISSILE BARRIERS AND LOADINGS 3.5-5 y 3.5.1.1 Missile Barriers Within Containment 3.5-5 3.5.1.2 Barriers For Missil's Generated
~. Outside of Plant Structures 3.5-6 3.5.1.3 Missile Barriers Within Plant Structures Other than containment 3.5-6
% ! 3.5.2 MISSILE SELECTION 3.5-8 w
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VNP TABLE OF CONTENTS (Continued)
Section Title Page j 3.5.2.1 General 3.5-8 3.5.2.2 Rotating Component Failure Missiles 3.5-9 3.5.2.3 Pressurized Component Failure Missiles 3.5-21 J' 3.5.2.4 Tornado Generated Missiles 3.5-23 3.5.2.5 Site Proximity Missiles 3.5-27 3.5.3 SELECTED MISSILES 3.5-30 ,
3.5.3.1 Nuclear Steam Supply System Missiles 3.5-30 3.5.3.2 Missiles from System Other than NSSS 3.5-30 3.5.3.3 Tornado Missiles SS 3.5-32 l 3.5.3.4 Protection of Structures, Systems and Components Against External hiss 11es S5 3.5-32 3.5.3.3 Tornado Missiles 3.5-32 3.5.3.4 Protection of Structures, Systems and Components Against External Missiles 3.5-32 3.5.3.5 Plant Layout Considerations Relative to Probable Trajectory of Low Pressure Turbine Missiles 3.5-33 b 3.5.4 BARRIER DESIGN PROCEDURES 3.5-34 e
gE 3.5.5 MISSILE BARRIER FEATURES
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5.6 REFERENCES
3.5-35 h
I 3.6 PROTECTION AGAINST DYNAMIC EFFECTS k ASSOCIATED WITH THE POSTULATED RUPTURE
$ OF PIPING 3.6-1 1
s 3.6.1 SYSTEMS IN WHICH DESIGN BASIS PIPING BREAKS l ih 3.6-la ARE POSTULATED TO OCCUR
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- Design Basis Piping Break Criteria for 5-fi . -
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VNP translational and rotational velocities. Tornado wind loads, and differential atmospheric pressure changes, with associated time intervals, are established and the supporting data furnished in the .above referenced sections.
The rationale used for assuming an average wind velocity of 300 mps is applicable to loadings on large structures. Associated with localized phenomena such as suction swaths, the wind velccity of 350 mph corresponds to loadings on small structures. See section 3 of Appendix 3M.
3.3.2.2 Determinhtion of Forces on Structures The load combinations and methods used to convert tornado loadings into forces and distribution across the structures are outlined in sections 3 and 4 of Appendix 3M. Loading combinations are stated in section 3.8. A load factor of unity for a tornado is used as given in Section 3.8.
3.3.2.3 Ability of Category I Structures to Perform Despite Failure of Structures Not Designed for Tornado Loads All permanent category II structures, systems and components and all category I structures, systems and components not ~
designed for tornado loadings are designed so that no missiles will be generated in a tornado event that will have more severe effects than those referred to in section 3.5. Each structural part or component is analyzed for design integrity and is determined mathematically to meet the above criteria. This is to assure that category I structures, systems, and components required for safe shutdown after a tornado will perform their intended functions.
\
l t Non-category I structures whose collapse could result in loss of required function category I structures, equipment, or systems required for safe shutdown after a tornado will be analytically checked to determine that they will not collapse when subjected to extreme environmental loads. The bases for analytical procedures that will be used are discussed in paragraph 3.8.4.4.
3.1.2.4 Exceptions to General Tornado Criteria 7
i Category II structures are not designed to the tornado criteria in paragraph 3.3.2.1.
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( $ 3.7.2.1.1.5 Description of Mathematical Medels. The mathematical models used for the major strdctures are described in sequence. Furthermore, the dynamic testing procedure used by the manufacturer in lieu of a dynamic analysis is described. The types of mathematical models used for plant i structures, systems and components are listed in table 3.7-5.
-f A. Model for Containment, Interior Structure, and Enclosure Building The seismic analysis of the containment building and interior structures, including soil-structure 1 interaction, is described in paragraph 3.7.2.5.
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- The lumped parameter model of the containment and the interior structures used in this analysis includes the inertial effects of the enclosure building. The stiffness characteristics of the enclosure building are not modeled because they are insignificant compared to the stiffness of the containment building.
The effect of deletion of the enclosure building above el. 270' on the lumped parameter model of the containment building and interior stuctures is S8 aslight reduction in the lumped mass values at
- certain nodes, which does not significantly affect
} the frequency characteristics of the model, thus with no significant impact on the acceleration values experienced by the containment and interior structures.
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The resulting effect, if any, of the deletion of the enclosure building on the containment and internal structures would be a slight reduction in i the seismic forces due to the slight reduction in mass values. l l
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B. Models for the Auxiliary, Control and Fuel Handling Buildings The seismic analysis for the auxiliary, control, and fuel handling buildings is included in the finite-element soil-structure seismic analysis, as described in paragraph 3.7.2.5.
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( ) 3.8.3.6.5 Construction Procedure The construction procedures are the same as described in paragraph 3.8.1.6.6.
'g 3.8.3.6.6 Quality Control 4
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The quality control requirements will be met as described in paragraph 3.8.l.6.7 and Chapter 17.
3.8.3.7 Testing and Inrervice surveillance Requiraments *
, 'y A formal program of testing and inservice surveillance is not planned for the internal structures. The internal structures are not directly related to the functioning of the containment concept, hence, no testing or surveillance is required.
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VNP 3.8.4 OTHER CATECORY I STRUCTURES y Other Category I structures include equipment building, auxiliary building, control building, fuel handling building, diesel generator building, auxiliary feedwater pump building, nuclear service cooling tower, emergency cooling water wells, condensate storage water tank, reactor makeup storage tank, h refueling water storage tank, diesel oil fuel storage tank, pipe and electrical cable tunnels and electrical duct banks.
. 3.S.4.1 Description of the Structure
/
3.S.4.1.1 Equipment Building )
The equipment building (EB) is a group of five steel-framed -
structures with uninsulated metal siding and roof deck. The EB completely surrounds the containment from grade to 270 foot level. There is no safety function associated with the EB except that its failure shall not damage any safety-related
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equipment. The approximate center line to center line column dimensions of the EB are: length, 176 feet, width, 176 feet, and height, 50 feet above grade.
With the removal of the enclosure building, the plant Vogtle J containment is essentially similar to other prestressed, post-
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tensioned, reinforced concrete containments. To incorporate the original enclosure building, reinforcing steel patterns in the i containment shell were modified slightly to accept embedments for the supporting structural steel; their pattern will not be altered after removal of the enclosure building since their S 8 k, affect on containment integrity is negligible.
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& Moreover, the removal of the upper enclosure building enhances r containment building integrity since the overall load of the enclosure building is no longer transmitted to the shell and concentrated loads occurring at the concrete to structural steel interface are eliminated. i
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compound. A thick membrane is placed on the soil structure
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upon which the basin mat is poured. The waterproofing material is used te prevent any intrusion of ground water through the basin shell and mat. The interior surfaces are also waterproofed using a thermoplastic membrane cast in the concrete. The cooling tower and basin are embedded to a h considerable depth. This embedment gives rise to additional a" resistance to overturning from lateral soil pressure. The soil structure interaction is accomplished by the LUSH program.
- 3.8.5.1.10 Equipment Building I The equipment building is supported on the adjacent buildings 3
p' except for the equipment hatch area which is supported at grade level by means cf a conventional slab tad grade beam founda-tion system or a mat foundation.
F The removal of the upper porti_on of the enclosure' building, s and its attendant structural steel and metal siding, results S8
> in approximately 2% lower overall containment building weight L and a -slight reduction in net soil bearing pressures.
b 3.8.1.1.11 Pipe and Electrical Cable Tunnels and Electrical Duct Banks lY The design for the seismic Category I pipe and electrical cable
-' tunnels and electrical duct banks is presently in development p stage as stated in paragraph 3.8.4.4.8. For preliminary layout
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of the underground systems, refer to figures 3.5-1 and 3.5-2.
The discussion as pertains to buried structures in this paragraph will be covered in the FSAR.
3.8.5.2 Applicable Codes, Standards and Specifications Refer to paragraph 3.8.1.2 for the containment and to paragraph
- 3.8.3.2 for other Category I Structures.
- 3.8.5.3 Loads and Loading Conditions g Containment foundation loads and loading combinations are
.T. - !. i discussed paragraphs 3.8.1.3 and 3.8.3.3.
Foundation loads and loading combinations for other Category I F'. structures are discussed in paragraph 3.8.4.3.
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- '/ % 23.8.5.4 Design and Analysis Procedures -
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Design and analysis procedures for the containment including F the base slab, are discussed in appendix 3R.
f.
The basic techniques for analyzing the foundations of Category K p- I structures are by the conventional methods, involving
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- POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980 U. .
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concrete structures practice. Stresses resulting from local moments, torques, and concentrated reactions, and uniform loading are computed by these methods. These methods are further discussed in paragraph 3.8.3.4.
~3.8.5.5 Structural Acceptance Criteria )
t The foundations of all Category I buildings are designed to meet the same structural acceptance criteria as the buildings themselves. These criteria are discussed in paragraphs 3.8.1.5, 3.8.3.5 and 3.8.4.5. The limiting conditions for the foundation medium together with a comparison actual capacity and estimated structure loads are found in paragraphs 2.5.4.10 )
and 2.5.4.11. .
3.8.5.6 Materials Quality control and special Construction Techniques The foundations and concrete supports are ccnstructed of '"
concrete using proven methods common to heavy industrial construction. For further discussion, refer to paragraph 3.8.3.6.
3.8.5.7 Testing ard Inservice Surveillance Requirements Testing and inservice surveillance are not required and are not J ,
i planned for foundations of structures or for concrete supports.
A discussion of the test program which serves as the basis for the Soils Investigation and Foundamion Report in Chapter 2, appendix 2A.
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VNP TABLE OF CONTENTS
' CHAPTER 15 Title Page Section 15.1-1 15 ACCIDENT ANALYSIS 15.1 CONDITION I - NORMAL OPERATION AND 15.1-1 OPERATIONAL iRANSIENTS 15.2~ CONDITION II - FAULTS OF MODERATE 15.2-1 FREQUENCY
.{. 15.2-1 15.2.8 LOSS OF NORMAL FEEDWATER 15.2.8.1 Identification of Causes and Accident 15.2-1 Description Analaysis of Effects and Consequences 15.2-2 15.2.8.2 15.2.9.4 Environmental Consequences of a Posutulated 15.2-4 Loss of AC Power to the Station Auxiliaries 15.2-6 15.2.14 REFERENCES 15.3-1 CONDITION III - INFREQUENT FAULTS.
( 15.3 15.3.5.3 En"ironmental Consequences of a Postulated Waste Gas Decay Tank Rupture 15.3-1 15.3-4 15.3.5.4 References 15.4-1 15.4 CONDITION IV - LIMITING FAULTS 15.4.1.3 Environment al Consequences of a Postulated S8 Loss of Coolanc Accident S8 15.4-2 (see Sub-section 6.2. C3T 15.4.1.3 Environmental Const!quences of a Postulated 15.4-2
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Loss of Coolant Accident 15.4.2.4 Environmental Consequences of a Postulated Steam Line Break 15.4-17 e
15.4.3.4 Environmental Consequences of a Postulated
'77 Steam Generator Tube Rupture 15.4-19
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15.4-22 g 15.4.5 FUEL EANDLING ACCIDENT Environmental Consequences 15.4-22 15.4.5.4 C
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S8 gg POST CONSTRUCTIOF PERMIT SUPPLEMENTA3Y INFORMATION - DECEMBER 30, 1980 1(. 7
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s TABLE OF CONTENTS (Continued)
Section Title Page 15.4.6 ENVIRONMENTAL CONSEQUENCES OF A POSTULATED ROD EJECTION ACCIDENT 15.4-24
- 15.4.6.1 Model 15.4-24 15.4.6.2 Assumptions 15.4-25 i
15.4.6.3 Ultra Conservative Analysis 15.4-28 15.4.6.4 Results 15.4-29
)y 15.
4.7 REFERENCES
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TABIE OF CONTENTS (Continued)
LIST OF TABLES Table Title Page 15.3-1 Waste Gas Decay Tank Inventory 15.3-2 15.4-11 Parameters Used in LOCA Analysis S8 15.4-5 lS8 15.4-11 Parameters Used in LOCA Analysis S6 15.4-5 15.4-llA Iodine and Noble Gas Inventory in Reactor Core and Fuel Rod Gaps S6 15.4-6 15.4-llB Noble Gas and Iodine Inventory in the Containment Atmosphere Immediately After LOCA and Available for Leakage S6 15.4-7 15.4-llc Activity Releases to Atmosphere from Loss o:' Coolant Accident S6 15.4-9 15.4-11 Parameters Used in LOCA Analysis 15.4-5 15.4,lla Parameters Used to Determine Hydrogen Purging Activity Release 15.4-13 15.4-11b Doses From Containment Purging to Control Hydrogen 15.4-14 15.4-12A Containment Penetration Locations Relative to the Penetration Area Filter Boundary S8 15.4-llb l
15.4-12B Deleted S8 15.4-11c S8 15.4-12 Potential Offsite Doses Due to LOCA Vogtle Nuclear Plant S8 15.4-10 15.4-12 Loss of Coolant Accident 15.4-10 15.4-12A LOCA Cases Analyzed 15.4-llb 15.4-12B LOCA Off-Site Doses 15.4-11c
( 15.4-13 Parameters Used in Rod Ejection _ _ _f" . . _ ,
i Accident Analysis 15.4-26 i
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i i S8 15-111 POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION ' ECEMBER 30, 1980
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VNP t TABLE OF CONTENTS (Contin'aed) )
LIST OF FIGURES Ficure Title 15.2-29 Transient Response Following a Loss of f Normal Feedwater 15.2-52 Loss of AC Power - Thyroid Doses 15.2-53 Loss of AC Power - Whole Body Gamma and Beta Doses j 15.4-59 Deleted S8 S8 s 15.4-60 Deleted S8 15.4-59 Schematic of Leakage Path for 90%
Mixing Case 15.4-60 Schematic of Leakage Path for Short Circuit Case 15.4-61 Steam Line Break - Whole Body Ga=ma Beta Dose s
15.4-62 Steam Line Break Accident - Thyroid Doses 15.4-63 Steam Generator Tube Rupture - Whole Body Gamma and Seta Doses 15.4-64 Steam Generator Tube Rupture - Thyroid Doses >
15.4-65 Rod Ejection Accident Thyroid Dose - No Core Melting 15.4-66 Rod Ejection Accident Beta Doses - No Core Melting
)
15.4-67 Rod Ejection Accident Gamma Doses -
No Core Melting 15.4-68 Rod Ejection Accident Thyroid Doses -
0.25% Core Melting f 15.4-69 Rod Ejection Accident Beta' Dose -
0.25% Core Melting 15.4-70 Rod Ejectior Accident Gamma Dose -
0.25% Co.c r.elting i
S8 15-iv POST CONSTRUCTION PEPMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
.m.
p I
VNP
- s l
I
( 15.4 CONDITION IV -- LIMITING FAULTS Unless otherwise indicated, the applicable data for this section are presented in RESAR-3, section 15.4. The information for the four-loop plant is applicable to the VNP.
f
(_ NOTE: The following paragraph should be read in place of the first paragraph on page 15.4-23 of RESAR-3.
Figure 15.4-29 shows the required total peaking factor at the license application power rating to meet the AEC Interim g- Acceptance Criteria for ECCS as a function of calculated peak containment pressure for a double ended cold leg break. The
\- calculated peak containment pressure for the double ended cold leg break is reported in section 6.2.1. Using this pressure in figure 15.4-29, the maximum allowable linear power and total peaking factor at a core power of 3411 MWt for which the ECCS will meet the AEC Interim Acceptance Criteria can be obtained.
Note that the peaking factor in figure 15.4-29 is based on ECCS analysis at containment pressure values of 90 percent of the calculated peak containment pressure for blowdown and 80 percent of the peak for reflood, as specified by the AEC Interim Policy Statement.
NOTE: The first paragraph in RESAR-3, page 15.4-42 should r; read as follows:
s Core Power and Reactor Coolant System Transient Figure 15.4-37 shows the reactor coolant system transient and core heat flux following a main steam pipe rupture (complete severance of a pipe) outside the containment, downstream of the flow measuring nozzle at initial no load condition (case a) .
The break assumed is the largest break which can occur anywhere outside the containment either upstream or downstream of the isolation valves. Offsite power is assumed available such that full reactor coolant flow exists. The transient shown assumes a steam release from only one steam ge erator. Should the core be critical at near zero power when the rupture occurs the
(" initiation of safety injection by high differential pressure T. between any steam line and the remaining steam lines or by high steam flow signals in coincidence with either low-low reactor
'l coolant system temperature or low steam line pressure will trip the reactor. Steam release from more than one steam generator F-will be prevented by automatic trip of the bi-directional isolation valves in the steam lines by the high steam flow signals in coincidence with either low reactor coolant system temperature or low steam line pressure. Even with the failure of one valve, release is limited to no more than 5 seconds for i
4 ,
S8 15.4-1
h VNP the other steam generators while the one generator blows down. }
The steam line isolation valves are designed to be fully closed in less than 5 seconds.
The following paragraphs on environmental consequences of postulated accidents are added to RESAR-3, section 15.4.
15.4.1.3 Radiological Consecuences of a Postulated Loss Of Coolant Accident (See Subsection 6. 2.1. 3) .
The results of analyses presented in this section demonstrate that the amount of radioactivity released to the environment in the event of a loss-of-coolant accident results in doses )
S8 less than the guideline values specified in 10 CFR 100 and less than or comparable to the calculated doses acceptable at the construction permit state for Plant Vogtle.
The analyses performed is based on Regulatory Guide 1.4, Revision 2. The parameters used for this analysis are listed in table 15.4-11. In addition, an evaluation of the offsite dose resulting from purging the containment for hydrogen con-trol, and an evaluation of the offsite doses resulting from recirculation loop leakage, are presented in subsection 15.4.1.3.9.
15.4.:.3.1 Fission Product Release to the Containment
}
The calculation of potential offsite doses re.alting frcm a loss of coolant accident are based on the conservative fission product releases recommended by Regulatory Guide 1.4.
One hundred percent of the core noble gas inventory and 25 per-cent of the core iodine inventory is assumed to be immediately available for leakage from the primary containment. Ninety one percent of the halogen activity available for release is assumed to be in elemental form, 4 percent in methyl form, and 5 percent in particulate form. The total core noble gas and iodine inventories are given in table 15.4-11A, while the activity in the containment atmosphere immediately following the LOCA and available for leakage is shown in table 15.4-11B. )
15.4.1.3.2 Containment Model The activity released from the containment was calculated with a two-volume model to represent sprayed and unsprayed ragions of the containment.
O . -L. -
)
N' h
y S8 15.4-2 l POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980 b .:.a_.__.a_ -._%~ ~ ~ -m m .. - -h m- - -- ;-e - -
3._.g. -, _m = . ~ ~ - - - - - - - - - - - - ~~- -
VNP h
( ) The initial instantaneous release of fission products to the containment is assumed to be the only source of activity in the containment. The activity change with respect to time in each of the two well-mixed volumes is described by the following equations:
O ky ~
,/ da y 7 Q 12 Q 21 i
dt
=-
)
7_
A 1, ]. a1 V
a 1+V 2 a 2 (15. 4-1) 7
]=1 L.
- 1. k 2
- da 2 i
0 21 0 12
"~
dt -
A 2,j"2 ~
V 2
2+V y ay US.4-2) 3=1 where:
p a,ay are the fission product activities of'a species 2
in volumes 1 and 2, respectively (Curies)
Q12' 02"1 are the transfer rates between the two volumes t
(ft3/hr) .
( .i
-* V,V y 2 are the volumes of the unsprayed and sprayed regions of the containment, respectively (ft3)
, A 1,j, A
2,3 are the removal coefficients due to the jth removal process in volume 1 and 2, respectively (hrs-1) '
k,k are the number of removal processes applicable l 2 j in val ;7ec 1 -nd E c- rr-tively The transfer rate between the sprayed and unsprayed regions was assumed to be limited to the forced convection induced by the fan-cooler units. The flow rate per fan-cooler unit, and the
'1 number of units assumed in operation are summarized in table r
/ 15.4-11. This assumed minimum flow rate conservatively neglects the effect of natural convection, steam condensation, and
' ' ' ~ diffusion although these effects are expected to enhance the -
mixing rate between the sprayed and unsprayed volumes.
v
.~
T I
j :
)
S8 15.4-3 POST CONSTRUCTION PERMIT-SUPPLEMENTARY INFORMATION - DECEMBER' 30, 1980
m.~. ,._-,...y.m.n~,..,. , , . - . _ ,, . ,.. _ ,__. ,,.... _ __ .._, _
VNP t
3 15.4.1.3.5 Penetration Rooms and Associated Safety Grade Filtration and Exhaust Systems The majority of penetrations through the containment w311 lS8 occur within the electrical and piping penecration rooms.
sg
-These rooms are contained within the fuel handling building,
& auxiliary building, and control building and are serviced by J .the safety-grade electrical and piping penetration room filtration and exhaust systems, and the rooms are at a negative pressure when the filtration systems are in operation.
Containment penetration locations relative to the penetration S8 y area filter boundary are listed in table 15.4-12A. All ECCS 2 piping which potentially recirculates contaminated fluids g-
/ following an accident are completely routed within the ecntain-ment or the piping i2netration room. Any airborne radio-
~
activity released as a result of leakage from these piping S8 systems outside the containment will be collected and filtered prior . to release to the envirc nment.
30 In the calculation of offsite doses, no credit has been taken for filtration of airborne containment leakage.
15.4.1.3.6 Deleted 15.4.1.3.7 Deleted 3,
5 is G
b c N E 2:
. m -..~ , ,
L-p
,4 -
- ( 332 f
3
)
s
- S8 15.4-4a POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
VNP l Table 15.4-11 PARAMETERS USED IN LOCA ANALYSES REGULATURY GUIDE 1.4 ANALYSIS Core thermal power 3565 MWt 0 3 Containment free volume 2.75 x 10 ft 6 3 Sprayed containment free volume 2.145 x 10 ft 5 3 Unsprayed containment free volume 6.05 x 10 ft Primary containment deck fan flow rate 85,000 cfm Number cm deck fans assumed operating 4 of 8 Activity released to containment and available for release Noble gases 100% of core inventory
- Iodines 25% of core inventory Form of iodine activity in primary containment available for release .
Element iodine 91%
Methyl iodine 4%
- Particulate iodine 5%
Containment leak rate assumed in conjunc- .20% per day tion with implementation of the modified (0-24 hours) S8 enclosure buidling design 0.10% per day (1-30 days)
Containment spray removal
-1 h Elemental iodine 10.0 hr 9 S8 0.4 hr-1
~
f Particulates L
Meteorology See Table 15.B-2 l
i
- The activity available for release reflects the assumption of 50 percent plateout on centainment surfaces.
t h S8 15.4-5 7
L POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980 w . _ sac _a&d49 . w m.w a c m a z y w w .u w - g en e w .47-
_,_v-__.___-.. _., m, I. . VNP E
?;
f Table 15.4-11A f
E IODINE AND NOBLE GAS INVENTORY IN REACTOR CORE b
CORE ACTIVITY
... ISOTOPE (Curies 1 I-131 8.80 x 10 7 8
I-132 1.34 x 10
'. 8 I-133 1.97 x 10 8 >
I-134 2.31 x 10 8
I-135 1.79 x 10 Xe-131m 5 6.68 x 10 0
Xe-133 2.03 x 10 6
Xe-133m 5.16 x 10
, -- Xe-135 . 5.55 x 10 a,-
7 1 Xe-135m 5.46 x 10 8
Xe-138 1.79 x 10 Kr-83m 1.64 x 10 Kr-85 9.99 x 10'
~
. . _ . Kr-85m 3.95 x 10 Kr-87 7.59 x 10 0
Kr-88 1.08 x 10 )
0
, Kr-89 1.40 x 10 *
., m at p -
_- ,, ~.w eq : ; - ; -. _
C
)
7 S8 15.4-6
. POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30,-1980 ty:T~ ~~4y _ .
_. . - , ~ ym g . ._. ; - . ._ .
. .-,--..,..,_,,r,.,.r.~,-~ -_ . - . _ . . - . .
VNP The effectiveness of the secondary containment is shown by the results obtained for both the realistic and the hypothetical !
cases analyzed, which describe all possible leakage paths through the annulus and penetration room. The results, given in table 15.4-12, also demonstrate that the effectiveness of 3
o k
l F
b p _
g , _, .
F g -
Is-E C #,
f 5
+
b , . , _ . ._ _ - .-
1.
p -
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[ '
!~
1 I
u i -
if
'Ir S8 '15.4-9a
,i1 POST CONSTRUCTION PERMIT SUPPIE.MENTARY INFORMATION - DECEMBER 30, 1980 3 -
A- - - -
~
..- l. ' x
. n, .+ - . .
~
m l 0 a
Q Table 15.4-12 E! POTENTIAL OFFSITE DOSES DUE TO LOCA y VOGTLE NUCLEAR PLANT C
o Present Calcuation II Z SER Calculation (1) (No Leakage m 10 CFR 100 (Original EB, Filtration Credit M
-L Guidelines 0.3%/ Day) 0.2%/ Day) 5 y.
F f m: Exclusion Area Boundary l' 0w- Thyroid (rem) 300 122 111 h4 m .Whole Body (rem) 25 7 3 y w *- Low Population Zone a
S8 5
- m. i
[ Thyroid (rem) 300 70 73 o
Whole Body (rem) 25 8 1 a
.8
, Notes:
(1) These are the NRC Safety Evaluation Report doses at the construction permit stage with the original enclosure building, a 0.3% per day con-
',E- tainment leak rate, and 37.5% bypass leakage.
IIl, n. h m
(2) These are doses for the proposed design utilizing the NRC calculational methodology and a 0.2% per day-containment leak rate with no leakage i . 38 -
filtration credit.
..p.
O O
v v ,-
s W -
.m. - _ _ _ _ m.___ . ,_y-. m. . , . . . . . . . - _.
VNP s
s
} 15.4.1.3.9 Environmental Consequences of Containment Purging to Control Hydrogen After a Loss-of-Coolant Accident Post lost-of-coolant accident purging provides a backup method to containment recombiners for controlling the potential hydrogen accumulation in the containment. This analysis is based on AEC Safety Guide 7 and the Westinghouse model for hydrogen production and accumulation discussed in Section 15.4.1.2 of RESAR-3. (See figure 6.2-14. )
The Post-LOCA system normally utilizes a differential pressure
't between'the containment and the outside atmosphere to permit
) purging. This analysis is based on a pressure of 3 psig in the containment. The containment is pressurized to 3 psig with diluent air when the hydrogen reaches 3 volume percent after the~ loss-of-coolant accident. The hydrogen concentration is
. reduced by this pressurization. Purging is thus delayed until the hydrogen concentration in the containment has once again increased to 3 volume percent. It is assumed that purging is continuous. The 3 percent hydrogen level was selected as the point of starting the purge because of the following factors:
A. This level allows a sufficient margin of safety below
- . the lower flammability limit of 4.1 percent.
. ~..
[ B. It provides a sufficient margin so that purging could be delayed a few days if desired. With neither containment purging nor recombiner operation the hydrogen generation rate is sufficiently low so more r
t 21 .,
- s. .
I
- 4
- f E, ,,3
- y. , u. - , ,
y ;.wi 1'
).
fd n Y i'l'l .
A V . -
58 15.4-11a POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980 t ,;r. -
s -c.g.f-M 56.h ld2Og . , - ,~,.- y A ,[ , , , , _ _ ,
VNP TABLE 15.4-12A )
CONTAINMENT PENETRATION LOCATIONS RELATIVE TO THE PENETRATION AREA FILTER BOUNDARY PenetratiLn No. System Location 1 Main steam Outside 2 Main steam Outside 3 Main steam Outside
)
4 Main steam Outside 5 (spare) Inside 6 (not used) -
7 SG blowdown Inside 8 SG blowdown Inside 9 SG blowdown Inside 58 10 SG blowdown Inside ,
llA Chemical addition Inside 11B SG sample Inside llc SG sample Inside 12A Chemical addition Inside 12B SG sampic Inside 12C SG sample Inside 13A Contain, air radioactivity Inside ;
13B Contain. air radioactivity Inside 14 (spare) Inside
. 15 Refueling cavity purification Inside h 16 (spare) Inside i
17 (spare) Inside 18 Feedwater Outside
- k S8 15.4-llb POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980 k~ m - _ _ _ _
I - ..
VEP TABLE 15.4-12A (Continund)
Penetration No. System Location 19 Feedwater Outside 20 Feedwater Outside 21 Feedwater Outside 22 Demin. water Inside
/ 23 (spare) Inside 24 RCS sample Inside 25 (spare) Inside 26 (not used) --
27 (not used) -
28 ACCW Outside 29 ACCW Outside 30 Safety injection Inside S8 31 Safety injection Inside 32 Boron injection Inside 33 Safety injection Inside 34 Containment spray Inside 35 Containment spray Inside 36 RHR recirculation Inside f
( 37 RER recirculation Inside 38 Containment spray Inside 39 Containment spray Inside 40 Fire protection -
.Inside
{
41 Accumulator test & drain Inside 42 Accumulator N 2 Outside 43 NSCW Inside S8 15.4-llb-1 POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMbrR 30, 1980
VNP l
TABLE 15.4-12A (Continued) \
9 Penetration No. System Location 44 NSCW Inside 45 NSCW Inside 46 NSCW Inside 7
r 47 (spare) Inside 48 Letdown Inside ,
49 Excess letdown Inside 50 Charging Inside 4
51 RCP seal water Inside 52 RCP seal water Inside 53 RCP seal water Inside 54 RCP seal water Inside
[
S8' 55 (spare) Outside }
t 56 RER Inside
- 57 RER Inside 58 RER Inside 59 RER Inside 60 RER Inside 61 (spare) Inside 62- PRT sample Inside )
63 PRT makeup Inside 64 (spare) Inside 65 (not used) --
g 66 (spare) Inside 67A Pressurizer steam sample Inside b 67B Pressurizer licuid sample Inside }
S8 15.4-ilb-2 POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
m , - - m .,m - z-,m.- m ,--- - ,.. . . - - .- - - - - - - - . - . . . - ,
VNP e
( ,
TABLE 15.4-12A (Continued)
I Penetration No. System Location 68 Leak rate test Inside "s .
- c 69A Chemical addition Inside
.A 69? Chemical addition Inside
( 69C Containment pressure Inside 70A Contain. H 2 monitor Inside f '7 70B Contain. H 2 monitor Inside j 70C Contain. pressure Inside 71A, Contain. H 2 m nitor Inside 71B Contain. H 2 m nitor Inside 71C Contain, pressure Inside 72A Accumulator sample Inside 72B Accumulator sample 80 p Inside 73A Accumulator sample Inside 73B Accumulator sample Inside 74 (not used) -
75 (spare) Inside i 76 (spare) Inside 77 RCDT pump Inside i
- \ y
( 81 Instrument air Inside r
0 82 (spare) Inside 56 1; 83 Contain. purge Outside p
84 Contain. purge Outside
(
E S8 15.4-llb-3 v.
POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
~ ~
is: . a ,
- f. VNP h
i h
e I
TABLE 15.4-12A (Continued) F f Penetration No. System Location f;
85 A&B (spare) Inside
- BSC Contain. pressure Inside =
86 (spare) Inside 87 Leak rate test Inside 88 (spare) Inside
[
[ 89 Fuel transfer tute Inside 90 (spare) Inside L 91 NSCW Inside
! 92 NSCW Inside
?-
3 S8 93 NSCW Inside 94 NSCW Inside 95 NSCW Inside
100 Post LOCA purge Outside Equip. hatch Outside i
Personnel lock Outside Escape lock Outside Electrical penetrations Inside
- ~c S8 15.4-llb-4 POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
VNP Table 15.4-12B Deleted lS8
.~
t l
l l
1 S8 15.4-11c l
POST CONSTRUCTION PERMIT SUPPIEMENTARY INFORMATION - DECEMBER 30, 1980 L
VNP than 30 days are required for the hydrogen concen- }
tration in the containment to increase from 3 percent to 4.1 percent.
C. The optimum starting time for the purge, from the standpoint of minimizing the doses, is the latest time. )
The purge rate was selected to match the rate of hydrogen generation at the time of initiation of the purge. The hydrogen concentration in the containment will slowly decrease from 3 percent as purging centinues at a constant rate, as the production rate decreases with time. The required continuous purge rate is obtained from the hydrogen production rate at the )
time of initiation of purging.
The dose analysis is based on the activity released from the containment after the time of the postulated LOCA until all the activity in the containment is either removed or released. The infinite-time thyroid, whole body beta, and whole body gamma doses at the site boundary and the low population zone due to activity release from containment leakage following the postulated LOCA are computed using the core activity release model described in paragraph 15.4.1.3.1. Then the analysis is repeated except that the doses are based on uutivity released from both containment leai: age and purging. The doses due to purging are then determined by subtraction df the doses due to '
}
containment leakage from the doses due to both containment and purging. The parameters used to compute the activity releases from containment leakage and from purging are given in tables 15.4-11 and 15.4-11A. .
The dose models discussed in appendix 15B, including the atmo-spheric dilution factors given in table 15.B-2, are used in determining doses.
In the evaluation of doses from activity released to the atmosphere after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (30 days), annual average diffusion factors at the site boundary and low population zone, 2.0 x 10-6 sec/m3 and 4.2 x 10-7 sec/m3 respectively, are used. )
The whole body gamma, whole body beta, and thyroid doses due to containment purging at the site bou.ndary and low population zone are given in table 15.4-llB.
The whole body gamma, whole body beta, and thyroid doses due to g
containment purging to control hydrogen were analyzed using.the ]
assumptions outlined in AEC Safety Guide 7, " Control of S8 15.4-12 POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
%- .. . = - adL.w a.m w -xmm =m" :w c -
Yh?
( '
s
(. ' .
( i.-
This figure has been deleted by Supple:nent 8. lS8 s
- (-
P r
L
[ i.
4
( I
'i
- Figure 15.4-59
Yh7 This figure has been deleted by Supplement 8. lSS i
1 l ;;
(
1 I
l l Figure 15.4-60 t
i POST-cot 2STRUCTION _ PERM _I_T SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
VNP k a
- s. TABLE OF CONTENTS APPENDIX 15B Section Title Page i
ISB DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B.1 INTRODUCTION S6 15B-1 15B.2 ASSUMPTIONS 56 15B-1 3'
15B.3 S6 15B-1 GAMMA DOSE 15B.4 THYROID INHALATION DOSE S6 15B-2 15B.5 REFERENCES S6 15B-3 f
15B DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15B.1 INTRODUCTION 15B-1 15B.2 ASSUMPTIONS 15B-1
) 15B.3 G?AMA AND BETA DOSE 15B-2 15B.4 THYR ID INHALATION DOSE 15B-3 15B.5 REFERENCES 15B-6 LIST OF TABLES Table Title Page 15B-1 PHYSICAL DATA FOR ISOTOPES S6 15B-4 15B-2 ACCIDENT ATMOSPHERIC DILUTION FACTORS (X/Q) AT EXCLUSION AREA
~
gg
/ BOUNDARY AND LOW POPULATION ZONE FOR THE VOGTLE NUCLEAR PLANT S8 15B-4 15B-2 ACCIDENT ATMOSPHERIC DILUTION p FACTORS (X/Q) AT EXCLUSION AREA e BOUNDARY AND LOW POPULATION ZONE
( ) FOR THE VOGTLE NUCLEAR PLANT S6 15B-5 15B-1 PHYSICAL DATA FOR ISOTOPES 15B-4 15B-2 ACCIDENT ATMOSPHERIC DILUTION FACTORS 15B-5 i
S8 ISB-i POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
' ' ' - ' ~
.s.&_._.._.___ -- --
. v. _ ..
-.--~~-n- ~. . ..c.~~,.-,,,-.,-
. .,n ,,
VNP Table 15B-2 ACCIDENT ATMOSPHERIC DILUTION FACTORS (X/Q)
- AT EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE FOR THE VOGTLE NUCLEAR PLANT
~ ,. ,.
~'
e Exclusion Low Population Area Boundary Zone Time Period (3218 Meters)
(hours) (1060 Meters) 0-2 1. 8 (-4) S8
- 3. 3 (-5) 0-8
- 2. 2 (-5) 8-24
- 9. 2 (-6 )
24-96 S8
- 2. 7 (-6)96-720 3
's
- X/O values, expressed in sec/m , are based on theThese
' guidelines set forth in Regulatory Guide 1.145.
'~
values are based on three years of site data.
-4 NOTE: 1. 8 (-4) = 1.8 x 10
. 't.
i $
r L-d
(.i '
8 -
F - ~ *. 4,',
r
.f*
j:-
'E- S8 15B-5 30, 1980
. POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER
{
VNP i
,7-
) TABLE OF CONTENTS (Continued)
Section Title Page 1
5
~~
3.13 CONTAINMENT POLAR CRANE 16.3.13-1
'l
/ 3.13.1 APPLICABILITY 16.3.13-1 3.13.2 OBJE^ .'IVE 16.3.13-1 3.13.3 SPECIFICATION 16.3.13-1 N
i 3.13.4 BASIS 16.3.13-1 4 SURVEILLANCE REQUIREMENTS 16.4.l-1 4.1 OPERATIONAL SAFETY REVIEW 16.4.1-1
~
4.1.1 APPLICABILITY 16.4.1-1 4.1.2 OBJECTIVE 16.4.1-1 4.1.3 ;5PECIFICATION 10.4.1-1 4.1.4 BASIS 16.4.1-1 4.1.4.1 Check -
16.4.1-1 4.1.4.2 Calibration 16.4.1-7 4.1.4.3 Testing 16.4.1-7 4.2 PRIMARY SYSTEM SURVEILLANCE 16.4.2-1 4.2.1 APPLICABILITY 16.4.2-1 4.2.2 OBJECTIVE 16.4.2-1
, 4.2.3 SPECIFICATION 16.4.2-1
+
i 4.2.4 BASIS 16.4.2-2 4.3 PRIMARY SYSTEM TESTING 16.4.3-1 4.3.1 APPLICABILITY 16.4.3-1 4.3.2 OBJECTIVE .16.4.3-1 4.3.3 SPECIFICATION 16.4.3-1 t i
)
58 16-vii POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980 ywwa.w a, a:.-+ _ u .- .s h w ~,s,;. aaa : - - _
_. _ _ m V61 TABLE OF CONTENTS (Continued) )
Section Tit.1e Page 4.3.4 BASIS 16.4.3-1 4.3.5 REFERENCE 16.4.3-2 4.4 CONTAINMENT 16.4.4-1 4.4.1 APPLICABILITY 16.4.4-1 1
4.4.2 OBJECTIVE 16.4.4-1 4.4.3 SPECIFICATION 16.4.4-1
- 4. 4.3.1 Integrated Leakage Rate Tests S8 16.4.4-1 SSl 4.4.3.1 Integrated Leakage Rate Tests 16.4.4-1 4.4.3.2 Local Leakage Rate Tests 16.4.4-4 4.4.3.3 Isolation valve Functional Tests 16.4.4-6 4.4.3.4 Annual Inspection 16.4.4-6 4.4.3.5 Containment Modifications 16.4.4-6 4.4.4 BASIS 16.4.4-6 4.5 EMERGENCY CORE COOLING SYSTEM, CONTAINMENT COOLING SYSTEM, CONTAINMENT SPRAY SYSTEM, PENETRATION ROOM, CONTROL ROOM, AND ENCLOSURE BUILDING FILTRATION SYSTEM TESTS 16.4.5-1 4.5.1 APPLICABILITY 16.4.5-1 4.5.2 OBJECTIVE 16.4.5-1 4.5.3 SPECIFICATIONS 16.4.5-1 4.5.3.1 System Tests 16.4.5-1 4.5.3.2 Component Tests 16.4.5-4 h 4.5.4 BASIS 16.4.5-5 h 4.5.5 REFERENCE 16.4.5-7 b
S8 16-viii POST CONSTRUCTION PERMIT SUPPLEMENTARY.INFORMATION - DECEMBER 30, 1980
TECHNICAL SPECIFICATIONS t
4.4 CONTAINMENT VNP 4.4.1 APPLICA.BILITY Applies to initial and periodic reactor building leakage testing.
4.4.2 OBJECTIVE To verify that leakage from the containment is maintained within allowable limits.
4.4.3 SPECIFICATION 4.4.3.1 Integrated Leakage Rate Tests 4.4.3 SPECIFICATION A. Calculated Peak Pressure Leakage Rate In conjunction with implementation of the modified enclosure building design, the maximum allowable integrated leakage rate (L ) from the contain-ment at the 43 psig calcula,ted peak containment S8 internal pressure (P ) shall not exceed 0.20 percent of weight of Pthe original content of containment air at that pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B. Testing at Reduced Pressure.
The periodic integrated leak rate test may be performed at a test pressure, Pt, of not less than 21 psig. The maximum allowable test leakage rate, t
Lt, is determined as follows:
- 1. Prior to reactor operation the initial value of i the integrated leakage rate of the containment
! shall be measured at the maximum calculated l pressure Pp and at the reduced test pressure (P )t l to be used in the pericdic integrated leakage l rate tests. The leakage rates thus measured shall be identified as L pm and L tm respectively.
- 2. L p shall be equal to L a f#
L pm, values ofi fLtm \ below 0.7 and greater than 0.3 (Lpgj l
I S8 16.4.4-1 POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 198d
TECHNICAL SPECZFICATIONS VNP p -1/2 )
- 3. L t shall be equal to L a f#
P
- ~ ~
values of above 0.7.
pm
- 4. If L tm/Lpm is less than 0.3, the initial f integrated test results shall be subject to review by the AEC to establish an acceptre :e ,
value of L t*
The test is acceptable if the measured leakage rate L m t does not exceed the maximum allowable leakage rate Lt )
determined above.
C. Conduct of Tests
- 1. The test duration shall be at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless experience from at least two prior tests providen evidence of the adequacy of a shorter
, test duration.
- 2. Test accuracy shall be verified by supplementary means, such as measuring the quantity of air required to return to the starting point or by
- imposing a known leak rate to demonstrate the validity of measurements.
- 3. Containment isolation valves for the purpoae of the test shall be closed by the means provided for normal operation of the valves, without preliminary exercises or adjustment.
D. Frequency of Test After the initial preoperational leakage rate test, two integrated leakage rate tests shall be made at approximately equal intervals between each major shutdown for inservice inspection (to be made at 10-year intervals). In addition, an integrated test )
shall be made at each 10-year interval, coinciding with the inservice inspection shutdown. The test shall coincide with a shutdown for major fuel reloading. The leakage rate measured during these tests shall be called L tm
- l' S8 16.4.4-2
,B POST CONSTRUCTION PERMIT SUPPLEMENTARY INFORMATION - DECEMBER 30, 1980
5 INSTRUCTION SHEET f SUPPLEMENT NO. 8
. ALVIN W. VOGTLE NUCLEAR PLANT I
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